  Dec. 30, 2019

      Title 10
      Energy
      Parts 51 to 199
      Revised as of January 1, 2020
      Containing a codification of documents of general applicability and future effect
      As of January 1, 2020

        Published by the Office of the Federal Register National Archives and Records Administration as a Special Edition of the Federal Register

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      Table of Contents
      Page

        Explanation
        v

        Title 10:

          Chapter I—Nuclear Regulatory Commission (Continued)
          3

        Finding Aids:
        Table of CFR Titles and Chapters
        887
        Alphabetical List of Agencies Appearing in the CFR
        907
        List of CFR Sections Affected
        917

      Cite this Code: CFR

      To cite the regulations in this volume use title, part and section number. Thus, 10 CFR 51.1 refers to title 10, part 51, section 1.

      Explanation
      The Code of Federal Regulations is a codification of the general and permanent rules published in the Federal Register by the Executive departments and agencies of the Federal Government. The Code is divided into 50 titles which represent broad areas subject to Federal regulation. Each title is divided into chapters which usually bear the name of the issuing agency. Each chapter is further subdivided into parts covering specific regulatory areas.
      Each volume of the Code is revised at least once each calendar year and issued on a quarterly basis approximately as follows:

        Title 1 through Title 16
        as of January 1
        Title 17 through Title 27
        as of April 1
        Title 28 through Title 41
        as of July 1
        Title 42 through Title 50
        as of October 1

      The appropriate revision date is printed on the cover of each volume.

        LEGAL STATUS
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        EFFECTIVE AND EXPIRATION DATES
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        OMB CONTROL NUMBERS

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        PAST PROVISIONS OF THE CODE
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        “[RESERVED]” TERMINOLOGY
        The term “[Reserved]” is used as a place holder within the Code of Federal Regulations. An agency may add regulatory information at a “[Reserved]” location at any time. Occasionally “[Reserved]” is used editorially to indicate that a portion of the CFR was left vacant and not dropped in error.

        INCORPORATION BY REFERENCE

          What is incorporation by reference? Incorporation by reference was established by statute and allows Federal agencies to meet the requirement to publish regulations in the Federal Register by referring to materials already published elsewhere. For an incorporation to be valid, the Director of the Federal Register must approve it. The legal effect of incorporation by reference is that the material is treated as if it were published in full in the Federal Register (5 U.S.C. 552(a)). This material, like any other properly issued regulation, has the force of law.

          What is a proper incorporation by reference? The Director of the Federal Register will approve an incorporation by reference only when the requirements of 1 CFR part 51 are met. Some of the elements on which approval is based are:
        (a) The incorporation will substantially reduce the volume of material published in the Federal Register.
        (b) The matter incorporated is in fact available to the extent necessary to afford fairness and uniformity in the administrative process.
        (c) The incorporating document is drafted and submitted for publication in accordance with 1 CFR part 51.

          What if the material incorporated by reference cannot be found? If you have any problem locating or obtaining a copy of material listed as an approved incorporation by reference, please contact the agency that issued the regulation containing that incorporation. If, after contacting the agency, you find the material is not available, please notify the Director of the Federal Register, National Archives and Records Administration, 8601 Adelphi Road, College Park, MD 20740-6001, or call 202-741-6010.

        CFR INDEXES AND TABULAR GUIDES

        A subject index to the Code of Federal Regulations is contained in a separate volume, revised annually as of January 1, entitled CFR Index and Finding Aids. This volume contains the Parallel Table of Authorities and Rules. A list of CFR titles, chapters, subchapters, and parts and an alphabetical list of agencies publishing in the CFR are also included in this volume.

        An index to the text of “Title 3—The President” is carried within that volume.

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        REPUBLICATION OF MATERIAL
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        INQUIRIES
        For a legal interpretation or explanation of any regulation in this volume, contact the issuing agency. The issuing agency's name appears at the top of odd-numbered pages.

        For inquiries concerning CFR reference assistance, call 202-741-6000 or write to the Director, Office of the Federal Register, National Archives and Records Administration, 8601 Adelphi Road, College Park, MD 20740-6001 or e-mail fedreg.info@nara.gov.

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        ELECTRONIC SERVICES

        The full text of the Code of Federal Regulations, the LSA (List of CFR Sections Affected), The United States Government Manual, the Federal Register, Public Laws, Public Papers of the Presidents of the United States, Compilation of Presidential Documents and the Privacy Act Compilation are available in electronic format via www.govinfo.gov. For more information, contact the GPO Customer Contact Center, U.S. Government Publishing Office. Phone 202-512-1800, or 866-512-1800 (toll-free). E-mail, ContactCenter@gpo.gov.

        The Office of the Federal Register also offers a free service on the National Archives and Records Administration's (NARA) website for public law numbers, Federal Register finding aids, and related information. Connect to NARA's website at www.archives.gov/federal-register.

        The e-CFR is a regularly updated, unofficial editorial compilation of CFR material and Federal Register amendments, produced by the Office of the Federal Register and the Government Publishing Office. It is available at www.ecfr.gov.

          Oliver A. Potts,

          Director,

          Office of the Federal Register

          January 1, 2020

      THIS TITLE
      Title 10—Energy is composed of four volumes. The parts in these volumes are arranged in the following order: Parts 1-50, 51-199, 200-499 and part 500-end. The first and second volumes containing parts 1-199 are comprised of chapter I—Nuclear Regulatory Commission. The third and fourth volumes containing part 200-end are comprised of chapters II, III and X—Department of Energy, chapter XIII—Nuclear Waste Technical Review Board, chapter XVII—Defense Nuclear Facilities Safety Board, and chapter XVIII—Northeast Interstate Low-Level Radioactive Waste Commission. The contents of these volumes represent all current regulations codified under this title of the CFR as of January 1, 2020.
      For this volume, Cheryl E. Sirofchuck was Chief Editor. The Code of Federal Regulations publication program is under the direction of John Hyrum Martinez, assisted by Stephen J. Frattini.

    10 CFR Ch. I (1-1-20 Edition)
    Nuclear Regulatory Commission

        Title 10—Energy
        (This book contains parts 51 to 199)

        Part

            chapter i—Nuclear Regulatory Commission (Continued)
          51

          CHAPTER I—NUCLEAR REGULATORY COMMISSION (CONTINUED)

          Editorial Note:
          Nomenclature changes to chapter I appear at 70 FR 69421, Nov. 16, 2005, and at 72 FR 33386, June 18, 2007.

        Part
        Page

          51
          Environmental protection regulations for domestic licensing and related regulatory functions
          5
          52
          Licenses, certifications, and approvals for nuclear power plants
          67
          53
          [Reserved]
          54
          Requirements for renewal of operating licenses for nuclear power plants
          158
          55
          Operators' licenses
          164
          60
          Disposal of high-level radioactive wastes in geologic repositories
          181
          61
          Licensing requirements for land disposal of radioactive waste
          216
          62
          Criteria and procedures for emergency access to non-federal and regional low-level waste disposal facilities
          243
          63
          Disposal of high-level radioactive wastes in a geologic repository at Yucca Mountain, Nevada
          252
          70
          Domestic licensing of special nuclear material
          295
          71
          Packaging and transportation of radioactive material
          345
          72
          Licensing requirements for the independent storage of spent nuclear fuel, high-level radioactive waste, and reactor-related greater than Class C waste
          402
          73
          Physical protection of plants and materials
          468
          74
          Material control and accounting of special nuclear material
          597
          75
          Safeguards on nuclear material—implementation of safeguards agreements between the United States and the International Atomic Energy Agency
          620
          76
          Certification of gaseous diffusion plants
          639

          81
          Standard specifications for the granting of patent licenses
          665
          95
          Facility security clearance and safeguarding of national security information and restricted data
          673
          100
          Reactor site criteria
          691
          110
          Export and import of nuclear equipment and material
          706
          140
          Financial protection requirements and indemnity agreements
          762
          150
          Exemptions and continued regulatory authority in Agreement States and in offshore waters under section 274
          834
          160
          Trespassing on Commission property
          848
          170
          Fees for facilities, materials, import and export licenses, and other regulatory services under the Atomic Energy Act of 1954, as amended
          849
          171
          Annual fees for reactor licenses and fuel cycle licenses and materials licenses, including holders of certificates of compliance, registrations, and quality assurance program approvals and government agencies licensed by the NRC
          866
          172-199
          [Reserved]

        Pt. 51
        PART 51—ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND RELATED REGULATORY FUNCTIONS

          Sec.
          51.1
          Scope.
          51.2
          Subparts.
          51.3
          Resolution of conflict.
          51.4
          Definitions.
          51.5
          Interpretations.
          51.6
          Specific exemptions.

            Subpart A—National Environmental Policy Act—Regulations Implementing Section 102(2)
            51.10
            Purpose and scope of subpart; application of regulations of Council on Environmental Quality.
            51.11
            Relationship to other subparts. [Reserved]
            51.12
            Application of subpart to ongoing environmental work.
            51.13
            Emergencies.
            51.14
            Definitions.
            51.15
            Time schedules.
            51.16
            Proprietary information.
            51.17
            Information collection requirements; OMB approval.

              Preliminary Procedures

              classification of licensing and regulatory actions
              51.20
              Criteria for and identification of licensing and regulatory actions requiring environmental impact statements.
              51.21
              Criteria for and identification of licensing and regulatory actions requiring environmental assessments.
              51.22
              Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review.
              51.23
              Environmental impacts of continued storage of spent nuclear fuel beyond the licensed life for operation of a reactor.

              determinations to prepare environmental impact statements, environmental assessments or findings of no significant impact, and related procedures
              51.25
              Determination to prepare environmental impact statement or environmental assessment; eligibility for categorical exclusion.
              51.26
              Requirement to publish notice of intent and conduct scoping process.
              51.27
              Notice of intent.

              scoping
              51.28
              Scoping—participants.
              51.29
              Scoping-environmental impact statement and supplement to environmental impact statement.

              environmental assessment
              51.30
              Environmental assessment.
              51.31
              Determinations based on environmental assessment.

              finding of no significant impact
              51.32
              Finding of no significant impact.
              51.33
              Draft finding of no significant impact; distribution.
              51.34
              Preparation of finding of no significant impact.
              51.35
              Requirement to publish finding of no significant impact; limitation on Commission action.

              Environmental Reports and Information—Requirements Applicable to Applicants and Petitioners for Rulemaking

              general
              51.40
              Consultation with NRC staff.
              51.41
              Requirement to submit environmental information.

              environmental reports—general requirements
              51.45
              Environmental report.

              environmental reports—production and utilization facilities
              51.49
              Environmental report—limited work authorization.
              51.50
              Environmental report—construction permit, early site permit, or combined license stage.
              51.51
              Uranium fuel cycle environmental data—Table S-3.
              51.52
              Environmental effects of transportation of fuel and waste—Table S-4.
              51.53
              Postconstruction environmental reports.
              51.54
              Environmental report—manufacturing license.
              51.55
              Environmental report—standard design certification.
              51.58
              Environmental report—number of copies; distribution.

              environmental reports—materials licenses
              51.60
              Environmental report—materials licenses.
              51.61
              Environmental report—independent spent fuel storage installation (ISFSI) or monitored retrievable storage installation (MRS) license.
              51.62
              Environmental report—land disposal of radioactive waste licensed under 10 CFR part 61.
              51.66

              Environmental report—number of copies; distribution.

              51.67
              Environmental information concerning geologic repositories.

              environmental reports—rulemaking
              51.68
              Environmental report—rulemaking.

              Environmental Impact Statements

              draft environmental impact statements—general requirements
              51.70
              Draft environmental impact statement—general.
              51.71
              Draft environmental impact statement—contents.
              51.72
              Supplement to draft environmental impact statement.
              51.73
              Request for comments on draft environmental impact statement.
              51.74
              Distribution of draft environmental impact statement and supplement to draft environmental impact statement; news releases.

              draft environmental impact statements—production and utilization facilities
              51.75
              Draft environmental impact statement—construction permit, early site permit, or combined license.
              51.76
              Draft environmental impact statement—limited work authorization.
              51.77
              Distribution of draft environmental impact statement.

              draft environmental impact statements—materials licenses
              51.80
              Draft environmental impact statement—materials license.
              51.81
              Distribution of draft environmental impact statement.

              draft environmental impact statements—rulemaking
              51.85
              Draft environmental impact statement—rulemaking.
              51.86
              Distribution of draft environmental impact statement.

              legislative environmental impact statements—proposals for legislation
              51.88
              Proposals for legislation.

              final environmental impact statements—general requirements
              51.90
              Final environmental impact statement—general.
              51.91
              Final environmental impact statement—contents.
              51.92
              Supplement to the final environmental impact statement.
              51.93
              Distribution of final environmental impact statement and supplement to final environmental impact statement; news releases.
              51.94
              Requirement to consider final environmental impact statement.

              final environmental impact statements—production and utilization facilities
              51.95
              Postconstruction environmental impact statements.

              final environmental impact statements—materials licenses
              51.97
              Final environmental impact statement—materials license.

              final environmental impact statements—rulemaking
              51.99
              [Reserved]

              NEPA Procedure and Administrative Action

              general
              51.100
              Timing of Commission action.
              51.101
              Limitations on actions.
              51.102
              Requirement to provide a record of decision; preparation.
              51.103
              Record of decision—general.
              51.104
              NRC proceeding using public hearings; consideration of environmental impact statement.

              production and utilization facilities
              51.105
              Public hearings in proceedings for issuance of construction permits or early site permits; limited work authorizations.
              51.105a
              Public hearings in proceedings for issuance of manufacturing licenses.
              51.106
              Public hearings in proceedings for issuance of operating licenses.
              51.107
              Public hearings in proceedings for issuance of combined licenses; limited work authorizations.
              51.108
              Public hearings on Commission findings that inspections, tests, analyses, and acceptance criteria of combined licenses are met.

              materials licenses
              51.109
              Public hearings in proceedings for issuance of materials license with respect to a geologic repository.

              rulemaking
              51.110
              [Reserved]

              Public Notice of and Access to Environmental Documents
              51.116
              Notice of intent.
              51.117
              Draft environmental impact statement—notice of availability.
              51.118
              Final environmental impact statement—notice of availability.
              51.119
              Publication of finding of no significant impact; distribution.
              51.120
              Availability of environmental documents for public inspection.
              51.121
              Status of NEPA actions.

              51.122
              List of interested organizations and groups.
              51.123
              Charges for environmental documents; distribution to public; distribution to governmental agencies.

              Commenting
              51.124
              Commission duty to comment.

              Responsible Official
              51.125
              Responsible official.
              Appendix A to Subpart A of Part 51—Format for Presentation of Material in Environmental Impact Statements
              Appendix B to Subpart A of Part 51—Environmental Effect of Renewing the Operating License of a Nuclear Power Plant

            Subpart B [Reserved]

          Authority:
          Atomic Energy Act of 1954, secs. 161, 193 (42 U.S.C. 2201, 2243); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); National Environmental Policy Act of 1969 (42 U.S.C. 4332, 4334, 4335); Nuclear Waste Policy Act of 1982, secs. 144(f), 121, 135, 141, 148 (42 U.S.C. 10134(f), 10141, 10155, 10161, 10168); 44 U.S.C. 3504 note.
          Sections 51.20, 51.30, 51.60, 51.80, and 51.97 also issued under Nuclear Waste Policy Act secs. 135, 141, 148 (42 U.S.C. 10155, 10161, 10168).
          Section 51.22 also issued under Atomic Energy Act sec. 274 (42 U.S.C. 2021) and under Nuclear Waste Policy Act sec. 121 (42 U.S.C. 10141).
          Sections 51.43, 51.67, and 51.109 also issued under Nuclear Waste Policy Act sec. 114(f) (42 U.S.C. 10134(f)).

          Source:
          49 FR 9381, Mar. 12, 1984, unless otherwise noted.

          Editorial Note:
          Nomenclature changes to part 51 appear at 80 FR 74980, Dec. 1, 2015.

          § 51.1
          Scope.
          This part contains environmental protection regulations applicable to NRC's domestic licensing and related regulatory functions. These regulations do not apply to export licensing matters within the scope of part 110 of this chapter or to any environmental effects which NRC's domestic licensing and related regulatory functions may have upon the environment of foreign nations. Subject to these limitations, the regulations in this part implement:
          (a) Section 102(2) of the National Environmental Policy Act of 1969, as amended.

          § 51.2
          Subparts.
          (a) The regulations in subpart A of this part implement section 102(2) of the National Environmental Policy Act of 1969, as amended.

          § 51.3
          Resolution of conflict.
          In any conflict between a general rule in subpart A of this part and a special rule in another subpart of this part or another part of this chapter applicable to a particular type of proceeding, the special rule governs.

          § 51.4
          Definitions.
          As used in this part:

            Act means the Atomic Energy Act of 1954 (Pub. L. 83-703, 68 Stat. 919) including any amendments thereto.

            Commission means the Nuclear Regulatory Commission or its authorized representatives.

            Construction means:
          (1) For production and utilization facilities, the activities in paragraph (1)(i) of this definition, and does not mean the activities in paragraph (1)(ii) of this definition.
          (i) Activities constituting construction are the driving of piles, subsurface preparation, placement of backfill, concrete, or permanent retaining walls within an excavation, installation of foundations, or in-place assembly, erection, fabrication, or testing, which are for:
          (A) Safety-related structures, systems, or components (SSCs) of a facility, as defined in 10 CFR 50.2;
          (B) SSCs relied upon to mitigate accidents or transients or used in plant emergency operating procedures;
          (C) SSCs whose failure could prevent safety-related SSCs from fulfilling their safety-related function;
          (D) SSCs whose failure could cause a reactor scram or actuation of a safety-related system;
          (E) SSCs necessary to comply with 10 CFR part 73;
          (F) SSCs necessary to comply with 10 CFR 50.48 and criterion 3 of 10 CFR part 50, appendix A; and
          (G) Onsite emergency facilities (i.e., technical support and operations support centers), necessary to comply with 10 CFR 50.47 and 10 CFR part 50, appendix E.
          (ii) Construction does not include:

          (A) Changes for temporary use of the land for public recreational purposes;
          (B) Site exploration, including necessary borings to determine foundation conditions or other preconstruction monitoring to establish background information related to the suitability of the site, the environmental impacts of construction or operation, or the protection of environmental values;
          (C) Preparation of a site for construction of a facility, including clearing of the site, grading, installation of drainage, erosion and other environmental mitigation measures, and construction of temporary roads and borrow areas;
          (D) Erection of fences and other access control measures that are not safety or security related, and do not pertain to radiological controls;
          (E) Excavation;
          (F) Erection of support buildings (e.g., construction equipment storage sheds, warehouse and shop facilities, utilities, concrete mixing plants, docking and unloading facilities, and office buildings) for use in connection with the construction of the facility;
          (G) Building of service facilities (e.g., paved roads, parking lots, railroad spurs, exterior utility and lighting systems, potable water systems, sanitary sewerage treatment facilities, and transmission lines);
          (H) Procurement or fabrication of components or portions of the proposed facility occurring at other than the final, in-place location at the facility;
          (I) Manufacture of a nuclear power reactor under a manufacturing license under subpart F of part 52 of this chapter to be installed at the proposed site and to be part of the proposed facility; or
          (J) With respect to production or utilization facilities, other than testing facilities and nuclear power plants, required to be licensed under section 104.a or section 104.c of the Act, the erection of buildings which will be used for activities other than operation of a facility and which may also be used to house a facility (e.g., the construction of a college laboratory building with space for installation of a training reactor).
          (2) For materials licenses, taking any site-preparation activity at the site of a facility subject to the regulations in 10 CFR parts 30, 36, 40, and 70 that has a reasonable nexus to radiological health and safety or the common defense and security; provided, however, that construction does not mean:
          (i) Those actions or activities listed in paragraphs (1)(ii)(A)-(H) of this definition; or
          (ii) Taking any other action that has no reasonable nexus to radiological health and safety or the common defense and security.

            NRC means the Nuclear Regulatory Commission, the agency established by Title II of the Energy Reorganization Act of 1974, as amended.

            NRC staff means any NRC officer or employee or his/her authorized representative, except a Commissioner, a member of a Commissioner's immediate staff, an Atomic Safety and Licensing Board, a presiding officer, an administrative judge, an administrative law judge, or any other officer or employee of the Commission who performs adjudicatory functions.

            NRC Staff Director means:

            Executive Director for Operations;
            Director, Office of Nuclear Reactor Regulation;
            Director, Office of Nuclear Material Safety and Safeguards;
            Director, Office of Nuclear Regulatory Research;
            Director, Office of Governmental and Public Affairs; and
            The designee of any NRC staff director.

          [49 FR 9381, Mar. 12, 1984, as amended at 51 FR 35999, Oct. 8, 1986; 52 FR 31612, Aug. 21, 1987; 72 FR 57443, Oct. 9, 2007; 73 FR 5723, Jan. 31, 2008; 76 FR 56964, Sept. 15, 2011; 77 FR 46599, Aug. 3, 2012; 79 FR 75740, Dec. 19, 2014; 84 FR 65644, Nov. 29, 2019]

          § 51.5
          Interpretations.
          Except as specifically authorized by the Commission in writing, no interpretation of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

          § 51.6
          Specific exemptions.

          The Commission may, upon application of any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and are otherwise in the public interest.

          Subpart A—National Environmental Policy Act—Regulations Implementing Section 102(2)

            § 51.10
            Purpose and scope of subpart; application of regulations of Council on Environmental Quality.
            (a) The National Environmental Policy Act of 1969, as amended (NEPA) directs that, to the fullest extent possible: (1) The policies, regulations, and public laws of the United States shall be interpreted and administered in accordance with the policies set forth in NEPA, and (2) all agencies of the Federal Government shall comply with the procedures in section 102(2) of NEPA except where compliance would be inconsistent with other statutory requirements. The regulations in this subpart implement section 102(2) of NEPA in a manner which is consistent with the NRC's domestic licensing and related regulatory authority under the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and the Uranium Mill Tailings Radiation Control Act of 1978, and which reflects the Commission's announced policy to take account of the regulations of the Council on Environmental Quality published November 29, 1978 (43 FR 55978-56007) voluntarily, subject to certain conditions. This subpart does not apply to export licensing matters within the scope of part 110 of this chapter nor does it apply to any environmental effects which NRC's domestic licensing and related regulatory functions may have upon the environment of foreign nations.
            (b) The Commission recognizes a continuing obligation to conduct its domestic licensing and related regulatory functions in a manner which is both receptive to environmental concerns and consistent with the Commission's responsibility as an independent regulatory agency for protecting the radiological health and safety of the public. Accordingly, the Commission will:
            (1) Examine any future interpretation or change to the Council's NEPA regulations;
            (2) Follow the provisions of 40 CFR 1501.5 and 1501.6 relating to lead agencies and cooperating agencies, except that the Commission reserves the right to prepare an independent environmental impact statement whenever the NRC has regulatory jurisdiction over an acitivity even though the NRC has not been designated as lead agency for preparation of the statement; and
            (3) Reserve the right to make a final decision on any matter within the NRC's regulatory authority even though another agency has made a predecisional referral of an NRC action to the Council under the procedures of 40 CFR part 1504.
            (c) The regulations in this subpart 1

               also address the limitations imposed on NRC's authority and responsibility under the National Environmental Policy Act of 1969, as amended, by the Federal Water Pollution Control Act Amendments of 1972, Pub. L. 92-500, 86 Stat. 816 et seq. (33 U.S.C. 1251 et seq.) In accordance with section 511(c)(2) of the Federal Water Pollution Control Act (86 Stat. 893, 33 U.S.C 1371(c)(2)) the NRC recognizes that responsibility for Federal regulation of nonradiological pollutant discharges 2
               into receiving waters rests by statute with the Environmental Protection Agency.

                1 See also Second Memorandum of Understanding Regarding Implementation of Certain NRC and EPA Responsibilities and Policy Statement on Implementation of Section 511 of the Federal Water Pollution Control Act (FWPCA) attached as Appendix A thereto, which were published in the Federal Register on December 31, 1975 (40 FR 60115) and became effective January 30, 1976.

                2 On June 1, 1976, the U.S. Supreme Court held that “ ‘pollutants’ subject to regulation under the FWPCA [Federal Water Pollution Control Act] do not include source, byproduct, and special nuclear materials, . . .” Train v. Colorado PIRG, 426 U.S. 1 at 25.

            (d) Commission actions initiating or relating to administrative or judicial civil or criminal enforcement actions or proceedings are not subject to Section 102(2) of NEPA. These actions include issuance of notices of violation, orders, and denials of requests for action pursuant to subpart B of part 2 of this chapter; matters covered by part 15 and part 160 of this chapter; and issuance of confirmatory action letters, bulletins, generic letters, notices of deviation, and notices of nonconformance.
            [49 FR 9381, Mar. 12, 1984, as amended at 54 FR 43578, Oct. 26, 1989; 61 FR 43408, Aug. 22, 1996]

            § 51.11
            Relationship to other subparts. [Reserved]

            § 51.12
            Application of subpart to ongoing environmental work.
            (a) Except as otherwise provided in this section, the regulations in this subpart shall apply to the fullest extent practicable to NRC's ongoing environmental work.
            (b) No environmental report or any supplement to an environmental report filed with the NRC and no environmental assessment, environmental impact statement or finding of no significant impact or any supplement to any of the foregoing issued by the NRC before June 7, 1984, need be redone and no notice of intent to prepare an environmental impact statement or notice of availability of these environmental documents need be republished solely by reason of the promulgation on March 12, 1984, of this revision of part 51.
            [49 FR 9381, Mar. 12, 1984, as amended at 49 FR 24513, June 14, 1984]

            § 51.13
            Emergencies.
            Whenever emergency circumstances make it necessary and whenever, in other situations, the health and safety of the public may be adversely affected if mitigative or remedial actions are delayed, the Commission may take an action with significant environmental impact without observing the provisions of these regulations. In taking an action covered by this section, the Commission will consult with the Council as soon as feasible concerning appropriate alternative NEPA arrangements.

            § 51.14
            Definitions.
            (a) As used in this subpart:

              Categorical Exclusion means a category of actions which do not individually or cumulatively have a significant effect on the human environment and which the Commission has found to have no such effect in accordance with procedures set out in § 51.22, and for which, therefore, neither an environmental assessment nor an environmental impact statement is required.

              Cooperating Agency means any Federal agency other than the NRC which has jurisdiction by law or special expertise with respect to any environmental impact involved in a proposal (or a reasonable alternative) for legislation or other major Federal action significantly affecting the quality of the human environment. By agreement with the Commission, a State or local agency of similar qualifications or, when the effects are on a reservation, an Indian Tribe, may become a cooperating agency.

              Council means the Council on Environmental Quality (CEQ) established by Title II of NEPA.

              DOE means the U.S. Department of Energy or its duly authorized representatives.

              Environmental Assessment means a concise public document for which the Commission is responsible that serves to:
            (1) Briefly provide sufficient evidence and analysis for determining whether to prepare an environmental impact statement or a finding of no significant impact.
            (2) Aid the Commission's compliance with NEPA when no environmental impact statement is necessary.
            (3) Facilitate preparation of an environmental impact statement when one is necessary.

              Environmental document includes an environmental assessment, an environmental impact statement, a finding of no significant impact, an environmental report and any supplements to or comments upon those documents, and a notice of intent.

              Environmental Impact Statement means a detailed written statement as required by section 102(2)(C) of NEPA.

              Environmental report means a document submitted to the Commission by an applicant for a permit, license, or other form of permission, or an amendment to or renewal of a permit, license or other form of permission, or by a petitioner for rulemaking, in order to aid the Commission in complying with section 102(2) of NEPA.

              Finding of No Significant Impact means a concise public document for which the Commission is responsible that briefly states the reasons why an action, not otherwise excluded, will not have a significant effect on the human environment and for which therefore an environmental impact statement will not be prepared.

              NEPA means the National Environmental Policy Act of 1969, as amended (Pub. L. 91-190, 83 Stat. 852, 856, as amended by Pub. L. 94-83, 89 Stat. 424, 42 U.S.C. 4321, et seq.).

              Notice of Intent means a notice that an environmental impact statement will be prepared and considered.

              Uranium enrichment facility means:
            (1) Any facility used for separating the isotopes for uranium or enriching uranium in the isotope 235, except laboratory scale facilities designed or used for experimental or analytical purposes only; or
            (2) Any equipment or device, or important component part especially designed for such equipment or device, capable of separating the isotopes of uranium or enriching uranium in the isotope 235.
            (b) The definitions in 40 CFR 1508.3, 1508.7, 1508.8, 1508.14, 1508.15, 1508.16, 1508.17, 1508.18, 1508.20, 1508.23, 1508.25, 1508.26, and 1508.27, will also be used in implementing section 102(2) of NEPA.
            [49 FR 9381, Mar. 12, 1984, as amended at 57 FR 18391, Apr. 30, 1992]

            § 51.15
            Time schedules.
            Consistent with the purposes of NEPA, the Administrative Procedure Act, the Commission's rules of practice in part 2 of this chapter, §§ 51.100 and 51.101, and with other essential considerations of national policy:
            (a) The appropriate NRC staff director may, and upon the request of an applicant for a proposed action or a petitioner for rulemaking shall, establish a time schedule for all or any constituent part of the NRC staff NEPA process. To the maximum extent practicable, the NRC staff will conduct its NEPA review in accordance with any time schedule established under this section.
            (b) As specified in 10 CFR part 2, the presiding officer, the Atomic Safety and Licensing Board or the Commissioners acting as a collegial body may establish a time schedule for all or any part of an adjudicatory or rulemaking proceeding to the extent that each has jurisdiction.
            [49 FR 9381, Mar. 12, 1984, as amended at 69 FR 2276, Jan. 14, 2004]

            § 51.16
            Proprietary information.
            (a) Proprietary information, such as trade secrets or privileged or confidential commercial or financial information, will be treated in accordance with the procedures provided in § 2.390 of this chapter.
            (b) Any proprietary information which a person seeks to have withheld from public disclosure shall be submitted in accordance with § 2.390 of this chapter. When submitted, the proprietary information should be clearly identified and accompanied by a request, containing detailed reasons and justifications, that the proprietary information be withheld from public disclosure. A non-proprietary summary describing the general content of the proprietary information should also be provided.
            [69 FR 2276, Jan. 14, 2004]

            § 51.17
            Information collection requirements; OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0021.
            (b) The approved information collection requirements in this part appear in §§ 51.6, 51.16, 51.41, 51.45, 51.49, 51.50, 51.51, 51.52, 51.53, 51.54, 51.55, 51.58, 51.60, 51.61, 51.62, 51.66, 51.68, and 51.69.
            [49 FR 24513, June 14, 1984, as amended at 62 FR 52188, Oct. 6, 1997; 67 FR 67100, Nov. 4, 2002; 72 FR 57443, Oct. 9, 2007]

            Preliminary Procedures

            classification of licensing and regulatory actions

              § 51.20
              Criteria for and identification of licensing and regulatory actions requiring environmental impact statements.
              (a) Licensing and regulatory actions requiring an environmental impact statement shall meet at least one of the following criteria:
              (1) The proposed action is a major Federal action significantly affecting the quality of the human environment.
              (2) The proposed action involves a matter which the Commission, in the exercise of its discretion, has determined should be covered by an environmental impact statement.
              (b) The following types of actions require an environmental impact statement or a supplement to an environmental impact statement:
              (1) Issuance of a limited work authorization or a permit to construct a nuclear power reactor, testing facility, or fuel reprocessing plant under part 50 of this chapter, or issuance of an early site permit under part 52 of this chapter.
              (2) Issuance or renewal of a full power or design capacity license to operate a nuclear power reactor, testing facility, or fuel reprocessing plant under part 50 of this chapter, or a combined license under part 52 of this chapter.
              (3) Issuance of a permit to construct or a design capacity license to operate or renewal of a design capacity license to operate an isotopic enrichment plant pursuant to part 50 of this chapter.
              (4) Conversion of a provisional operating license for a nuclear power reactor, testing facility or fuel reprocessing plant to a full term or design capacity license pursuant to part 50 of this chapter if a final environmental impact statement covering full term or design capacity operation has not been previously prepared.
              (5)-(6) [Reserved]
              (7) Issuance of a license to possess and use special nuclear material for processing and fuel fabrication, scrap recovery, or conversion of uranium hexafluoride pursuant to part 70 of this chapter.
              (8) Issuance of a license to possess and use source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter.
              (9) Issuance of a license pursuant to part 72 of this chapter for the storage of spent fuel in an independent spent fuel storage installation (ISFSI) at a site not occupied by a nuclear power reactor, or for the storage of spent fuel or high-level radioactive waste in a monitored retrievable storage installation (MRS).
              (10) Issuance of a license for a uranium enrichment facility.
              (11) Issuance of renewal of a license authorizing receipt and disposal of radioactive waste from other persons pursuant to part 61 of this chapter.
              (12) Issuance of a license amendment pursuant to part 61 of this chapter authorizing (i) closure of a land disposal site, (ii) transfer of the license to the disposal site owner for the purpose of institutional control, or (iii) termination of the license at the end of the institutional control period.
              (13) Issuance of a construction authorization and license pursuant to part 60 or part 63 of this chapter.
              (14) Any other action which the Commission determines is a major Commission action significantly affecting the quality of the human environment. As provided in § 51.22(b), the Commission may, in special circumstances, prepare an environmental impact statement on an action covered by a categorical exclusion.
              [49 FR 9381, Mar. 12, 1984, as amended at 53 FR 31681, Aug. 19, 1988; 53 FR 24052, June 27, 1988; 54 FR 15398, Apr. 18, 1989; 54 FR 27870, July 3, 1989; 57 FR 18392, Apr. 30, 1992; 66 FR 55790, Nov. 2, 2001; 72 FR 49509, Aug. 28, 2007]

              § 51.21
              Criteria for and identification of licensing and regulatory actions requiring environmental assessments.

              All licensing and regulatory actions subject to this subpart require an environmental assessment except those identified in § 51.20(b) as requiring an environmental impact statement, those identified in § 51.22(c) as categorical exclusions, and those identified in § 51.22(d) as other actions not requiring environmental review. As provided in § 51.22(b), the Commission may, in special circumstances, prepare an environmental assessment on an action covered by a categorical exclusion.
              [54 FR 27870, July 3, 1989]

              § 51.22
              Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review.
              (a) Licensing, regulatory, and administrative actions eligible for categorical exclusion shall meet the following criterion: The action belongs to a category of actions which the Commission, by rule or regulation, has declared to be a categorical exclusion, after first finding that the category of actions does not individually or cumulatively have a significant effect on the human environment.
              (b) Except in special circumstances, as determined by the Commission upon its own initiative or upon request of any interested person, an environmental assessment or an environmental impact statement is not required for any action within a category of actions included in the list of categorical exclusions set out in paragraph (c) of this section. Special circumstances include the circumstance where the proposed action involves unresolved conflicts concerning alternative uses of available resources within the meaning of section 102(2)(E) of NEPA.
              (c) The following categories of actions are categorical exclusions:
              (1) Amendments to parts 1, 2, 4, 5, 7, 8, 9, 10, 11, 12, 13, 15, 16, 19, 21, 25, 26, 55, 75, 95, 110, 140, 150, 160, 170, or 171 of this chapter, and actions on petitions for rulemaking relating to parts 1, 2, 4, 5, 7, 9, 10, 11, 12, 13, 14, 15, 16, 19, 21, 25, 26, 55, 75, 95, 110, 140, 150, 160, 170, or 171 of this chapter.
              (2) Amendments to the regulations in this chapter which are corrective or of a minor or nonpolicy nature and do not substantially modify existing regulations, and actions on petitions for rulemaking relating to these amendments.
              (3) Amendments to parts 20, 30, 31, 32, 33, 34, 35, 37, 39, 40, 50, 51, 52, 54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and 100 of this chapter which relate to—
              (i) Procedures for filing and reviewing applications for licenses or construction permits or early site permits or other forms of permission or for amendments to or renewals of licenses or construction permits or early site permits or other forms of permission;
              (ii) Recordkeeping requirements;
              (iii) Reporting requirements;
              (iv) Education, training, experience, qualification or other employment suitability requirements or
              (v) Actions on petitions for rulemaking relating to these amendments.
              (4) Entrance into or amendment, suspension, or termination of all or part of an agreement with a State pursuant to section 274 of the Atomic Energy Act of 1954, as amended, providing for assumption by the State and discontinuance by the Commission of certain regulatory authority of the Commission.
              (5) Procurement of general equipment and supplies.
              (6) Procurement of technical assistance, confirmatory research provided that the confirmatory research does not involve any significant construction impacts, and personal services relating to the safe operation and protection of commercial reactors, other facilities, and materials subject to NRC licensing and regulation.
              (7) Personnel actions.
              (8) Issuance, amendment, or renewal of operators' licenses pursuant to part 55 of this chapter.
              (9) Issuance of an amendment to a permit or license for a reactor under part 50 or part 52 of this chapter that changes a requirement or issuance of an exemption from a requirement, with respect to installation or use of a facility component located within the restricted area, as defined in part 20 of this chapter; or the issuance of an amendment to a permit or license for a reactor under part 50 or part 52 of this chapter that changes an inspection or a surveillance requirement; provided that:
              (i) The amendment or exemption involves no significant hazards consideration;

              (ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and

              (iii) There is no significant increase in individual or cumulative occupational radiation exposure.
              (10) Issuance of an amendment to a permit or license issued under this chapter which—
              (i) Changes surety, insurance and/or indemnity requirements;
              (ii) Changes recordkeeping, reporting, or administrative procedures or requirements;
              (iii) Changes the licensee's or permit holder's name, phone number, business or e-mail address;
              (iv) Changes the name, position, or title of an officer of the licensee or permit holder, including but not limited to, the radiation safety officer or quality assurance manager; or
              (v) Changes the format of the license or permit or otherwise makes editorial, corrective or other minor revisions, including the updating of NRC approved references.
              (11) Issuance of amendments to licenses for fuel cycle plants and radioactive waste disposal sites and amendments to materials licenses identified in § 51.60(b)(1) which are administrative, organizational, or procedural in nature, or which result in a change in process operations or equipment, provided that (i) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, (ii) there is no significant increase in individual or cumulative occupational radiation exposure, (iii) there is no significant construction impact, and (iv) there is no significant increase in the potential for or consequences from radiological accidents.

              (12) Issuance of an amendment to a license under parts 50, 52, 60, 61, 63, 70, 72, or 75 of this chapter relating solely to safeguards matters (i.e., protection against sabotage or loss or diversion of special nuclear material) or issuance of an approval of a safeguards plan submitted under parts 50, 52, 70, 72, and 73 of this chapter, provided that the amendment or approval does not involve any significant construction impacts. These amendments and approvals are confined to—
              (i) Organizational and procedural matters;
              (ii) Modifications to systems used for security and/or materials accountability;
              (iii) Administrative changes; and
              (iv) Review and approval of transportation routes pursuant to 10 CFR 73.37.
              (13) Approval of package designs for packages to be used for the transportation of licensed materials.
              (14) Issuance, amendment, or renewal of materials licenses issued pursuant to 10 CFR parts 30, 31, 32, 33, 34, 35, 36, 39, 40 or part 70 authorizing the following types of activities:
              (i) Distribution of radioactive material and devices or products containing radioactive material to general licensees and to persons exempt from licensing.
              (ii) Distribution of radiopharmaceuticals, generators, reagent kits and/or sealed sources to persons licensed pursuant to 10 CFR 35.18.
              (iii) Nuclear pharmacies.
              (iv) Medical and veterinary.
              (v) Use of radioactive materials for research and development and for educational purposes.
              (vi) Industrial radiography.
              (vii) Irradiators.
              (viii) Use of sealed sources and use of gauging devices, analytical instruments and other devices containing sealed sources.
              (ix) Use of uranium as shielding material in containers or devices.
              (x) Possession of radioactive material incident to performing services such as installation, maintenance, leak tests and calibration.
              (xi) Use of sealed sources and/or radioactive tracers in well-logging procedures.
              (xii) Acceptance of packaged radioactive wastes from others for transfer to licensed land burial facilities provided the interim storage period for any package does not exceed 180 days and the total possession limit for all packages held in interim storage at the same time does not exceed 50 curies.
              (xiii) Manufacturing or processing of source, byproduct, or special nuclear materials for distribution to other licensees, except processing of source material for extraction of rare earth and other metals.
              (xiv) Nuclear laundries.

              (xv) Possession, manufacturing, processing, shipment, testing, or other use of depleted uranium military munitions.
              (xvi) Any use of source, byproduct, or special nuclear material not listed above which involves quantities and forms of source, byproduct, or special nuclear material similar to those listed in paragraphs (c)(14) (i) through (xv) of this section (Category 14).
              (15) Issuance, amendment or renewal of licenses for import of nuclear facilities and materials pursuant to part 110 of this chapter, except for import of spent power reactor fuel.
              (16) Issuance or amendment of guides for the implementation of regulations in this chapter, and issuance or amendment of other informational and procedural documents that do not impose any legal requirements.
              (17) Issuance of an amendment to a permit or license under parts 30, 40, 50, 52, or part 70 of this chapter which deletes any limiting condition of operation or monitoring requirement based on or applicable to any matter subject to the provisions of the Federal Water Pollution Control Act.
              (18) Issuance of amendments or orders authorizing licensees of production or utilization facilities to resume operation, provided the basis for the authorization rests solely on a determination or redetermination by the Commission that applicable emergency planning requirements are met.
              (19) Issuance, amendment, modification, or renewal of a certificate of compliance of gaseous diffusion enrichment facilities pursuant to 10 CFR part 76.
              (20) Decommissioning of sites where licensed operations have been limited to the use of—
              (i) Small quantities of short-lived radioactive materials;
              (ii) Radioactive materials in sealed sources, provided there is no evidence of leakage of radioactive material from these sealed sources; or
              (iii) Radioactive materials in such a manner that a decommissioning plan is not required by 10 CFR 30.36(g)(1), 40.42(g)(1), or 70.38(g)(1), and the NRC has determined that the facility meets the radiological criteria for unrestricted use in 10 CFR 20.1402 without further remediation or analysis.
              (21) Approvals of direct or indirect transfers of any license issued by NRC and any associated amendments of license required to reflect the approval of a direct or indirect transfer of an NRC license.
              (22) Issuance of a standard design approval under part 52 of this chapter.
              (23) The Commission finding for a combined license under § 52.103(g) of this chapter.
              (24) Grants to institutions of higher education in the United States, to fund scholarships, fellowships, and stipends for the study of science, engineering, or another field of study that the NRC determines is in a critical skill area related to its regulatory mission, to support faculty and curricular development in such fields, and to support other domestic educational, technical assistance, or training programs (including those of trade schools) in such fields, except to the extent that such grants or programs include activities directly affecting the environment, such as:
              (i) The construction of facilities;
              (ii) A major disturbance brought about by blasting, drilling, excavating or other means;
              (iii) Field work, except that which only involves noninvasive or non-harmful techniques such as taking water or soil samples or collecting non-protected species of flora and fauna; or
              (iv) The release of radioactive material.
              (25) Granting of an exemption from the requirements of any regulation of this chapter, provided that—
              (i) There is no significant hazards consideration;
              (ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite;
              (iii) There is no significant increase in individual or cumulative public or occupational radiation exposure;
              (iv) There is no significant construction impact;
              (v) There is no significant increase in the potential for or consequences from radiological accidents; and
              (vi) The requirements from which an exemption is sought involve:
              (A) Recordkeeping requirements;
              (B) Reporting requirements;

              (C) Inspection or surveillance requirements;
              (D) Equipment servicing or maintenance scheduling requirements;
              (E) Education, training, experience, qualification, requalification or other employment suitability requirements;
              (F) Safeguard plans, and materials control and accounting inventory scheduling requirements;
              (G) Scheduling requirements;
              (H) Surety, insurance or indemnity requirements; or
              (I) Other requirements of an administrative, managerial, or organizational nature.
              (d) In accordance with section 121 of the Nuclear Waste Policy Act of 1982 (42 U.S.C. 10141), the promulgation of technical requirements and criteria that the Commission will apply in approving or disapproving applications under part 60 or 63 of this chapter shall not require an environmental impact statement, an environmental assessment, or any environmental review under subparagraph (E) or (F) of section 102(2) of NEPA.
              [49 FR 9381, Mar. 12, 1984]

                Editorial Note:
                For Federal Register citations affecting § 51.22, see the List of CFR Sections Affected, which appears in the Finding Aids section of the printed volume and at www.govinfo.gov.

              § 51.23
              Environmental impacts of continued storage of spent nuclear fuel beyond the licensed life for operation of a reactor.
              (a) The Commission has generically determined that the environmental impacts of continued storage of spent nuclear fuel beyond the licensed life for operation of a reactor are those impacts identified in NUREG-2157, “Generic Environmental Impact Statement for Continued Storage of Spent Nuclear Fuel.”
              (b) The environmental reports described in §§ 51.50, 51.53, and 51.61 are not required to discuss the environmental impacts of spent nuclear fuel storage in a reactor facility storage pool or an ISFSI for the period following the term of the reactor operating license, reactor combined license, or ISFSI license. The impact determinations in NUREG-2157 regarding continued storage shall be deemed incorporated into the environmental impact statements described in §§ 51.75, 51.80(b), 51.95, and 51.97(a). The impact determinations in NUREG-2157 regarding continued storage shall be considered in the environmental assessments described in §§ 51.30(b) and 51.95(d), if the impacts of continued storage of spent fuel are relevant to the proposed action.
              (c) This section does not alter any requirements to consider the environmental impacts of spent fuel storage during the term of a reactor operating license or combined license, or a license for an ISFSI in a licensing proceeding.
              [49 FR 34694, Aug. 31, 1984, as amended at 55 FR 38474, Sept. 18, 1990; 72 FR 49509, Aug. 28, 2007; 75 FR 81037, Dec. 23, 2010; 79 FR 56260, Sept. 19, 2014]

            determinations to prepare environmental impact statements, environmental assessments or findings of no significant impact, and related procedures

              § 51.25
              Determination to prepare environmental impact statement or environmental assessment; eligibility for categorical exclusion.
              Before taking a proposed action subject to the provisions of this subpart, the appropriate NRC staff director will determine on the basis of the criteria and classifications of types of actions in §§ 51.20, 51.21 and 51.22 of this subpart whether the proposed action is of the type listed in § 51.22(c) as a categorical exclusion or whether an environmental impact statement or an environmental assessment should be prepared. An environmental assessment is not necessary if it is determined that an environmental impact statement will be prepared.

              § 51.26
              Requirement to publish notice of intent and conduct scoping process.

              (a) Whenever the appropriate NRC staff director determines that an environmental impact statement will be prepared by NRC in connection with a proposed action, a notice of intent will be prepared as provided in § 51.27, and will be published in the Federal Register as provided in § 51.116, and an appropriate scoping process (see §§ 51.27, 51.28, and 51.29) will be conducted.
              (b) The scoping process may include a public scoping meeting.
              (c) Upon receipt of an application and accompanying environmental impact statement under § 60.22 or § 63.22 of this chapter (pertaining to geologic repositories for high-level radioactive waste), the appropriate NRC staff director will include in the notice of docketing required to be published by § 2.101(f)(8) of this chapter a statement of Commission intention to adopt the environmental impact statement to the extent practicable. However, if the appropriate NRC staff director determines, at the time of such publication or at any time thereafter, that NRC should prepare a supplemental environmental impact statement in connection with the Commission's action on the license application, the NRC shall follow the procedures set out in paragraph (a) of this section.

              (d) Whenever the appropriate NRC staff director determines that a supplement to an environmental impact statement will be prepared by the NRC, a notice of intent will be prepared as provided in § 51.27, and will be published in the Federal Register as provided in § 51.116. The NRC staff need not conduct a scoping process (see §§ 51.27, 51.28, and 51.29), provided, however, that if scoping is conducted, then the scoping must be directed at matters to be addressed in the supplement. If scoping is conducted in a proceeding for a combined license referencing an early site permit under part 52, then the scoping must be directed at matters to be addressed in the supplement as described in § 51.92(e).
              [49 FR 9381, Mar. 12, 1984, as amended at 54 FR 27870, July 3, 1989; 66 FR 55791, Nov. 2, 2001; 72 FR 49510, Aug. 28, 2007]

              § 51.27
              Notice of intent.
              (a) The notice of intent required by § 51.26(a) shall:
              (1) State that an environmental impact statement will be prepared;
              (2) Describe the proposed action and, to the extent sufficient information is available, possible alternatives;
              (3) State whether the applicant or petitioner for rulemaking has filed an environmental report, and, if so, where copies are available for public inspection;
              (4) Describe the proposed scoping process, including the role of participants, whether written comments will be accepted, the last date for submitting comments and where comments should be sent, whether a public scoping meeting will be held, the time and place of any scoping meeting or when the time and place of the meeting will be announced; and
              (5) State the name, address and telephone number of an individual in NRC who can provide information about the proposed action, the scoping process, and the environmental impact statement.
              (b) The notice of intent required by § 51.26(d) shall:
              (1) State that a supplement to a final environmental impact statement will be prepared in accordance with § 51.72 or § 51.92. For a combined license application that references an early site permit, the supplement to the early site permit environmental impact statement will be prepared in accordance with § 51.92(e);
              (2) Describe the proposed action and, to the extent required, possible alternatives. For the case of a combined license referencing an early site permit, identify the proposed action as the issuance of a combined license for the construction and operation of a nuclear power plant as described in the combined license application at the site described in the early site permit referenced in the combined license application;
              (3) Identify the environmental report prepared by the applicant and information on where copies are available for public inspection;
              (4) Describe the matters to be addressed in the supplement to the final environmental impact statement;

              (5) Describe any proposed scoping process that the NRC staff may conduct, including the role of participants, whether written comments will be accepted, the last date for submitting comments and where comments should be sent, whether a public scoping meeting will be held, the time and place of any scoping meeting or when the time and place of the meeting will be announced; and
              (6) State the name, address, and telephone number of an individual in NRC who can provide information about the proposed action, the scoping process, and the supplement to the environmental impact statement.
              [49 FR 9381, Mar. 12, 1984, as amended at 72 FR 49510, Aug. 28, 2007]

            scoping

              § 51.28
              Scoping—participants.
              (a) The appropriate NRC staff director shall invite the following persons to participate in the scoping process:
              (1) The applicant or the petitioner for rulemaking;
              (2) Any person who has petitioned for leave to intervene in the proceeding or who has been admitted as a party to the proceeding;
              (3) Any other Federal agency which has jurisdiction by law or special expertise with respect to any environmental impact involved or which is authorized to develop and enforce relevant environmental standards;
              (4) Affected State and local agencies, including those authorized to develop and enforce relevant environmental standards;
              (5) Any affected Indian Tribe; and
              (6) Any person who has requested an opportunity to participate in the scoping process.
              (b) The appropriate NRC staff director may also invite any other appropriate person to participate in the scoping process.
              (c) Participation in the scoping process for an environmental impact statement does not entitle the participant to become a party to the proceeding to which the environmental impact statement relates. Participation in an adjudicatory proceeding is governed by the procedures in §§ 2.309 and 2.315 of this chapter. Participation in a rulemaking proceeding in which the Commission has decided to have a hearing is governed by the provisions in the notice of hearing.
              [49 FR 9381, Mar. 12, 1984, as amended at 74 FR 62682, Dec. 1, 2009]

              § 51.29
              Scoping-environmental impact statement and supplement to environmental impact statement.
              (a) The scoping process for an environmental impact statement shall begin as soon as practicable after publication of the notice of intent as provided in § 51.116, and shall be used to:
              (1) Define the proposed action which is to be the subject of the statement or supplement. For environmental impact statements other than a supplement to an early site permit final environmental impact statement prepared for a combined license application, the provisions of 40 CFR 1502.4 will be used for this purpose. For a supplement to an early site permit final environmental impact statement prepared for a combined license application, the proposed action shall be as set forth in the relevant provisions of § 51.92(e).
              (2) Determine the scope of the statement and identify the significant issues to be analyzed in depth.
              (3) Identify and eliminate from detailed study issues which are peripheral or are not significant or which have been covered by prior environmental review. Discussion of these issues in the statement will be limited to a brief presentation of why they are peripheral or will not have a significant effect on the quality of the human environment or a reference to their coverage elsewhere.
              (4) Identify any environmental assessments and other environmental impact statements which are being or will be prepared that are related to but are not part of the scope of the statement under consideration.
              (5) Identify other environmental review and consultation requirements related to the proposed action so that other required analyses and studies may be prepared concurrently and integrated with the environmental impact statement.
              (6) Indicate the relationship between the timing of the preparation of environmental analyses and the Commission's tentative planning and decision-making schedule.

              (7) Identify any cooperating agencies, and as appropriate, allocate assignments for preparation and schedules for completion of the statement to the NRC and any cooperating agencies.

              (8) Describe the means by which the environmental impact statement will be prepared, including any contractor assistance to be used.
              (b) At the conclusion of the scoping process, the appropriate NRC staff director will prepare a concise summary of the determinations and conclusions reached, including the significant issues identified, and will send a copy of the summary to each participant in the scoping process.
              (c) At any time prior to issuance of the draft environmental impact statement, the appropriate NRC staff director may revise the determinations made under paragraph (b) of this section, as appropriate, if substantial changes are made in the proposed action, or if significant new circumstances or information arise which bear on the proposed action or its impacts.
              [49 FR 9381, Mar. 12, 1984, as amended at 72 FR 49510, Aug. 28, 2007]

            environmental assessment

              § 51.30
              Environmental assessment.
              (a) An environmental assessment for proposed actions, other than those for a standard design certification under 10 CFR part 52 or a manufacturing license under part 52, shall identify the proposed action and include:
              (1) A brief discussion of:
              (i) The need for the proposed action;
              (ii) Alternatives as required by section 102(2)(E) of NEPA;
              (iii) The environmental impacts of the proposed action and alternatives as appropriate; and
              (2) A list of agencies and persons consulted, and identification of sources used.
              (b) As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be considered in the environmental assessment, if the impacts of continued storage of spent fuel are relevant to the proposed action.
              (c) An environmental assessment for a proposed action regarding a monitored retrievable storage installation (MRS) will not address the need for the MRS or any alternative to the design criteria for an MRS set forth in section 141(b)(1) of the Nuclear Waste Policy Act of 1982 (96 Stat. 2242, 42 U.S.C. 10161(b)(1)).
              (d) An environmental assessment for a standard design certification under subpart B of part 52 of this chapter must identify the proposed action, and will be limited to the consideration of the costs and benefits of severe accident mitigation design alternatives and the bases for not incorporating severe accident mitigation design alternatives in the design certification. An environmental assessment for an amendment to a design certification will be limited to the consideration of whether the design change which is the subject of the proposed amendment renders a severe accident mitigation design alternative previously rejected in the earlier environmental assessment to become cost beneficial, or results in the identification of new severe accident mitigation design alternatives, in which case the costs and benefits of new severe accident mitigation design alternatives and the bases for not incorporating new severe accident mitigation design alternatives in the design certification must be addressed.

              (e) An environmental assessment for a manufacturing license under subpart F of part 52 of this chapter must identify the proposed action, and will be limited to the consideration of the costs and benefits of severe accident mitigation design alternatives and the bases for not incorporating severe accident mitigation design alternatives in the manufacturing license. An environmental assessment for an amendment to a manufacturing license will be limited to consideration of whether the design change which is the subject of the proposed amendment either renders a severe accident mitigation design alternative previously rejected in an environmental assessment to become cost beneficial, or results in the identification of new severe accident mitigation design alternatives, in which case the costs and benefits of new severe accident mitigation design alternatives and the bases for not incorporating new severe accident mitigation design alternatives in the manufacturing license must be addressed. In either case, the environmental assessment will not address the environmental impacts associated with manufacturing the reactor under the manufacturing license.
              [49 FR 9381, Mar. 12, 1984, as amended at 49 FR 34694, Aug. 31, 1984; 53 FR 31681, Aug. 19, 1988; 72 FR 49510, Aug. 28, 2007; 79 FR 56260, Sept. 19, 2014]

              § 51.31
              Determinations based on environmental assessment.
              (a) General. Upon completion of an environmental assessment for proposed actions other than those involving a standard design certification or a manufacturing license under part 52 of this chapter, the appropriate NRC staff director will determine whether to prepare an environmental impact statement or a finding of no significant impact on the proposed action. As provided in § 51.33, a determination to prepare a draft finding of no significant impact may be made.
              (b) Standard design certification. (1) For actions involving the issuance or amendment of a standard design certification, the Commission shall prepare a draft environmental assessment for public comment as part of the proposed rule. The proposed rule must state that:
              (i) The Commission has determined in § 51.32 that there is no significant environmental impact associated with the issuance of the standard design certification or its amendment, as applicable; and
              (ii) Comments on the environmental assessment will be limited to the consideration of SAMDAs as required by § 51.30(d).
              (2) The Commission will prepare a final environmental assessment following the close of the public comment period for the proposed standard design certification.
              (c) Manufacturing license. (1) Upon completion of the environmental assessment for actions involving issuance or amendment of a manufacturing license (manufacturing license environmental assessment), the appropriate NRC staff director will determine the costs and benefits of severe accident mitigation design alternatives and the bases for not incorporating severe accident mitigation design alternatives in the design of the reactor to be manufactured under the manufacturing license. The NRC staff director may determine to prepare a draft environmental assessment.
              (2) The manufacturing license environmental assessment must state that:
              (i) The Commission has determined in § 51.32 that there is no significant environmental impact associated with the issuance of a manufacturing license or an amendment to a manufacturing license, as applicable;
              (ii) The environmental assessment will not address the environmental impacts associated with manufacturing the reactor under the manufacturing license; and
              (iii) Comments on the environmental assessment will be limited to the consideration of severe accident mitigation design alternatives as required by § 51.30(e).
              (3) If the NRC staff director makes a determination to prepare and issue a draft environmental assessment for public review and comment before making a final determination on the manufacturing license application, the assessment will be marked, “Draft.” The NRC notice of availability on the draft environmental assessment will include a request for comments which specifies where comments should be submitted and when the comment period expires. The notice will state that copies of the environmental assessment and any related environmental documents are available for public inspection and where inspections can be made. A copy of the final environmental assessment will be sent to the U.S. Environmental Protection Agency, the applicant, any party to a proceeding, each commenter, and any other Federal, State, and local agencies, and Indian Tribes, State, regional, and metropolitan clearinghouses expressing an interest in the action. Additional copies will be made available in accordance with § 51.123.

              (4) When a hearing is held under the regulations in part 2 of this chapter on the proposed issuance of the manufacturing license or amendment, the NRC staff director will prepare a final environmental assessment which may be subject to modification as a result of review and decision as appropriate to the nature and scope of the proceeding.

              (5) Only a party admitted into the proceeding with respect to a contention on the environmental assessment, or an entity participating in the proceeding pursuant to § 2.315(c) of this chapter, may take a position and offer evidence on the matters within the scope of the environmental assessment.
              [72 FR 49510, Aug. 28, 2007]

            finding of no significant impact

              § 51.32
              Finding of no significant impact.
              (a) A finding of no significant impact will:
              (1) Identify the proposed action;
              (2) State that the Commission has determined not to prepare an environmental impact statement for the proposed action;
              (3) Briefly present the reasons why the proposed action will not have a significant effect on the quality of the human environment;
              (4) Include the environmental assessment or a summary of the environmental assessment. If the assessment is included, the finding need not repeat any of the discussion in the assessment but may incorporate it by reference;
              (5) Note any other related environmental documents; and
              (6) State that the finding and any related environmental documents are available for public inspection and where the documents may be inspected.
              (b) The Commission finds that there is no significant environmental impact associated with the issuance of:
              (1) A standard design certification under subpart B of part 52 of this chapter;
              (2) An amendment to a design certification;
              (3) A manufacturing license under subpart F of part 52 of this chapter; or
              (4) An amendment to a manufacturing license.
              [49 FR 9381, Mar. 12, 1984, as amended at 72 FR 49511, Aug. 28, 2007]

              § 51.33
              Draft finding of no significant impact; distribution.
              (a) As provided in paragraph (b) of this section, the appropriate NRC staff director may make a determination to prepare and issue a draft finding of no significant impact for public review and comment before making a final determination whether to prepare an environmental impact statement or a final finding of no significant impact on the proposed action.
              (b) Circumstances in which a draft finding of no significant impact may be prepared will ordinarily include the following:
              (1) A finding of no significant impact appears warranted for the proposed action but the proposed action is (i) closely similar to one which normally requires the preparation of an environmental impact statement, or (ii) without precedent; and
              (2) The appropriate NRC staff director determines that preparation of a draft finding of no significant impact will further the purposes of NEPA.

              (c) A draft finding of no significant impact will (1) be marked “Draft”, (2) contain the information specified in § 51.32, (3) be accompanied by or include a request for comments on the proposed action and on the draft finding within thirty (30) days, or such longer period as may be specified in the notice of the draft finding, and (4) be published in the Federal Register as required by §§ 51.35 and 51.119.
              (d) A draft finding will be distributed as provided in § 51.74(a). Additional copies will be made available in accordance with § 51.123.
              (e) When a draft finding of no significant impact is issued for a proposed action, a final determination to prepare an environmental impact statement or a final finding of no significant impact for that action shall not be made until the last day of the public comment period has expired.

              § 51.34
              Preparation of finding of no significant impact.
              (a) Except as provided in paragraph (b) of this section, the finding of no significant impact will be prepared by the NRC staff director authorized to take the action.

              (b) When a hearing is held on the proposed action under the regulations in subpart G of part 2 of this chapter or when the action can only be taken by the Commissioners acting as a collegial body, the appropriate NRC staff director will prepare a proposed finding of no significant impact, which may be subject to modification as a result of review and decision as appropriate to the nature and scope of the proceeding. In such cases, the presiding officer, or the Commission acting as a collegial body, as appropriate, will issue the final finding of no significant impact.
              [49 FR 9381, Mar. 12, 1984, as amended at 77 FR 46600, Aug. 3, 2012; 79 FR 66604, Nov. 10, 2014]

              § 51.35
              Requirement to publish finding of no significant impact; limitation on Commission action.

              (a) Whenever the Commission makes a draft or final finding of no significant impact on a proposed action, the finding will be published in the Federal Register as provided in § 51.119.

              (b) Except as provided in § 51.13, the Commission shall not take the proposed action until after the final finding has been published in the Federal Register.

            Environmental Reports and Information—Requirements Applicable to Applicants and Petitioners for Rulemaking

            general

              § 51.40
              Consultation with NRC staff.
              (a) A prospective applicant or petitioner for rulemaking is encouraged to confer with NRC staff as early as possible in its planning process before submitting environmental information or filing an environmental report.
              (b) Requests for guidance or information on environmental matters may include inquiries relating to:
              (1) Applicable NRC rules and regulations;
              (2) Format, content and procedures for filing environmental reports and other environmental information, including the type and quantity of environmental information likely to be needed to address issues and concerns identified in the scoping process described in § 51.29 in a manner appropriate to their relative significance;
              (3) Availability of relevant environmental studies and environmental information;
              (4) Need for, appropriate level and scope of any environmental studies or information which the Commission may require to be submitted in connection with an application or petition for rulemaking;
              (5) Public meetings with NRC staff.
              (c) Questions concerning environmental matters should be addressed to the following NRC staff offices as appropriate:
              (1) Utilization facilities: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-1270, e-mail RidsNrrOd@nrc.gov.

              (2) Production facilities: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-7800, e-mail RidsNmssOd@nrc.gov.

              (3) Materials licenses: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-7800, e-mail RidsNmssOd@nrc.gov.

              (4) Rulemaking: ATTN: Chief, Regulatory Analysis and Rulemaking Support Branch, Division of Rulemaking, Environmental, and Financial Support, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (800) 368-5642.
              (5) General Environmental Matters: Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 415-1700.
              [49 FR 9381, Mar. 12, 1984, as amended at 53 FR 13399, Apr. 25, 1988; 60 FR 24552, May 9, 1995; 68 FR 58810, Oct. 10, 2003; 73 FR 5724, Jan. 31, 2008; 84 FR 65644, Nov. 29, 2019]

              § 51.41
              Requirement to submit environmental information.

              The Commission may require an applicant for a permit, license, or other form of permission, or amendment to or renewal of a permit, license or other form of permission, or a petitioner for rulemaking to submit such information to the Commission as may be useful in aiding the Commission in complying with section 102(2) of NEPA. The Commission will independently evaluate and be responsible for the reliability of any information which it uses.

            environmental reports—general requirements

              § 51.45
              Environmental report.
              (a) General. As required by §§ 51.50, 51.53, 51.54, 51.55, 51.60, 51.61, 51.62, or 51.68, as appropriate, each applicant or petitioner for rulemaking shall submit with its application or petition for rulemaking one signed original of a separate document entitled “Applicant's” or “Petitioner's Environmental Report,” as appropriate. An applicant or petitioner for rulemaking may submit a supplement to an environmental report at any time.
              (b) Environmental considerations. The environmental report shall contain a description of the proposed action, a statement of its purposes, a description of the environment affected, and discuss the following considerations:
              (1) The impact of the proposed action on the environment. Impacts shall be discussed in proportion to their significance;
              (2) Any adverse environmental effects which cannot be avoided should the proposal be implemented;
              (3) Alternatives to the proposed action. The discussion of alternatives shall be sufficiently complete to aid the Commission in developing and exploring, pursuant to section 102(2)(E) of NEPA, “appropriate alternatives to recommended courses of action in any proposal which involves unresolved conflicts concerning alternative uses of available resources.” To the extent practicable, the environmental impacts of the proposal and the alternatives should be presented in comparative form;
              (4) The relationship between local short-term uses of man's environment and the maintenance and enhancement of long-term productivity; and
              (5) Any irreversible and irretrievable commitments of resources which would be involved in the proposed action should it be implemented.
              (c) Analysis. The environmental report must include an analysis that considers and balances the environmental effects of the proposed action, the environmental impacts of alternatives to the proposed action, and alternatives available for reducing or avoiding adverse environmental effects. An environmental report required for materials licenses under § 51.60 must also include a description of those site preparation activities excluded from the definition of construction under § 51.4 which have been or will be undertaken at the proposed site (i.e., those activities listed in paragraphs (2)(i) and (2)(ii) in the definition of construction contained in § 51.4); a description of the impacts of such excluded site preparation activities; and an analysis of the cumulative impacts of the proposed action when added to the impacts of such excluded site preparation activities on the human environment. An environmental report prepared at the early site permit stage under § 51.50(b), limited work authorization stage under § 51.49, construction permit stage under § 51.50(a), or combined license stage under § 51.50(c) must include a description of impacts of the preconstruction activities performed by the applicant at the proposed site (i.e., those activities listed in paragraph (1)(ii) in the definition of “construction” contained in § 51.4), necessary to support the construction and operation of the facility which is the subject of the early site permit, limited work authorization, construction permit, or combined license application. The environmental report must also contain an analysis of the cumulative impacts of the activities to be authorized by the limited work authorization, construction permit, or combined license in light of the preconstruction impacts described in the environmental report. Except for an environmental report prepared at the early site permit stage, or an environmental report prepared at the license renewal stage under § 51.53(c), the analysis in the environmental report should also include consideration of the economic, technical, and other benefits and costs of the proposed action and its alternatives. Environmental reports prepared at the license renewal stage under § 51.53(c) need not discuss the economic or technical benefits and costs of either the proposed action or alternatives except if these benefits and costs are either essential for a determination regarding the inclusion of an alternative in the range of alternatives considered or relevant to mitigation. In addition, environmental reports prepared under § 51.53(c) need not discuss issues not related to the environmental effects of the proposed action and its alternatives. The analyses for environmental reports shall, to the fullest extent practicable, quantify the various factors considered. To the extent that there are important qualitative considerations or factors that cannot be quantified, those considerations or factors shall be discussed in qualitative terms. The environmental report should contain sufficient data to aid the Commission in its development of an independent analysis.
              (d) Status of compliance. The environmental report shall list all Federal permits, licenses, approvals and other entitlements which must be obtained in connection with the proposed action and shall describe the status of compliance with these requirements. The environmental report shall also include a discussion of the status of compliance with applicable environmental quality standards and requirements including, but not limited to, applicable zoning and land-use regulations, and thermal and other water pollution limitations or requirements which have been imposed by Federal, State, regional, and local agencies having responsibility for environmental protection. The discussion of alternatives in the report shall include a discussion of whether the alternatives will comply with such applicable environmental quality standards and requirements.
              (e) Adverse information. The information submitted pursuant to paragraphs (b) through (d) of this section should not be confined to information supporting the proposed action but should also include adverse information.
              [49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28486, June 5, 1996; 61 FR 66542, Dec. 18, 1996; 68 FR 58810, Oct. 10, 2003; 72 FR 49511, Aug. 28, 2007; 72 FR 57443, Oct. 9, 2007; 73 FR 22787, Apr. 28, 2008; 76 FR 56965, Sept. 15, 2011]

            environmental reports—production and utilization facilities

              § 51.49
              Environmental report—limited work authorization.
              (a) Limited work authorization submitted as part of complete construction permit or combined license application. Each applicant for a construction permit or combined license applying for a limited work authorization under § 50.10(d) of this chapter in a complete application under 10 CFR 2.101(a)(1) through (a)(4), shall submit with its application a separate document, entitled, “Applicant's Environmental Report—Limited Work Authorization Stage,” which is in addition to the environmental report required by § 51.50 of this part. Each environmental report must also contain the following information:
              (1) A description of the activities proposed to be conducted under the limited work authorization;
              (2) A statement of the need for the activities; and
              (3) A description of the environmental impacts that may reasonably be expected to result from the activities, the mitigation measures that the applicant proposes to implement to achieve the level of environmental impacts described, and a discussion of the reasons for rejecting mitigation measures that could be employed by the applicant to further reduce environmental impacts.
              (b) Phased application for limited work authorization and construction permit or combined license. If the construction permit or combined license application is filed in accordance with § 2.101(a)(9) of this chapter, then the environmental report for part one of the application may be limited to a discussion of the activities proposed to be conducted under the limited work authorization. If the scope of the environmental report for part one is so limited, then part two of the application must include the information required by § 51.50, as applicable.
              (c) Limited work authorization submitted as part of an early site permit application. Each applicant for an early site permit under subpart A of part 52 of this chapter requesting a limited work authorization shall submit with its application the environmental report required by § 51.50(b). Each environmental report must contain the following information:
              (1) A description of the activities proposed to be conducted under the limited work authorization;
              (2) A statement of the need for the activities; and
              (3) A description of the environmental impacts that may reasonably be expected to result from the activities, the mitigation measures that the applicant proposes to implement to achieve the level of environmental impacts described, and a discussion of the reasons for rejecting mitigation measures that could be employed by the applicant to further reduce environmental impacts.
              (d) Limited work authorization request submitted by early site permit holder. Each holder of an early site permit requesting a limited work authorization shall submit with its application a document entitled, “Applicant's Environmental Report—Limited Work Authorization under Early Site Permit,” containing the following information:
              (1) A description of the activities proposed to be conducted under the limited work authorization;
              (2) A statement of the need for the activities;
              (3) A description of the environmental impacts that may reasonably be expected to result from the activities, the mitigation measures that the applicant proposes to implement to achieve the level of environmental impacts described, and a discussion of the reasons for rejecting mitigation measures that could be employed by the applicant to further reduce environmental impacts; and
              (4) Any new and significant information for issues related to the impacts of construction of the facility that were resolved in the early site permit proceeding with respect to the environmental impacts of the activities to be conducted under the limited work authorization.
              (5) A description of the process used to identify new and significant information regarding NRC's conclusions in the early site permit environmental impact statement. The process must be a reasonable methodology for identifying this new and significant information.
              (e) Limited work authorization for a site where an environmental impact statement was prepared, but the facility construction was not completed. If the limited work authorization is for activities to be conducted at a site for which the Commission has previously prepared an environmental impact statement for the construction and operation of a nuclear power plant, and a construction permit was issued but construction of the plant was never completed, then the applicant's environmental report may incorporate by reference the earlier environmental impact statement. In the event of such referencing, the environmental report must identify:
              (1) Any new and significant information material to issues related to the impacts of construction of the facility that were resolved in the construction permit proceeding for the matters required to be addressed in paragraph (a) of this section; and
              (2) A description of the process used to identify new and significant information regarding the NRC's conclusions in the construction permit environmental impact statement. The process must use a reasonable methodology for identifying this new and significant information.
              (f) Environmental report. An environmental report submitted in accordance with this section must separately evaluate the environmental impacts and proposed alternatives attributable to the activities proposed to be conducted under the limited work authorization. At the option of the applicant, the “Applicant's Environmental Report—Limited Work Authorization Stage,” may contain the information required to be submitted in the environmental report required under § 51.50, which addresses the impacts of construction and operation for the proposed facility (including the environmental impacts attributable to the limited work authorization), and discusses the overall costs and benefits balancing for the proposed action.
              [72 FR 57444, Oct. 9, 2007]

              § 51.50
              Environmental report—construction permit, early site permit, or combined license stage.
              (a) Construction permit stage. Each applicant for a permit to construct a production or utilization facility covered by § 51.20 shall submit with its application a separate document, entitled “Applicant's Environmental Report—Construction Permit Stage,” which shall contain the information specified in §§ 51.45, 51.51, and 51.52. Each environmental report shall identify procedures for reporting and keeping records of environmental data, and any conditions and monitoring requirements for protecting the non-aquatic environment, proposed for possible inclusion in the license as environmental conditions in accordance with § 50.36b of this chapter. As stated in § 51.23, no discussion of the environmental impacts of the continued storage of spent fuel is required in this report.
              (b) Early site permit stage. Each applicant for an early site permit shall submit with its application a separate document, entitled “Applicant's Environmental Report—Early Site Permit Stage,” which shall contain the information specified in §§ 51.45, 51.51, and 51.52, as modified in this paragraph.
              (1) The environmental report must include an evaluation of alternative sites to determine whether there is any obviously superior alternative to the site proposed.
              (2) The environmental report may address one or more of the environmental effects of construction and operation of a reactor, or reactors, which have design characteristics that fall within the site characteristics and design parameters for the early site permit application, provided however, that the environmental report must address all environmental effects of construction and operation necessary to determine whether there is any obviously superior alternative to the site proposed. The environmental report need not include an assessment of the economic, technical, or other benefits (for example, need for power) and costs of the proposed action or an evaluation of alternative energy sources. As stated in § 51.23, no discussion of the environmental impacts of the continued storage of spent fuel is required in this report.
              (3) For other than light-water-cooled nuclear power reactors, the environmental report must contain the basis for evaluating the contribution of the environmental effects of fuel cycle activities for the nuclear power reactor.
              (4) Each environmental report must identify the procedures for reporting and keeping records of environmental data, and any conditions and monitoring requirements for protecting the non-aquatic environment, proposed for possible inclusion in the license as environmental conditions in accordance with § 50.36b of this chapter.
              (c) Combined license stage. Each applicant for a combined license shall submit with its application a separate document, entitled “Applicant's Environmental Report—Combined License Stage.” Each environmental report shall contain the information specified in §§ 51.45, 51.51, and 51.52, as modified in this paragraph. For other than light-water-cooled nuclear power reactors, the environmental report shall contain the basis for evaluating the contribution of the environmental effects of fuel cycle activities for the nuclear power reactor. Each environmental report shall identify procedures for reporting and keeping records of environmental data, and any conditions and monitoring requirements for protecting the non-aquatic environment, proposed for possible inclusion in the license as environmental conditions in accordance with § 50.36b of this chapter. The combined license environmental report may reference information contained in a final environmental document previously prepared by the NRC staff. As stated in § 51.23, no discussion of the environmental impacts of the continued storage of spent fuel is required in this report.
              (1) Application referencing an early site permit. If the combined license application references an early site permit, then the “Applicant's Environmental Report—Combined License Stage” need not contain information or analyses submitted to the Commission in “Applicant's Environmental Report—Early Site Permit Stage,” or resolved in the Commission's early site permit environmental impact statement, but must contain, in addition to the environmental information and analyses otherwise required:
              (i) Information to demonstrate that the design of the facility falls within the site characteristics and design parameters specified in the early site permit;
              (ii) Information to resolve any significant environmental issue that was not resolved in the early site permit proceeding;
              (iii) Any new and significant information for issues related to the impacts of construction and operation of the facility that were resolved in the early site permit proceeding;
              (iv) A description of the process used to identify new and significant information regarding the NRC's conclusions in the early site permit environmental impact statement. The process must use a reasonable methodology for identifying such new and significant information; and
              (v) A demonstration that all environmental terms and conditions that have been included in the early site permit will be satisfied by the date of issuance of the combined license. Any terms or conditions of the early site permit that could not be met by the time of issuance of the combined license, must be set forth as terms or conditions of the combined license.
              (2) Application referencing standard design certification. If the combined license references a standard design certification, then the combined license environmental report may incorporate by reference the environmental assessment previously prepared by the NRC for the referenced design certification. If the design certification environmental assessment is referenced, then the combined license environmental report must contain information to demonstrate that the site characteristics for the combined license site fall within the site parameters in the design certification environmental assessment.
              (3) Application referencing a manufactured reactor. If the combined license application proposes to use a manufactured reactor, then the combined license environmental report may incorporate by reference the environmental assessment previously prepared by the NRC for the underlying manufacturing license. If the manufacturing license environmental assessment is referenced, then the combined license environmental report must contain information to demonstrate that the site characteristics for the combined license site fall within the site parameters in the manufacturing license environmental assessment. The environmental report need not address the environmental impacts associated with manufacturing the reactor under the manufacturing license.
              [72 FR 49511, Aug. 28, 2007, as amended at 79 FR 56260, Sept. 19, 2014]

              § 51.51
              Uranium fuel cycle environmental data—Table S-3.
              (a) Under § 51.50, every environmental report prepared for the construction permit stage or early site permit stage or combined license stage of a light-water-cooled nuclear power reactor, and submitted on or after September 4, 1979, shall take Table S-3, Table of Uranium Fuel Cycle Environmental Data, as the basis for evaluating the contribution of the environmental effects of uranium mining and milling, the production of uranium hexafluoride, isotopic enrichment, fuel fabrication, reprocessing of irradiated fuel, transportation of radioactive materials and management of low-level wastes and high-level wastes related to uranium fuel cycle activities to the environmental costs of licensing the nuclear power reactor. Table S-3 shall be included in the environmental report and may be supplemented by a discussion of the environmental significance of the data set forth in the table as weighed in the analysis for the proposed facility.
              (b) Table S-3.

                Table S-3—Table of Uranium Fuel Cycle Environmental Data 1

                [Normalized to model LWR annual fuel requirement [WASH-1248] or reference reactor year [NUREG-0116]]
                [See footnotes at end of this table]

                  Environmental considerations
                  Total
                  Maximum effect per annual fuel requirement or reference reactor year of model 1,000 MWe LWR

                    Natural Resource Use

                  Land (acres):

                  Temporarily committed 2

                  100

                  Undisturbed area
                  79

                  Disturbed area
                  22
                  Equivalent to a 110 MWe coal-fired power plant.

                  Permanently committed
                  13

                  Overburden moved (millions of MT)
                  2.8
                  Equivalent to 95 MWe coal-fired power plant.

                  Water (millions of gallons):

                  Discharged to air
                  160
                   = 2 percent of model 1,000 MWe LWR with cooling tower.

                  Discharged to water bodies
                  11,090

                  Discharged to ground
                  127

                  Total
                  11,377
                  <4 percent of model 1,000 MWe LWR with once-through cooling.

                  Fossil fuel:

                  Electrical energy (thousands of MW-hour)
                  323
                  <5 percent of model 1,000 MWe LWR output.

                  Equivalent coal (thousands of MT)
                  118
                  Equivalent to the consumption of a 45 MWe coal-fired power plant.

                  Natural gas (millions of scf)
                  135

                  <0.4 percent of model 1,000 MWe energy output.

                    Effluents—Chemical (MT)

                  Gases (including entrainment): 3

                  SOX

                  4,400

                  NOX
                    4

                  1,190
                  Equivalent to emissions from 45 MWe coal-fired plant for a year.

                  Hydrocarbons
                  14

                  CO
                  29.6

                  Particulates
                  1,154

                  Other gases:

                  F
                  .67
                  Principally from UF6 production, enrichment, and reprocessing. Concentration within range of state standards—below level that has effects on human health.

                  HCl
                  .014

                  Liquids:

                  SO −
                    4
                    NO −
                      3

                    Fluoride
                    Ca +

                    C1 −

                    Na +

                    NH 3

                    Fe

                  9.925.8
                    12.9
                    5.4
                    8.5
                    12.1
                    10.0
                    .4

                  From enrichment, fuel fabrication, and reprocessing steps. Components that constitute a potential for adverse environmental effect are present in dilute concentrations and receive additional dilution by receiving bodies of water to levels below permissible standards. The constituents that require dilution and the flow of dilution water are: NH 3—600 cfs., NO 3—20 cfs., Fluoride—70 cfs.

                  Tailings solutions (thousands of MT)
                  240
                  From mills only—no significant effluents to environment.

                  Solids
                  91,000
                  Principally from mills—no significant effluents to environment.

                  Effluents—Radiological (curies)

                  Gases (including entrainment):

                  Rn-222

                  Presently under reconsideration by the Commission.

                  Ra-226
                  .02

                  Th-230
                  .02

                  Uranium
                  .034

                  Tritium (thousands)
                  18.1

                  C-14
                  24

                  Kr-85 (thousands)
                  400

                  Ru-106
                  .14
                  Principally from fuel reprocessing plants.

                  I-129
                  1.3

                  I-131
                  .83

                  Tc-99

                  Presently under consideration by the Commission.

                  Fission products and transuranics
                  .203

                  Liquids:

                  Uranium and daughters
                  2.1
                  Principally from milling—included tailings liquor and returned to ground—no effluents; therefore, no effect on environment.

                  Ra-226
                  .0034
                  From UF6 production.

                  Th-230
                  .0015

                  Th-234
                  .01
                  From fuel fabrication plants—concentration 10 percent of 10 CFR 20 for total processing 26 annual fuel requirements for model LWR.

                  Fission and activation products
                  5.9 × 10−6

                  Solids (buried on site):

                  Other than high level (shallow)
                  11,300
                  9,100 Ci comes from low level reactor wastes and 1,500 Ci comes from reactor decontamination and decommissioning—buried at land burial facilities. 600 Ci comes from mills—included in tailings returned to ground. Approximately 60 Ci comes from conversion and spent fuel storage. No significant effluent to the environment.

                  TRU and HLW (deep)
                  1.1 × 107

                  Buried at Federal Repository.

                  Effluents—thermal (billions of British thermal units)
                  4,063
                  <5 percent of model 1,000 MWe LWR.

                  Transportation (person-rem):

                  Exposure of workers and general public
                  2.5

                  Occupational exposure (person-rem)
                  22.6
                  From reprocessing and waste management.

                  1 In some cases where no entry appears it is clear from the background documents that the matter was addressed and that, in effect, the Table should be read as if a specific zero entry had been made. However, there are other areas that are not addressed at all in the Table. Table S-3 does not include health effects from the effluents described in the Table, or estimates of releases of Radon-222 from the uranium fuel cycle or estimates of Technetium-99 released from waste management or reprocessing activities. These issues may be the subject of litigation in the individual licensing proceedings.
                Data supporting this table are given in the “Environmental Survey of the Uranium Fuel Cycle,” WASH-1248, April 1974; the “Environmental Survey of the Reprocessing and Waste Management Portion of the LWR Fuel Cycle,” NUREG-0116 (Supp.1 to WASH-1248); the “Public Comments and Task Force Responses Regarding the Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle,” NUREG-0216 (Supp. 2 to WASH-1248); and in the record of the final rulemaking pertaining to Uranium Fuel Cycle Impacts from Spent Fuel Reprocessing and Radioactive Waste Management, Docket RM-50-3. The contributions from reprocessing, waste management and transportation of wastes are maximized for either of the two fuel cycles (uranium only and no recycle). The contribution from transportation excludes transportation of cold fuel to a reactor and of irradiated fuel and radioactive wastes from a reactor which are considered in Table S-4 of § 51.20(g). The contributions from the other steps of the fuel cycle are given in columns A-E of Table S-3A of WASH-1248.

                  2 The contributions to temporarily committed land from reprocessing are not prorated over 30 years, since the complete temporary impact accrues regardless of whether the plant services one reactor for one year or 57 reactors for 30 years.

                  3 Estimated effluents based upon combustion of equivalent coal for power generation.

                  4 1.2 percent from natural gas use and process.

              [49 FR 9381, Mar. 12, 1984; 49 FR 10922, Mar. 23, 1984, as amended at 67 FR 77652, Dec. 19, 2002; 72 FR 49512, Aug. 28, 2007]

              § 51.52
              Environmental effects of transportation of fuel and waste—Table S-4.
              Under § 51.50, every environmental report prepared for the construction permit stage or early site permit stage or combined license stage of a light-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement concerning transportation of fuel and radioactive wastes to and from the reactor. That statement shall indicate that the reactor and this transportation either meet all of the conditions in paragraph (a) of this section or all of the conditions of paragraph (b) of this section.
              (a)(1) The reactor has a core thermal power level not exceeding 3,800 megawatts;
              (2) The reactor fuel is in the form of sintered uranium dioxide pellets having a uranium-235 enrichment not exceeding 4% by weight, and the pellets are encapsulated in zircaloy rods;
              (3) The average level of irradiation of the irradiated fuel from the reactor does not exceed 33,000 megawatt-days per metric ton, and no irradiated fuel assembly is shipped until at least 90 days after it is discharged from the reactor;
              (4) With the exception of irradiated fuel, all radioactive waste shipped from the reactor is packaged and in a solid form;
              (5) Unirradiated fuel is shipped to the reactor by truck; irradiated fuel is shipped from the reactor by truck, rail, or barge; and radioactive waste other than irradiated fuel is shipped from the reactor by truck or rail; and

              (6) The environmental impacts of transportation of fuel and waste to and from the reactor, with respect to normal conditions of transport and possible accidents in transport, are as set forth in Summary Table S-4 in paragraph (c) of this section; and the values in the table represent the contribution of the transportation to the environmental costs of licensing the reactor.
              (b) For reactors not meeting the conditions of paragraph (a) of this section, the statement shall contain a full description and detailed analysis of the environmental effects of transportation of fuel and wastes to and from the reactor, including values for the environmental impact under normal conditions of transport and for the environmental risk from accidents in transport. The statement shall indicate that the values determined by the analysis represent the contribution of such effects to the environmental costs of licensing the reactor.
              (c)

                Summary Table S-4—Environmental Impact of Transportation of Fuel and Waste to and From One Light-Water-Cooled Nuclear Power Reactor 1

                Normal Conditions of Transport

                  Environmental impact

                  Heat (per irradiated fuel cask in transit)
                  250,000 Btu/hr.

                  Weight (governed by Federal or State restrictions)
                  73,000 lbs. per truck; 100 tons per cask per rail car.

                  Traffic density:

                  Truck
                  Less than 1 per day.

                  Rail
                  Less than 3 per month

                  Exposed population
                  Estimated number of persons exposed
                  Range of doses to exposed individuals 2 (per reactor year)
                  Cumulative dose to exposed population (per reactor year) 3

                  Transportation workers
                  200
                  0.01 to 300 millirem
                  4 man-rem.

                  General public:

                  Onlookers
                  1,100
                  0.003 to 1.3 millirem
                  3 man-rem.

                  Along Route
                  600,000
                  0.0001 to 0.06 millirem

                Accidents in Transport

                  Environmental risk

                  Radiological effects
                  Small 4

                  Common (nonradiological) causes
                  1 fatal injury in 100 reactor years; 1 nonfatal injury in 10 reactor years; $475 property damage per reactor year.

                  1 Data supporting this table are given in the Commission's “Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants,” WASH-1238, December 1972; and Supp. 1 of NUREG-75/038, April 1975. Both documents are available for inspection and copying at the Commission's Public Document Room, One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852 and may be obtained from National Technical Information Service, Springfield, VA 22161. The WASH-1238 is available from NTIS at a cost of $5.45 (microfiche, $2.25) and NUREG-75/038 is available at a cost of $3.25 (microfiche, $2.25).

                  2 The Federal Radiation Council has recommended that the radiation doses from all sources of radiation other than natural background and medical exposures should be limited to 5,000 millirem per year for individuals as a result of occupational exposure and should be limited to 500 millirem per year for individuals in the general population. The dose to individuals due to average natural background radiation is about 130 millirem per year.

                  3 Man-rem is an expression for the summation of whole body doses to individuals in a group. Thus, if each member of a population group of 1,000 people were to receive a dose of 0.001 rem (1 millirem), or if 2 people were to receive a dose of 0.5 rem (500 millirem) each, the total man-rem dose in each case would be 1 man-rem.

                  4 Athough the environmental risk of radiological effects stemming from transportation accidents is currently incapable of being numerically quantified, the risk remains small regardless of whether it is being appiled to a single reactor or a multireactor site.

              [49 FR 9381, Mar. 12, 1984; 49 FR 10922, Mar. 23, 1984, as amended at 53 FR 43420, Oct. 27, 1988; 72 FR 49512, Aug. 28, 2007; 79 FR 66604, Nov. 10, 2014]

              § 51.53
              Postconstruction environmental reports.
              (a) General. Any environmental report prepared under the provisions of this section may incorporate by reference any information contained in a prior environmental report or supplement thereto that relates to the production or utilization facility or site, or any information contained in a final environmental document previously prepared by the NRC staff that relates to the production or utilization facility or site. Documents that may be referenced include, but are not limited to, the final environmental impact statement; supplements to the final environmental impact statement, including supplements prepared at the license renewal stage; NRC staff-prepared final generic environmental impact statements; and environmental assessments and records of decisions prepared in connection with the construction permit, operating license, early site permit, combined license and any license amendment for that facility.
              (b) Operating license stage. Each applicant for a license to operate a production or utilization facility covered by § 51.20 shall submit with its application a separate document entitled “Supplement to Applicant's Environmental Report—Operating License Stage,” which will update “Applicant's Environmental Report—Construction Permit Stage.” Unless otherwise required by the Commission, the applicant for an operating license for a nuclear power reactor shall submit this report only in connection with the first licensing action authorizing full-power operation. In this report, the applicant shall discuss the same matters described in §§ 51.45, 51.51, and 51.52, but only to the extent that they differ from those discussed or reflect new information in addition to that discussed in the final environmental impact statement prepared by the Commission in connection with the construction permit. No discussion of need for power, or of alternative energy sources, or of alternative sites for the facility, is required in this report. As stated in § 51.23, no discussion of the environmental impacts of the continued storage of spent fuel is required in this report.
              (c) Operating license renewal stage. (1) Each applicant for renewal of a license to operate a nuclear power plant under part 54 of this chapter shall submit with its application a separate document entitled “Applicant's Environmental Report—Operating License Renewal Stage.”
              (2) The report must contain a description of the proposed action, including the applicant's plans to modify the facility or its administrative control procedures as described in accordance with § 54.21 of this chapter. This report must describe in detail the affected environment around the plant, the modifications directly affecting the environment or any plant effluents, and any planned refurbishment activities. In addition, the applicant shall discuss in this report the environmental impacts of alternatives and any other matters described in § 51.45. The report is not required to include discussion of need for power or the economic costs and economic benefits of the proposed action or of alternatives to the proposed action except insofar as such costs and benefits are either essential for a determination regarding the inclusion of an alternative in the range of alternatives considered or relevant to mitigation. The environmental report need not discuss other issues not related to the environmental effects of the proposed action and the alternatives. As stated in § 51.23, no discussion of the environmental impacts of the continued storage of spent fuel is required in this report.
              (3) For those applicants seeking an initial renewed license and holding an operating license, construction permit, or combined license as of June 30, 1995, the environmental report shall include the information required in paragraph (c)(2) of this section subject to the following conditions and considerations:
              (i) The environmental report for the operating license renewal stage is not required to contain analyses of the environmental impacts of the license renewal issues identified as Category 1 issues in appendix B to subpart A of this part.
              (ii) The environmental report must contain analyses of the environmental impacts of the proposed action, including the impacts of refurbishment activities, if any, associated with license renewal and the impacts of operation during the renewal term, for those issues identified as Category 2 issues in appendix B to subpart A of this part. The required analyses are as follows:

              (A) If the applicant's plant utilizes cooling towers or cooling ponds and withdraws makeup water from a river, an assessment of the impact of the proposed action on water availability and competing water demands, the flow of the river, and related impacts on stream (aquatic) and riparian (terrestrial) ecological communities must be provided. The applicant shall also provide an assessment of the impacts of the withdrawal of water from the river on alluvial aquifers during low flow.
              (B) If the applicant's plant utilizes once-through cooling or cooling pond heat dissipation systems, the applicant shall provide a copy of current Clean Water Act 316(b) determinations and, if necessary, a 316(a) variance in accordance with 40 CFR part 125, or equivalent State permits and supporting documentation. If the applicant cannot provide these documents, it shall assess the impact of the proposed action on fish and shellfish resources resulting from thermal changes and impingement and entrainment.
              (C) If the applicant's plant pumps more than 100 gallons (total onsite) of groundwater per minute, an assessment of the impact of the proposed action on groundwater must be provided.
              (D) If the applicant's plant is located at an inland site and utilizes cooling ponds, an assessment of the impact of the proposed action on groundwater quality must be provided.
              (E) All license renewal applicants shall assess the impact of refurbishment, continued operations, and other license-renewal-related construction activities on important plant and animal habitats. Additionally, the applicant shall assess the impact of the proposed action on threatened or endangered species in accordance with Federal laws protecting wildlife, including but not limited to, the Endangered Species Act, and essential fish habitat in accordance with the Magnuson-Stevens Fishery Conservation and Management Act.
              (F) [Reserved]
              (G) If the applicant's plant uses a cooling pond, lake, or canal or discharges into a river, an assessment of the impact of the proposed action on public health from thermophilic organisms in the affected water must be provided.
              (H) If the applicant's transmission lines that were constructed for the specific purpose of connecting the plant to the transmission system do not meet the recommendations of the National Electric Safety Code for preventing electric shock from induced currents, an assessment of the impact of the proposed action on the potential shock hazard from the transmission lines must be provided.
              (I)-(J) [Reserved]
              (K) All applicants shall identify any potentially affected historic or archaeological properties and assess whether any of these properties will be affected by future plant operations and any planned refurbishment activities in accordance with the National Historic Preservation Act.
              (L) If the staff has not previously considered severe accident mitigation alternatives for the applicant's plant in an environmental impact statement or related supplement or in an environmental assessment, a consideration of alternatives to mitigate severe accidents must be provided.
              (M) [Reserved]
              (N) Applicants shall provide information on the general demographic composition of minority and low-income populations and communities (by race and ethnicity) residing in the immediate vicinity of the plant that could be affected by the renewal of the plant's operating license, including any planned refurbishment activities, and ongoing and future plant operations.
              (O) Applicants shall provide information about other past, present, and reasonably foreseeable future actions occurring in the vicinity of the nuclear plant that may result in a cumulative effect.
              (P) An applicant shall assess the impact of any documented inadvertent releases of radionuclides into groundwater. The applicant shall include in its assessment a description of any groundwater protection program used for the surveillance of piping and components containing radioactive liquids for which a pathway to groundwater may exist. The assessment must also include a description of any past inadvertent releases and the projected impact to the environment (e.g., aquifers, rivers, lakes, ponds, ocean) during the license renewal term.

              (iii) The report must contain a consideration of alternatives for reducing adverse impacts, as required by § 51.45(c), for all Category 2 license renewal issues in appendix B to subpart A of this part. No such consideration is required for Category 1 issues in appendix B to subpart A of this part.
              (iv) The environmental report must contain any new and significant information regarding the environmental impacts of license renewal of which the applicant is aware.
              (d) Postoperating license stage. Each applicant for a license amendment authorizing decommissioning activities for a production or utilization facility either for unrestricted use or based on continuing use restrictions applicable to the site; and each applicant for a license amendment approving a license termination plan or decommissioning plan under § 50.82 of this chapter either for unrestricted use or based on continuing use restrictions applicable to the site; and each applicant for a license or license amendment to store spent fuel at a nuclear power reactor after expiration of the operating license for the nuclear power reactor shall submit with its application a separate document, entitled “Supplement to Applicant's Environmental Report—Post Operating License Stage,” which will update “Applicant's Environmental Report—Operating License Stage,” as appropriate, to reflect any new information or significant environmental change associated with the applicant's proposed decommissioning activities or with the applicant's proposed activities with respect to the planned storage of spent fuel. As stated in § 51.23, no discussion of the environmental impacts of the continued storage of spent fuel is required in this report. The “Supplement to Applicant's Environmental Report—Post Operating License Stage” may incorporate by reference any information contained in “Applicant's Environmental Report—Construction Permit Stage.”
              [61 FR 66543, Dec. 18, 1996, as amended at 64 FR 48506, Sept. 3, 1999; 68 FR 58810, Oct. 10, 2003; 72 FR 49513, Aug. 28, 2007; 78 FR 37316, June 20, 2013; 79 FR 56260, Sept. 19, 2014; 79 FR 66604, Nov. 10, 2014]

              § 51.54
              Environmental report—manufacturing license.
              (a) Each applicant for a manufacturing license under subpart F of part 52 of this chapter shall submit with its application a separate document entitled, “Applicant's Environmental Report—Manufacturing License.” The environmental report must address the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives into the design of the reactor to be manufactured. The environmental report need not address the environmental impacts associated with manufacturing the reactor under the manufacturing license, the benefits and impacts of utilizing the reactor in a nuclear power plant, or an evaluation of alternative energy sources.
              (b) Each applicant for an amendment to a manufacturing license shall submit with its application a separate document entitled, “Applicant's Supplemental Environmental Report—Amendment to Manufacturing License.” The environmental report must address whether the design change which is the subject of the proposed amendment either renders a severe accident mitigation design alternative previously rejected in an environmental assessment to become cost beneficial, or results in the identification of new severe accident mitigation design alternatives that may be reasonably incorporated into the design of the manufactured reactor. The environmental report need not address the environmental impacts associated with manufacturing the reactor under the manufacturing license.
              [72 FR 49513, Aug. 28, 2007]

              § 51.55
              Environmental report—standard design certification.
              (a) Each applicant for a standard design certification under subpart B of part 52 of this chapter shall submit with its application a separate document entitled, “Applicant's Environmental Report—Standard Design Certification.” The environmental report must address the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives in the design to be certified.

              (b) Each applicant for an amendment to a design certification shall submit with its application a separate document entitled, “Applicant's Supplemental Environmental Report—Amendment to Standard Design Certification.” The environmental report must address whether the design change which is the subject of the proposed amendment either renders a severe accident mitigation design alternative previously rejected in an environmental assessment to become cost beneficial, or results in the identification of new severe accident mitigation design alternatives that may be reasonably incorporated into the design certification.
              [72 FR 49513, Aug. 28, 2007]

              § 51.58
              Environmental report—number of copies; distribution.

              (a) Each applicant for a license or permit to site, construct, manufacture, or operate a production or utilization facility covered by §§ 51.20(b)(1), (b)(2), (b)(3), or (b)(4), each applicant for renewal of an operating or combined license for a nuclear power plant, each applicant for a license amendment authorizing the decommissioning of a production or utilization facility covered by § 51.20, and each applicant for a license or license amendment to store spent fuel at a nuclear power plant after expiration of the operating license or combined license for the nuclear power plant shall submit a copy to the Director of the Office of Nuclear Reactor Regulation, or the Director of the Office of Nuclear Material Safety and Safeguards, as appropriate, of an environmental report or any supplement to an environmental report. These reports must be sent either by mail addressed: ATTN: Document Control Desk; U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland, between the hours of 7:30 a.m. and 4:15 p.m. eastern time; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the communication is on paper, the signed original must be sent. If a submission due date falls on a Saturday, Sunday, or Federal holiday, the next Federal working day becomes the official due date. The applicant shall maintain the capability to generate additional copies of the environmental report or any supplement to the environmental report for subsequent distribution to parties and Boards in the NRC proceedings; Federal, State, and local officials; and any affected Indian Tribes, in accordance with written instructions issued by the Director of the Office of Nuclear Reactor Regulation or the Director of the Office of Nuclear Material Safety and Safeguards, as appropriate.
              (b) Each applicant for a license to manufacture a nuclear power reactor, or for an amendment to a license to manufacture, seeking approval of the final design of the nuclear power reactor under subpart F of part 52 of this chapter, shall submit to the Commission an environmental report or any supplement to an environmental report in the manner specified in § 50.3 of this chapter. The applicant shall maintain the capability to generate additional copies of the environmental report or any supplement to the environmental report for subsequent distribution to parties and Boards in the NRC proceeding; Federal, State, and local officials; and any affected Indian Tribes, in accordance with written instructions issued by the Director of the Office of Nuclear Reactor Regulation.
              [72 FR 49513, Aug. 28, 2007, as amended at 74 FR 62682, Dec. 1, 2009; 84 FR 65645, Nov. 29, 2019]

            environmental reports—materials licenses

              § 51.60
              Environmental report—materials licenses.
              (a) Each applicant for a license or other form of permission, or an amendment to or renewal of a license or other form of permission issued pursuant to parts 30, 32, 33, 34, 35, 36, 39, 40, 61, 70 and/or 72 of this chapter, and covered by paragraphs (b)(1) through (b)(5) of this section, shall submit with its application to: ATTN: Document Control Desk, Director, Nuclear Material Safety and Safeguards, a separate document, entitled “Applicant's Environmental Report” or “Supplement to Applicant's Environmental Report,” as appropriate. The “Applicant's Environmental Report” shall contain the information specified in § 51.45. If the application is for an amendment to or a renewal of a license or other form of permission for which the applicant has previously submitted an environmental report, the supplement to applicant's environmental report may be limited to incorporating by reference, updating or supplementing the information previously submitted to reflect any significant environmental change, including any significant environmental change resulting from operational experience or a change in operations or proposed decommissioning activities. If the applicant is the U.S. Department of Energy, the environmental report may be in the form of either an environmental impact statement or an environmental assessment, as appropriate.
              (b) As required by paragraph (a) of this section, each applicant shall prepare an environmental report for the following types of actions:
              (1) Issuance or renewal of a license or other form of permission for:
              (i) Possession and use of special nuclear material for processing and fuel fabrication, scrap recovery, or conversion of uranium hexafluoride pursuant to part 70 of this chapter.
              (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter.
              (iii) Storage of spent fuel in an independent spent fuel storage installation (ISFSI) or the storage of spent fuel or high-level radio-active waste in a monitored retrievable storage installation (MRS) pursuant to part 72 of this chapter.
              (iv) Receipt and disposal of radioactive waste from other persons pursuant to part 61 of this chapter.
              (v) Processing of source material for extraction of rare earth and other metals.
              (vi) Use of radioactive tracers in field flood studies involving secondary and tertiary oil and gas recovery.
              (vii) Construction and operation of a uranium enrichment facility.
              (2) Issuance of an amendment that would authorize or result in (i) a significant expansion of a site, (ii) a significant change in the types of effluents, (iii) a significant increase in the amounts of effluents, (iv) a significant increase in individual or cumulative occupational radiation exposure, (v) a significant increase in the potential for or consequences from radiological accidents, or (vi) a significant increase in spent fuel storage capacity, in a license or other form of permission to conduct an activity listed in paragraph (b)(1) of this section.
              (3) Amendment of a license to authorize the decommissioning of an independent spent fuel storage installation (ISFSI) or a monitored retrievable storage installation (MRS) pursuant to part 72 of this chapter.
              (4) Issuance of a license amendment pursuant to part 61 of this chapter authorizing (i) closure of a land disposal site, (ii) transfer of the license to the disposal site owner for the purpose of institutional control, or (iii) termination of the license at the end of the institutional control period.
              (5) Any other licensing action for which the Commission determines an Environmental Report is necessary.
              [49 FR 9381, Mar. 12, 1984, as amended at 53 FR 31681, Aug. 19, 1988; 57 FR 18392, Apr. 30, 1992; 58 FR 7737, Feb. 9, 1993; 62 FR 26732, May 14, 1997; 68 FR 58811, Oct. 10, 2003]

              § 51.61
              Environmental report—independent spent fuel storage installation (ISFSI) or monitored retrievable storage installation (MRS) license.

              Each applicant for issuance of a license for storage of spent fuel in an independent spent fuel storage installation (ISFSI) or for the storage of spent fuel and high-level radioactive waste in a monitored retrievable storage installation (MRS) pursuant to part 72 of this chapter shall submit with its application to: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, a separate document entitled “Applicant's Environmental Report—ISFSI License” or “Applicant's Environmental Report—MRS License,” as appropriate. If the applicant is the U.S. Department of Energy, the environmental report may be in the form of either an environmental impact statement or an environmental assessment, as appropriate. The environmental report shall contain the information specified in § 51.45 and shall address the siting evaluation factors contained in subpart E of part 72 of this chapter. As stated in § 51.23, no discussion of the environmental impacts of the continued storage of spent fuel in an ISFSI is required in this report.
              [79 FR 56261, Sept. 19, 2014]

              § 51.62
              Environmental report—land disposal of radioactive waste licensed under 10 CFR part 61.

              (a) Each applicant for issuance of a license for land disposal of radioactive waste pursuant to part 61 of this chapter shall submit with its application to: ATTN: Document Control Desk, Director of Nuclear Material Safety and Safeguards, a separate document, entitled “Applicant's Environmental Report—License for Land Disposal of Radioactive Waste.” The environmental report and any supplement to the environmental report may incorporate by reference information contained in the application or in any previous application, statement or report filed with the Commission provided that such references are clear and specific and that copies of the information so incorporated are available at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public Document Room.
              (b) The environmental report shall contain the information specified in § 51.45, shall address the applicant's environmental monitoring program required by §§ 61.12(l), 61.53 and 61.59(b) of this chapter, and shall be as complete as possible in the light of information that is available at the time the environmental report is submitted.
              (c) The applicant shall supplement the environmental report in a timely manner as necessary to permit the Commission to review, prior to issuance, amendment or renewal of a license, new information regarding the environmental impact of previously proposed activities, information regarding the environmental impact of any changes in previously proposed activities, or any significant new information regarding the environmental impact of closure activities and long-term performance of the disposal site.
              [49 FR 9381, Mar. 12, 1984, as amended at 53 FR 43420, Oct. 27, 1988; 64 FR 48952, Sept. 9, 1999; 68 FR 58811, Oct. 10, 2003]

              § 51.66
              Environmental report—number of copies; distribution.
              Each applicant for a license or other form of permission, or an amendment to or renewal of a license or other form of permission issued under parts 30, 32, 33, 34, 35, 36, 39, 40, 61, 70, and/or 72 of this chapter, and covered by §§ 51.60(b)(1) through (6); or by §§ 51.61 or 51.62 shall submit to the Director of Nuclear Material Safety and Safeguards an environmental report or any supplement to an environmental report in the manner specified in § 51.58(a). The applicant shall maintain the capability to generate additional copies of the environmental report or any supplement to the environmental report for subsequent distribution to Federal, State, and local officials, and any affected Indian Tribes in accordance with written instructions issued by the Director of Nuclear Material Safety and Safeguards.
              [72 FR 49514, Aug. 28, 2007]

              § 51.67
              Environmental information concerning geologic repositories.

              (a) In lieu of an environmental report, the Department of Energy, as an applicant for a license or license amendment pursuant to part 60 or 63 of this chapter, shall submit to the Commission any final environmental impact statement which the Department prepares in connection with any geologic repository developed under Subtitle A of Title I, or under Title IV, of the Nuclear Waste Policy Act of 1982, as amended. (See § 60.22 or § 63.22 of this chapter as to the required time and manner of submission.) The statement shall include, among the alternatives under consideration, denial of a license or construction authorization by the Commission.
              (b) Under applicable provisions of law, the Department of Energy may be required to supplement its final environmental impact statement if it makes a substantial change in its proposed action that is relevant to environmental concerns or determines that there are significant new circumstances or information relevant to environmental concerns and bearing on the proposed action or its impacts. The Department shall submit any supplement to its final environmental impact statement to the Commission. (See § 60.22 or § 63.22 of this chapter as to the required time and manner of submission.)
              (c) Whenever the Department of Energy submits a final environmental impact statement, or a final supplement to an environmental impact statement, to the Commission pursuant to this section, it shall also inform the Commission of the status of any civil action for judicial review initiated pursuant to section 119 of the Nuclear Waste Policy Act of 1982. This status report, which the Department shall update from time to time to reflect changes in status, shall:
              (1) State whether the environmental impact statement has been found by the courts of the United States to be adequate or inadequate; and
              (2) Identify any issues relating to the adequacy of the environmental impact statement that may remain subject to judicial review.
              [54 FR 27870, July 3, 1989, as amended at 66 FR 55791, Nov. 2, 2001]

            environmental reports—rulemaking

              § 51.68
              Environmental report—rulemaking.
              Petitioners for rulemaking requesting amendments of parts 30, 31, 32, 33, 34, 35, 36, 39, 40 or part 70 of this chapter concerning the exemption from licensing and regulatory requirements of or authorizing general licenses for any equipment, device, commodity or other product containing byproduct material, source material or special nuclear material shall submit with the petition a separate document entitled “Petitioner's Environmental Report,” which shall contain the information specified in § 51.45.
              [68 FR 58811, Oct. 10, 2003]

            Environmental Impact Statements

            draft environmental impact statements—general requirements

              § 51.70
              Draft environmental impact statement—general.
              (a) The NRC staff will prepare a draft environmental impact statement as soon as practicable after publication of the notice of intent to prepare an environmental impact statement and completion of the scoping process. To the fullest extent practicable, environmental impact statements will be prepared concurrently or integrated with environmental impact analyses and related surveys and studies required by other Federal law.
              (b) The draft environmental impact statement will be concise, clear and analytic, will be written in plain language with appropriate graphics, will state how alternatives considered in it and decisions based on it will or will not achieve the requirements of sections 101 and 102(1) of NEPA and of any other relevant and applicable environmental laws and policies, will identify any methodologies used and sources relied upon, and will be supported by evidence that the necessary environmental analyses have been made. The format provided in section 1(a) of appendix A of this subpart should be used. The NRC staff will independently evaluate and be responsible for the reliability of all information used in the draft environmental impact statement.
              (c) The Commission will cooperate with State and local agencies to the fullest extent possible to reduce duplication between NEPA and State and local requirements, in accordance with 40 CFR 1506.2 (b) and (c).

              § 51.71
              Draft environmental impact statement—contents.
              (a) Scope. The draft environmental impact statement will be prepared in accordance with the scope decided upon in the scoping process required by §§ 51.26 and 51.29. As appropriate and to the extent required by the scope, the draft statement will address the topics in paragraphs (b), (c), (d) and (e) of this section and the matters specified in §§ 51.45, 51.50, 51.51, 51.52, 51.53, 51.54, 51.61 and 51.62.
              (b) Analysis of major points of view. To the extent sufficient information is available, the draft environmental impact statement will include consideration of major points of view concerning the environmental impacts of the proposed action and the alternatives, and contain an analysis of significant problems and objections raised by other Federal, State, and local agencies, by any affected Indian Tribes, and by other interested persons.
              (c) Status of compliance. The draft environmental impact statement will list all Federal permits, licenses, approvals, and other entitlements which must be obtained in implementing the proposed action and will describe the status of compliance with those requirements. If it is uncertain whether a Federal permit, license, approval, or other entitlement is necessary, the draft environmental impact statement will so indicate.
              (d) Analysis. Unless excepted in this paragraph or § 51.75, the draft environmental impact statement will include a preliminary analysis that considers and weighs the environmental effects, including any cumulative effects, of the proposed action; the environmental impacts of alternatives to the proposed action; and alternatives available for reducing or avoiding adverse environmental effects. Additionally, the draft environmental impact statement will include a consideration of the economic, technical, and other benefits and costs of the proposed action and alternatives. The draft environmental impact statement will indicate what other interests and considerations of Federal policy, including factors not related to environmental quality, if applicable, are relevant to the consideration of environmental effects of the proposed action identified under paragraph (a) of this section. The draft supplemental environmental impact statement prepared at the license renewal stage under § 51.95(c) need not discuss the economic or technical benefits and costs of either the proposed action or alternatives except if benefits and costs are either essential for a determination regarding the inclusion of an alternative in the range of alternatives considered or relevant to mitigation. In addition, the supplemental environmental impact statement prepared at the license renewal stage need not discuss other issues not related to the environmental effects of the proposed action and associated alternatives. The draft supplemental environmental impact statement for license renewal prepared under § 51.95(c) will rely on conclusions as amplified by the supporting information in the GEIS for issues designated as Category 1 in appendix B to subpart A of this part. The draft supplemental environmental impact statement must contain an analysis of those issues identified as Category 2 in appendix B to subpart A of this part that are open for the proposed action. The analysis for all draft environmental impact statements will, to the fullest extent practicable, quantify the various factors considered. To the extent that there are important qualitative considerations or factors that cannot be quantified, these considerations or factors will be discussed in qualitative terms. Consideration will be given to compliance with environmental quality standards and requirements that have been imposed by Federal, State, regional, and local agencies having responsibility for environmental protection, including applicable zoning and land-use regulations and water pollution limitations or requirements issued or imposed under the Federal Water Pollution Control Act. The environmental impact of the proposed action will be considered in the analysis with respect to matters covered by environmental quality standards and requirements irrespective of whether a certification or license from the appropriate authority has been obtained. 3
                 While satisfaction of Commission standards and criteria pertaining to radiological effects will be necessary to meet the licensing requirements of the Atomic Energy Act, the analysis will, for the purposes of NEPA, consider the radiological effects of the proposed action and alternatives.

                  3 Compliance with the environmental quality standards and requirements of the Federal Water Pollution Control Act (imposed by EPA or designated permitting states) is not a substitute for, and does not negate the requirement for NRC to weigh all environmental effects of the proposed action, including the degradation, if any, of water quality, and to consider alternatives to the proposed action that are available for reducing adverse effects. Where an environmental assessment of aquatic impact from plant discharges is available from the permitting authority, the NRC will consider the assessment in its determination of the magnitude of environmental impacts for striking an overall cost-benefit balance at the construction permit and operating license and early site permit and combined license stages, and in its determination of whether the adverse environmental impacts of license renewal are so great that preserving the option of license renewal for energy planning decision-makers would be unreasonable at the license renewal stage. When no such assessment of aquatic impacts is available from the permitting authority, NRC will establish on its own, or in conjunction with the permitting authority and other agencies having relevant expertise, the magnitude of potential impacts for striking an overall cost-benefit balance for the facility at the construction permit and operating license and early site permit and combined license stages, and in its determination of whether the adverse environmental impacts of license renewal are so great that preserving the option of license renewal for energy planning decision-makers would be unreasonable at the license renewal stage.

              (e) Effect of limited work authorization. If a limited work authorization was issued either in connection with or subsequent to an early site permit, or in connection with a construction permit or combined license application, then the environmental impact statement for the construction permit or combined license application will not address or consider the sunk costs associated with the limited work authorization.
              (f) Preliminary recommendation. The draft environmental impact statement normally will include a preliminary recommendation by the NRC staff respecting the proposed action. This preliminary recommendation will be based on the information and analysis described in paragraphs (a) through (d) of this section and §§ 51.75, 51.76, 51.80, 51.85, and 51.95, as appropriate, and will be reached after considering the environmental effects of the proposed action and reasonable alternatives, 4
                 and, except for supplemental environmental impact statements for the operating license renewal stage prepared pursuant to § 51.95(c), after weighing the costs and benefits of the proposed action. In lieu of a recommendation, the NRC staff may indicate in the draft statement that two or more alternatives remain under consideration.

                  4 The consideration of reasonable alternatives to a proposed action involving nuclear power reactors (e.g., alternative energy sources) is intended to assist the NRC in meeting its NEPA obligations and does not preclude any State authority from making separate determinations with respect to these alternatives and in no way preempts, displaces, or affects the authority of States or other Federal agencies to address these issues.

              [49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28488, June 5, 1996; 61 FR 66544, Dec. 18, 1996; 72 FR 49514, Aug. 28, 2007; 72 FR 57445, Oct. 9, 2007; 78 FR 37317, June 20, 2013]

              § 51.72
              Supplement to draft environmental impact statement.

              (a) The NRC staff will prepare a supplement to a draft environmental impact statement for which a notice of availability has been published in the Federal Register as provided in § 51.117, if:
              (1) There are substantial changes in the proposed action that are relevant to environmental concerns; or
              (2) There are significant new circumstances or information relevant to environmental concerns and bearing on the proposed action or its impacts.

              (b) The NRC staff may prepare a supplement to a draft environmental impact statement when, in its opinion, preparation of a supplement will further the purposes of NEPA.

              (c) The supplement to a draft environmental impact statement will be prepared and noticed in the same manner as the draft environmental impact statement except that a scoping process need not be used.

              § 51.73
              Request for comments on draft environmental impact statement.

              Each draft environmental impact statement and each supplement to a draft environmental impact statement distributed in accordance with § 51.74, and each news release provided pursuant to § 51.74(d) will be accompanied by or include a request for comments on the proposed action and on the draft environmental impact statement or any supplement to the draft environmental impact statement and will state where comments should be submitted and the date on which the comment period closes. A minimum comment period of 45 days will be provided. The comment period will be calculated from the date on which the Environmental Protection Agency notice stating that the draft statement or the supplement to the draft statement has been filed with EPA is published in the Federal Register. If no comments are provided within the time specified, it will be presumed, unless the agency or person requests an extension of time, that the agency or person has no comment to make. To the extent practicable, NRC staff will grant reasonable requests for extensions of time of up to fifteen (15) days.

              § 51.74
              Distribution of draft environmental impact statement and supplement to draft environmental impact statement; news releases.
              (a) A copy of the draft environmental impact statement will be distributed to:
              (1) The Environmental Protection Agency.
              (2) Any other Federal agency which has special expertise or jurisdiction by law with respect to any environmental impact involved or which is authorized to develop and enforce relevant environmental standards.
              (3) The applicant or petitioner for rulemaking and any other party to the proceeding.
              (4) Appropriate State and local agencies authorized to develop and enforce relevant environmental standards.
              (5) Appropriate State, regional and metropolitan clearinghouses.
              (6) Appropriate Indian Tribes when the proposed action may have an environmental impact on a reservation.
              (7) Upon written request, any organization or group included in the master list of interested organizations and groups maintained under § 51.122.
              (8) Upon written request, any other person to the extent available.
              (b) Additional copies will be made available in accordance with § 51.123.
              (c) A supplement to a draft environmental impact statement will be distributed in the same manner as the draft environmental impact statement to which it relates.
              (d) News releases stating the availability for comment and place for obtaining or inspecting a draft environmental statement or supplement will be provided to local newspapers and other appropriate media.

              (e) A notice of availability will be published in the Federal Register in accordance with § 51.117.

            draft environmental impact statements—production and utilization facilities

              § 51.75
              Draft environmental impact statement—construction permit, early site permit, or combined license.
              (a) Construction permit stage. A draft environmental impact statement relating to issuance of a construction permit for a production or utilization facility will be prepared in accordance with the procedures and measures described in §§ 51.70, 51.71, 51.72, and 51.73. The contribution of the environmental effects of the uranium fuel cycle activities specified in § 51.51 shall be evaluated on the basis of impact values set forth in Table S-3, Table of Uranium Fuel Cycle Environmental Data, which shall be set out in the draft environmental impact statement. With the exception of radon-222 and technetium-99 releases, no further discussion of fuel cycle release values and other numerical data that appear explicitly in the table shall be required.5
                 The impact statement shall take account of dose commitments and health effects from fuel cycle effluents set forth in Table S-3 and shall in addition take account of economic, socioeconomic, and possible cumulative impacts and other fuel cycle impacts as may reasonably appear significant. As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated into the environmental impact statement.

                  5 Values for releases of Rn-222 and Tc-99 are not given in the table. The amount and significance of Rn-222 releases from the fuel cycle and Tc-99 releases from waste management or reprocessing activities shall be considered in the draft environmental impact statement and may be the subject of litigation in individual licensing proceedings.

              (b) Early site permit stage. A draft environmental impact statement relating to issuance of an early site permit for a production or utilization facility will be prepared in accordance with the procedures and measures described in §§ 51.70, 51.71, 51.72, 51.73, and this section. The contribution of the environmental effects of the uranium fuel cycle activities specified in § 51.51 shall be evaluated on the basis of impact values set forth in Table S-3, Table of Uranium Fuel Cycle Environmental Data, which shall be set out in the draft environmental impact statement. With the exception of radon-222 and technetium-99 releases, no further discussion of fuel cycle release values and other numerical data that appear explicitly in the table shall be required.5 The impact statement shall take account of dose commitments and health effects from fuel cycle effluents set forth in Table S-3 and shall in addition take account of economic, socioeconomic, and possible cumulative impacts and other fuel cycle impacts as may reasonably appear significant. As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated into the environmental impact statement. The draft environmental impact statement must include an evaluation of alternative sites to determine whether there is any obviously superior alternative to the site proposed. The draft environmental impact statement must also include an evaluation of the environmental effects of construction and operation of a reactor, or reactors, which have design characteristics that fall within the site characteristics and design parameters for the early site permit application, but only to the extent addressed in the early site permit environmental report or otherwise necessary to determine whether there is any obviously superior alternative to the site proposed. The draft environmental impact statement must not include an assessment of the economic, technical, or other benefits (for example, need for power) and costs of the proposed action or an evaluation of alternative energy sources, unless these matters are addressed in the early site permit environmental report.
              (c) Combined license stage. A draft environmental impact statement relating to issuance of a combined license that does not reference an early site permit will be prepared in accordance with the procedures and measures described in §§ 51.70, 51.71, 51.72, and 51.73. The contribution of the environmental effects of the uranium fuel cycle activities specified in § 51.51 shall be evaluated on the basis of impact values set forth in Table S-3, Table of Uranium Fuel Cycle Environmental Data, which shall be set out in the draft environmental impact statement. With the exception of radon-222 and technetium-99 releases, no further discussion of fuel cycle release values and other numerical data that appear explicitly in the table shall be required.5 The impact statement shall take account of dose commitments and health effects from fuel cycle effluents set forth in Table S-3 and shall in addition take account of economic, socioeconomic, and possible cumulative impacts and other fuel cycle impacts as may reasonably appear significant. As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated into the environmental impact statement.
              (1) Combined license application referencing an early site permit. If the combined license application references an early site permit, then the NRC staff shall prepare a draft supplement to the early site permit environmental impact statement. The supplement must be prepared in accordance with § 51.92(e).
              (2) Combined license application referencing a standard design certification. If the combined license application references a standard design certification and the site characteristics of the combined license's site fall within the site parameters specified in the design certification environmental assessment, then the draft combined license environmental impact statement shall incorporate by reference the design certification environmental assessment, and summarize the findings and conclusions of the environmental assessment with respect to severe accident mitigation design alternatives.
              (3) Combined license application referencing a manufactured reactor. If the combined license application proposes to use a manufactured reactor and the site characteristics of the combined license's site fall within the site parameters specified in the manufacturing license environmental assessment, then the draft combined license environmental impact statement shall incorporate by reference the manufacturing license environmental assessment, and summarize the findings and conclusions of the environmental assessment with respect to severe accident mitigation design alternatives. The combined license environmental impact statement report will not address the environmental impacts associated with manufacturing the reactor under the manufacturing license.
              [72 FR 49514, Aug. 28, 2007, as amended at 79 FR 56261, Sept. 19, 2014]

              § 51.76
              Draft environmental impact statement—limited work authorization.
              The NRC will prepare a draft environmental impact statement relating to issuance of a limited work authorization in accordance with the procedures and measures described in §§ 51.70, 51.71, and 51.73, as further supplemented or modified in the following paragraphs.
              (a) Limited work authorization submitted as part of complete construction permit or combined license application. If the application for a limited work authorization is submitted as part of a complete construction permit or combined license application, then the NRC may prepare a partial draft environmental impact statement. The analysis called for by § 51.71(d) must be limited to the activities proposed to be conducted under the limited work authorization. Alternatively, the NRC may prepare a complete draft environmental impact statement prepared in accordance with § 51.75(a) or (c), as applicable.
              (b) Phased application for limited work authorization under § 2.101(a)(9) of this chapter. If the application for a limited work authorization is submitted in accordance with § 2.101(a)(9) of this chapter, then the draft environmental impact statement for part one of the application may be limited to consideration of the activities proposed to be conducted under the limited work authorization, and the proposed redress plan. However, if the environmental report contains the full set of information required to be submitted under § 51.50(a) or (c), then a draft environmental impact statement must be prepared in accordance with § 51.75(a) or (c), as applicable. Siting issues, including whether there is an obviously superior alternative site, or issues related to operation of the proposed nuclear power plant at the site, including need for power, may not be considered. After part two of the application is docketed, the NRC will prepare a draft supplement to the final environmental impact statement for part two of the application under § 51.72. No updating of the information contained in the final environmental impact statement prepared for part one is necessary in preparation of the supplemental environmental impact statement. The draft supplement must consider all environmental impacts associated with the prior issuance of the limited work authorization, but may not address or consider the sunk costs associated with the limited work authorization.
              (c) Limited work authorization submitted as part of an early site permit application. If the application for a limited work authorization is submitted as part of an application for an early site permit, then the NRC will prepare an environmental impact statement in accordance with § 51.75(b). However, the analysis called for by § 51.71(d) must also address the activities proposed to be conducted under the limited work authorization.
              (d) Limited work authorization request submitted by an early site permit holder. If the application for a limited work authorization is submitted by a holder of an early site permit, then the NRC will prepare a draft supplement to the environmental impact statement for the early site permit. The supplement is limited to consideration of the activities proposed to be conducted under the limited work authorization, the adequacy of the proposed redress plan, and whether there is new and significant information identified with respect to issues related to the impacts of construction of the facility that were resolved in the early site permit proceeding with respect to the environmental impacts of the activities to be conducted under the limited work authorization. No other updating of the information contained in the final environmental impact statement prepared for the early site permit is required.
              (e) Limited work authorization for a site where an environmental impact statement was prepared, but the facility construction was not completed. If the limited work authorization is for activities to be conducted at a site for which the Commission has previously prepared an environmental impact statement for the construction and operation of a nuclear power plant, and a construction permit was issued but construction of the plant was not completed, then the draft environmental impact statement shall incorporate by reference the earlier environmental impact statement. The draft environmental impact statement must be limited to a consideration of whether there is significant new information with respect to the environmental impacts of construction, relevant to the activities to be conducted under the limited work authority, so that the conclusion of the referenced environmental impact statement on the impacts of construction would, when analyzed in accordance with § 51.71, lead to the conclusion that the limited work authorization should not be issued or should be issued with appropriate conditions.
              (f) Draft environmental impact statement. A draft environmental impact statement prepared under this section must separately evaluate the environmental impacts and proposed alternatives attributable to the activities proposed to be conducted under the limited work authorization. However, if the “Applicant's Environmental Report—Limited Work Authorization Stage,” also contains the information required to be submitted in the environmental report required under § 51.50, then the environmental impact statement must address the impacts of construction and operation for the proposed facility (including the environmental impacts attributable to the limited work authorization), and discuss the overall costs and benefits balancing for the underlying proposed action, in accordance with § 51.71, and § 51.75(a) or (c), as applicable.
              [72 FR 57445, Oct. 9, 2007]

              § 51.77
              Distribution of draft environmental impact statement.
              (a) In addition to the distribution authorized by § 51.74, a copy of a draft environmental statement for a licensing action for a production or utilization facility, except an action authorizing issuance, amendment or renewal of a license to manufacture a nuclear power reactor pursuant to 10 CFR part 52, appendix M will also be distributed to:
              (1) The chief executive of the municipality or county identified in the draft environmental impact statement as the preferred site for the proposed facility or activity.
              (2) Upon request, the chief executive of each municipality or county identified in the draft environmental impact statement as an alternative site.
              (b) Additional copies will be made available in accordance with § 51.123.
              [49 FR 9381, Mar. 12, 1984, as amended at 54 FR 15398, Apr. 18, 1989]

            draft environmental impact statements—materials licenses

              § 51.80
              Draft environmental impact statement—materials license.
              (a) The NRC staff will either prepare a draft environmental impact statement or as provided in § 51.92, a supplement to a final environmental impact statement for each type of action identified in § 51.20(b) (7) through (12). Except as the context may otherwise require, procedures and measures similar to those described in §§ 51.70, 51.71, 51.72 and 51.73 will be followed.
              (b)(1) Independent spent fuel storage installation (ISFSI). As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated in the environmental impact statement.
              (2) Monitored retrievable storage installation (MRS). As provided in sections 141 (c), (d), and (e) and 148 (a) and (c) of the Nuclear Waste Policy Act of 1982, as amended (NWPA) (96 Stat. 2242, 2243, 42 U.S.C. 10161 (c), (d), (e); 101 Stat. 1330-235, 1330-236, 42 U.S.C. 10168 (a) and (c)), a draft environmental impact statement for the construction of a monitored retrievable storage installation (MRS) will not address the need for the MRS or any alternative to the design criteria for an MRS set forth in section 141(b)(1) of the NWPA (96 Stat. 2242, 42 U.S.C. 10161(b)(1)) but may consider alternative facility designs which are consistent with these design criteria.
              [49 FR 34695, Aug. 31, 1984, as amended at 53 FR 31682, Aug. 19, 1988; 79 FR 56262, Sept. 19, 2014]

              § 51.81
              Distribution of draft environmental impact statement.
              Copies of the draft environmental impact statement and any supplement to the draft environmental impact statement will be distributed in accordance with the provisions of § 51.74.

            draft environmental impact statements—rulemaking

              § 51.85
              Draft environmental impact statement—rulemaking.
              Except as the context may otherwise require, procedures and measures similar to those described in §§ 51.70, 51.71, 51.72 and 51.73 will be followed in proceedings for rulemaking for which the Commission has determined to prepare an environmental impact statement.

              § 51.86
              Distribution of draft environmental impact statement.
              Copies of the draft environmental impact statement and any supplement to the draft environmental impact statement will be distributed in accordance with the provisions of § 51.74.

            legislative environmental impact statements—proposals for legislation

              § 51.88
              Proposals for legislation.
              The Commission will, as a matter of policy, follow the provisions of 40 CFR 1506.8 regarding the NEPA process for proposals for legislation.

            final environmental impact statements—general requirements

              § 51.90
              Final environmental impact statement—general.
              After receipt and consideration of comments requested pursuant to §§ 51.73 and 51.117, the NRC staff will prepare a final environmental impact statement in accordance with the requirements in §§ 51.70(b) and 51.71 for a draft environmental impact statement. The format provided in section 1(a) of appendix A of this subpart should be used.

              § 51.91
              Final environmental impact statement—contents.
              (a)(1) The final environmental impact statement will include responses to any comments on the draft environmental impact statement or on any supplement to the draft environmental impact statement. Responses to comments may include:
              (i) Modification of alternatives, including the proposed action;
              (ii) Development and evaluation of alternatives not previously given serious consideration;
              (iii) Supplementation or modification of analyses;
              (iv) Factual corrections;

              (v) Explanation of why comments do not warrant further response, citing sources, authorities or reasons which support this conclusion.
              (2) All substantive comments received on the draft environmental impact statement or any supplement to the draft environmental impact statement (or summaries thereof where the response has been exceptionally voluminous) will be attached to the final statement, whether or not each comment is discussed individually in the text of the statement.
              (3) If changes in the draft environmental impact statement in response to comments are minor and are confined either to factual corrections or to explanations of why the comments do not warrant further response, the changes may be made by attaching errata sheets to the draft statement. The entire document with a new cover may then be issued as the final environmental impact statement.
              (b) The final environmental impact statement will discuss any relevant responsible opposing view not adequately discussed in the draft environmental impact statement or in any supplement to the draft environmental impact statement, and respond to the issues raised.
              (c) The final environmental impact statement will state how the alternatives considered in it and decisions based on it will or will not achieve the requirements of sections 101 and 102(1) of NEPA and of any other relevant and applicable environmental laws and policies.
              (d) The final environmental impact statement will include a final analysis and a final recommendation on the action to be taken.

              § 51.92
              Supplement to the final environmental impact statement.

              (a) If the proposed action has not been taken, the NRC staff will prepare a supplement to a final environmental impact statement for which a notice of availability has been published in the Federal Register as provided in § 51.118, if:
              (1) There are substantial changes in the proposed action that are relevant to environmental concerns; or
              (2) There are new and significant circumstances or information relevant to environmental concerns and bearing on the proposed action or its impacts.
              (b) In a proceeding for a combined license application under 10 CFR part 52 referencing an early site permit under part 52, the NRC staff shall prepare a supplement to the final environmental impact statement for the referenced early site permit in accordance with paragraph (e) of this section.
              (c) The NRC staff may prepare a supplement to a final environmental impact statement when, in its opinion, preparation of a supplement will further the purposes of NEPA.
              (d) The supplement to a final environmental impact statement will be prepared in the same manner as the final environmental impact statement except that a scoping process need not be used.
              (e) The supplement to an early site permit final environmental impact statement which is prepared for a combined license application in accordance with § 51.75(c)(1) and paragraph (b) of this section must:
              (1) Identify the proposed action as the issuance of a combined license for the construction and operation of a nuclear power plant as described in the combined license application at the site described in the early site permit referenced in the combined license application;
              (2) Incorporate by reference the final environmental impact statement prepared for the early site permit;
              (3) Contain no separate discussion of alternative sites;
              (4) Include an analysis of the economic, technical, and other benefits and costs of the proposed action, to the extent that the final environmental impact statement prepared for the early site permit did not include an assessment of these benefits and costs;
              (5) Include an analysis of other energy alternatives, to the extent that the final environmental impact statement prepared for the early site permit did not include an assessment of energy alternatives;

              (6) Include an analysis of any environmental issue related to the impacts of construction or operation of the facility that was not resolved in the proceeding on the early site permit; and

              (7) Include an analysis of the issues related to the impacts of construction and operation of the facility that were resolved in the early site permit proceeding for which new and significant information has been identified, including, but not limited to, new and significant information demonstrating that the design of the facility falls outside the site characteristics and design parameters specified in the early site permit.

              (f)(1) A supplement to a final environmental impact statement will be accompanied by or will include a request for comments as provided in § 51.73 and a notice of availability will be published in the Federal Register as provided in § 51.117 if paragraphs (a) or (b) of this section applies.

              (2) If comments are not requested, a notice of availability of a supplement to a final environmental impact statement will be published in the Federal Register as provided in § 51.118.
              [72 FR 49515, Aug. 28, 2007]

              § 51.93
              Distribution of final environmental impact statement and supplement to final environmental impact statement; news releases.
              (a) A copy of the final environmental impact statement will be distributed to:
              (1) The Environmental Protection Agency.
              (2) The applicant or petitioner for rulemaking and any other party to the proceeding.
              (3) Appropriate State, regional and metropolitan clearinghouses.
              (4) Each commenter.
              (b) Additional copies will be made available in accordance with § 51.123.
              (c) If the final environmental impact statement is unusually long or there are so many comments on a draft environmental impact statement or any supplement to a draft environmental impact statement that distribution of the entire final statement to all commenters is impracticable, a summary of the final statement and the substantive comments will be distributed. When the final environmental impact statement has been prepared by adding errata sheets to the draft environmental impact statement as provided in § 51.91(a)(3), only the comments, the responses to the comments and the changes to the environmental impact statement will be distributed.
              (d) A supplement to a final environmental impact statement will be distributed in the same manner as the final environmental impact statement to which it relates.
              (e) News releases stating the availability and place for obtaining or inspecting a final environmental impact statement or supplement will be provided to local newspapers and other appropriate media.

              (f) A notice of availability will be published in the Federal Register in accordance with § 51.118.

              § 51.94
              Requirement to consider final environmental impact statement.
              The final environmental impact statement, together with any comments and any supplement, will accompany the application or petition for rulemaking through, and be considered in, the Commission's decisionmaking process. The final environmental impact statement, together with any comments and any supplement, will be made a part of the record of the appropriate adjudicatory or rulemaking proceeding.

            final environmental impact statements—production and utilization facilities

              § 51.95
              Postconstruction environmental impact statements.
              (a) General. Any supplement to a final environmental impact statement or any environmental assessment prepared under the provisions of this section may incorporate by reference any information contained in a final environmental document previously prepared by the NRC staff that relates to the same production or utilization facility. Documents that may be referenced include, but are not limited to, the final environmental impact statement; supplements to the final environmental impact statement, including supplements prepared at the operating license stage; NRC staff-prepared final generic environmental impact statements; environmental assessments and records of decisions prepared in connection with the construction permit, the operating license, the early site permit, or the combined license and any license amendment for that facility. A supplement to a final environmental impact statement will include a request for comments as provided in § 51.73.
              (b) Initial operating license stage. In connection with the issuance of an operating license for a production or utilization facility, the NRC staff will prepare a supplement to the final environmental impact statement on the construction permit for that facility, which will update the prior environmental review. The supplement will only cover matters that differ from the final environmental impact statement or that reflect significant new information concerning matters discussed in the final environmental impact statement. Unless otherwise determined by the Commission, a supplement on the operation of a nuclear power plant will not include a discussion of need for power, or of alternative energy sources, or of alternative sites, and will only be prepared in connection with the first licensing action authorizing full-power operation. As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated into the environmental impact statement.
              (c) Operating license renewal stage. In connection with the renewal of an operating license or combined license for a nuclear power plant under 10 CFR parts 52 or 54 of this chapter, the Commission shall prepare an environmental impact statement, which is a supplement to the Commission's NUREG-1437, “Generic Environmental Impact Statement for License Renewal of Nuclear Plants” (June 2013), which is available in the NRC's Public Document Room, 11555 Rockville Pike, Rockville, Maryland 20852.
              (1) The supplemental environmental impact statement for the operating license renewal stage shall address those issues as required by § 51.71. In addition, the NRC staff must comply with 40 CFR 1506.6(b)(3) in conducting the additional scoping process as required by § 51.71(a).
              (2) The supplemental environmental impact statement for license renewal is not required to include discussion of need for power or the economic costs and economic benefits of the proposed action or of alternatives to the proposed action except insofar as such benefits and costs are either essential for a determination regarding the inclusion of an alternative in the range of alternatives considered or relevant to mitigation. In addition, the supplemental environmental impact statement prepared at the license renewal stage need not discuss other issues not related to the environmental effects of the proposed action and the alternatives. The analysis of alternatives in the supplemental environmental impact statement should be limited to the environmental impacts of such alternatives and should otherwise be prepared in accordance with § 51.71 and appendix A to subpart A of this part. As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated into the supplemental environmental impact statement.
              (3) The supplemental environmental impact statement shall be issued as a final impact statement in accordance with §§ 51.91 and 51.93 after considering any significant new information relevant to the proposed action contained in the supplement or incorporated by reference.

              (4) The supplemental environmental impact statement must contain the NRC staff's recommendation regarding the environmental acceptability of the license renewal action. In order to make recommendations and reach a final decision on the proposed action, the NRC staff, adjudicatory officers, and Commission shall integrate the conclusions in the generic environmental impact statement for issues designated as Category 1 with information developed for those Category 2 issues applicable to the plant under § 51.53(c)(3)(ii) and any new and significant information. Given this information, the NRC staff, adjudicatory officers, and Commission shall determine whether or not the adverse environmental impacts of license renewal are so great that preserving the option of license renewal for energy planning decisionmakers would be unreasonable.

              (d) Postoperating license stage. In connection with the amendment of an operating or combined license authorizing decommissioning activities at a production or utilization facility covered by § 51.20, either for unrestricted use or based on continuing use restrictions applicable to the site, or with the issuance, amendment or renewal of a license to store spent fuel at a nuclear power reactor after expiration of the operating or combined license for the nuclear power reactor, the NRC staff will prepare a supplemental environmental impact statement for the post operating or post combined license stage or an environmental assessment, as appropriate, which will update the prior environmental documentation prepared by the NRC for compliance with NEPA under the provisions of this part. The supplement or assessment may incorporate by reference any information contained in the final environmental impact statement—for the operating or combined license stage, as appropriate, or in the records of decision prepared in connection with the early site permit, construction permit, operating license, or combined license for that facility. The supplement will include a request for comments as provided in § 51.73. As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated into the supplemental environmental impact statement or shall be considered in the environmental assessment, if the impacts of continued storage of spent fuel are applicable to the proposed action.
              [61 FR 66545, Dec. 18, 1996, as amended at 72 FR 49516, Aug. 28, 2007; 78 FR 37317, June 20, 2013; 79 FR 56262, Sept. 19, 2014]

            final environmental impact statements—materials licenses

              § 51.97
              Final environmental impact statement—materials license.
              (a) Independent spent fuel storage installation (ISFSI). As stated in § 51.23, the generic impact determinations regarding the continued storage of spent fuel in NUREG-2157 shall be deemed incorporated into the environmental impact statement.
              (b) Monitored retrievable storage facility (MRS). As provided in sections 141 (c), (d), and (e) and 148 (a) and (c) of the Nuclear Waste Policy Act of 1982, as amended (NWPA) (96 Stat. 2242, 2243, 42 U.S.C. 10161 (c), (d), (e); 101 Stat. 1330-235, 1330-236, 42 U.S.C. 10168 (a), (c)) a final environmental impact statement for the construction of a monitored retrievable storage installation (MRS) will not address the need for the MRS or any alternative to the design criteria for an MRS set forth in section 141(b)(1) of the NWPA (96 Stat. 2242, 42 U.S.C. 10161(b)(1)) but may consider alternative facility designs which are consistent with these design criteria.
              (c) Uranium enrichment facility. As provided in section 5(e) of the Solar, Wind, Waste, and Geothermal Power Production Incentives Act of 1990 (104 Stat. 2834 at 2835, 42 U.S.C. 2243), a final environmental impact statement must be prepared before the hearing on the issuance of a license for a uranium enrichment facility is completed.
              [49 FR 34695, Aug. 31, 1984, as amended at 53 FR 31682, Aug. 19, 1988; 57 FR 18392, Apr. 30, 1992; 79 FR 56262, Sept. 19, 2014]

            final environmental impact statements—rulemaking

              § 51.99
              [Reserved]

            NEPA Procedure and Administrative Action

            general

              § 51.100
              Timing of Commission action.
              (a)(1) Except as provided in § 51.13 and paragraph (b) of this section, no decision on a proposed action, including the issuance of a permit, license, or other form of permission, or amendment to or renewal of a permit, license, or other form of permission, or the issuance of an effective regulation, for which an environmental impact statement is required, will be made and no record of decision will be issued until the later of the following dates:

              (i) Ninety (90) days after publication by the Environmental Protection Agency of a Federal Register notice stating that the draft environmental impact statement has been filed with EPA.

              (ii) Thirty (30) days after publication by the Environmental Protection Agency of a Federal Register notice stating that the final environmental impact statement has been filed with EPA.
              (2) If a notice of filing of a final environmental impact statement is published by the Environmental Protection Agency within ninety (90) days after a notice of filing of a draft environmental impact statement has been published by EPA, the minimum thirty (30) day period and the minimum ninety (90) day period may run concurrently to the extent they overlap.

              (b) In any rulemaking proceeding for the purpose of protecting the public health or safety or the common defense and security, the Commission may make and publish the decision on the final rule at the same time that the Environmental Protection Agency publishes the Federal Register notice of filing of the final environmental impact statement.

              § 51.101
              Limitations on actions.
              (a) Until a record of decision is issued in connection with a proposed licensing or regulatory action for which an environmental impact statement is required under § 51.20, or until a final finding of no significant impact is issued in connection with a proposed licensing or regulatory action for which an environmental assessment is required under § 51.21:
              (1) No action concerning the proposal may be taken by the Commission which would (i) have an adverse environmental impact, or (ii) limit the choice of reasonable alternatives.
              (2) Any action concerning the proposal taken by an applicant which would (i) have an adverse environmental impact, or (ii) limit the choice of reasonable alternatives may be grounds for denial of the license. In the case of an application covered by §§ 30.32(f), 40.31(f), 50.10(c), 70.21(f), or §§ 72.16 and 72.34 of this chapter, the provisions of this paragraph will be applied in accordance with §§ 30.33(a)(5), 40.32(e), 50.10 (c) and (e), 70.23(a)(7) or § 72.40(b) of this chapter, as appropriate.
              (b) While work on a required program environmental impact statement is in progress, the Commission will not undertake in the interim any major Federal action covered by the program which may significantly affect the quality of the human environment unless such action:
              (1) Is justified independently of the program;
              (2) Is itself accompanied by an adequate environmental impact statement; and
              (3) Will not prejudice the ultimate decision on the program. Absent any satisfactory explanation to the contrary, interim action which tends to determine subsequent development or limit reasonable alternatives, will be considered prejudicial.
              (c) This section does not preclude any applicant for an NRC permit, license, or other form of permission, or amendment to or renewal of an NRC permit, license, or other form of permission, (1) from developing any plans or designs necessary to support an application; or (2) after prior notice and consultation with NRC staff, (i) from performing any physical work necessary to support an application, or (ii) from performing any other physical work relating to the proposed action if the adverse environmental impact of that work is de minimis.
              [49 FR 9381, Mar. 12, 1984, as amended at 53 FR 31682, Aug. 19, 1988]

              § 51.102
              Requirement to provide a record of decision; preparation.
              (a) A Commission decision on any action for which a final environmental impact statement has been prepared shall be accompanied by or include a concise public record of decision.
              (b) Except as provided in paragraph (c) of this section, the record of decision will be prepared by the NRC staff director authorized to take the action.

              (c) When a hearing is held on the proposed action under the regulations in subpart G of part 2 of this chapter or when the action can only be taken by the Commissioners acting as a collegial body, the initial decision of the presiding officer or the final decision of the Commissioners acting as a collegial body will constitute the record of decision. An initial or final decision constituting the record of decision will be distributed as provided in § 51.93.
              [49 FR 9381, Mar. 12, 1984, as amended at 77 FR 46600, Aug. 3, 2012; 79 FR 66604, Nov. 10, 2014]

              § 51.103
              Record of decision—general.
              (a) The record of decision required by § 51.102 shall be clearly identified and shall:
              (1) State the decision.
              (2) Identify all alternatives considered by the Commission in reaching the decision, state that these alternatives were included in the range of alternatives discussed in the environmental impact statement, and specify the alternative or alternatives which were considered to be environmentally preferable.
              (3) Discuss preferences among alternatives based on relevant factors, including economic and technical considerations where appropriate, the NRC's statutory mission, and any essential considerations of national policy, which were balanced by the Commission in making the decision and state how these considerations entered into the decision.
              (4) State whether the Commission has taken all practicable measures within its jurisdiction to avoid or minimize environmental harm from the alternative selected, and if not, to explain why those measures were not adopted. Summarize any license conditions and monitoring programs adopted in connection with mitigation measures.
              (5) In making a final decision on a license renewal action pursuant to part 54 of this chapter, the Commission shall determine whether or not the adverse environmental impacts of license renewal are so great that preserving the option of license renewal for energy planning decisionmakers would be unreasonable.
              (6) In a construction permit or a combined license proceeding where a limited work authorization under 10 CFR 50.10 was issued, the Commission's decision on the construction permit or combined license application will not address or consider the sunk costs associated with the limited work authorization in determining the proposed action.
              (b) The record of decision may be integrated into any other record prepared by the Commission in connection with the action.
              (c) The record of decision may incorporate by reference material contained in a final environmental impact statement.
              [49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28490, June 5, 1996; 61 FR 66546, Dec. 18, 1996; 61 FR 68543, Dec. 30, 1996; 72 FR 57445, Oct. 9, 2007]

              § 51.104
              NRC proceeding using public hearings; consideration of environmental impact statement.
              (a)(1) In any proceeding in which (i) a hearing is held on the proposed action, (ii) a final environmental impact statement has been prepared in connection with the proposed action, and (iii) matters within the scope of NEPA and this subpart are in issue, the NRC staff may not offer the final environmental impact statement in evidence or present the position of the NRC staff on matters within the scope of NEPA and this subpart until the final environmental impact statement is filed with the Environmental Protection Agency, furnished to commenting agencies and made available to the public.
              (2) Any party to the proceeding may take a position and offer evidence on the aspects of the proposed action within the scope of NEPA and this subpart in accordance with the provisions of part 2 of this chapter applicable to that proceeding or in accordance with the terms of the notice of hearing.
              (3) In the proceeding the presiding officer will decide those matters in controversy among the parties within the scope of NEPA and this subpart.

              (b) In any proceeding in which a hearing is held where the NRC staff has determined that no environmental impact statement need be prepared for the proposed action, unless the Commission orders otherwise, any party to the proceeding may take a position and offer evidence on the aspects of the proposed action within the scope of NEPA and this subpart in accordance with the provisions of part 2 of this chapter applicable to that proceeding or in accordance with the terms of the notice of hearing. In the proceeding, the presiding officer will decide any such matters in controversy among the parties.
              (c) In any proceeding in which a limited work authorization is requested, unless the Commission orders otherwise, a party to the proceeding may take a position and offer evidence only on the aspects of the proposed action within the scope of NEPA and this subpart which are within the scope of that party's admitted contention, in accordance with the provisions of part 2 of this chapter applicable to the limited work authorization or in accordance with the terms of any notice of hearing applicable to the limited work authorization. In the proceeding, the presiding officer will decide all matters in controversy among the parties.
              [49 FR 9381, Mar. 12, 1984, as amended at 72 FR 57445, Oct. 9, 2007]

            production and utilization facilities

              § 51.105
              Public hearings in proceedings for issuance of construction permits or early site permits; limited work authorizations.
              (a) In addition to complying with applicable requirements of § 51.104, in a proceeding for the issuance of a construction permit or early site permit for a nuclear power reactor, testing facility, fuel reprocessing plant or isotopic enrichment plant, the presiding officer will:
              (1) Determine whether the requirements of Sections 102(2) (A), (C), and (E) of NEPA and the regulations in this subpart have been met;
              (2) Independently consider the final balance among conflicting factors contained in the record of the proceeding with a view to determining the appropriate action to be taken;
              (3) Determine, after weighing the environmental, economic, technical, and other benefits against environmental and other costs, and considering reasonable alternatives, whether the construction permit or early site permit should be issued, denied, or appropriately conditioned to protect environmental values;
              (4) Determine, in an uncontested proceeding, whether the NEPA review conducted by the NRC staff has been adequate; and
              (5) Determine, in a contested proceeding, whether in accordance with the regulations in this subpart, the construction permit or early site permit should be issued as proposed by the NRC's Director, Office of Nuclear Reactor Regulation.
              (b) The presiding officer in an early site permit hearing shall not admit contentions proffered by any party concerning the benefits assessment (e.g., need for power) or alternative energy sources if those issues were not addressed by the applicant in the early site permit application.
              (c)(1) In addition to complying with the applicable provisions of § 51.104, in any proceeding for the issuance of a construction permit for a nuclear power plant or an early site permit under part 52 of this chapter, where the applicant requests a limited work authorization under § 50.10(d) of this chapter, the presiding officer shall—
              (i) Determine whether the requirements of Section 102(2)(A), (C), and (E) of NEPA and the regulations in the subpart have been met, with respect to the activities to be conducted under the limited work authorization;
              (ii) Independently consider the balance among conflicting factors with respect to the limited work authorization which is contained in the record of the proceeding, with a view to determining the appropriate action to be taken;
              (iii) Determine whether the redress plan will adequately redress the activities performed under the limited work authorization, should limited work activities be terminated by the holder or the limited work authorization be revoked by the NRC, or upon effectiveness of the Commission's final decision denying the associated construction permit or early site permit, as applicable;
              (iv) In an uncontested proceeding, determine whether the NEPA review conducted by the NRC staff for the limited work authorization has been adequate; and

              (v) In a contested proceeding, determine whether, in accordance with the regulations in this subpart, the limited work authorization should be issued as proposed.

              (2) If the limited work authorization is for activities to be conducted at a site for which the Commission has previously prepared an environmental impact statement for the construction and operation of a nuclear power plant, and a construction permit was issued but construction of the plant was never completed, then in making the determinations in paragraph (c)(1) of this section, the presiding officer shall be limited to a consideration whether there is, with respect to construction activities encompassed by the environmental impact statement which are analogous to the activities to be conducted under the limited work authorization, new and significant information on the environmental impacts of those activities, such that the limited work authorization should not be issued as proposed.
              (3) The presiding officer's determination in this paragraph shall be made in a partial initial decision to be issued separately from, and in advance of, the presiding officer's decision in paragraph (a) of this section.
              [72 FR 49516, Aug. 28, 2007, as amended at 72 FR 57446, Oct. 9, 2007; 73 FR 5724, Jan. 31, 2008; 84 FR 65645, Nov. 29, 2019]

              § 51.105a
              Public hearings in proceedings for issuance of manufacturing licenses.
              In addition to complying with applicable requirements of § 51.31(c), in a proceeding for the issuance of a manufacturing license, the presiding officer will determine whether, in accordance with the regulations in this subpart, the manufacturing license should be issued as proposed by the NRC's Director, Office of Nuclear Reactor Regulation.
              [73 FR 5724, Jan. 31, 2008, as amended at 84 FR 65645, Nov. 29, 2019]

              § 51.106
              Public hearings in proceedings for issuance of operating licenses.
              (a) Consistent with the requirements of this section and as appropriate, the presiding officer in an operating license hearing shall comply with any applicable requirements of §§ 51.104 and 51.105.
              (b) During the course of a hearing on an application for issuance of an operating license for a nuclear power reactor, or a testing facility, the presiding officer may authorize, pursuant to § 50.57(c) of this chapter, the loading of nuclear fuel in the reactor core and limited operation within the scope of § 50.57(c) of this chapter, upon compliance with the procedures described therein. In any such hearing, where any party opposes such authorization on the basis of matters covered by subpart A of this part, the provisions of §§ 51.104 and 51.105 will apply, as appropriate.
              (c) The presiding officer in an operating license hearing shall not admit contentions proffered by any party concerning need for power or alternative energy sources or alternative sites for the facility for which an operating license is requested.

              (d) The presiding officer in an operating license hearing shall not raise issues concerning alternative sites for the facility for which an operating license is requested sua sponte.

              § 51.107
              Public hearings in proceedings for issuance of combined licenses; limited work authorizations.
              (a) In addition to complying with the applicable requirements of § 51.104, in a proceeding for the issuance of a combined license for a nuclear power reactor under part 52 of this chapter, the presiding officer will:
              (1) Determine whether the requirements of Sections 102(2) (A), (C), and (E) of NEPA and the regulations in this subpart have been met;
              (2) Independently consider the final balance among conflicting factors contained in the record of the proceeding with a view to determining the appropriate action to be taken;
              (3) Determine, after weighing the environmental, economic, technical, and other benefits against environmental and other costs, and considering reasonable alternatives, whether the combined license should be issued, denied, or appropriately conditioned to protect environmental values;

              (4) Determine, in an uncontested proceeding, whether the NEPA review conducted by the NRC staff has been adequate; and

              (5) Determine, in a contested proceeding, whether in accordance with the regulations in this subpart, the combined license should be issued as proposed by the NRC's Director, Office of Nuclear Reactor Regulation.
              (b) If a combined license application references an early site permit, then the presiding officer in the combined license hearing shall not admit any contention proffered by any party on environmental issues which have been accorded finality under § 52.39 of this chapter, unless the contention:
              (1) Demonstrates that the nuclear power reactor proposed to be built does not fit within one or more of the site characteristics or design parameters included in the early site permit;
              (2) Raises any significant environmental issue that was not resolved in the early site permit proceeding; or
              (3) Raises any issue involving the impacts of construction and operation of the facility that was resolved in the early site permit proceeding for which new and significant information has been identified.
              (c) If the combined license application references a standard design certification, or proposes to use a manufactured reactor, then the presiding officer in a combined license hearing shall not admit contentions proffered by any party concerning severe accident mitigation design alternatives unless the contention demonstrates that the site characteristics fall outside of the site parameters in the standard design certification or underlying manufacturing license for the manufactured reactor.
              (d)(1) In any proceeding for the issuance of a combined license where the applicant requests a limited work authorization under § 50.10(d) of this chapter, the presiding officer, in addition to complying with any applicable provision of § 51.104, shall:
              (i) Determine whether the requirements of Section 102(2)(A), (C), and (E) of NEPA and the regulations in this subpart have been met, with respect to the activities to be conducted under the limited work authorization;
              (ii) Independently consider the balance among conflicting factors with respect to the limited work authorization which is contained in the record of the proceeding, with a view to determining the appropriate action to be taken;
              (iii) Determine whether the redress plan will adequately redress the activities performed under the limited work authorization, should limited work activities be terminated by the holder or the limited work authorization be revoked by the NRC, or upon effectiveness of the Commission's final decision denying the combined license application;
              (iv) In an uncontested proceeding, determine whether the NEPA review conducted by the NRC staff for the limited work authorization has been adequate; and
              (v) In a contested proceeding, determine whether, in accordance with the regulations in this subpart, the limited work authorization should be issued as proposed by the Director, Office of Nuclear Reactor Regulation.
              (2) If the limited work authorization is for activities to be conducted at a site for which the Commission has previously prepared an environmental impact statement for the construction and operation of a nuclear power plant, and a construction permit was issued but construction of the plant was never completed, then in making the determinations in paragraph (c)(1) of this section, the presiding officer shall be limited to a consideration whether there is, with respect to construction activities encompassed by the environmental impact statement which are analogous to the activities to be conducted under the limited work authorization, new and significant information on the environmental impacts of those activities, so that the limited work authorization should not be issued as proposed by the Director, Office of Nuclear Reactor Regulation.
              (3) In making the determination required by this section, the presiding officer may not address or consider the sunk costs associated with the limited work authorization.

              (4) The presiding officer's determination in this paragraph shall be made in a partial initial decision to be issued separately from, and in advance of, the presiding officer's decision in paragraph (a) of this section on the combined license.
              [72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57446, Oct. 9, 2007; 73 FR 5724, Jan. 31, 2008; 84 FR 65645, Nov. 29, 2019]

              § 51.108
              Public hearings on Commission findings that inspections, tests, analyses, and acceptance criteria of combined licenses are met.
              In any public hearing requested under 10 CFR 52.103(b), the Commission will not admit any contentions on environmental issues, the adequacy of the environmental impact statement for the combined license issued under subpart C of part 52, or the adequacy of any other environmental impact statement or environmental assessment referenced in the combined license application. The Commission will not make any environmental findings in connection with the finding under 10 CFR 52.103(g).
              [72 FR 49517, Aug. 28, 2007]

            materials licenses

              § 51.109
              Public hearings in proceedings for issuance of materials license with respect to a geologic repository.

              (a)(1) In a proceeding for issuance of a construction authorization for a high-level radioactive waste repository at a geologic repository operations area under parts 60 and 63 of this chapter, and in a proceeding for issuance of a license to receive and possess source, special nuclear, and byproduct material at a geologic repository operations area under parts 60 and 63 of this chapter, the NRC staff shall, upon the publication of the notice of hearing in the Federal Register, present its position on whether it is practicable to adopt, without further supplementation, the environmental impact statement (including any supplement thereto) prepared by the Secretary of Energy. If the position of the staff is that supplementation of the environmental impact statement by NRC is required, it shall file its final supplemental environmental impact statement with the Environmental Protection Agency, furnish that statement to commenting agencies, and make it available to the public, before presenting its position, or as soon thereafter as may be practicable. In discharging its responsibilities under this paragraph, the staff shall be guided by the principles set forth in paragraphs (c) and (d) of this section.

              (2) Any other party to the proceeding who contends that it is not practicable to adopt the DOE environmental impact statement, as it may have been supplemented, shall file a contention to that effect within thirty (30) days after the publication of the notice of hearing in the Federal Register. Such contention must be accompanied by one or more affidavits which set forth factual and/or technical bases for the claim that, under the principles set forth in paragraphs (c) and (d) of this section, it is not practicable to adopt the DOE environmental impact statement, as it may have been supplemented. The presiding officer shall resolve disputes concerning adoption of the DOE environmental impact statement by using, to the extent possible, the criteria and procedures that are followed in ruling on motions to reopen under § 2.326 of this chapter.
              (b) In any such proceeding, the presiding officer will determine those matters in controversy among the parties within the scope of NEPA and this subpart, specifically including whether, and to what extent, it is practicable to adopt the environmental impact statement prepared by the Secretary of Energy in connection with the issuance of a construction authorization and license for such repository.
              (c) The presiding officer will find that it is practicable to adopt any environmental impact statement prepared by the Secretary of Energy in connection with a geologic repository proposed to be constructed under Title I of the Nuclear Waste Policy Act of 1982, as amended, unless:
              (1)(i) The action proposed to be taken by the Commission differs from the action proposed in the license application submitted by the Secretary of Energy; and
              (ii) The difference may significantly affect the quality of the human environment; or

              (2) Significant and substantial new information or new considerations render such environmental impact statement inadequate.
              (d) To the extent that the presiding officer determines it to be practicable, in accordance with paragraph (c) of this section, to adopt the environmental impact statement prepared by the Secretary of Energy, such adoption shall be deemed to satisfy all responsibilities of the Commission under NEPA and no further consideration under NEPA or this subpart shall be required.
              (e) To the extent that it is not practicable, in accordance with paragraph (c) of this section, to adopt the environmental impact statement prepared by the Secretary of Energy, the presiding officer will:
              (1) Determine whether the requirements of section 102(2) (A), (C), and (E) of NEPA and the regulations in this subpart have been met;
              (2) Independently consider the final balance among conflicting factors contained in the record of the proceeding with a view to determining the appropriate action to be taken;
              (3) Determine, after weighing the environmental, economic, technical and other benefits against environmental and other costs, whether the construction authorization or license should be issued, denied, or appropriately conditioned to protect environmental values;
              (4) Determine, in an uncontested proceeding, whether the NEPA review conducted by the NRC staff has been adequate; and
              (5) Determine, in a contested proceeding, whether in accordance with the regulations in this subpart, the construction authorization or license should be issued as proposed.
              (f) In making the determinations described in paragraph (e) of this section, the environmental impact statement will be deemed modified to the extent that findings and conclusions differ from those in the final statement prepared by the Secretary of Energy, as it may have been supplemented. The initial decision will be distributed to any persons not otherwise entitled to receive it who responded to the request in the notice of docketing, as described in § 51.26(c). If the Commission reaches conclusions different from those of the presiding officer with respect to such matters, the final environmental impact statement will be deemed modified to that extent and the decision will be similarly distributed.
              (g) The provisions of this section shall be followed, in place of those set out in § 51.104, in any proceedings for the issuance of a license to receive and possess source, special nuclear, and byproduct material at a geologic repository operations area.
              [54 FR 27870, July 3, 1989, as amended at 69 FR 2276, Jan. 14, 2004; 77 FR 46600, Aug. 3, 2012]

            rulemaking

              § 51.110
              [Reserved]

            Public Notice of and Access to Environmental Documents

              § 51.116
              Notice of intent.

              (a) In accordance with § 51.26, the appropriate NRC staff director will publish in the Federal Register a notice of intent stating that an environmental impact statement will be prepared. The notice will contain the information specified in § 51.27.
              (b) Copies of the notice will be sent to appropriate Federal, State, and local agencies, and Indian Tribes, appropriate State, regional, and metropolitan clearinghouses and to interested persons upon request. A public announcement of the notice of intent will also be made.

              § 51.117
              Draft environmental impact statement—notice of availability.

              (a) Upon completion of a draft environmental impact statement or any supplement to a draft environmental impact statement, the appropriate NRC staff director will publish a notice of availability of the statement in the Federal Register.

              (b) The notice will request comments on the proposed action and on the draft statement or any supplement to the draft statement and will specify where comments should be submitted and when the comment period expires.

              (c) The notice will (1) state that copies of the draft statement or any supplement to the draft statement are available for public inspection; (2) state where inspection may be made, and (3) state that any comments of Federal, State, and local agencies, Indian Tribes or other interested persons will be made available for public inspection when received.
              (d) Copies of the notice will be sent to appropriate Federal, State, and local agencies, and Indian Tribes, appropriate State, regional, and metropolitan clearinghouses, and to interested persons upon request.

              § 51.118
              Final environmental impact statement—notice of availability.

              (a) Upon completion of a final environmental impact statement or any supplement to a final environmental impact statement, the appropriate NRC staff director will publish a notice of availability of the statement in the Federal Register. The notice will state that copies of the final statement or any supplement to the final statement are available for public inspection and where inspection may be made. Copies of the notice will be sent to appropriate Federal, State, and local agencies, and Indian Tribes, appropriate State, regional, and metropolitan clearinghouses and to interested persons upon request.
              (b) Upon adoption of a final environmental impact statement or any supplement to a final environmental impact statement prepared by the Department of Energy with respect to a geologic repository that is subject to the Nuclear Waste Policy Act of 1982, the appropriate NRC staff director shall follow the procedures set out in paragraph (a) of this section.
              [49 FR 9381, Mar. 12, 1984, as amended at 54 FR 27871, July 3, 1989]

              § 51.119
              Publication of finding of no significant impact; distribution.

              (a) As required by § 51.35, the appropriate NRC staff director will publish the finding of no significant impact in the Federal Register. The finding of no significant impact will be identified as a draft or final finding, and will contain the information specified in §§ 51.32 or 51.33, as appropriate. A draft finding of no significant impact will include a request for comments which specifies where comments should be submitted and when the comment period expires.
              (b) The finding will state that copies of the finding, the environmental assessment setting forth the basis for the finding and any related environmental documents are available for public inspection and where inspection may be made.
              (c) A copy of a final finding will be sent to appropriate Federal, State, and local agencies, and Indian Tribes, appropriate State, regional, and metropolitan clearinghouses, the applicant or petitioner for rulemaking and any other party to the proceeding, and if a draft finding was issued, to each commenter. Additional copies will be made available in accordance with § 51.123.

              § 51.120
              Availability of environmental documents for public inspection.

              Copies of environmental reports, draft and final environmental impact statements, environmental assessments, and findings of no significant impact, together with any related comments and environmental documents, will be made available at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public Document Room.
              [64 FR 48952, Sept. 9, 1999]

              § 51.121
              Status of NEPA actions.
              Individuals or organizations desiring information on the NRC's NEPA process or on the status of specific NEPA actions should address inquiries to:
              (a) Utilization facilities: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-1270, e-mail RidsNrrOd@nrc.gov.

              (b) Production facilities: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-7800, e-mail RidsNmssOd@nrc.gov.

              (c) Materials licenses: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-7800, e-mail RidsNmssOd@nrc.gov.

              (d) Rulemaking: ATTN: Chief, Regulatory Analysis and Rulemaking Support Branch, Division of Rulemaking, Environmental, and Financial Support, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (800) 368-5642.
              (e) General environmental matters: Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 415-1700.
              [53 FR 13399, Apr. 25, 1988, as amended at 60 FR 24552, May 9, 1995; 68 FR 58811, Oct. 10, 2003; 73 FR 5724, Jan. 31, 2008; 77 FR 39907, July 6, 2012; 84 FR 65645, Nov. 29, 2019]

              § 51.122
              List of interested organizations and groups.
              The NRC Office of the Chief Information Officer will maintain a master list of organizations and groups, including relevant conservation commissions, known to be interested in the Commission's licensing and regulatory activities. The NRC Office of the Chief Information Officer with the assistance of the appropriate NRC staff director will select from this master list those organizations and groups that may have an interest in a specific NRC NEPA action and will promptly notify such organizations and groups of the availability of a draft environmental impact statement or a draft finding of no significant impact.
              [49 FR 9381, Mar. 12, 1984, as amended at 52 FR 31612, Aug. 21, 1987; 54 FR 53316, Dec. 28, 1989; 77 FR 39907, July 6, 2012]

              § 51.123
              Charges for environmental documents; distribution to public; distribution to governmental agencies.
              (a) Distribution to public. Upon written request to the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, e-mail distribution.resource@nrc.gov, and to the extent available, single copies of draft environmental impact statements and draft findings of no significant impact will be made available to interested persons without charge. Single copies of final environmental impact statements and final findings of no significant impact will also be provided without charge to the persons listed in §§ 51.93(a) and 51.119(c), respectively. When more than one copy of an environmental impact statement or a finding of no significant impact is requested or when available NRC copies have been exhausted, the requestor will be advised that the NRC will provide copies at the charges specified in § 9.35 of this chapter.
              (b) Distribution to governmental agencies. Upon written request to the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, e-mail distribution.resource@nrc.gov, and to the extent available, copies of draft and final environmental impact statements and draft final findings of no significant impact will be made available in the number requested to Federal, State and local agencies, Indian Tribes, and State, regional, and metropolitan clearinghouses. When available NRC copies have been exhausted, the requester will be advised that the NRC will provide copies at the charges specified in § 9.35 of this chapter.
              (c) Charges. Charges for the reproduction of environmental documents by the NRC at locations other than the NRC Public Document Room located in Washington, DC vary according to location.
              [50 FR 21037, May 22, 1985, as amended at 52 FR 31612, Aug. 21, 1987; 53 FR 43421, Oct. 27, 1988; 61 FR 9902, Mar. 12, 1996; 64 FR 48952, Sept. 9, 1999; 68 FR 58812, Oct. 10, 2003; 80 FR 74980, Dec. 1, 2015]

            Commenting

              § 51.124
              Commission duty to comment.
              It is the policy of the Commission to comment on draft environmental impact statements prepared by other Federal agencies, consistent with the provisions of 40 CFR 1503.2 and 1503.3.

            Responsible Official

              § 51.125
              Responsible official.
              The Executive Director for Operations shall be responsible for overall review of NRC NEPA compliance, except for matters under the jurisdiction of a presiding officer, administrative judge, administrative law judge, Atomic Safety and Licensing Board, or the Commission acting as a collegial body.
              [77 FR 46600, Aug. 3, 2012]

            Pt. 51, Subpt. A, App. A
            Appendix A to Subpart A of Part 51—Format for Presentation of Material in Environmental Impact Statements
            1. General
            2. Cover sheet
            3. Summary
            4. Purpose of and need for action
            5. Alternatives including the proposed action
            6. Affected environment
            7. Environmental consequences and mitigating actions
            8. List of preparers
            9. Appendices
            1. General.

            (a) The Commission will use a format for environmental impact statements which will encourage good analysis and clear presentation of the alternatives including the proposed action. The following standard format for environmental impact statements should be followed unless there is a compelling reason to do otherwise:
            (1) Cover sheet*
            (2) Summary*
            (3) Table of Contents
            (4) Purpose of and Need for Action*
            (5) Alternatives including the proposed action*
            (6) Affected Environment*
            (7) Environmental Consequences and Mitigating Actions*
            (8) List of Preparers*
            (9) List of Agencies, Organizations and Persons to Whom Copies of the Statement are Sent
            (10) Substantive Comments Received and NRC Staff Responses
            (11) Index
            (12) Appendices (if any)*
            If a different format is used, it shall include paragraphs (1), (2), (3), (8), (9), (10), and (11) of this section and shall include the substance of paragraphs (4), (5), (6), (7), and (12) of this section, in any appropriate format.
            Additional guidance on the presentation of material under the format headings identified by an asterisk is set out in sections 2.-9. of this appendix.
            (b) The techniques of tiering and incorporation by reference described respectively in 40 CFR 1502.20 and 1508.28 and 40 CFR 1502.21 1
               of CEQ's NEPA regulations may be used as appropriate to aid in the presentation of issues, eliminate repetition or reduce the size of an environmental impact statement. In appropriate circumstances, draft or final environmental impact statements prepared by other Federal agencies may be adopted in whole or in part in accordance with the procedures outlined in 40 CFR 1506.3 2
               of CEQ's NEPA regulations. In final environmental impact statements, material under the following format headings will normally be presented in less than 150 pages: Purpose of and Need for Action, Alternatives Including the Proposed Action, Affected Environment, and Environmental Consequences and Mitigating Actions. For proposals of unusual scope or complexity, the material presented under these format headings may extend to 300 pages.

                1 Tiering—40 CFR 1502.20, 40 CFR 1508.28; Incorporation by reference—40 CFR 1502.21.

                2 Adoption—40 CFR 1506.3.

            2. Cover sheet.

            The cover sheet will not exceed one page. It will include:
            (a) The name of the NRC office responsible for preparing the statement and a list of any cooperating agencies.
            (b) The title of the proposed action that is the subject of the statement with a list of the states, counties or municipalities where the facility or other subject of the action is located, as appropriate.
            (c) The name, address, and telephone number of the individual in NRC who can supply further information.
            (d) A designation of the statement as a draft or final statement, or a draft or final supplement.
            (e) A one paragraph abstract of the statement.
            (f) For draft environmental impact statements, the date by which comments must be received. This date may be specified in the form of the following or a substantially similar statement:
            “Comments should be filed no later than 3

               days after the date on which the Environmental Protection Agency notice stating that the draft environmental impact statement has been filed with EPA is published in the Federal Register. Comments received after the expiration of the comment period will be considered if it is practical to do so but assurance of consideration of late comments cannot be given.”

                3 The number of days in the comment period should be inserted. The minimum comment period is 45 days (see § 51.73.)

            3. Summary.

            Each environmental impact statement will contain a summary which adequately and accurately summarizes the statement. The summary will stress the major issues considered. The summary will discuss the areas of controversy, will identify any remaining issues to be resolved, and will present the major conclusions and recommendations. The summary will normally not exceed 15 pages.

            4. Purpose of and need for action.

            The statement will briefly describe and specify the need for the proposed action. The alternative of no action will be discussed. In the case of nuclear power plant construction or siting, consideration will be given to the potential impact of conservation measures in determining the demand for power and consequent need for additional generating capacity.
            5. Alternatives including the proposed action.

            This section is the heart of the environmental impact statement. It will present the environmental impacts of the proposal and the alternatives in comparative form. Where important to the comparative evaluation of alternatives, appropriate mitigating measures of the alternatives will be discussed. All reasonable alternatives will be identified. The range of alternatives discussed will encompass those proposed to be considered by the ultimate decisionmaker. An otherwise reasonable alternative will not be excluded from discussion solely on the ground that it is not within the jurisdiction of the NRC. 4
               The discussion of alternatives will take into accounts, without duplicating, the environmental information and analyses included in sections, 4., 6. and 7. of this appendix.

                4 With respect to limitations on NRC's NEPA authority and responsibility imposed by the Federal Water Pollution Control Act Amendments of 1972, see §§ 51.10(c), 51.22(c)(17) and 51.71(d).

            In the draft environmental impact statement, this section will either include a preliminary recommendation on the action to be taken, or identify the alternatives under consideration.
            In the final environmental impact statement, this section will include a final recommendation on the action to be taken.
            6. Affected environment.

            The environmental impact statement will succinctly describe the environment to be affected by the proposed action. Data and analyses in the statement will be commensurate with the importance of the impact, with less important material summarized, consolidated, or simply referenced. Effort and attention will be concentrated on important issues; useless bulk will be eliminated.
            7. Environmental consequences and mitigating actions.

            This section discusses the environmental consequences of alternatives, including the proposed actions and any mitigating actions which may be taken. Alternatives eliminated from detailed study will be identified and a discussion of those alternatives will be confined to a brief statement of the reasons why the alternatives were eliminated. The level of information for each alternative considered in detail will reflect the depth of analysis required for sound decisionmaking.
            The discussion will include any adverse environmental effects which cannot be avoided should the alternative be implemented, the relationship between short-term uses of man's environment and the maintenance and enhancement of long-term productivity, and any irreversible or irretrievable commitments of resources which would be involved in the alternative should it be implemented. This section will include discussions of:
            (a) Direct effects and their significance.
            (b) Indirect effects and their significance.
            (c) Possible conflicts between the alternative and the objectives of Federal, regional, State, and local (and in the case of a reservation, Indian Tribe) land use plans, policies and controls for the area concerned.
            (d) Means to mitigate adverse environmental impacts.
            8. List of preparers.

            The environmental impact statement will list the names and qualifications (expertise, experience, professional disciplines), of the persons who were primarily responsible for preparing the environmental impact statement or significant background papers. Persons responsible for making an independent evaluation of information submitted by the applicant or petitioner for rulemaking or others will be included in the list. Where possible, the persons who are responsible for a particular analysis, including analyses in background papers, will be identified.
            9. Appendices.

            An appendix to an environmental impact statement will:
            (a) Consist of material prepared in connection with an environmental impact statement (as distinct from material which is not so prepared and which is incorporated by reference (40 CFR 1502.21)).
            (b) Normally consist of material which substantiates any analysis fundamental to the impact statement. Discussion of methodology used may be placed in an appendix.
            (c) Normally be analytic.
            (d) Be relevant to the decision to be made.
            (e) Be circulated with the environmental impact statement or be readily available on request.
            Discussion of Footnotes
            1. Tiering.

            40 CFR 1502.20 states:

            “Agencies are encouraged to tier their environmental impact statements to eliminate repetitive discussions of the same issues and to focus on the actual issues ripe for decision at each level of environmental review (§ 1508.28). Whenever a broad environmental impact statement has been prepared (such as a program or policy statement) and a subsequent statement or environmental assessment is then prepared on an action included within the entire program or policy (such as a site specific action) the subsequent statement or environmental assessment need only summarize the issues discussed in the broader statement and incorporate discussions from the broader statement by reference and shall concentrate on the issues specific to the subsequent action. The subsequent document shall state where the earlier document is available. Tiering may also be appropriate for different stages of actions. (Sec. 1508.28).”
            40 CFR 1508.28 states:
            “ ‘Tiering’ refers to the coverage of general matters in broader environmental impact statements (such as national program or policy statements) with subsequent narrower statements or environmental analyses (such as regional or basinwide program statements or ultimately site-specific statements) incorporating by reference the general discussions and concentrating solely on the issues specific to the statement subsequently prepared. Tiering is appropriate when the sequence of statements or analyses is:
            “(a) From a program, plan, or policy environmental impact statement to a program, plan, or policy statement or analysis of lesser scope or to a site-specific statement or analysis.
            “(b) From an environmental impact statement on a specific action at an early stage (such as need and site selection) to a supplement (which is preferred) or a subsequent statement or analysis at a later stage (such as environmental mitigation). Tiering in such cases is appropriate when it helps the lead agency to focus on the issues which are ripe for decision and exclude from consideration issues already decided or not yet ripe.”

              Incorporation by reference. 40 CFR 1502.21 states:
            “Agencies shall incorporate material into an environmental impact statement by reference when the effect will be to cut down on bulk without impeding agency and public review of the action. The incorporated material shall be cited in the statement and its content briefly described. No material may be incorporated by reference unless it is reasonably available for inspection by potentially interested persons within the time allowed for comment. Material based on proprietary data which is itself not available for review and comment shall not be incorporated by reference.”
            2. Adoption.

            40 CFR 1506.3 states:
            “(a) An agency may adopt a Federal draft or final environmental impact statement or portion thereof provided that the statement or portion thereof meets the standards for an adequate statement under these regulations.
            “(b) If the actions covered by the original environmental impact statement and the proposed action are substantially the same, the agency adopting another agency's statement is not required to recirculate it except as a final statement. Otherwise the adopting agency shall treat the statement as a draft and recirculate it (except as provided in paragraph (c) of this section).
            “(c) A cooperating agency may adopt without recirculating the environmental impact statement of a lead agency when, after an independent review of the statement, the cooperating agency concludes that its comments and suggestions have been satisfied.
            “(d) When an agency adopts a statement which is not final within the agency that prepared it, or when the action it assesses is the subject of a referral under part 1504, or when the statement's adequacy is the subject of a judicial action which is not final, the agency shall so specify.”
            [49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28490, June 5, 1996; 61 FR 66546, Dec. 18, 1996]

            Pt. 51, Subpt. A, App. B
            Appendix B to Subpart A of Part 51—Environmental Effect of Renewing the Operating License of a Nuclear Power Plant

            The Commission has assessed the environmental impacts associated with granting a renewed operating license for a nuclear power plant to a licensee who holds either an operating license or construction permit as of June 30, 1995. Table B-1 summarizes the Commission's findings on the scope and magnitude of environmental impacts of renewing the operating license for a nuclear power plant as required by section 102(2) of the National Environmental Policy Act of 1969, as amended. Table B-1, subject to an evaluation of those issues identified in Category 2 as requiring further analysis and possible significant new information, represents the analysis of the environmental impacts associated with renewal of any operating license and is to be used in accordance with § 51.95(c). On a 10-year cycle, the Commission intends to review the material in this appendix and update it if necessary. A scoping notice must be published in the Federal Register indicating the results of the NRC's review and inviting public comments and proposals for other areas that should be updated.

              Table B-1—Summary of Findings on NEPA Issues for License Renewal of Nuclear Power Plants 1

                Issue
                Category 2

                Finding 3

                  Land Use

                Onsite land use
                1
                SMALL. Changes in onsite land use from continued operations and refurbishment associated with license renewal would be a small fraction of the nuclear power plant site and would involve only land that is controlled by the licensee.

                Offsite land use
                1
                SMALL. Offsite land use would not be affected by continued operations and refurbishment associated with license renewal.

                Offsite land use in transmission line right-of-ways (ROWs) 4

                1
                SMALL. Use of transmission line ROWs from continued operations and refurbishment associated with license renewal would continue with no change in land use restrictions.

                  Visual Resources

                Aesthetic impacts
                1
                SMALL. No important changes to the visual appearance of plant structures or transmission lines are expected from continued operations and refurbishment associated with license renewal.

                  Air Quality

                Air quality impacts (all plants)
                1

                SMALL. Air quality impacts from continued operations and refurbishment associated with license renewal are expected to be small at all plants. Emissions resulting from refurbishment activities at locations in or near air quality nonattainment or maintenance areas would be short-lived and would cease after these refurbishment activities are completed. Operating experience has shown that the scale of refurbishment activities has not resulted in exceedance of the de minimis thresholds for criteria pollutants, and best management practices including fugitive dust controls and the imposition of permit conditions in State and local air emissions permits would ensure conformance with applicable State or Tribal Implementation Plans.

                Emissions from emergency diesel generators and fire pumps and routine operations of boilers used for space heating would not be a concern, even for plants located in or adjacent to nonattainment areas. Impacts from cooling tower particulate emissions even under the worst-case situations have been small.

                Air quality effects of transmission lines 4

                1
                SMALL. Production of ozone and oxides of nitrogen is insignificant and does not contribute measurably to ambient levels of these gases.

                  Noise

                Noise impacts
                1
                SMALL. Noise levels would remain below regulatory guidelines for offsite receptors during continued operations and refurbishment associated with license renewal.

                  Geologic Environment

                Geology and soils
                1
                SMALL. The effect of geologic and soil conditions on plant operations and the impact of continued operations and refurbishment activities on geology and soils would be small for all nuclear power plants and would not change appreciably during the license renewal term.

                  Surface Water Resources

                Surface water use and quality (non-cooling system impacts)
                1
                SMALL. Impacts are expected to be small if best management practices are employed to control soil erosion and spills. Surface water use associated with continued operations and refurbishment associated with license renewal would not increase significantly or would be reduced if refurbishment occurs during a plant outage.

                Altered current patterns at intake and discharge structures
                1
                SMALL. Altered current patterns would be limited to the area in the vicinity of the intake and discharge structures. These impacts have been small at operating nuclear power plants.

                Altered salinity gradients
                1
                SMALL. Effects on salinity gradients would be limited to the area in the vicinity of the intake and discharge structures. These impacts have been small at operating nuclear power plants.

                Altered thermal stratification of lakes
                1
                SMALL. Effects on thermal stratification would be limited to the area in the vicinity of the intake and discharge structures. These impacts have been small at operating nuclear power plants.

                Scouring caused by discharged cooling water
                1
                SMALL. Scouring effects would be limited to the area in the vicinity of the intake and discharge structures. These impacts have been small at operating nuclear power plants.

                Discharge of metals in cooling system effluent
                1
                SMALL. Discharges of metals have not been found to be a problem at operating nuclear power plants with cooling-tower-based heat dissipation systems and have been satisfactorily mitigated at other plants. Discharges are monitored and controlled as part of the National Pollutant Discharge Elimination System (NPDES) permit process.

                Discharge of biocides, sanitary wastes, and minor chemical spills
                1
                SMALL. The effects of these discharges are regulated by Federal and State environmental agencies. Discharges are monitored and controlled as part of the NPDES permit process. These impacts have been small at operating nuclear power plants.

                Surface water use conflicts (plants with once-through cooling systems)
                1
                SMALL. These conflicts have not been found to be a problem at operating nuclear power plants with once-through heat dissipation systems.

                Surface water use conflicts (plants with cooling ponds or cooling towers using makeup water from a river)
                2
                SMALL or MODERATE. Impacts could be of small or moderate significance, depending on makeup water requirements, water availability, and competing water demands.

                Effects of dredging on surface water quality
                1
                SMALL. Dredging to remove accumulated sediments in the vicinity of intake and discharge structures and to maintain barge shipping has not been found to be a problem for surface water quality. Dredging is performed under permit from the U.S. Army Corps of Engineers, and possibly, from other State or local agencies.

                Temperature effects on sediment transport capacity
                1
                SMALL. These effects have not been found to be a problem at operating nuclear power plants and are not expected to be a problem.

                  Groundwater Resources

                Groundwater contamination and use (non-cooling system impacts)
                1
                SMALL. Extensive dewatering is not anticipated from continued operations and refurbishment associated with license renewal. Industrial practices involving the use of solvents, hydrocarbons, heavy metals, or other chemicals, and/or the use of wastewater ponds or lagoons have the potential to contaminate site groundwater, soil, and subsoil. Contamination is subject to State or Environmental Protection Agency regulated cleanup and monitoring programs. The application of best management practices for handling any materials produced or used during these activities would reduce impacts.

                Groundwater use conflicts (plants that withdraw less than 100 gallons per minute [gpm])
                1
                SMALL. Plants that withdraw less than 100 gpm are not expected to cause any groundwater use conflicts.

                Groundwater use conflicts (plants that withdraw more than 100 gallons per minute [gpm])
                2
                SMALL, MODERATE, or LARGE. Plants that withdraw more than 100 gpm could cause groundwater use conflicts with nearby groundwater users.

                Groundwater use conflicts (plants with closed-cycle cooling systems that withdraw makeup water from a river)
                2
                SMALL, MODERATE, or LARGE. Water use conflicts could result from water withdrawals from rivers during low-flow conditions, which may affect aquifer recharge. The significance of impacts would depend on makeup water requirements, water availability, and competing water demands.

                Groundwater quality degradation resulting from water withdrawals
                1
                SMALL. Groundwater withdrawals at operating nuclear power plants would not contribute significantly to groundwater quality degradation.

                Groundwater quality degradation (plants with cooling ponds in salt marshes)
                1
                SMALL. Sites with closed-cycle cooling ponds could degrade groundwater quality. However, groundwater in salt marshes is naturally brackish and thus, not potable. Consequently, the human use of such groundwater is limited to industrial purposes.

                Groundwater quality degradation (plants with cooling ponds at inland sites)
                2
                SMALL, MODERATE, or LARGE. Inland sites with closed-cycle cooling ponds could degrade groundwater quality. The significance of the impact would depend on cooling pond water quality, site hydrogeologic conditions (including the interaction of surface water and groundwater), and the location, depth, and pump rate of water wells.

                Radionuclides released to groundwater
                2
                SMALL or MODERATE. Leaks of radioactive liquids from plant components and pipes have occurred at numerous plants. Groundwater protection programs have been established at all operating nuclear power plants to minimize the potential impact from any inadvertent releases. The magnitude of impacts would depend on site-specific characteristics.

                  Terrestrial Resources

                Effects on terrestrial resources (non-cooling system impacts)
                2
                SMALL, MODERATE, or LARGE. Impacts resulting from continued operations and refurbishment associated with license renewal may affect terrestrial communities. Application of best management practices would reduce the potential for impacts. The magnitude of impacts would depend on the nature of the activity, the status of the resources that could be affected, and the effectiveness of mitigation.

                Exposure of terrestrial organisms to radionuclides
                1
                SMALL. Doses to terrestrial organisms from continued operations and refurbishment associated with license renewal are expected to be well below exposure guidelines developed to protect these organisms.

                Cooling system impacts on terrestrial resources (plants with once-through cooling systems or cooling ponds)
                1
                SMALL. No adverse effects to terrestrial plants or animals have been reported as a result of increased water temperatures, fogging, humidity, or reduced habitat quality. Due to the low concentrations of contaminants in cooling system effluents, uptake and accumulation of contaminants in the tissues of wildlife exposed to the contaminated water or aquatic food sources are not expected to be significant issues.

                Cooling tower impacts on vegetation (plants with cooling towers)
                1
                SMALL. Impacts from salt drift, icing, fogging, or increased humidity associated with cooling tower operation have the potential to affect adjacent vegetation, but these impacts have been small at operating nuclear power plants and are not expected to change over the license renewal term.

                Bird collisions with plant structures and transmission lines 4

                1
                SMALL. Bird collisions with cooling towers and other plant structures and transmission lines occur at rates that are unlikely to affect local or migratory populations and the rates are not expected to change.

                Water use conflicts with terrestrial resources (plants with cooling ponds or cooling towers using makeup water from a river)
                2
                SMALL or MODERATE. Impacts on terrestrial resources in riparian communities affected by water use conflicts could be of moderate significance.

                Transmission line right-of-way (ROW) management impacts on terrestrial resources 4

                1
                SMALL. Continued ROW management during the license renewal term is expected to keep terrestrial communities in their current condition. Application of best management practices would reduce the potential for impacts.

                Electromagnetic fields on flora and fauna (plants, agricultural crops, honeybees, wildlife, livestock) 4

                1
                SMALL. No significant impacts of electromagnetic fields on terrestrial flora and fauna have been identified. Such effects are not expected to be a problem during the license renewal term.

                  Aquatic Resources

                Impingement and entrainment of aquatic organisms (plants with once-through cooling systems or cooling ponds)
                2
                SMALL, MODERATE, or LARGE. The impacts of impingement and entrainment are small at many plants but may be moderate or even large at a few plants with once-through and cooling-pond cooling systems, depending on cooling system withdrawal rates and volumes and the aquatic resources at the site.

                Impingement and entrainment of aquatic organisms (plants with cooling towers)
                1
                SMALL. Impingement and entrainment rates are lower at plants that use closed-cycle cooling with cooling towers because the rates and volumes of water withdrawal needed for makeup are minimized.

                Entrainment of phytoplankton and zooplankton (all plants)
                1
                SMALL. Entrainment of phytoplankton and zooplankton has not been found to be a problem at operating nuclear power plants and is not expected to be a problem during the license renewal term.

                Thermal impacts on aquatic organisms (plants with once-through cooling systems or cooling ponds)
                2
                SMALL, MODERATE, or LARGE. Most of the effects associated with thermal discharges are localized and are not expected to affect overall stability of populations or resources. The magnitude of impacts, however, would depend on site-specific thermal plume characteristics and the nature of aquatic resources in the area.

                Thermal impacts on aquatic organisms (plants with cooling towers)
                1
                SMALL. Thermal effects associated with plants that use cooling towers are expected to be small because of the reduced amount of heated discharge.

                Infrequently reported thermal impacts (all plants)
                1
                SMALL. Continued operations during the license renewal term are expected to have small thermal impacts with respect to the following:

                Cold shock has been satisfactorily mitigated at operating nuclear plants with once-through cooling systems, has not endangered fish populations or been found to be a problem at operating nuclear power plants with cooling towers or cooling ponds, and is not expected to be a problem.

                Thermal plumes have not been found to be a problem at operating nuclear power plants and are not expected to be a problem.

                Thermal discharge may have localized effects but is not expected to affect the larger geographical distribution of aquatic organisms.

                Premature emergence has been found to be a localized effect at some operating nuclear power plants but has not been a problem and is not expected to be a problem.

                Stimulation of nuisance organisms has been satisfactorily mitigated at the single nuclear power plant with a once-through cooling system where previously it was a problem. It has not been found to be a problem at operating nuclear power plants with cooling towers or cooling ponds and is not expected to be a problem.

                Effects of cooling water discharge on dissolved oxygen, gas supersaturation, and eutrophication
                1
                SMALL. Gas supersaturation was a concern at a small number of operating nuclear power plants with once-through cooling systems but has been mitigated. Low dissolved oxygen was a concern at one nuclear power plant with a once-through cooling system but has been mitigated. Eutrophication (nutrient loading) and resulting effects on chemical and biological oxygen demands have not been found to be a problem at operating nuclear power plants.

                Effects of non-radiological contaminants on aquatic organisms
                1
                SMALL. Best management practices and discharge limitations of NPDES permits are expected to minimize the potential for impacts to aquatic resources during continued operations and refurbishment associated with license renewal. Accumulation of metal contaminants has been a concern at a few nuclear power plants but has been satisfactorily mitigated by replacing copper alloy condenser tubes with those of another metal.

                Exposure of aquatic organisms to radionuclides
                1
                SMALL. Doses to aquatic organisms are expected to be well below exposure guidelines developed to protect these aquatic organisms.

                Effects of dredging on aquatic organisms
                1
                SMALL. Dredging at nuclear power plants is expected to occur infrequently, would be of relatively short duration, and would affect relatively small areas. Dredging is performed under permit from the U.S. Army Corps of Engineers, and possibly, from other State or local agencies.

                Water use conflicts with aquatic resources (plants with cooling ponds or cooling towers using makeup water from a river)
                2
                SMALL or MODERATE. Impacts on aquatic resources in stream communities affected by water use conflicts could be of moderate significance in some situations.

                Effects on aquatic resources (non-cooling system impacts)
                1
                SMALL. Licensee application of appropriate mitigation measures is expected to result in no more than small changes to aquatic communities from their current condition.

                Impacts of transmission line right-of-way (ROW) management on aquatic resources 4

                1
                SMALL. Licensee application of best management practices to ROW maintenance is expected to result in no more than small impacts to aquatic resources.

                Losses from predation, parasitism, and disease among organisms exposed to sublethal stresses
                1
                SMALL. These types of losses have not been found to be a problem at operating nuclear power plants and are not expected to be a problem during the license renewal term.

                  Special Status Species and Habitats

                Threatened, endangered, and protected species and essential fish habitat
                2
                The magnitude of impacts on threatened, endangered, and protected species, critical habitat, and essential fish habitat would depend on the occurrence of listed species and habitats and the effects of power plant systems on them. Consultation with appropriate agencies would be needed to determine whether special status species or habitats are present and whether they would be adversely affected by continued operations and refurbishment associated with license renewal.

                  Historic and Cultural Resources

                Historic and cultural resources 4

                2
                Continued operations and refurbishment associated with license renewal are expected to have no more than small impacts on historic and cultural resources located onsite and in the transmission line ROW because most impacts could be mitigated by avoiding those resources. The National Historic Preservation Act (NHPA) requires the Federal agency to consult with the State Historic Preservation Officer (SHPO) and appropriate Native American Tribes to determine the potential effects on historic properties and mitigation, if necessary.

                  Socioeconomics

                Employment and income, recreation and tourism
                1
                SMALL. Although most nuclear plants have large numbers of employees with higher than average wages and salaries, employment, income, recreation, and tourism impacts from continued operations and refurbishment associated with license renewal are expected to be small.

                Tax revenues
                1
                SMALL. Nuclear plants provide tax revenue to local jurisdictions in the form of property tax payments, payments in lieu of tax (PILOT), or tax payments on energy production. The amount of tax revenue paid during the license renewal term as a result of continued operations and refurbishment associated with license renewal is not expected to change.

                Community services and education
                1
                SMALL. Changes resulting from continued operations and refurbishment associated with license renewal to local community and educational services would be small. With little or no change in employment at the licensee's plant, value of the power plant, payments on energy production, and PILOT payments expected during the license renewal term, community and educational services would not be affected by continued power plant operations.

                Population and housing
                1
                SMALL. Changes resulting from continued operations and refurbishment associated with license renewal to regional population and housing availability and value would be small. With little or no change in employment at the licensee's plant expected during the license renewal term, population and housing availability and values would not be affected by continued power plant operations.

                Transportation
                1
                SMALL. Changes resulting from continued operations and refurbishment associated with license renewal to traffic volumes would be small.

                  Human Health

                Radiation exposures to the public
                1
                SMALL. Radiation doses to the public from continued operations and refurbishment associated with license renewal are expected to continue at current levels, and would be well below regulatory limits.

                Radiation exposures to plant workers
                1
                SMALL. Occupational doses from continued operations and refurbishment associated with license renewal are expected to be within the range of doses experienced during the current license term, and would continue to be well below regulatory limits.

                Human health impact from chemicals
                1
                SMALL. Chemical hazards to plant workers resulting from continued operations and refurbishment associated with license renewal are expected to be minimized by the licensee implementing good industrial hygiene practices as required by permits and Federal and State regulations. Chemical releases to the environment and the potential for impacts to the public are expected to be minimized by adherence to discharge limitations of NPDES and other permits.

                Microbiological hazards to the public (plants with cooling ponds or canals or cooling towers that discharge to a river)
                2
                SMALL, MODERATE, or LARGE. These organisms are not expected to be a problem at most operating plants except possibly at plants using cooling ponds, lakes, or canals, or that discharge into rivers. Impacts would depend on site-specific characteristics.

                Microbiological hazards to plant workers
                1
                SMALL. Occupational health impacts are expected to be controlled by continued application of accepted industrial hygiene practices to minimize worker exposures as required by permits and Federal and State regulations.

                Chronic effects of electromagnetic fields (EMFs) 4 6

                N/A 5

                Uncertain impact. Studies of 60-Hz EMFs have not uncovered consistent evidence linking harmful effects with field exposures. EMFs are unlike other agents that have a toxic effect (e.g., toxic chemicals and ionizing radiation) in that dramatic acute effects cannot be forced and longer-term effects, if real, are subtle. Because the state of the science is currently inadequate, no generic conclusion on human health impacts is possible.

                Physical occupational hazards
                1
                SMALL. Occupational safety and health hazards are generic to all types of electrical generating stations, including nuclear power plants, and are of small significance if the workers adhere to safety standards and use protective equipment as required by Federal and State regulations.

                Electric shock hazards 4

                2
                SMALL, MODERATE, or LARGE. Electrical shock potential is of small significance for transmission lines that are operated in adherence with the National Electrical Safety Code (NESC). Without a review of conformance with NESC criteria of each nuclear power plant's in-scope transmission lines, it is not possible to determine the significance of the electrical shock potential.

                  Postulated Accidents

                Design-basis accidents
                1
                SMALL. The NRC staff has concluded that the environmental impacts of design-basis accidents are of small significance for all plants.

                Severe accidents
                2
                SMALL. The probability-weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to groundwater, and societal and economic impacts from severe accidents are small for all plants. However, alternatives to mitigate severe accidents must be considered for all plants that have not considered such alternatives.

                  Environmental Justice

                Minority and low-income populations
                2
                Impacts to minority and low-income populations and subsistence consumption resulting from continued operations and refurbishment associated with license renewal will be addressed in plant-specific reviews. See NRC Policy Statement on the Treatment of Environmental Justice Matters in NRC Regulatory and Licensing Actions (69 FR 52040; August 24, 2004).

                  Waste Management

                Low-level waste storage and disposal
                1
                SMALL. The comprehensive regulatory controls that are in place and the low public doses being achieved at reactors ensure that the radiological impacts to the environment would remain small during the license renewal term.

                Onsite storage of spent nuclear fuel
                1
                During the license renewal term, SMALL. The expected increase in the volume of spent nuclear fuel from an additional 20 years of operation can be safely accommodated onsite during the license renewal term with small environmental impacts through dry or pool storage at all plants.

                For the period after the licensed life for reactor operations, the impacts of onsite storage of spent nuclear fuel during the continued storage period are discussed in NUREG-2157 and as stated in § 51.23(b), shall be deemed incorporated into this issue.

                Offsite radiological impacts of spent nuclear fuel and high-level waste disposal
                1
                For the high-level waste and spent-fuel disposal component of the fuel cycle, the EPA established a dose limit of 0.15 mSv (15 millirem) per year for the first 10,000 years and 1.0 mSv (100 millirem) per year between 10,000 years and 1 million years for offsite releases of radionuclides at the proposed repository at Yucca Mountain, Nevada.The Commission concludes that the impacts would not be sufficiently large to require the NEPA conclusion, for any plant, that the option of extended operation under 10 CFR part 54 should be eliminated. Accordingly, while the Commission has not assigned a single level of significance for the impacts of spent fuel and high level waste disposal, this issue is considered Category 1.

                Mixed-waste storage and disposal
                1
                SMALL. The comprehensive regulatory controls and the facilities and procedures that are in place ensure proper handling and storage, as well as negligible doses and exposure to toxic materials for the public and the environment at all plants. License renewal would not increase the small, continuing risk to human health and the environment posed by mixed waste at all plants. The radiological and nonradiological environmental impacts of long-term disposal of mixed waste from any individual plant at licensed sites are small.

                Nonradioactive waste storage and disposal
                1
                SMALL. No changes to systems that generate nonradioactive waste are anticipated during the license renewal term. Facilities and procedures are in place to ensure continued proper handling, storage, and disposal, as well as negligible exposure to toxic materials for the public and the environment at all plants.

                  Cumulative Impacts

                Cumulative impacts
                2
                Cumulative impacts of continued operations and refurbishment associated with license renewal must be considered on a plant-specific basis. Impacts would depend on regional resource characteristics, the resource-specific impacts of license renewal, and the cumulative significance of other factors affecting the resource.

                  Uranium Fuel Cycle

                Offsite radiological impacts—individual impacts from other than the disposal of spent fuel and high-level waste
                1
                SMALL. The impacts to the public from radiological exposures have been considered by the Commission in Table S-3 of this part. Based on information in the GEIS, impacts to individuals from radioactive gaseous and liquid releases, including radon-222 and technetium-99, would remain at or below the NRC's regulatory limits.

                Offsite radiological impacts—collective impacts from other than the disposal of spent fuel and high-level waste
                1
                There are no regulatory limits applicable to collective doses to the general public from fuel-cycle facilities. The practice of estimating health effects on the basis of collective doses may not be meaningful. All fuel-cycle facilities are designed and operated to meet the applicable regulatory limits and standards. The Commission concludes that the collective impacts are acceptable.

                The Commission concludes that the impacts would not be sufficiently large to require the NEPA conclusion, for any plant, that the option of extended operation under 10 CFR part 54 should be eliminated. Accordingly, while the Commission has not assigned a single level of significance for the collective impacts of the uranium fuel cycle, this issue is considered Category 1.

                Nonradiological impacts of the uranium fuel cycle
                1
                SMALL. The nonradiological impacts of the uranium fuel cycle resulting from the renewal of an operating license for any plant would be small.

                Transportation
                1
                SMALL. The impacts of transporting materials to and from uranium-fuel-cycle facilities on workers, the public, and the environment are expected to be small.

                  Termination of Nuclear Power Plant Operations and Decommissioning

                Termination of plant operations and decommissioning
                1
                SMALL. License renewal is expected to have a negligible effect on the impacts of terminating operations and decommissioning on all resources.

                1 Data supporting this table are contained in NUREG-1437, Revision 1, “Generic Environmental Impact Statement for License Renewal of Nuclear Plants” (June 2013).

                2 The numerical entries in this column are based on the following category definitions:
              Category 1: For the issue, the analysis reported in the Generic Environmental Impact Statement has shown:
              (1) The environmental impacts associated with the issue have been determined to apply either to all plants or, for some issues, to plants having a specific type of cooling system or other specified plant or site characteristic;
              (2) A single significance level (i.e., small, moderate, or large) has been assigned to the impacts (except for Offsite radiological impacts—collective impacts from other than the disposal of spent fuel and high-level waste); and
              (3) Mitigation of adverse impacts associated with the issue has been considered in the analysis, and it has been determined that additional plant-specific mitigation measures are not likely to be sufficiently beneficial to warrant implementation.
              The generic analysis of the issue may be adopted in each plant-specific review.
              Category 2: For the issue, the analysis reported in the Generic Environmental Impact Statement has shown that one or more of the criteria of Category 1 cannot be met, and therefore additional plant-specific review is required.

                3 The impact findings in this column are based on the definitions of three significance levels. Unless the significance level is identified as beneficial, the impact is adverse, or in the case of “small,” may be negligible. The definitions of significance follow:
              SMALL—For the issue, environmental effects are not detectable or are so minor that they will neither destabilize nor noticeably alter any important attribute of the resource. For the purposes of assessing radiological impacts, the Commission has concluded that those impacts that do not exceed permissible levels in the Commission's regulations are considered small as the term is used in this table.
              MODERATE—For the issue, environmental effects are sufficient to alter noticeably, but not to destabilize, important attributes of the resource.
              LARGE—For the issue, environmental effects are clearly noticeable and are sufficient to destabilize important attributes of the resource.

              For issues where probability is a key consideration (i.e., accident consequences), probability was a factor in determining significance.

                4 This issue applies only to the in-scope portion of electric power transmission lines, which are defined as transmission lines that connect the nuclear power plant to the substation where electricity is fed into the regional power distribution system and transmission lines that supply power to the nuclear plant from the grid.

                5 NA (not applicable). The categorization and impact finding definitions do not apply to these issues.

                6 If, in the future, the Commission finds that, contrary to current indications, a consensus has been reached by appropriate Federal health agencies that there are adverse health effects from electromagnetic fields, the Commission will require applicants to submit plant-specific reviews of these health effects as part of their license renewal applications. Until such time, applicants for license renewal are not required to submit information on this issue.

            [61 FR 66546, Dec. 18, 1996, as amended at 62 FR 59276, Nov. 3, 1997; 64 FR 48507, Sept. 3, 1999; 66 FR 39278, July 30, 2001; 78 FR 37317, June 20, 2013; 79 FR 56262, Sept. 19, 2014]

          Subpart B [Reserved]

        Pt. 52
        PART 52—LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS

            General Provisions
            Sec.
            52.0
            Scope; applicability of 10 CFR Chapter I provisions.
            52.1
            Definitions.
            52.2
            Interpretations.
            52.3
            Written communications.
            52.4
            Deliberate misconduct.
            52.5
            Employee protection.
            52.6
            Completeness and accuracy of information.
            52.7
            Specific exemptions.
            52.8
            Combining licenses; elimination of repetition.
            52.9
            Jurisdictional limits.
            52.10
            Attacks and destructive acts.
            52.11
            Information collection requirements: OMB approval.

            Subpart A—Early Site Permits
            52.12
            Scope of subpart.
            52.13
            Relationship to other subparts.
            52.15
            Filing of applications.
            52.16

            Contents of applications; general information.

            52.17
            Contents of applications; technical information.
            52.18
            Standards for review of applications.
            52.21
            Administrative review of applications; hearings.
            52.23
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            52.24
            Issuance of early site permit.
            52.25
            Extent of activities permitted.
            52.26
            Duration of permit.
            52.27
            Limited work authorization after issuance of early site permit.
            52.28
            Transfer of early site permit.
            52.29
            Application for renewal.
            52.31
            Criteria for renewal.
            52.33
            Duration of renewal.
            52.35
            Use of site for other purposes.
            52.39
            Finality of early site permit determinations.

            Subpart B—Standard Design Certifications
            52.41
            Scope of subpart.
            52.43
            Relationship to other subparts.
            52.45
            Filing of applications.
            52.46
            Contents of applications; general information.
            52.47
            Contents of applications; technical information.
            52.48
            Standards for review of applications.
            52.51
            Administrative review of applications.
            52.53
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            52.54
            Issuance of standard design certification.
            52.55
            Duration of certification.
            52.57
            Application for renewal.
            52.59
            Criteria for renewal.
            52.61
            Duration of renewal.
            52.63
            Finality of standard design certifications.

            Subpart C—Combined Licenses
            52.71
            Scope of subpart.
            52.73
            Relationship to other subparts.
            52.75
            Filing of applications.
            52.77
            Contents of applications; general information.
            52.79
            Contents of applications; technical information in final safety analysis report.
            52.80
            Contents of applications; additional technical information.
            52.81
            Standards for review of applications.
            52.83
            Finality of referenced NRC approvals; partial initial decision on site suitability.
            52.85
            Administrative review of applications; hearings.
            52.87
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            52.89
            [Reserved]
            52.91
            Authorization to conduct limited work authorization activities.
            52.93
            Exemptions and variances.
            52.97
            Issuance of combined licenses.
            52.98
            Finality of combined licenses; information requests.
            52.99
            Inspection during construction; ITAAC schedules and notifications; NRC notices.
            52.103
            Operation under a combined license.
            52.104
            Duration of combined license.
            52.105
            Transfer of combined license.
            52.107
            Application for renewal.
            52.109
            Continuation of combined license.
            52.110
            Termination of license.

            Subpart D [Reserved]

            Subpart E—Standard Design Approvals
            52.131
            Scope of subpart.
            52.133
            Relationship to other subparts.
            52.135
            Filing of applications.
            52.136
            Contents of applications; general information.
            52.137
            Contents of applications; technical information.
            52.139
            Standards for review of applications.
            52.141
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            52.143
            Staff approval of design.
            52.145
            Finality of standard design approvals; information requests.
            52.147
            Duration of design approval.

            Subpart F—Manufacturing Licenses
            52.151
            Scope of subpart.
            52.153
            Relationship to other subparts.
            52.155
            Filing of applications.
            52.156
            Contents of applications; general information.
            52.157
            Contents of applications; technical information in final safety analysis report.
            52.158
            Contents of application; additional technical information.
            52.159
            Standards for review of application.
            52.161
            [Reserved]
            52.163
            Administrative review of applications; hearings.
            52.165
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            52.167
            Issuance of manufacturing license.
            52.169
            [Reserved]
            52.171
            Finality of manufacturing licenses; information requests.
            52.173
            Duration of manufacturing license.
            52.175
            Transfer of manufacturing license.
            52.177
            Application for renewal.
            52.179
            Criteria for renewal.
            52.181
            Duration of renewal.

            Subpart G [Reserved]

            Subpart H—Enforcement
            52.301
            Violations.
            52.303
            Criminal penalties.
            Appendix A to Part 52—Design Certification Rule for the U.S. Advanced Boiling Water Reactor

            Appendix B to Part 52—Design Certification Rule for the System 80 + Design

            Appendix C to Part 52—Design Certification Rule for the AP600 Design
            Appendix D to Part 52—Design Certification Rule for the AP1000 Design
            Appendix E to Part 52—Design Certification Rule for the ESBWR Design
            Appendix F to Part 52—Design Certification Rule for the APR1400 Design
            Appendixes G-M to Part 52 [Reserved]
            Appendix N to Part 52—Standardization of Nuclear Power Plant Designs: Combined Licenses to Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites

          Authority:
          Atomic Energy Act of 1954, secs. 103, 104, 147, 149, 161, 181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2133, 2134, 2167, 2169, 2201, 2231, 2232, 2233, 2235, 2236, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.

          Source:
          72 FR 49517, Aug. 28, 2007, unless otherwise noted.

          General Provisions

            § 52.0
            Scope; applicability of 10 CFR Chapter I provisions.
            (a) This part governs the issuance of early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities licensed under Section 103 of the Atomic Energy Act of 1954, as amended (68 Stat. 919), and Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242). This part also gives notice to all persons who knowingly provide to any holder of or applicant for an approval, certification, permit, or license, or to a contractor, subcontractor, or consultant of any of them, components, equipment, materials, or other goods or services that relate to the activities of a holder of or applicant for an approval, certification, permit, or license, subject to this part, that they may be individually subject to NRC enforcement action for violation of the provisions in 10 CFR 52.4.
            (b) Unless otherwise specifically provided for in this part, the regulations in 10 CFR Chapter I apply to a holder of or applicant for an approval, certification, permit, or license. A holder of or applicant for an approval, certification, permit, or license issued under this part shall comply with all requirements in 10 CFR Chapter I that are applicable. A license, approval, certification, or permit issued under this part is subject to all requirements in 10 CFR Chapter I which, by their terms, are applicable to early site permits, design certifications, combined licenses, design approvals, or manufacturing licenses.

            § 52.1
            Definitions.
            (a) As used in this part—

              Combined license means a combined construction permit and operating license with conditions for a nuclear power facility issued under subpart C of this part.

              Decommission means to remove a facility or site safely from service and reduce residual radioactivity to a level that permits—
            (i) Release of the property for unrestricted use and termination of the license; or
            (ii) Release of the property under restricted conditions and termination of the license.

              Design characteristics are the actual features of a reactor or reactors. Design characteristics are specified in a standard design approval, a standard design certification, a combined license application, or a manufacturing license.

              Design parameters are the postulated features of a reactor or reactors that could be built at a proposed site. Design parameters are specified in an early site permit.

              Early site permit means a Commission approval, issued under subpart A of this part, for a site for one or more nuclear power facilities. An early site permit is a partial construction permit.

              License means a license, including an early site permit, combined license or manufacturing license under this part or a renewed license issued by the Commission under this part or part 54 of this chapter.

              Licensee means a person who is authorized to conduct activities under a license issued by the Commission.

              Limited work authorization means the authorization provided by the Director of the Office of Nuclear Reactor Regulation under § 50.10 of this chapter.

              Major feature of the emergency plans means an aspect of those plans necessary to:

            (i) Address in whole or part one or more of the 16 standards in 10 CFR 50.47(b); or
            (ii) Describe the emergency planning zones as required in 10 CFR 50.33(g).

              Manufacturing license means a license, issued under subpart F of this part, authorizing the manufacture of nuclear power reactors but not their construction, installation, or operation at the sites on which the reactors are to be operated.

              Modular design means a nuclear power station that consists of two or more essentially identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated independent of the state of completion or operating condition of any other module co-located on the same site, even though the nuclear power station may have some shared or common systems.

              Prototype plant means a nuclear power plant that is used to test new safety features, such as the testing required under 10 CFR 50.43(e). The prototype plant is similar to a first-of-a-kind or standard plant design in all features and size, but may include additional safety features to protect the public and the plant staff from the possible consequences of accidents during the testing period.

              Site characteristics are the actual physical, environmental and demographic features of a site. Site characteristics are specified in an early site permit or in a final safety analysis report for a combined license.

              Site parameters are the postulated physical, environmental and demographic features of an assumed site. Site parameters are specified in a standard design approval, standard design certification, or manufacturing license.

              Standard design means a design which is sufficiently detailed and complete to support certification or approval in accordance with subpart B or E of this part, and which is usable for a multiple number of units or at a multiple number of sites without reopening or repeating the review.

              Standard design approval or design approval means an NRC staff approval, issued under subpart E of this part, of a final standard design for a nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final design for the entire reactor facility or the final design of major portions thereof.

              Standard design certification or design certification means a Commission approval, issued under subpart B of this part, of a final standard design for a nuclear power facility. This design may be referred to as a certified standard design.
            (b) All other terms in this part have the meaning set out in 10 CFR 50.2, or Section 11 of the Atomic Energy Act, as applicable.
            [72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57446, Oct. 9, 2007; 79 FR 66604, Nov. 10, 2014; 84 FR 65645, Nov. 29, 2019; 84 FR 68781, Dec. 17, 2019]

            § 52.2
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

            § 52.3
            Written communications.
            (a) General requirements. All correspondence, reports, applications, and other written communications from an applicant, licensee, or holder of a standard design approval to the Nuclear Regulatory Commission concerning the regulations in this part, individual license conditions, or the terms and conditions of an early site permit or standard design approval, must be sent either by mail addressed: ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland, between the hours of 7:30 a.m. and 4:15 p.m. eastern time; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, e-mail, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the communication is on paper, the signed original must be sent. If a submission due date falls on a Saturday, Sunday, or Federal holiday, the next Federal working day becomes the official due date.
            (b) Distribution requirements. Copies of all correspondence, reports, and other written communications concerning the regulations in this part or individual license conditions, or the terms and conditions of an early site permit or standard design approval, must be submitted to the persons listed in paragraph (b)(1) of this section (addresses for the NRC Regional Offices are listed in appendix D to part 20 of this chapter).
            (1) Applications for amendment of permits and licenses; reports; and other communications. All written communications (including responses to: generic letters, bulletins, information notices, regulatory information summaries, inspection reports, and miscellaneous requests for additional information) that are required of holders of early site permits, standard design approvals, combined licenses, or manufacturing licenses issued under this part must be submitted as follows, except as otherwise specified in paragraphs (b)(2) through (b)(7) of this section: to the NRC's Document Control Desk (if on paper, the signed original), with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector, if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under subpart F of this part.
            (2) Applications and amendments to applications. Applications for early site permits, standard design approvals, combined licenses, manufacturing licenses and amendments to any of these types of applications must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector, if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under subpart F of this part, except as otherwise specified in paragraphs (b)(3) through (b)(7) of this section. If the application or amendment is on paper, the submission to the Document Control Desk must be the signed original.
            (3) Acceptance review application. Written communications required for an application for determination of suitability for docketing must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office. If the communication is on paper, the submission to the Document Control Desk must be the signed original.
            (4) Security plan and related submissions. Written communications, as defined in paragraphs (b)(4)(i) through (iv) of this section, must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office. If the communication is on paper, the submission to the Document Control Desk must be the signed original.
            (i) Physical security plan under § 52.79 of this chapter;
            (ii) Safeguards contingency plan under § 52.79 of this chapter;
            (iii) Change to security plan, guard training and qualification plan, or safeguards contingency plan made without prior Commission approval under § 50.54(p) of this chapter;
            (iv) Application for amendment of physical security plan, guard training and qualification plan, or safeguards contingency plan under § 50.90 of this chapter.
            (5) Emergency plan and related submissions. Written communications as defined in paragraphs (b)(5)(i) through (iii) of this section must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility. If the communication is on paper, the submission to the Document Control Desk must be the signed original.

            (i) Emergency plan under § 52.17(b) or § 52.79(a);
            (ii) Change to an emergency plan under § 50.54(q) of this chapter;
            (iii) Emergency implementing procedures under appendix E, Section V of part 50 of this chapter.
            (6) Updated FSAR. An updated final safety analysis report (FSAR) or replacement pages under § 50.71(e) of this chapter, or the regulations in this part must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under subpart F of this part. Paper copy submissions may be made using replacement pages; however, if a licensee chooses to use electronic submission, all subsequent updates or submissions must be performed electronically on a total replacement basis. If the communication is on paper, the submission to the Document Control Desk must be the signed original. If the communications are submitted electronically, see Guidance for Electronic Submissions to the Commission.
            (7) Quality assurance related submissions. (i) A change to the safety analysis report quality assurance program description under § 50.54(a)(3) or § 50.55(f)(4) of this chapter, or a change to a licensee's NRC-accepted quality assurance topical report under § 50.54(a)(3) or § 50.55(f)(4) of this chapter, must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility. If the communication is on paper, the submission to the Document Control Desk must be the signed original.

            (ii) A change to an NRC-accepted quality assurance topical report from nonlicensees (i.e., architect/engineers, NSSS suppliers, fuel suppliers, constructors, etc.) must be submitted to the NRC's Document Control Desk. If the communication is on paper, the signed original must be sent.
            (8) Certification of permanent cessation of operations. The licensee's certification of permanent cessation of operations under § 52.110(a)(1), must state the date on which operations have ceased or will cease, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation.
            (9) Certification of permanent fuel removal. The licensee's certification of permanent fuel removal under § 52.110(a)(1), must state the date on which the fuel was removed from the reactor vessel and the disposition of the fuel, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation.
            (c) Form of communications. All paper copies submitted to meet the requirements set forth in paragraph (b) of this section must be typewritten, printed or otherwise reproduced in permanent form on unglazed paper. Exceptions to these requirements imposed on paper submissions may be granted for the submission of micrographic, photographic, or similar forms.
            (d) Regulation governing submission. Applicants, licensees, and holders of standard design approvals submitting correspondence, reports, and other written communications under the regulations of this part are requested but not required to cite whenever practical, in the upper right corner of the first page of the submission, the specific regulation or other basis requiring submission.
            [72 FR 49517, Aug. 28, 2007, as amended at 74 FR 62682, Dec. 1, 2009; 80 FR 74980, Dec. 1, 2015]

            § 52.4
            Deliberate misconduct.
            (a) Applicability. This section applies to any:
            (1) Licensee;
            (2) Holder of a standard design approval;
            (3) Applicant for a standard design certification;
            (4) Applicant for a license or permit;
            (5) Applicant for a standard design approval;
            (6) Employee of a licensee;
            (7) Employee of an applicant for a license, a standard design certification, or a standard design approval;

            (8) Any contractor (including a supplier or consultant), subcontractor, or employee of a contractor or subcontractor of any licensee; or
            (9) Any contractor (including a supplier or consultant), subcontractor, or employee of a contractor or subcontractor of any applicant for a license, a standard design certification, or a standard design approval.
            (b) Definitions. For purposes of this section:

              Deliberate misconduct means an intentional act or omission that a person or entity knows:
            (i) Would cause a licensee or an applicant for a license, standard design certification, or standard design approval to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license, standard design certification, or standard design approval; or
            (ii) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, holder of a standard design approval, applicant for a license, standard design certification, or standard design approval, or contractor, or subcontractor.
            (c) Prohibition against deliberate misconduct. Any person or entity subject to this section, who knowingly provides to any licensee, any applicant for a license, standard design certification or standard design approval, or a contractor, or subcontractor of a person or entity subject to this section, any components, equipment, materials, or other goods or services that relate to a licensee's or applicant's activities under this part, may not:
            (1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee, holder of a standard design approval, or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission, any standard design approval, or standard design certification; or
            (2) Deliberately submit to the NRC; a licensee, an applicant for a license, standard design certification or standard design approval; or a licensee's, standard design approval holder's, or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.
            (d) A person or entity who violates paragraph (c)(1) or (c)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.

            § 52.5
            Employee protection.
            (a) Discrimination by a Commission licensee, holder of a standard design approval, an applicant for a license, standard design certification, or standard design approval, a contractor or subcontractor of a Commission licensee, holder of a standard design approval, applicant for a license, standard design certification, or standard design approval, against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in Section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act or the Energy Reorganization Act.
            (1) The protected activities include but are not limited to:
            (i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in the introductory text of paragraph (a) of this section or possible violations of requirements imposed under either of those statutes;
            (ii) Refusing to engage in any practice made unlawful under either of the statutes named in the introductory text of paragraph (a) of this section or under these requirements if the employee has identified the alleged illegality to the employer;
            (iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;

            (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in the introductory text of paragraph (a) of this section; and

            (v) Assisting or participating in, or is about to assist or participate in, these activities.
            (2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.
            (3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.
            (b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.
            (c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, a holder of a standard design approval, an applicant for a Commission license, standard design certification, or a standard design approval, or a contractor or subcontractor of a Commission licensee, holder of a standard design approval, or any applicant may be grounds for—
            (1) Denial, revocation, or suspension of the license or standard design approval;
            (2) Withdrawal or revocation of a proposed or final standard design certification;
            (3) Imposition of a civil penalty on the licensee, holder of a standard design approval, or applicant (including an applicant for a standard design certification under this part following Commission adoption of final design certification rule) or a contractor or subcontractor of the licensee, holder of a standard design approval, or applicant.
            (4) Other enforcement action.
            (d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.
            (e)(1) Each licensee, each holder of a standard design approval, and each applicant for a license, standard design certification, or standard design approval, shall prominently post the revision of NRC Form 3, “Notice to Employees,” referenced in 10 CFR 19.11(e). This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted not later than thirty (30) days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, standard design certification, or standard design approval under 10 CFR part 52, and for 30 days following license termination or the expiration or termination of the standard design certification or standard design approval under 10 CFR part 52.

            (2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, via email to Forms.Resource@nrc.gov, or by visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.

            (f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor under Section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.
            (g) Part 19 of this chapter sets forth requirements and regulatory provisions applicable to licensees, holders of a standard design approval, applicants for a license, standard design certification, or standard design approval, and contractors or subcontractors of a Commission licensee, or holder of a standard design approval, and are in addition to the requirements in this section.
            [72 FR 49517, Aug. 28, 2007, as amended at 72 FR 63974, Nov. 14, 2007; 73 FR 30458, May 28, 2008; 79 FR 66604, Nov. 10, 2014]

            § 52.6
            Completeness and accuracy of information.
            (a) Information provided to the Commission by a licensee (including an early site permit holder, a combined license holder, and a manufacturing license holder), a holder of a standard design approval under this part, and an applicant for a license or an applicant for a standard design certification or a standard design approval under this part, and information required by statute or by the Commission's regulations, orders, license conditions, or terms and conditions of a standard design approval to be maintained by the licensee, the holder of a standard design approval under this part, the applicant for a standard design certification under this part following Commission adoption of a final design certification rule, and an applicant for a license, a standard design certification, or a standard design approval under this part shall be complete and accurate in all material respects.
            (b) Each applicant or licensee, each holder of a standard design approval under this part, and each applicant for a standard design certification under this part following Commission adoption of a final design certification regulation, shall notify the Commission of information identified by the applicant or the licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant, licensee, or holder violates this paragraph only if the applicant, licensee, or holder fails to notify the Commission of information that the applicant, licensee, or holder has been identified as having a significant implication for public health and safety or common defense and security. Notification shall be provided to the Administrator of the appropriate Regional Office within 2 working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.

            § 52.7
            Specific exemptions.
            The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part. The Commission's consideration will be governed by § 50.12 of this chapter, unless other criteria are provided for in this part, in which case the Commission's consideration will be governed by the criteria in this part. Only if those criteria are not met will the Commission's consideration be governed by § 50.12 of this chapter. The Commission's consideration of requests for exemptions from requirements of the regulations of other parts in this chapter, which are applicable by virtue of this part, shall be governed by the exemption requirements of those parts.

            § 52.8
            Combining licenses; elimination of repetition.
            (a) An applicant for a license under this part may combine in its application several applications for different kinds of licenses under the regulations of this chapter.
            (b) An applicant may incorporate by reference in its application information contained in previous applications, statements or reports filed with the Commission, provided, however, that such references are clear and specific.
            (c) The Commission may combine in a single license the activities of an applicant which would otherwise be licensed separately.

            § 52.9
            Jurisdictional limits.
            No permit, license, standard design approval, or standard design certification under this part shall be deemed to have been issued for activities which are not under or within the jurisdiction of the United States.

            § 52.10
            Attacks and destructive acts.
            Neither an applicant for a license to manufacture, construct, and operate a utilization facility under this part, nor for an amendment to this license, or an applicant for an early site permit, a standard design certification, or standard design approval under this part, or for an amendment to the early site permit, standard design certification, or standard design approval, is required to provide for design features or other measures for the specific purpose of protection against the effects of—
            (a) Attacks and destructive acts, including sabotage, directed against the facility by an enemy of the United States, whether a foreign government or other person; or
            (b) Use or deployment of weapons incident to U.S. defense activities.

            § 52.11
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under Control Number 3150-0151.
            (b) The approved information collection requirements contained in this part appear in §§ 52.7, 52.15, 52.16, 52.17, 52.29, 52.35, 52.39, 52.45, 52.46, 52.47, 52.57, 52.63, 52.75, 52.77, 52.79, 52.80, 52.93, 52.99, 52.110, 52.135, 52.136, 52.137, 52.155, 52.156, 52.157, 52.158, 52.171, 52.177, and appendices A, B, C, D, E, F, and N of this part.
            [72 FR 49517, Aug. 28, 2007, as amended at 79 FR 61983, Nov. 14, 2014; 84 FR 23452, May 22, 2019]

          Subpart A—Early Site Permits

            § 52.12
            Scope of subpart.
            This subpart sets out the requirements and procedures applicable to Commission issuance of an early site permit for approval of a site for one or more nuclear power facilities separate from the filing of an application for a construction permit or combined license for the facility.

            § 52.13
            Relationship to other subparts.
            This subpart applies when any person who may apply for a construction permit under 10 CFR part 50, or for a combined license under this part seeks an early site permit from the Commission separately from an application for a construction permit or a combined license.

            § 52.15
            Filing of applications.
            (a) Any person who may apply for a construction permit under 10 CFR part 50, or for a combined license under this part, may file an application for an early site permit with the Director, Office of Nuclear Reactor Regulation. An application for an early site permit may be filed notwithstanding the fact that an application for a construction permit or a combined license has not been filed in connection with the site for which a permit is sought.
            (b) The application must comply with the applicable filing requirements of §§ 52.3 and 50.30 of this chapter.
            (c) The fees associated with the filing and review of an application for the initial issuance or renewal of an early site permit are set forth in 10 CFR part 170.
            [72 FR 49517, Aug. 28, 2007, as amended at 84 FR 65645, Nov. 29, 2019]

            § 52.16
            Contents of applications; general information.
            The application must contain all of the information required by 10 CFR 50.33(a) through (d) and (j) of this chapter.

            § 52.17
            Contents of applications; technical information.
            (a) For applications submitted before September 27, 2007, the rule provisions in effect at the date of docketing apply unless otherwise requested by the applicant in writing. The application must contain:
            (1) A site safety analysis report. The site safety analysis report shall include the following:
            (i) The specific number, type, and thermal power level of the facilities, or range of possible facilities, for which the site may be used;
            (ii) The anticipated maximum levels of radiological and thermal effluents each facility will produce;
            (iii) The type of cooling systems, intakes, and outflows that may be associated with each facility;
            (iv) The boundaries of the site;
            (v) The proposed general location of each facility on the site;
            (vi) The seismic, meteorological, hydrologic, and geologic characteristics of the proposed site with appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area and with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated;
            (vii) The location and description of any nearby industrial, military, or transportation facilities and routes;
            (viii) The existing and projected future population profile of the area surrounding the site;
            (ix) A description and safety assessment of the site on which a facility is to be located. The assessment must contain an analysis and evaluation of the major structures, systems, and components of the facility that bear significantly on the acceptability of the site under the radiological consequence evaluation factors identified in paragraphs (a)(1)(ix)(A) and (a)(1)(ix)(B) of this section. In performing this assessment, an applicant shall assume a fission product release 1
               from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable site characteristics, including site meteorology, to evaluate the offsite radiological consequences. Site characteristics must comply with part 100 of this chapter. The evaluation must determine that:

                1 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

            (A) An individual located at any point on the boundary of the exclusion area for any 2 hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem 2
               total effective dose equivalent (TEDE).

                2 A whole body dose of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations at the time could be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, its use is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, to assure that these designs provide assurance of low risk of public exposure to radiation, in the event of an accidents.

            (B) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE;
            (x) Information demonstrating that site characteristics are such that adequate security plans and measures can be developed;

            (xi) For applications submitted after September 27, 2007, a description of the quality assurance program applied to site-related activities for the future design, fabrication, construction, and testing of the structures, systems, and components of a facility or facilities that may be constructed on the site. Appendix B to 10 CFR part 50 sets forth the requirements for quality assurance programs for nuclear power plants. The description of the quality assurance program for a nuclear power plant site shall include a discussion of how the applicable requirements of appendix B to part 50 of this chapter will be satisfied; and
            (xii) An evaluation of the site against applicable sections of the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. The evaluation required by this section shall include an identification and description of all differences in analytical techniques and procedural measures proposed for a site and those corresponding techniques and measures given in the SRP acceptance criteria. Where such a difference exists, the evaluation shall discuss how the proposed alternative provides an acceptable method of complying with the Commission's regulations, or portions thereof, that underlie the corresponding SRP acceptance criteria. The SRP is not a substitute for the regulations, and compliance is not a requirement.
            (2) A complete environmental report as required by 10 CFR 51.50(b).
            (b)(1) The site safety analysis report must identify physical characteristics of the proposed site, such as egress limitations from the area surrounding the site, that could pose a significant impediment to the development of emergency plans. If physical characteristics are identified that could pose a significant impediment to the development of emergency plans, the application must identify measures that would, when implemented, mitigate or eliminate the significant impediment.
            (2) The site safety analysis report may also:
            (i) Propose major features of the emergency plans, in accordance with the pertinent standards of § 50.47 of this chapter and the requirements of appendix E to part 50 of this chapter, such as the exact size and configuration of the emergency planning zones, for review and approval by the NRC, in consultation with the Federal Emergency Management Agency (FEMA) in the absence of complete and integrated emergency plans; or
            (ii) Propose complete and integrated emergency plans for review and approval by the NRC, in consultation with FEMA, in accordance with the applicable standards of § 50.47 of this chapter and the requirements of appendix E to part 50 of this chapter. To the extent approval of emergency plans is sought, the application must contain the information required by § 50.33(g) and (j) of this chapter.
            (3) Emergency plans submitted under paragraph (b)(2)(ii) of this section must include the proposed inspections, tests, and analyses that the holder of a combined license referencing the early site permit shall perform, and the acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, the facility has been constructed and will be operated in conformity with the emergency plans, the provisions of the Act, and the Commission's rules and regulations. Major features of an emergency plan submitted under paragraph (b)(2)(i) of this section may include proposed inspections, tests, analyses, and acceptance criteria.

            (4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the site safety analysis report must include a description of contacts and arrangements made with Federal, State, and local governmental agencies with emergency planning responsibilities. The site safety analysis report must contain any certifications that have been obtained. If these certifications cannot be obtained, the site safety analysis report must contain information, including a utility plan, sufficient to show that the proposed plans provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency at the site. Under the option set forth in paragraph (b)(2)(ii) of this section, the applicant shall make good faith efforts to obtain from the same governmental agencies certifications that:

            (i) The proposed emergency plans are practicable;
            (ii) These agencies are committed to participating in any further development of the plans, including any required field demonstrations, and
            (iii) That these agencies are committed to executing their responsibilities under the plans in the event of an emergency.
            (c) An applicant may request that a limited work authorization under 10 CFR 50.10 be issued in conjunction with the early site permit. The application must include the information otherwise required by 10 CFR 50.10(d)(3). Applications submitted before, and pending as of November 8, 2007, must include the information required by § 52.17(c) effective on the date of docketing.
            (d) Each applicant for an early site permit under this part shall protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable.
            [72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57447, Oct. 9, 2007; 73 FR 63571, Oct. 24, 2008; 78 FR 34249, June 7, 2013; 78 FR 75450, Dec. 12, 2013]

            § 52.18
            Standards for review of applications.
            Applications filed under this subpart will be reviewed according to the applicable standards set out in 10 CFR part 50 and its appendices and 10 CFR part 100. In addition, the Commission shall prepare an environmental impact statement during review of the application, in accordance with the applicable provisions of 10 CFR part 51. The Commission shall determine, after consultation with FEMA, whether the information required of the applicant by § 52.17(b)(1) shows that there is not significant impediment to the development of emergen cy plans that cannot be mitigated or eliminated by measures proposed by the applicant, whether any major features of emergency plans submitted by the applicant under § 52.17(b)(2)(i) are acceptable in accordance with the applicable standards of § 50.47 of this chapter and the requirements of appendix E to part 50 of this chapter, and whether any emergency plans submitted by the applicant under § 52.17(b)(2)(ii) provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
            [72 FR 49517, Aug. 28, 2007, as amended at 78 FR 34249, June 7, 2013; 78 FR 75450, Dec. 12, 2013]

            § 52.21
            Administrative review of applications; hearings.
            An early site permit is subject to all procedural requirements in 10 CFR part 2, including the requirements for docketing in § 2.101(a)(1) through (4) of this chapter, and the requirements for issuance of a notice of hearing in §§ 2.104(a) and (d) of this chapter, provided that the designated sections may not be construed to require that the environmental report, or draft or final environmental impact statement include an assessment of the benefits of construction and operation of the reactor or reactors, or an analysis of alternative energy sources. The presiding officer in an early site permit hearing shall not admit contentions proffered by any party concerning an assessment of the benefits of construction and operation of the reactor or reactors, or an analysis of alternative energy sources if those issues were not addressed by the applicant in the early site permit application. All hearings conducted on applications for early site permits filed under this part are governed by the procedures contained in subparts C, G, L, and N of 10 CFR part 2, as applicable.

            § 52.23
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            The Commission shall refer a copy of the application for an early site permit to the ACRS. The ACRS shall report on those portions of the application which concern safety.

            § 52.24
            Issuance of early site permit.
            (a) After conducting a hearing under § 52.21 and receiving the report to be submitted by the ACRS under § 52.23, the Commission may issue an early site permit, in the form the Commission deems appropriate, if the Commission finds that:

            (1) An application for an early site permit meets the applicable standards and requirements of the Act and the Commission's regulations;
            (2) Notifications, if any, to other agencies or bodies have been duly made;
            (3) There is reasonable assurance that the site is in conformity with the provisions of the Act, and the Commission's regulations;
            (4) The applicant is technically qualified to engage in any activities authorized;
            (5) The proposed inspections, tests, analyses and acceptance criteria, including any on emergency planning, are necessary and sufficient, within the scope of the early site permit, to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's regulations;
            (6) Issuance of the permit will not be inimical to the common defense and security or to the health and safety of the public;
            (7) Any significant adverse environmental impact resulting from activities requested under § 52.17(c) can be redressed; and
            (8) The findings required by subpart A of 10 CFR part 51 have been made.
            (b) The early site permit must specify the site characteristics, design parameters, and terms and conditions of the early site permit the Commission deems appropriate. Before issuance of either a construction permit or combined license referencing an early site permit, the Commission shall find that any relevant terms and conditions of the early site permit have been met. Any terms or conditions of the early site permit that could not be met by the time of issuance of the construction permit or combined license, must be set forth as terms or conditions of the construction permit or combined license.
            (c) The early site permit shall specify those 10 CFR 50.10 activities requested under § 52.17(c) that the permit holder is authorized to perform.
            [72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57447, Oct. 9, 2007]

            § 52.25
            Extent of activities permitted.
            If the activities authorized by § 52.24(c) are performed and the site is not referenced in an application for a construction permit or a combined license issued under subpart C of this part while the permit remains valid, then the early site permit remains in effect solely for the purpose of site redress, and the holder of the permit shall redress the site in accordance with the terms of the site redress plan required by § 52.17(c). If, before redress is complete, a use not envisaged in the redress plan is found for the site or parts thereof, the holder of the permit shall carry out the redress plan to the greatest extent possible consistent with the alternate use.

            § 52.26
            Duration of permit.
            (a) Except as provided in paragraph (b) of this section, an early site permit issued under this subpart may be valid for not less than 10, nor more than 20 years from the date of issuance.
            (b) An early site permit continues to be valid beyond the date of expiration in any proceeding on a construction permit application or a combined license application that references the early site permit and is docketed before the date of expiration of the early site permit, or, if a timely application for renewal of the permit has been docketed, before the Commission has determined whether to renew the permit.
            (c) An applicant for a construction permit or combined license may, at its own risk, reference in its application a site for which an early site permit application has been docketed but not granted.
            (d) Upon issuance of a construction permit or combined license, a referenced early site permit is subsumed, to the extent referenced, into the construction permit or combined license.
            [72 FR 49517, Aug. 28, 2007. Redesignated at 72 FR 57447, Oct. 9, 2007]

            § 52.27
            Limited work authorization after issuance of early site permit.
            A holder of an early site permit may request a limited work authorization in accordance with § 50.10 of this chapter.
            [72 FR 57447, Oct. 9, 2007]

            § 52.28
            Transfer of early site permit.
            An application to transfer an early site permit will be processed under 10 CFR 50.80.

            § 52.29
            Application for renewal.
            (a) Not less than 12, nor more than 36 months before the expiration date stated in the early site permit, or any later renewal period, the permit holder may apply for a renewal of the permit. An application for renewal must contain all information necessary to bring up to date the information and data contained in the previous application.
            (b) Any person whose interests may be affected by renewal of the permit may request a hearing on the application for renewal. The request for a hearing must comply with 10 CFR 2.309. If a hearing is granted, notice of the hearing will be published in accordance with 10 CFR 2.309.
            (c) An early site permit, either original or renewed, for which a timely application for renewal has been filed, remains in effect until the Commission has determined whether to renew the permit. If the permit is not renewed, it continues to be valid in certain proceedings in accordance with the provisions of § 52.27(b).
            (d) The Commission shall refer a copy of the application for renewal to the ACRS. The ACRS shall report on those portions of the application which concern safety and shall apply the criteria set forth in § 52.31.

            § 52.31
            Criteria for renewal.
            (a) The Commission shall grant the renewal if it determines that:
            (1) The site complies with the Act, the Commission's regulations, and orders applicable and in effect at the time the site permit was originally issued; and
            (2) Any new requirements the Commission may wish to impose are:
            (i) Necessary for adequate protection to public health and safety or common defense and security;
            (ii) Necessary for compliance with the Commission's regulations, and orders applicable and in effect at the time the site permit was originally issued; or
            (iii) A substantial increase in overall protection of the public health and safety or the common defense and security to be derived from the new requirements, and the direct and indirect costs of implementation of those requirements are justified in view of this increased protection.
            (b) A denial of renewal for failure to comply with the provisions of § 52.31(a) does not bar the permit holder or another applicant from filing a new application for the site which proposes changes to the site or the way that it is used to correct the deficiencies cited in the denial of the renewal.

            § 52.33
            Duration of renewal.
            Each renewal of an early site permit may be for not less than 10, nor more than 20 years, plus any remaining years on the early site permit then in effect before renewal.

            § 52.35
            Use of site for other purposes.
            A site for which an early site permit has been issued under this subpart may be used for purposes other than those described in the permit, including the location of other types of energy facilities. The permit holder shall inform the Director, Office of Nuclear Reactor Regulation (Director), of any significant uses for the site which have not been approved in the early site permit. The information about the activities must be given to the Director at least 30 days in advance of any actual construction or site modification for the activities. The information provided could be the basis for imposing new requirements on the permit, in accordance with the provisions of § 52.39. If the permit holder informs the Director that the holder no longer intends to use the site for a nuclear power plant, the Director may terminate the permit.
            [73 FR 5724, Jan. 31, 2008, as amended at 84 FR 65645, Nov. 29, 2019; 84 FR 68781, Dec. 17, 2019]

            § 52.39
            Finality of early site permit determinations.
            (a) Commission finality. (1) Notwithstanding any provision in 10 CFR 50.109, while an early site permit is in effect under §§ 52.27 or 52.33, the Commission may not change or impose new site characteristics, design parameters, or terms and conditions, including emergency planning requirements, on the early site permit unless the Commission:
            (i) Determines that a modification is necessary to bring the permit or the site into compliance with the Commission's regulations and orders applicable and in effect at the time the permit was issued;
            (ii) Determines the modification is necessary to assure adequate protection of the public health and safety or the common defense and security;
            (iii) Determines that a modification is necessary based on an update under paragraph (b) of this section; or
            (iv) Issues a variance requested under paragraph (d) of this section.
            (2) In making the findings required for issuance of a construction permit or combined license, or the findings required by § 52.103, or in any enforcement hearing other than one initiated by the Commission under paragraph (a)(1) of this section, if the application for the construction permit or combined license references an early site permit, the Commission shall treat as resolved those matters resolved in the proceeding on the application for issuance or renewal of the early site permit, except as provided for in paragraphs (b), (c), and (d) of this section.
            (i) If the early site permit approved an emergency plan (or major features thereof) that is in use by a licensee of a nuclear power plant, the Commission shall treat as resolved changes to the early site permit emergency plan (or major features thereof) that are identical to changes made to the licensee's emergency plans in compliance with § 50.54(q) of this chapter occurring after issuance of the early site permit.
            (ii) If the early site permit approved an emergency plan (or major features thereof) that is not in use by a licensee of a nuclear power plant, the Commission shall treat as resolved changes that are equivalent to those that could be made under § 50.54(q) of this chapter without prior NRC approval had the emergency plan been in use by a licensee.
            (b) Updating of early site permit-emergency preparedness. An applicant for a construction permit, operating license, or combined license who has filed an application referencing an early site permit issued under this subpart shall update the emergency preparedness information that was provided under § 52.17(b), and discuss whether the updated information materially changes the bases for compliance with applicable NRC requirements.
            (c) Hearings and petitions. (1) In any proceeding for the issuance of a construction permit, operating license, or combined license referencing an early site permit, contentions on the following matters may be litigated in the same manner as other issues material to the proceeding:
            (i) The nuclear power reactor proposed to be built does not fit within one or more of the site characteristics or design parameters included in the early site permit;
            (ii) One or more of the terms and conditions of the early site permit have not been met;
            (iii) A variance requested under paragraph (d) of this section is unwarranted or should be modified;
            (iv) New or additional information is provided in the application that substantially alters the bases for a previous NRC conclusion or constitutes a sufficient basis for the Commission to modify or impose new terms and conditions related to emergency preparedness; or
            (v) Any significant environmental issue that was not resolved in the early site permit proceeding, or any issue involving the impacts of construction and operation of the facility that was resolved in the early site permit proceeding for which significant new information has been identified.

            (2) Any person may file a petition requesting that the site characteristics, design parameters, or terms and conditions of the early site permit should be modified, or that the permit should be suspended or revoked. The petition will be considered in accordance with § 2.206 of this chapter. Before construction commences, the Commission shall consider the petition and determine whether any immediate action is required. If the petition is granted, an appropriate order will be issued. Construction under the construction permit or combined license will not be affected by the granting of the petition unless the order is made immediately effective. Any change required by the Commission in response to the petition must meet the requirements of paragraph (a)(1) of this section.
            (d) Variances. An applicant for a construction permit, operating license, or combined license referencing an early site permit may include in its application a request for a variance from one or more site characteristics, design parameters, or terms and conditions of the early site permit, or from the site safety analysis report. In determining whether to grant the variance, the Commission shall apply the same technically relevant criteria applicable to the application for the original or renewed early site permit. Once a construction permit or combined license referencing an early site permit is issued, variances from the early site permit will not be granted for that construction permit or combined license.
            (e) Early site permit amendment. The holder of an early site permit may not make changes to the early site permit, including the site safety analysis report, without prior Commission approval. The request for a change to the early site permit must be in the form of an application for a license amendment, and must meet the requirements of 10 CFR 50.90 and 50.92.
            (f) Information requests. Except for information requests seeking to verify compliance with the current licensing basis of the early site permit, information requests to the holder of an early site permit must be evaluated before issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information. Each evaluation performed by the NRC staff must be in accordance with 10 CFR 50.54(f), and must be approved by the Executive Director for Operations or his or her designee before issuance of the request.

          Subpart B—Standard Design Certifications

            § 52.41
            Scope of subpart.
            (a) This subpart sets forth the requirements and procedures applicable to Commission issuance of rules granting standard design certifications for nuclear power facilities separate from the filing of an application for a construction permit or combined license for such a facility.
            (b)(1) Any person may seek a standard design certification for an essentially complete nuclear power plant design which is an evolutionary change from light water reactor designs of plants which have been licensed and in commercial operation before April 18, 1989.
            (2) Any person may also seek a standard design certification for a nuclear power plant design which differs significantly from the light water reactor designs described in paragraph (b)(1) of this section or uses simplified, inherent, passive, or other innovative means to accomplish its safety functions.

            § 52.43
            Relationship to other subparts.
            (a) This subpart applies to a person that requests a standard design certification from the NRC separately from an application for a combined license filed under subpart C of this part for a nuclear power facility. An applicant for a combined license may reference a standard design certification.
            (b) Subpart E of this part governs the NRC staff review and approval of a standard design. Subpart E may be used independently of the provisions in this subpart.
            (c) Subpart F of this part governs the issuance of licenses to manufacture nuclear power reactors to be installed and operated at sites not identified in the manufacturing license application. Subpart F may be used independently of the provisions in this subpart. However, an applicant for a manufacturing license under subpart F may reference a design certification.
            [72 FR 49517, Aug. 28, 2007, as amended at 84 FR 63568, Nov. 18, 2019]

            § 52.45
            Filing of applications.
            (a) An application for design certification may be filed notwithstanding the fact that an application for a construction permit, combined license, or manufacturing license for such a facility has not been filed.

            (b) The application must comply with the applicable filing requirements of §§ 52.3 and §§ 2.811 through 2.819 of this chapter.
            (c) The fees associated with the review of an application for the initial issuance or renewal of a standard design certification are set forth in 10 CFR part 170.

            § 52.46
            Contents of applications; general information.
            The application must contain all of the information required by 10 CFR 50.33(a) through (c) and (j).

            § 52.47
            Contents of applications; technical information.
            The application must contain a level of design information sufficient to enable the Commission to judge the applicant's proposed means of assuring that construction conforms to the design and to reach a final conclusion on all safety questions associated with the design before the certification is granted. The information submitted for a design certification must include performance requirements and design information sufficiently detailed to permit the preparation of acceptance and inspection requirements by the NRC, and procurement specifications and construction and installation specifications by an applicant. The Commission will require, before design certification, that information normally contained in certain procurement specifications and construction and installation specifications be completed and available for audit if the information is necessary for the Commission to make its safety determination.
            (a) The application must contain a final safety analysis report (FSAR) that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole, and must include the following information:
            (1) The site parameters postulated for the design, and an analysis and evaluation of the design in terms of those site parameters;
            (2) A description and analysis of the structures, systems, and components (SSCs) of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, upon which these requirements have been established, and the evaluations required to show that safety functions will be accomplished. It is expected that the standard plant will reflect through its design, construction, and operation an extremely low probability for accidents that could result in the release of significant quantities of radioactive fission products. The description shall be sufficient to permit understanding of the system designs and their relationship to the safety evaluations. Such items as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems shall be discussed insofar as they are pertinent. The following power reactor design characteristics will be taken into consideration by the Commission:
            (i) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials;
            (ii) The extent to which generally accepted engineering standards are applied to the design of the reactor;
            (iii) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials; and
            (iv) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. Special attention must be directed to plant design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release 3

               from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable postulated site parameters, including site meteorology, to evaluate the offsite radiological consequences. The evaluation must determine that:

                3 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. These accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

            (A) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem 4
               total effective dose equivalent (TEDE);

                4 A whole body dose of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations at the time could be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, its use is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. This dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, to assure that these designs provide assurance of low risk of public exposure to radiation, in the event of an accident.

            (B) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE;
            (3) The design of the facility including:
            (i) The principal design criteria for the facility. Appendix A to 10 CFR part 50, general design criteria (GDC), establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units;
            (ii) The design bases and the relation of the design bases to the principal design criteria;
            (iii) Information relative to materials of construction, general arrangement, and approximate dimensions, sufficient to provide reasonable assurance that the design will conform to the design bases with an adequate margin for safety;
            (4) An analysis and evaluation of the design and performance of structures, systems, and components with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. Analysis and evaluation of emergency core cooling system (ECCS) cooling performance and the need for high-point vents following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of §§ 50.46 and 50.46a of this chapter;
            (5) The kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in part 20 of this chapter;
            (6) The information required by § 20.1406 of this chapter;
            (7) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter;
            (8) The information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), and (f)(3)(v);

            (9) For applications for light-water-cooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. The evaluation required by this section shall include an identification and description of all differences in design features, analytical techniques, and procedural measures proposed for the design and those corresponding features, techniques, and measures given in the SRP acceptance criteria. Where a difference exists, the evaluation shall discuss how the proposed alternative provides an acceptable method of complying with the Commission's regulations, or portions thereof, that underlie the corresponding SRP acceptance criteria. The SRP is not a substitute for the regulations, and compliance is not a requirement.
            (10) The information with respect to the design of equipment to maintain control over radioactive materials in gaseous and liquid effluents produced during normal reactor operations described in 10 CFR 50.34a(e);
            (11) Proposed technical specifications prepared in accordance with the requirements of §§ 50.36 and 50.36a of this chapter;
            (12) An analysis and description of the equipment and systems for combustible gas control as required by 10 CFR 50.44;
            (13) The list of electric equipment important to safety that is required by 10 CFR 50.49(d);
            (14) A description of protection provided against pressurized thermal shock events, including projected values of the reference temperature for reactor vessel beltline materials as defined in 10 CFR 50.60 and 50.61;
            (15) Information demonstrating how the applicant will comply with requirements for reduction of risk from anticipated transients without scram events in § 50.62;
            (16) A coping analysis, and any design features necessary to address station blackout, as required by 10 CFR 50.63;
            (17) Information demonstrating how the applicant will comply with requirements for criticality accidents in § 50.68(b)(2)-(b)(4);
            (18) A description and analysis of the fire protection design features for the standard plant necessary to comply with 10 CFR part 50, appendix A, GDC 3, and § 50.48 of this chapter;
            (19) A description of the quality assurance program applied to the design of the structures, systems, and components of the facility. Appendix B to 10 CFR part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” sets forth the requirements for quality assurance programs for nuclear power plants. The description of the quality assurance program for a nuclear power plant shall include a discussion of how the applicable requirements of appendix B to 10 CFR part 50 were satisfied;
            (20) The information necessary to demonstrate that the standard plant complies with the earthquake engineering criteria in 10 CFR part 50, appendix S;
            (21) Proposed technical resolutions of those Unresolved Safety Issues and medium- and high-priority generic safety issues which are identified in the version of NUREG-0933 current on the date up to 6 months before the docket date of the application and which are technically relevant to the design;
            (22) The information necessary to demonstrate how operating experience insights have been incorporated into the plant design;
            (23) For light-water reactor designs, a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass;
            (24) A representative conceptual design for those portions of the plant for which the application does not seek certification, to aid the NRC in its review of the FSAR and to permit assessment of the adequacy of the interface requirements in paragraph (a)(25) of this section;
            (25) The interface requirements to be met by those portions of the plant for which the application does not seek certification. These requirements must be sufficiently detailed to allow completion of the FSAR;

            (26) Justification that compliance with the interface requirements of paragraph (a)(25) of this section is verifiable through inspections, tests, or analyses. The method to be used for verification of interface requirements must be included as part of the proposed ITAAC required by paragraph (b)(1) of this section; and
            (27) A description of the design-specific probabilistic risk assessment (PRA) and its results.
            (28) For applications for standard design certifications which are subject to 10 CFR 50.150(a), the information required by 10 CFR 50.150(b).
            (b) The application must also contain:
            (1) The proposed inspections, tests, analyses, and acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, a facility that incorporates the design certification has been constructed and will be operated in conformity with the design certification, the provisions of the Act, and the Commission's rules and regulations; and
            (2) An environmental report as required by 10 CFR 51.55.
            (c) This paragraph applies, according to its provisions, to particular applications:
            (1) An application for certification of a nuclear power reactor design that is an evolutionary change from light-water reactor designs of plants that have been licensed and in commercial operation before April 18, 1989, must provide an essentially complete nuclear power plant design except for site-specific elements such as the service water intake structure and the ultimate heat sink;
            (2) An application for certification of a nuclear power reactor design that differs significantly from the light-water reactor designs described in paragraph (c)(1) of this section or uses simplified, inherent, passive, or other innovative means to accomplish its safety functions must provide an essentially complete nuclear power reactor design except for site-specific elements such as the service water intake structure and the ultimate heat sink, and must meet the requirements of 10 CFR 50.43(e); and
            (3) An application for certification of a modular nuclear power reactor design must describe and analyze the possible operating configurations of the reactor modules with common systems, interface requirements, and system interactions. The final safety analysis must also account for differences among the configurations, including any restrictions that will be necessary during the construction and startup of a given module to ensure the safe operation of any module already operating.
            (d) Each applicant for a standard design certification under this part shall protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable.
            [72 FR 49517, Aug. 28, 2007, as amended at 73 FR 63571, Oct. 24, 2008; 74 FR 28147, June 12, 2009]

            § 52.48
            Standards for review of applications.
            Applications filed under this subpart will be reviewed for compliance with the standards set out in 10 CFR parts 20, 50 and its appendices, 51, 73, and 100.

            § 52.51
            Administrative review of applications.

            (a) A standard design certification is a rule that will be issued in accordance with the provisions of subpart H of 10 CFR part 2, as supplemented by the provisions of this section. The Commission shall initiate the rulemaking after an application has been filed under § 52.45 and shall specify the procedures to be used for the rulemaking. The notice of proposed rulemaking published in the Federal Register must provide an opportunity for the submission of comments on the proposed design certification rule. If, at the time a proposed design certification rule is published in the Federal Register under this paragraph (a), the Commission decides that a legislative hearing should be held, the information required by 10 CFR 2.1502(c) must be included in the Federal Register document for the proposed design certification.

            (b) Following the submission of comments on the proposed design certification rule, the Commission may, at its discretion, hold a legislative hearing under the procedures in subpart O of part 2 of this chapter. The Commission shall publish a document in the Federal Register of its decision to hold a legislative hearing. The document shall contain the information specified in paragraph (c) of this section, and specify whether the Commission or a presiding officer will conduct the legislative hearing.
            (c) Notwithstanding anything in 10 CFR 2.390 to the contrary, proprietary information will be protected in the same manner and to the same extent as proprietary information submitted in connection with applications for licenses, provided that the design certification shall be published in Chapter I of this title.

            § 52.53
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            The Commission shall refer a copy of the application to the ACRS. The ACRS shall report on those portions of the application which concern safety.

            § 52.54
            Issuance of standard design certification.
            (a) After conducting a rulemaking proceeding under § 52.51 on an application for a standard design certification and receiving the report to be submitted by the Advisory Committee on Reactor Safeguards under § 52.53, the Commission may issue a standard design certification in the form of a rule for the design which is the subject of the application, if the Commission determines that:
            (1) The application meets the applicable standards and requirements of the Atomic Energy Act and the Commission's regulations;
            (2) Notifications, if any, to other agencies or bodies have been duly made;
            (3) There is reasonable assurance that the standard design conforms with the provisions of the Act, and the Commission's regulations;
            (4) The applicant is technically qualified;
            (5) The proposed inspections, tests, analyses, and acceptance criteria are necessary and sufficient, within the scope of the standard design, to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, the facility has been constructed and will be operated in accordance with the design certification, the provisions of the Act, and the Commission's regulations;
            (6) Issuance of the standard design certification will not be inimical to the common defense and security or to the health and safety of the public;
            (7) The findings required by subpart A of part 51 of this chapter have been made; and
            (8) The applicant has implemented the quality assurance program described or referenced in the safety analysis report.
            (b) The design certification rule must specify the site parameters, design characteristics, and any additional requirements and restrictions of the design certification rule.
            (c) After the Commission has adopted a final design certification rule, the applicant shall not permit any individual to have access to or any facility to possess restricted data or classified National Security Information until the individual and/or facility has been approved for access under the provisions of 10 CFR parts 25 and/or 95, as applicable.

            § 52.55
            Duration of certification.
            (a) Except as provided in paragraph (b) of this section, a standard design certification issued under this subpart is valid for 15 years from the date of issuance.
            (b) A standard design certification continues to be valid beyond the date of expiration in any proceeding on an application for a combined license or an operating license that references the standard design certification and is docketed either before the date of expiration of the certification, or, if a timely application for renewal of the certification has been filed, before the Commission has determined whether to renew the certification. A design certification also continues to be valid beyond the date of expiration in any hearing held under § 52.103 before operation begins under a combined license that references the design certification.
            (c) An applicant for a construction permit or a combined license may, at its own risk, reference in its application a design for which a design certification application has been docketed but not granted.

            § 52.57
            Application for renewal.
            (a) Not less than 12 nor more than 36 months before the expiration of the initial 15-year period, or any later renewal period, any person may apply for renewal of the certification. An application for renewal must contain all information necessary to bring up to date the information and data contained in the previous application. The Commission will require, before renewal of certification, that information normally contained in certain procurement specifications and construction and installation specifications be completed and available for audit if this information is necessary for the Commission to make its safety determination. Notice and comment procedures must be used for a rulemaking proceeding on the application for renewal. The Commission, in its discretion, may require the use of additional procedures in individual renewal proceedings.
            (b) A design certification, either original or renewed, for which a timely application for renewal has been filed remains in effect until the Commission has determined whether to renew the certification. If the certification is not renewed, it continues to be valid in certain proceedings, in accordance with the provisions of § 52.55.
            (c) The Commission shall refer a copy of the application for renewal to the Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall report on those portions of the application which concern safety and shall apply the criteria set forth in § 52.59.

            § 52.59
            Criteria for renewal.
            (a) The Commission shall issue a rule granting the renewal if the design, either as originally certified or as modified during the rulemaking on the renewal, complies with the Atomic Energy Act and the Commission's regulations applicable and in effect at the time the certification was issued, provided, however, that the first time the Commission issues a rule granting the renewal for a standard design certification in effect on July 13, 2009, the Commission shall, in addition, find that the renewed design complies with the applicable requirements of 10 CFR 50.150.
            (b) The Commission may impose other requirements if it determines that:
            (1) They are necessary for adequate protection to public health and safety or common defense and security;
            (2) They are necessary for compliance with the Commission's regulations and orders applicable and in effect at the time the design certification was issued; or
            (3) There is a substantial increase in overall protection of the public health and safety or the common defense and security to be derived from the new requirements, and the direct and indirect costs of implementing those requirements are justified in view of this increased protection.
            (c) In addition, the applicant for renewal may request an amendment to the design certification. The Commission shall grant the amendment request if it determines that the amendment will comply with the Atomic Energy Act and the Commission's regulations in effect at the time of renewal. If the amendment request entails such an extensive change to the design certification that an essentially new standard design is being proposed, an application for a design certification must be filed in accordance with this subpart.
            (d) Denial of renewal does not bar the applicant, or another applicant, from filing a new application for certification of the design, which proposes design changes that correct the deficiencies cited in the denial of the renewal.
            [72 FR 49517, Aug. 28, 2007, as amended at 74 FR 28147, June 12, 2009]

            § 52.61
            Duration of renewal.
            Each renewal of certification for a standard design will be for not less than 10, nor more than 15 years.

            § 52.63
            Finality of standard design certifications.

            (a)(1) Notwithstanding any provision in 10 CFR 50.109, while a standard design certification rule is in effect under §§ 52.55 or 52.61, the Commission may not modify, rescind, or impose new requirements on the certification information, whether on its own motion, or in response to a petition from any person, unless the Commission determines in a rulemaking that the change:
            (i) Is necessary either to bring the certification information or the referencing plants into compliance with the Commission's regulations applicable and in effect at the time the certification was issued;
            (ii) Is necessary to provide adequate protection of the public health and safety or the common defense and security;
            (iii) Reduces unnecessary regulatory burden and maintains protection to public health and safety and the common defense and security;

            (iv) Provides the detailed design information to be verified under those inspections, tests, analyses, and acceptance criteria (ITAAC) which are directed at certification information (i.e., design acceptance criteria);
            (v) Is necessary to correct material errors in the certification information;
            (vi) Substantially increases overall safety, reliability, or security of facility design, construction, or operation, and the direct and indirect costs of implementation of the rule change are justified in view of this increased safety, reliability, or security; or
            (vii) Contributes to increased standardization of the certification information.
            (2)(i) In a rulemaking under § 52.63(a)(1), except for § 52.63(a)(1)(ii), the Commission will give consideration to whether the benefits justify the costs for plants that are already licensed or for which an application for a permit or license is under consideration.
            (ii) The rulemaking procedures for changes under § 52.63(a)(1) must provide for notice and opportunity for public comment.
            (3) Any modification the NRC imposes on a design certification rule under paragraph (a)(1) of this section will be applied to all plants referencing the certified design, except those to which the modification has been rendered technically irrelevant by action taken under paragraphs (a)(4) or (b)(1) of this section.
            (4) The Commission may not impose new requirements by plant-specific order on any part of the design of a specific plant referencing the design certification rule if that part was approved in the design certification while a design certification rule is in effect under § 52.55 or § 52.61, unless:
            (i) A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time the certification was issued, or to assure adequate protection of the public health and safety or the common defense and security; and
            (ii) Special circumstances as defined in 10 CFR 52.7 are present. In addition to the factors listed in § 52.7, the Commission shall consider whether the special circumstances which § 52.7 requires to be present outweigh any decrease in safety that may result from the reduction in standardization caused by the plant-specific order.
            (5) Except as provided in 10 CFR 2.335, in making the findings required for issuance of a combined license, construction permit, operating license, or manufacturing license, or for any hearing under § 52.103, the Commission shall treat as resolved those matters resolved in connection with the issuance or renewal of a design certification rule.
            (b)(1) An applicant or licensee who references a design certification rule may request an exemption from one or more elements of the certification information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of § 52.7. In addition to the factors listed in § 52.7, the Commission shall consider whether the special circumstances that § 52.7 requires to be present outweigh any decrease in safety that may result from the reduction in standardization caused by the exemption. The granting of an exemption on request of an applicant is subject to litigation in the same manner as other issues in the operating license or combined license hearing.

            (2) Subject to § 50.59 of this chapter, a licensee who references a design certification rule may make departures from the design of the nuclear power facility, without prior Commission approval, unless the proposed departure involves a change to the design as described in the rule certifying the design. The licensee shall maintain records of all departures from the facility and these records must be maintained and available for audit until the date of termination of the license.
            (c) The Commission will require, before granting a construction permit, combined license, operating license, or manufacturing license which references a design certification rule, that information normally contained in certain procurement specifications and construction and installation specifications be completed and available for audit if the information is necessary for the Commission to make its safety determinations, including the determination that the application is consistent with the certification information. This information may be acquired by appropriate arrangements with the design certification applicant.

          Subpart C—Combined Licenses

            § 52.71
            Scope of subpart.
            This subpart sets out the requirements and procedures applicable to Commission issuance of combined licenses for nuclear power facilities.

            § 52.73
            Relationship to other subparts.
            (a) An application for a combined license under this subpart may, but need not, reference a standard design certification, standard design approval, or manufacturing license issued under subparts B, E, or F of this part, respectively, or an early site permit issued under subpart A of this part. In the absence of a demonstration that an entity other than the one originally sponsoring and obtaining a design certification is qualified to supply a design, the Commission will entertain an application for a combined license that references a standard design certification issued under subpart B of this part only if the entity that sponsored and obtained the certification supplies the design for the applicant's use.
            (b) The Commission will require, before granting a combined license that references a standard design certification, that information normally contained in certain procurement specifications and construction and installation specifications be completed and available for audit if the information is necessary for the Commission to make its safety determinations, including the determination that the application is consistent with the certification information.

            § 52.75
            Filing of applications.
            (a) Any person except one excluded by § 50.38 of this chapter may file an application for a combined license for a nuclear power facility with the Director, Office of Nuclear Reactor Regulation.
            (b) The application must comply with the applicable filing requirements of §§ 52.3 and 50.30 of this chapter.
            (c) The fees associated with the filing and review of the application are set forth in 10 CFR part 170.
            [72 FR 49517, Aug. 28, 2007, as amended at 73 FR 5724, Jan. 31, 2008; 84 FR 65645, Nov. 29, 2019]

            § 52.77
            Contents of applications; general information.
            The application must contain all of the information required by 10 CFR 50.33.

            § 52.79
            Contents of applications; technical information in final safety analysis report.
            (a) The application must contain a final safety analysis report that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components of the facility as a whole. The final safety analysis report shall include the following information, at a level of information sufficient to enable the Commission to reach a final conclusion on all safety matters that must be resolved by the Commission before issuance of a combined license:
            (1)(i) The boundaries of the site;
            (ii) The proposed general location of each facility on the site;

            (iii) The seismic, meteorological, hydrologic, and geologic characteristics of the proposed site with appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area and with sufficient margin for the limited accuracy, quantity, and time in which the historical data have been accumulated;

            (iv) The location and description of any nearby industrial, military, or transportation facilities and routes;
            (v) The existing and projected future population profile of the area surrounding the site;
            (vi) A description and safety assessment of the site on which the facility is to be located. The assessment must contain an analysis and evaluation of the major structures, systems, and components of the facility that bear significantly on the acceptability of the site under the radiological consequence evaluation factors identified in paragraphs (a)(1)(vi)(A) and (a)(1)(vi)(B) of this section. In performing this assessment, an applicant shall assume a fission product release 5
               from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable site characteristics, including site meteorology, to evaluate the offsite radiological consequences. Site characteristics must comply with part 100 of this chapter. The evaluation must determine that:

                5 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. These accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

            (A) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem 6
               total effective dose equivalent (TEDE).

                6 A whole body dose of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations at the time could be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, its use is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, to assure that these designs provide assurance of low risk of public exposure to radiation, in the event of an accident.

            (B) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE; and
            (2) A description and analysis of the structures, systems, and components of the facility with emphasis upon performance requirements, the bases, with technical justification therefor, upon which these requirements have been established, and the evaluations required to show that safety functions will be accomplished. It is expected that reactors will reflect through their design, construction, and operation an extremely low probability for accidents that could result in the release of significant quantities of radioactive fission products. The descriptions shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems shall be discussed insofar as they are pertinent. The following power reactor design characteristics and proposed operation will be taken into consideration by the Commission:
            (i) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials;

            (ii) The extent to which generally accepted engineering standards are applied to the design of the reactor;

            (iii) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials;
            (iv) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. Special attention must be directed to plant design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release 7
               from the core into the containment assuming that the facility is operated at the ultimate power level contemplated;

                7 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. These accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

            (3) The kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in part 20 of this chapter;
            (4) The design of the facility including:
            (i) The principal design criteria for the facility. Appendix A to part 50 of this chapter, “General Design Criteria for Nuclear Power Plants,” establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units;
            (ii) The design bases and the relation of the design bases to the principal design criteria;
            (iii) Information relative to materials of construction, arrangement, and dimensions, sufficient to provide reasonable assurance that the design will conform to the design bases with adequate margin for safety.
            (5) An analysis and evaluation of the design and performance of structures, systems, and components with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. Analysis and evaluation of ECCS cooling performance and the need for high-point vents following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of §§ 50.46 and 50.46a of this chapter;
            (6) A description and analysis of the fire protection design features for the reactor necessary to comply with 10 CFR part 50, appendix A, GDC 3, and § 50.48 of this chapter;
            (7) A description of protection provided against pressurized thermal shock events, including projected values of the reference temperature for reactor vessel beltline materials as defined in §§ 50.60 and 50.61(b)(1) and (b)(2) of this chapter;
            (8) An analysis and description of the equipment and systems for combustible gas control as required by § 50.44 of this chapter;
            (9) The coping analyses, and any design features necessary to address station blackout, as described in § 50.63 of this chapter;
            (10) A description of the program, and its implementation, required by § 50.49(a) of this chapter for the environmental qualification of electric equipment important to safety and the list of electric equipment important to safety that is required by 10 CFR 50.49(d);

            (11) A description of the program(s), and their implementation, necessary to ensure that the systems and components meet the requirements of the ASME Boiler and Pressure Vessel Code and the ASME Code for Operation and Maintenance of Nuclear Power Plants in accordance with 50.55a of this chapter;

            (12) A description of the primary containment leakage rate testing program, and its implementation, necessary to ensure that the containment meets the requirements of appendix J to 10 CFR part 50;
            (13) A description of the reactor vessel material surveillance program required by appendix H to 10 CFR part 50 and its implementation;
            (14) A description of the operator training program, and its implementation, necessary to meet the requirements of 10 CFR part 55;
            (15) A description of the program, and its implementation, for monitoring the effectiveness of maintenance necessary to meet the requirements of § 50.65 of this chapter;
            (16)(i) The information with respect to the design of equipment to maintain control over radioactive materials in gaseous and liquid effluents produced during normal reactor operations, as described in § 50.34a(d) of this chapter;
            (ii) A description of the process and effluent monitoring and sampling program required by appendix I to 10 CFR part 50 and its implementation.
            (17) The information with respect to compliance with technically relevant positions of the Three Mile Island requirements in § 50.34(f) of this chapter, with the exception of § 50.34(f)(1)(xii), (f)(2)(ix), (f)(2)(xxv), and (f)(3)(v);
            (18) If the applicant seeks to use risk-informed treatment of SSCs in accordance with § 50.69 of this chapter, the information required by § 50.69(b)(2) of this chapter;
            (19) Information necessary to demonstrate that the plant complies with the earthquake engineering criteria in 10 CFR part 50, appendix S;
            (20) Proposed technical resolutions of those Unresolved Safety Issues and medium- and high-priority generic safety issues which are identified in the version of NUREG-0933 current on the date up to 6 months before the docket date of the application and which are technically relevant to the design;
            (21) Emergency plans complying with the requirements of § 50.47 of this chapter, and 10 CFR part 50, appendix E;
            (22)(i) All emergency plan certifications that have been obtained from the State and local governmental agencies with emergency planning responsibilities must state that:
            (A) The proposed emergency plans are practicable;
            (B) These agencies are committed to participating in any further development of the plans, including any required field demonstrations; and
            (C) These agencies are committed to executing their responsibilities under the plans in the event of an emergency;
            (ii) If certifications cannot be obtained after sustained, good faith efforts by the applicant, then the application must contain information, including a utility plan, sufficient to show that the proposed plans provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency at the site.
            (23) [Reserved]
            (24) If the application is for a nuclear power reactor design which differs significantly from light-water reactor designs that were licensed before 1997 or use simplified, inherent, passive, or other innovative means to accomplish their safety functions, the application must describe how the design meets the requirements in § 50.43(e) of this chapter;
            (25) A description of the quality assurance program, applied to the design, and to be applied to the fabrication, construction, and testing, of the structures, systems, and components of the facility. Appendix B to 10 CFR part 50 sets forth the requirements for quality assurance programs for nuclear power plants. The description of the quality assurance program for a nuclear power plant must include a discussion of how the applicable requirements of appendix B to 10 CFR part 50 have been and will be satisfied, including a discussion of how the quality assurance program will be implemented;
            (26) The applicant's organizational structure, allocations or responsibilities and authorities, and personnel qualifications requirements for operation;

            (27) Managerial and administrative controls to be used to assure safe operation. Appendix B to 10 CFR part 50 sets forth the requirements for these controls for nuclear power plants. The information on the controls to be used for a nuclear power plant shall include a discussion of how the applicable requirements of appendix B to 10 CFR part 50 will be satisfied;
            (28) Plans for preoperational testing and initial operations;
            (29)(i) Plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of structures, systems, and components;
            (ii) Plans for coping with emergencies, other than the plans required by § 52.79(a)(21);
            (30) Proposed technical specifications prepared in accordance with the requirements of §§ 50.36 and 50.36a of this chapter;
            (31) For nuclear power plants to be operated on multi-unit sites, an evaluation of the potential hazards to the structures, systems, and components important to safety of operating units resulting from construction activities, as well as a description of the managerial and administrative controls to be used to provide assurance that the limiting conditions for operation are not exceeded as a result of construction activities at the multi-unit sites;
            (32) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter;
            (33) A description of the training program required by § 50.120 of this chapter and its implementation;
            (34) A description and plans for implementation of an operator requalification program. The operator requalification program must as a minimum, meet the requirements for those programs contained in § 55.59 of this chapter;
            (35)(i) A physical security plan, describing how the applicant will meet the requirements of 10 CFR part 73 (and 10 CFR part 11, if applicable, including the identification and description of jobs as required by § 11.11(a) of this chapter, at the proposed facility). The plan must list tests, inspections, audits, and other means to be used to demonstrate compliance with the requirements of 10 CFR parts 11 and 73, if applicable;
            (ii) A description of the implementation of the physical security plan;
            (36)(i) A safeguards contingency plan in accordance with the criteria set forth in appendix C to 10 CFR part 73. The safeguards contingency plan shall include plans for dealing with threats, thefts, and radiological sabotage, as defined in part 73 of this chapter, relating to the special nuclear material and nuclear facilities licensed under this chapter and in the applicant's possession and control. Each application for this type of license shall include the information contained in the applicant's safeguards contingency plan. 8
               (Implementing procedures required for this plan need not be submitted for approval.)

                8 A physical security plan that contains all the information required in both § 73.55 of this chapter and appendix C to 10 CFR part 73 satisfies the requirement for a contingency plan.

            (ii) A training and qualification plan in accordance with the criteria set forth in appendix B to 10 CFR part 73.
            (iii) A cyber security plan in accordance with the criteria set forth in § 73.54 of this chapter;
            (iv) A description of the implementation of the safeguards contingency plan, training and qualification plan, and cyber security plan; and
            (v) Each applicant who prepares a physical security plan, a safeguards contingency plan, a training and qualification plan, or a cyber security plan, shall protect the plans and other related Safeguards Information against unauthorized disclosure in accordance with the requirements of § 73.21 of this chapter.
            (37) The information necessary to demonstrate how operating experience insights have been incorporated into the plant design;
            (38) For light-water reactor designs, a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass;
            (39) A description of the radiation protection program required by § 20.1101 of this chapter and its implementation.

            (40) A description of the fire protection program required by § 50.48 of this chapter and its implementation.

            (41) For applications for light-water-cooled nuclear power plant combined licenses, an evaluation of the facility against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. The evaluation required by this section shall include an identification and description of all differences in design features, analytical techniques, and procedural measures proposed for a facility and those corresponding features, techniques, and measures given in the SRP acceptance criteria. Where a difference exists, the evaluation shall discuss how the proposed alternative provides an acceptable method of complying with the Commission's regulations, or portions thereof, that underlie the corresponding SRP acceptance criteria. The SRP is not a substitute for the regulations, and compliance is not a requirement;
            (42) Information demonstrating how the applicant will comply with requirements for reduction of risk from anticipated transients without scram (ATWS) events in § 50.62 of this chapter;
            (43) Information demonstrating how the applicant will comply with requirements for criticality accidents in § 50.68 of this chapter;
            (44) A description of the fitness-for-duty program required by 10 CFR part 26 and its implementation.
            (45) The information required by § 20.1406 of this chapter.
            (46) A description of the plant-specific probabilistic risk assessment (PRA) and its results.
            (47) For applications for combined licenses which are subject to 10 CFR 50.150(a), the information required by 10 CFR 50.150(b).
            (b) If the combined license application references an early site permit, then the following requirements apply:

            (1) The final safety analysis report need not contain information or analyses submitted to the Commission in connection with the early site permit, provided, however, that the final safety analysis report must either include or incorporate by reference the early site permit site safety analysis report and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the design of the facility falls within the site characteristics and design parameters specified in the early site permit.
            (2) If the final safety analysis report does not demonstrate that design of the facility falls within the site characteristics and design parameters, the application shall include a request for a variance that complies with the requirements of §§ 52.39 and 52.93.
            (3) The final safety analysis report must demonstrate that all terms and conditions that have been included in the early site permit, other than those imposed under § 50.36b, will be satisfied by the date of issuance of the combined license. Any terms or conditions of the early site permit that could not be met by the time of issuance of the combined license, must be set forth as terms or conditions of the combined license.
            (4) If the early site permit approves complete and integrated emergency plans, or major features of emergency plans, then the final safety analysis report must include any new or additional information that updates and corrects the information that was provided under § 52.17(b), and discuss whether the new or additional information materially changes the bases for compliance with the applicable requirements. The application must identify changes to the emergency plans or major features of emergency plans that have been incorporated into the proposed facility emergency plans and that constitute or would constitute a reduction in effectiveness under § 50.54(q) of this chapter.
            (5) If complete and integrated emergency plans are approved as part of the early site permit, new certifications meeting the requirements of paragraph (a)(22) of this section are not required.
            (c) If the combined license application references a standard design approval, then the following requirements apply:

            (1) The final safety analysis report need not contain information or analyses submitted to the Commission in connection with the design approval, provided, however, that the final safety analysis report must either include or incorporate by reference the standard design approval final safety analysis report and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the characteristics of the site fall within the site parameters specified in the design approval. In addition, the plant-specific PRA information must use the PRA information for the design approval and must be updated to account for site-specific design information and any design changes or departures.
            (2) The final safety analysis report must demonstrate that all terms and conditions that have been included in the design approval will be satisfied by the date of issuance of the combined license.
            (d) If the combined license application references a standard design certification, then the following requirements apply:

            (1) The final safety analysis report need not contain information or analyses submitted to the Commission in connection with the design certification, provided, however, that the final safety analysis report must either include or incorporate by reference the standard design certification final safety analysis report and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the site characteristics fall within the site parameters specified in the design certification. In addition, the plant-specific PRA information must use the PRA information for the design certification and must be updated to account for site-specific design information and any design changes or departures.
            (2) The final safety analysis report must demonstrate that the interface requirements established for the design under § 52.47 have been met.
            (3) The final safety analysis report must demonstrate that all requirements and restrictions set forth in the referenced design certification rule, other than those imposed under § 50.36b, must be satisfied by the date of issuance of the combined license. Any requirements and restrictions set forth in the referenced design certification rule that could not be satisfied by the time of issuance of the combined license, must be set forth as terms or conditions of the combined license.
            (e) If the combined license application references the use of one or more manufactured nuclear power reactors licensed under subpart F of this part, then the following requirements apply:

            (1) The final safety analysis report need not contain information or analyses submitted to the Commission in connection with the manufacturing license, provided, however, that the final safety analysis report must either include or incorporate by reference the manufacturing license final safety analysis report and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the site characteristics fall within the site parameters specified in the manufacturing license. In addition, the plant-specific PRA information must use the PRA information for the manufactured reactor and must be updated to account for site-specific design information and any design changes or departures.
            (2) The final safety analysis report must demonstrate that the interface requirements established for the design have been met.
            (3) The final safety analysis report must demonstrate that all terms and conditions that have been included in the manufacturing license, other than those imposed under § 50.36b, will be satisfied by the date of issuance of the combined license. Any terms or conditions of the manufacturing license that could not be met by the time of issuance of the combined license, must be set forth as terms or conditions of the combined license.
            (f) Each applicant for a combined license under this subpart shall protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable.
            [72 FR 49517, Aug. 28, 2007, as amended at 73 FR 63571, Oct. 24, 2008; 74 FR 13970, Mar. 27, 2009; 74 FR 28147, June 12, 2009; 76 FR 72600, Nov. 23, 2011; 78 FR 34249, June 7, 2013; 84 FR 63568, Nov. 18, 2019]

            § 52.80
            Contents of applications; additional technical information.
            The application must contain:

            (a) The proposed inspections, tests, and analyses, including those applicable to emergency planning, that the licensee shall perform, and the acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, the facility has been constructed and will be operated in conformity with the combined license, the provisions of the Act, and the Commission's rules and regulations.
            (1) If the application references an early site permit with ITAAC, the early site permit ITAAC must apply to those aspects of the combined license which are approved in the early site permit.
            (2) If the application references a standard design certification, the ITAAC contained in the certified design must apply to those portions of the facility design which are approved in the design certification.

            (3) If the application references an early site permit with ITAAC or a standard design certification or both, the application may include a notification that a required inspection, test, or analysis in the ITAAC has been successfully completed and that the corresponding acceptance criterion has been met. The Federal Register notification required by § 52.85 must indicate that the application includes this notification.
            (b) An environmental report, either in accordance with 10 CFR 51.50(c) if a limited work authorization under 10 CFR 50.10 is not requested in conjunction with the combined license application, or in accordance with §§ 51.49 and 51.50(c) of this chapter if a limited work authorization is requested in conjunction with the combined license application.
            (c) If the applicant wishes to request that a limited work authorization under 10 CFR 50.10 be issued before issuance of the combined license, the application must include the information otherwise required by 10 CFR 50.10, in accordance with either 10 CFR 2.101(a)(1) through (a)(4), or 10 CFR 2.101(a)(9).
            (d) The applicant's plans for implementing the requirements of § 50.155 of this chapter including a schedule for achieving full compliance with these requirements, and a description of the equipment upon which the strategies and guidelines required by § 50.155(b)(1) of this chapter rely, including the planned locations of the equipment and how the equipment meets the requirements of § 50.155(c) of this chapter.
            [72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57447, Oct. 9, 2007; 74 FR 13970, Mar. 27, 2009; 84 FR 39719, Aug. 8, 2019]

            § 52.81
            Standards for review of applications.
            Applications filed under this subpart will be reviewed according to the standards set out in 10 CFR parts 20, 50, 51, 54, 55, 73, 100, and 140.

            § 52.83
            Finality of referenced NRC approvals; partial initial decision on site suitability.
            (a) If the application for a combined license under this subpart references an early site permit, design certification rule, standard design approval, or manufacturing license, the scope and nature of matters resolved for the application and any combined license issued are governed by the relevant provisions addressing finality, including §§ 52.39, 52.63, 52.98, 52.145, and 52.171.
            (b) While a partial decision on site suitability is in effect under 10 CFR 2.627(b)(2), the scope and nature of matters resolved in the proceeding are governed by the finality provisions in 10 CFR 2.629.
            [72 FR 49517, Aug. 28, 2007, as amended at 84 FR 63568, Nov. 18, 2019]

            § 52.85
            Administrative review of applications; hearings.
            A proceeding on a combined license is subject to all applicable procedural requirements contained in 10 CFR part 2, including the requirements for docketing (§ 2.101 of this chapter) and issuance of a notice of hearing (§ 2.104 of this chapter). If an applicant requests a Commission finding on certain ITAAC with the issuance of the combined license, then those ITAAC will be identified in the notice of hearing. All hearings on combined licenses are governed by the procedures contained in 10 CFR part 2.

            § 52.87
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            The Commission shall refer a copy of the application to the ACRS. The ACRS shall report on those portions of the application that concern safety and shall apply the standards referenced in § 52.81, in accordance with the finality provisions in § 52.83.

            § 52.89
            [Reserved]

            § 52.91
            Authorization to conduct limited work authorization activities.
            (a) If the application does not reference an early site permit which authorizes the holder to perform the activities under 10 CFR 50.10(d), the applicant may not perform those activities without obtaining the separate authorization required by 10 CFR 50.10(d). Authorization may be granted only after the presiding officer in the proceeding on the application has made the findings and determination required by 10 CFR 50.10(e), and the Director of the Office of Nuclear Reactor Regulation makes the determination required by 10 CFR 50.10(e).
            (b) If, after an applicant has performed the activities permitted by paragraph (a) of this section, the application for the combined license is withdrawn or denied, then the applicant shall implement the approved site redress plan.
            [72 FR 57447, Oct. 9, 2007, as amended at 84 FR 65645, Nov. 29, 2019]

            § 52.93
            Exemptions and variances.
            (a) Applicants for a combined license under this subpart, or any amendment to a combined license, may include in the application a request for an exemption from one or more of the Commission's regulations.
            (1) If the request is for an exemption from any part of a referenced design certification rule, the Commission may grant the request if it determines that the exemption complies with any exemption provisions of the referenced design certification rule, or with § 52.63 if there are no applicable exemption provisions in the referenced design certification rule.
            (2) For all other requests for exemptions, the Commission may grant a request if it determines that the exemption complies with § 52.7.
            (b) An applicant for a combined license who has filed an application referencing an early site permit issued under subpart A of this part may include in the application a request for a variance from one or more site characteristics, design parameters, or terms and conditions of the permit, or from the site safety analysis report. In determining whether to grant the variance, the Commission shall apply the same technically relevant criteria as were applicable to the application for the original or renewed site permit. Once a construction permit or combined license referencing an early site permit is issued, variances from the early site permit will not be granted for that construction permit or combined license.
            (c) An applicant for a combined license who has filed an application referencing a nuclear power reactor manufactured under a manufacturing license issued under subpart F of this part may include in the application a request for a departure from one or more design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor. The Commission may grant a request only if it determines that the departure will comply with the requirements of 10 CFR 52.7, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure.
            (d) Issuance of a variance under paragraph (b) or a departure under paragraph (c) of this section is subject to litigation during the combined license proceeding in the same manner as other issues material to that proceeding.

            § 52.97
            Issuance of combined licenses.
            (a)(1) After conducting a hearing in accordance with § 52.85 and receiving the report submitted by the ACRS, the Commission may issue a combined license if the Commission finds that:

            (i) The applicable standards and requirements of the Act and the Commission's regulations have been met;

            (ii) Any required notifications to other agencies or bodies have been duly made;
            (iii) There is reasonable assurance that the facility will be constructed and will operate in conformity with the license, the provisions of the Act, and the Commission's regulations.
            (iv) The applicant is technically and financially qualified to engage in the activities authorized; and
            (v) Issuance of the license will not be inimical to the common defense and security or to the health and safety of the public; and
            (vi) The findings required by subpart A of part 51 of this chapter have been made.
            (2) The Commission may also find, at the time it issues the combined license, that certain acceptance criteria in one or more of the inspections, tests, analyses, and acceptance criteria (ITAAC) in a referenced early site permit or standard design certification have been met. This finding will finally resolve that those acceptance criteria have been met, those acceptance criteria will be deemed to be excluded from the combined license, and findings under § 52.103(g) with respect to those acceptance criteria are unnecessary.
            (b) The Commission shall identify within the combined license the inspections, tests, and analyses, including those applicable to emergency planning, that the licensee shall perform, and the acceptance criteria that, if met, are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's rules and regulations.
            (c) A combined license shall contain the terms and conditions, including technical specifications, as the Commission deems necessary and appropriate.

            § 52.98
            Finality of combined licenses; information requests.
            (a) After issuance of a combined license, the Commission may not modify, add, or delete any term or condition of the combined license, the design of the facility, the inspections, tests, analyses, and acceptance criteria contained in the license which are not derived from a referenced standard design certification or manufacturing license, except in accordance with the provisions of § 52.103 or § 50.109 of this chapter, as applicable.
            (b) If the combined license does not reference a design certification or a reactor manufactured under a subpart F of this part manufacturing license, then a licensee may make changes in the facility as described in the final safety analysis report (as updated), make changes in the procedures as described in the final safety analysis report (as updated), and conduct tests or experiments not described in the final safety analysis report (as updated) under the applicable change processes in 10 CFR part 50 (e.g., §§ 50.54, 50.59, or 50.90 of this chapter).
            (c) If the combined license references a certified design, then—
            (1) Changes to or departures from information within the scope of the referenced design certification rule are subject to the applicable change processes in that rule; and
            (2) Changes that are not within the scope of the referenced design certification rule are subject to the applicable change processes in 10 CFR part 50, unless they also involve changes to or noncompliance with information within the scope of the referenced design certification rule. In these cases, the applicable provisions of this section and the design certification rule apply.
            (d) If the combined license references a reactor manufactured under a subpart F of this part manufacturing license, then—
            (1) Changes to or departures from information within the scope of the manufactured reactor's design are subject to the change processes in § 52.171; and
            (2) Changes that are not within the scope of the manufactured reactor's design are subject to the applicable change processes in 10 CFR part 50.

            (e) The Commission may issue and make immediately effective any amendment to a combined license upon a determination by the Commission that the amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. The amendment may be issued and made immediately effective in advance of the holding and completion of any required hearing. The amendment will be processed in accordance with the procedures specified in 10 CFR 50.91.
            (f) Any modification to, addition to, or deletion from the terms and conditions of a combined license, including any modification to, addition to, or deletion from the inspections, tests, analyses, or related acceptance criteria contained in the license is a proposed amendment to the license. There must be an opportunity for a hearing on the amendment.
            (g) Except for information sought to verify licensee compliance with the current licensing basis for that facility, information requests to the holder of a combined license must be evaluated before issuance to ensure that the burden to be imposed on the licensee is justified in view of the potential safety significance of the issue to be addressed in the requested information. Each evaluation performed by the NRC staff must be in accordance with 10 CFR 50.54(f) and must be approved by the Executive Director for Operations or his or her designee before issuance of the request.

            § 52.99
            Inspection during construction; ITAAC schedules and notifications; NRC notices.
            (a) Licensee schedule for completing inspections, tests, or analyses. The licensee shall submit to the NRC, no later than 1 year after issuance of the combined license or at the start of construction as defined at 10 CFR 50.10(a), whichever is later, its schedule for completing the inspections, tests, or analyses in the ITAAC. The licensee shall submit updates to the ITAAC schedules every 6 months thereafter and, within 1 year of its scheduled date for initial loading of fuel, the licensee shall submit updates to the ITAAC schedule every 30 days until the final notification is provided to the NRC under paragraph (c)(1) of this section.
            (b) Licensee and applicant conduct of activities subject to ITAAC. With respect to activities subject to an ITAAC, an applicant for a combined license may proceed at its own risk with design and procurement activities, and a licensee may proceed at its own risk with design, procurement, construction, and preoperational activities, even though the NRC may not have found that any one of the prescribed acceptance criteria are met.
            (c) Licensee notifications—(1) ITAAC closure notification. The licensee shall notify the NRC that prescribed inspections, tests, and analyses have been performed and that the prescribed acceptance criteria are met. The notification must contain sufficient information to demonstrate that the prescribed inspections, tests, and analyses have been performed and that the prescribed acceptance criteria are met.
            (2) ITAAC post-closure notifications. Following the licensee's ITAAC closure notifications under paragraph (c)(1) of this section until the Commission makes the finding under 10 CFR 52.103(g), the licensee shall notify the NRC, in a timely manner, of new information that materially alters the basis for determining that either inspections, tests, or analyses were performed as required, or that acceptance criteria are met. The notification must contain sufficient information to demonstrate that, notwithstanding the new information, the prescribed inspections, tests, or analyses have been performed as required, and the prescribed acceptance criteria are met.
            (3) Uncompleted ITAAC notification. If the licensee has not provided, by the date 225 days before the scheduled date for initial loading of fuel, the notification required by paragraph (c)(1) of this section for all ITAAC, then the licensee shall notify the NRC that the prescribed inspections, tests, or analyses for all uncompleted ITAAC will be performed and that the prescribed acceptance criteria will be met prior to operation. The notification must be provided no later than the date 225 days before the scheduled date for initial loading of fuel, and must provide sufficient information to demonstrate that the prescribed inspections, tests, or analyses will be performed and the prescribed acceptance criteria for the uncompleted ITAAC will be met, including, but not limited to, a description of the specific procedures and analytical methods to be used for performing the prescribed inspections, tests, and analyses and determining that the prescribed acceptance criteria are met.
            (4) All ITAAC complete notification. The licensee shall notify the NRC that all ITAAC are complete.
            (d) Licensee determination of non-compliance with ITAAC. (1) In the event that an activity is subject to an ITAAC derived from a referenced standard design certification and the licensee has not demonstrated that the prescribed acceptance criteria are met, the licensee may take corrective actions to successfully complete that ITAAC or request an exemption from the standard design certification ITAAC, as applicable. A request for an exemption must also be accompanied by a request for a license amendment under 10 CFR 52.98(f).
            (2) In the event that an activity is subject to an ITAAC not derived from a referenced standard design certification and the licensee has not demonstrated that the prescribed acceptance criteria are met, the licensee may take corrective actions to successfully complete that ITAAC or request a license amendment under 10 CFR 52.98(f).
            (e) NRC inspection, publication of notices, and availability of licensee notifications. The NRC shall ensure that the prescribed inspections, tests, and analyses in the ITAAC are performed.

            (1) At appropriate intervals until the last date for submission of requests for hearing under 10 CFR 52.103(a), the NRC shall publish notices in the Federal Register of the NRC staff's determination of the successful completion of inspections, tests, and analyses.
            (2) The NRC shall make publicly available the licensee notifications under paragraph (c) of this section. The NRC shall, no later than the date of publication of the notice of intended operation required by 10 CFR 52.103(a), make publicly available those licensee notifications under paragraph (c) of this section that have been submitted to the NRC at least seven (7) days before that notice.
            [77 FR 51892, Aug. 28, 2012]

            § 52.103
            Operation under a combined license.

            (a) The licensee shall notify the NRC of its scheduled date for initial loading of fuel no later than 270 days before the scheduled date and shall notify the NRC of updates to its schedule every 30 days thereafter. Not less than 180 days before the date scheduled for initial loading of fuel into a plant by a licensee that has been issued a combined license under this part, the Commission shall publish notice of intended operation in the Federal Register. The notice must provide that any person whose interest may be affected by operation of the plant may, within 60 days, request that the Commission hold a hearing on whether the facility as constructed complies, or on completion will comply, with the acceptance criteria in the combined license, except that a hearing shall not be granted for those ITAAC which the Commission found were met under § 52.97(a)(2).
            (b) A request for hearing under paragraph (a) of this section must show, prima facie, that—
            (1) One or more of the acceptance criteria of the ITAAC in the combined license have not been, or will not be, met; and
            (2) The specific operational consequences of nonconformance that would be contrary to providing reasonable assurance of adequate protection of the public health and safety.

            (c) The Commission, acting as the presiding officer, shall determine whether to grant or deny the request for hearing in accordance with the applicable requirements of 10 CFR 2.309. If the Commission grants the request, the Commission, acting as the presiding officer, shall determine whether during a period of interim operation there will be reasonable assurance of adequate protection to the public health and safety. The Commission's determination must consider the petitioner's prima facie showing and any answers thereto. If the Commission determines there is such reasonable assurance, it shall allow operation during an interim period under the combined license.

            (d) The Commission, in its discretion, shall determine appropriate hearing procedures, whether informal or formal adjudicatory, for any hearing under paragraph (a) of this section, and shall state its reasons therefore.

            (e) The Commission shall, to the maximum possible extent, render a decision on issues raised by the hearing request within 180 days of the publication of the notice provided by paragraph (a) of this section or by the anticipated date for initial loading of fuel into the reactor, whichever is later.
            (f) A petition to modify the terms and conditions of the combined license will be processed as a request for action in accordance with 10 CFR 2.206. The petitioner shall file the petition with the Secretary of the Commission. Before the licensed activity allegedly affected by the petition (fuel loading, low power testing, etc.) commences, the Commission shall determine whether any immediate action is required. If the petition is granted, then an appropriate order will be issued. Fuel loading and operation under the combined license will not be affected by the granting of the petition unless the order is made immediately effective.
            (g) The licensee shall not operate the facility until the Commission makes a finding that the acceptance criteria in the combined license are met, except for those acceptance criteria that the Commission found were met under § 52.97(a)(2). If the combined license is for a modular design, each reactor module may require a separate finding as construction proceeds.
            (h) After the Commission has made the finding in paragraph (g) of this section, the ITAAC do not, by virtue of their inclusion in the combined license, constitute regulatory requirements either for licensees or for renewal of the license; except for the specific ITAAC for which the Commission has granted a hearing under paragraph (a) of this section, all ITAAC expire upon final Commission action in the proceeding. However, subsequent changes to the facility or procedures described in the final safety analysis report (as updated) must comply with the requirements in §§ 52.98(e) or (f), as applicable.

            § 52.104
            Duration of combined license.
            A combined license is issued for a specified period not to exceed 40 years from the date on which the Commission makes a finding that acceptance criteria are met under § 52.103(g) or allowing operation during an interim period under the combined license under § 52.103(c).

            § 52.105
            Transfer of combined license.
            A combined license may be transferred in accordance with § 50.80 of this chapter.

            § 52.107
            Application for renewal.
            The filing of an application for a renewed license must be in accordance with 10 CFR part 54.

            § 52.109
            Continuation of combined license.
            Each combined license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated. During this period of continued effectiveness the licensee shall—
            (1) Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition; and
            (2) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC's regulations and the provisions of the combined license for the facility.

            § 52.110
            Termination of license.
            (a)(1) When a licensee has determined to permanently cease operations the licensee shall, within 30 days, submit a written certification to the NRC, consistent with the requirements of § 52.3(b)(8);
            (2) Once fuel has been permanently removed from the reactor vessel, the licensee shall submit a written certification to the NRC that meets the requirements of § 52.3(b)(9); and

            (3) For licensees whose licenses have been permanently modified to allow possession but not operation of the facility, before September 27, 2007, the certification required in paragraph (a)(1) of this section shall be deemed to have been submitted.

            (b) Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, or when a final legally effective order to permanently cease operations has come into effect, the 10 CFR part 52 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel.
            (c) Decommissioning will be completed within 60 years of permanent cessation of operations. Completion of decommissioning beyond 60 years will be approved by the Commission only when necessary to protect public health and safety. Factors that will be considered by the Commission in evaluating an alternative that provides for completion of decommissioning beyond 60 years of permanent cessation of operations include unavailability of waste disposal capacity and other site-specific factors affecting the licensee's capability to carry out decommissioning, including presence of other nuclear facilities at the site.
            (d)(1) Before or within 2 years following permanent cessation of operations, the licensee shall submit a post-shutdown decommissioning activities report (PSDAR) to the NRC, and a copy to the affected State(s). The report must include a description of the planned decommissioning activities along with a schedule for their accomplishment, an estimate of expected costs, and a discussion that provides the reasons for concluding that the environmental impacts associated with site-specific decommissioning activities will be bounded by appropriate previously issued environmental impact statements.

            (2) The NRC shall notice receipt of the PSDAR and make the PSDAR available for public comment. The NRC shall also schedule a public meeting in the vicinity of the licensee's facility upon receipt of the PSDAR. The NRC shall publish a document in the Federal Register and in a forum, such as local newspapers, that is readily accessible to individuals in the vicinity of the site, announcing the date, time and location of the meeting, along with a brief description of the purpose of the meeting.
            (e) Licensees shall not perform any major decommissioning activities, as defined in § 50.2 of this chapter, until 90 days after the NRC has received the licensee's PSDAR submittal and until certifications of permanent cessation of operations and permanent removal of fuel from the reactor vessel, as required under § 52.110(a)(1), have been submitted.
            (f) Licensees shall not perform any decommissioning activities, as defined in § 52.1, that—
            (1) Foreclose release of the site for possible unrestricted use;
            (2) Result in significant environmental impacts not previously reviewed; or
            (3) Result in there no longer being reasonable assurance that adequate funds will be available for decommissioning.
            (g) In taking actions permitted under § 50.59 of this chapter following submittal of the PSDAR, the licensee shall notify the NRC in writing and send a copy to the affected State(s), before performing any decommissioning activity inconsistent with, or making any significant schedule change from, those actions and schedules described in the PSDAR, including changes that significantly increase the decommissioning cost.
            (h)(1) Decommissioning trust funds may be used by licensees if—
            (i) The withdrawals are for expenses for legitimate decommissioning activities consistent with the definition of decommissioning in § 52.1;
            (ii) The expenditure would not reduce the value of the decommissioning trust below an amount necessary to place and maintain the reactor in a safe storage condition if unforeseen conditions or expenses arise and;
            (iii) The withdrawals would not inhibit the ability of the licensee to complete funding of any shortfalls in the decommissioning trust needed to ensure the availability of funds to ultimately release the site and terminate the license.

            (2) Initially, 3 percent of the generic amount specified in § 50.75 of this chapter may be used for decommissioning planning. For licensees that have submitted the certifications required under § 52.110(a) and commencing 90 days after the NRC has received the PSDAR, an additional 20 percent may be used. A site-specific decommissioning cost estimate must be submitted to the NRC before the licensee may use any funding in excess of these amounts.
            (3) Within 2 years following permanent cessation of operations, if not already submitted, the licensee shall submit a site-specific decommissioning cost estimate.
            (4) For decommissioning activities that delay completion of decommissioning by including a period of storage or surveillance, the licensee shall provide a means of adjusting cost estimates and associated funding levels over the storage or surveillance period.
            (i) All power reactor licensees must submit an application for termination of license. The application for termination of license must be accompanied or preceded by a license termination plan to be submitted for NRC approval.
            (1) The license termination plan must be a supplement to the FSAR or equivalent and must be submitted at least 2 years before termination of the license date.
            (2) The license termination plan must include—
            (i) A site characterization;
            (ii) Identification of remaining dismantlement activities;
            (iii) Plans for site remediation;
            (iv) Detailed plans for the final radiation survey;
            (v) A description of the end use of the site, if restricted;
            (vi) An updated site-specific estimate of remaining decommissioning costs;
            (vii) A supplement to the environmental report, under § 51.53 of this chapter, describing any new information or significant environmental change associated with the licensee's proposed termination activities; and
            (viii) Identification of parts, if any, of the facility or site that were released for use before approval of the license termination plan.

            (3) The NRC shall notice receipt of the license termination plan and make the license termination plan available for public comment. The NRC shall also schedule a public meeting in the vicinity of the licensee's facility upon receipt of the license termination plan. The NRC shall publish a document in the Federal Register and in a forum, such as local newspapers, which is readily accessible to individuals in the vicinity of the site, announcing the date, time and location of the meeting, along with a brief description of the purpose of the meeting.
            (j) If the license termination plan demonstrates that the remainder of decommissioning activities will be performed in accordance with the regulations in this chapter, will not be inimical to the common defense and security or to the health and safety of the public, and will not have a significant effect on the quality of the environment and after notice to interested persons, the Commission shall approve the plan, by license amendment, subject to terms and conditions as it deems appropriate and necessary and authorize implementation of the license termination plan.
            (k) The Commission shall terminate the license if it determines that—
            (1) The remaining dismantlement has been performed in accordance with the approved license termination plan; and
            (2) The final radiation survey and associated documentation, including an assessment of dose contributions associated with parts released for use before approval of the license termination plan, demonstrate that the facility and site have met the criteria for decommissioning in subpart E to 10 CFR part 20.
            (l) For a facility that has permanently ceased operation before the expiration of its license, the collection period for any shortfall of funds will be determined, upon application by the licensee, on a case-by-case basis taking into account the specific financial situation of each licensee.

          Subpart D [Reserved]

          Subpart E—Standard Design Approvals

            § 52.131
            Scope of subpart.

            This subpart sets out procedures for the filing, NRC staff review, and referral to the Advisory Committee on Reactor Safeguards of standard designs for a nuclear power reactor of the type described in § 50.22 of this chapter or major portions thereof.

            § 52.133
            Relationship to other subparts.
            (a) This subpart applies to a person that requests a standard design approval from the NRC staff separately from an application for a construction permit filed under 10 CFR part 50 or a combined license filed under subpart C of this part. An applicant for a construction permit or combined license may reference a standard design approval.
            (b) Subpart B of this part governs the certification by rulemaking of the design of a nuclear power plant. Subpart B may be used independently of the provisions in this subpart.
            (c) Subpart F of this part governs the issuance of licenses to manufacture nuclear power reactors to be installed and operated at sites not identified in the manufacturing license application. Subpart F of this part may be used independently of the provisions in this subpart.

            § 52.135
            Filing of applications.
            (a) Any person may submit a proposed standard design for a nuclear power reactor of the type described in 10 CFR 50.22 to the NRC staff for its review. The submittal may consist of either the final design for the entire facility or the final design of major portions thereof.
            (b) The submittal for review of the proposed standard design must be made in the same manner and in the same number of copies as provided in 10 CFR 50.30 and 52.3 for license applications.
            (c) The fees associated with the filing and review of the application are set forth in 10 CFR part 170.

            § 52.136
            Contents of applications; general information.
            The application must contain all of the information required by 10 CFR 50.33(a) through (d) and (j).

            § 52.137
            Contents of applications; technical information.
            If the applicant seeks review of a major portion of a standard design, the application need only contain the information required by this section to the extent the requirements are applicable to the major portion of the standard design for which NRC staff approval is sought.
            (a) The application must contain a final safety analysis report that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility, or major portion thereof, and must include the following information:
            (1) The site parameters postulated for the design, and an analysis and evaluation of the design in terms of those site parameters;
            (2) A description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification, upon which the requirements have been established, and the evaluations required to show that safety functions will be accomplished. It is expected that the standard plant will reflect through its design, construction, and operation an extremely low probability for accidents that could result in the release of significant quantities of radioactive fission products. The description shall be sufficient to permit understanding of the system designs and their relationship to the safety evaluations. Items such as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems shall be discussed insofar as they are pertinent. The following power reactor design characteristics will be taken into consideration by the Commission:
            (i) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials;

            (ii) The extent to which generally accepted engineering standards are applied to the design of the reactor;

            (iii) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials; and
            (iv) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. Special attention must be directed to plant design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release 9
               from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable postulated site parameters, including site meteorology, to evaluate the offsite radiological consequences. The evaluation must determine that:

                9 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. These accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

            (A) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem 10
               total effective dose equivalent (TEDE); and

                10 A whole body dose of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations at the time could be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, its use is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, to assure that these designs provide assurance of low risk of public exposure to radiation, in the event of an accident.

            (B) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE;
            (3) The design of the facility including:
            (i) The principal design criteria for the facility. Appendix A to 10 CFR part 50, general design criteria (GDC), establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units;
            (ii) The design bases and the relation of the design bases to the principal design criteria; and
            (iii) Information relative to materials of construction, general arrangement, and approximate dimensions, sufficient to provide reasonable assurance that the design will conform to the design bases with adequate margin for safety;

            (4) An analysis and evaluation of the design and performance of SSC with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents. Analysis and evaluation of ECCS cooling performance and the need for high-point vents following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of 10 CFR 50.46 and 50.46a;

            (5) The kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in part 20 of this chapter;
            (6) The information required by § 20.1406 of this chapter;
            (7) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter;
            (8) The information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), and (f)(3)(v) of 10 CFR 50.34(f);
            (9) For applications for light-water-cooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. The evaluation required by this section shall include an identification and description of all differences in design features, analytical techniques, and procedural measures proposed for the design and those corresponding features, techniques, and measures given in the SRP acceptance criteria. Where a difference exists, the evaluation shall discuss how the proposed alternative provides an acceptable method of complying with the Commission's regulations, or portions thereof, that underlie the corresponding SRP acceptance criteria. The SRP is not a substitute for the regulations, and compliance is not a requirement;
            (10) The information with respect to the design of equipment to maintain control over radioactive materials in gaseous and liquid effluents produced during normal reactor operations described in 10 CFR 50.34a(e);
            (11) The information pertaining to design features that affect plans for coping with emergencies in the operation of the reactor facility or a major portion thereof;
            (12) An analysis and description of the equipment and systems for combustible gas control as required by § 50.44 of this chapter;
            (13) The list of electric equipment important to safety that is required by 10 CFR 50.49(d);
            (14) A description of protection provided against pressurized thermal shock events, including projected values of the reference temperature for reactor vessel beltline materials as defined in 10 CFR 50.60 and 50.61;
            (15) Information demonstrating how the applicant will comply with requirements for reduction of risk from anticipated transients without scram (ATWS) events in § 50.62;
            (16) The coping analysis, and any design features necessary to address station blackout, as described in § 50.63 of this chapter;
            (17) Information demonstrating how the applicant will comply with requirements for criticality accidents in § 50.68(b)(2)-(b)(4);
            (18) A description and analysis of the fire protection design features for the standard plant necessary to comply with part 50, appendix A, GDC 3, and § 50.48 of this chapter;
            (19) A description of the quality assurance program applied to the design of the SSCs of the facility. Appendix B to 10 CFR part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” sets forth the requirements for quality assurance programs for nuclear power plants. The description of the quality assurance program for a nuclear power plant shall include a discussion of how the applicable requirements of appendix B to 10 CFR part 50 were satisfied;
            (20) The information necessary to demonstrate that the standard plant complies with the earthquake engineering criteria in 10 CFR part 50, appendix S;
            (21) Proposed technical resolutions of those Unresolved Safety Issues and medium- and high-priority generic safety issues which are identified in the version of NUREG-0933 current on the date up to 6 months before the docket date of the application and which are technically relevant to the design;

            (22) The information necessary to demonstrate how operating experience insights have been incorporated into the plant design;

            (23) For light-water reactor designs, a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass;
            (24) A description, analysis, and evaluation of the interfaces between the standard design and the balance of the nuclear power plant; and
            (25) A description of the design-specific probabilistic risk assessment and its results.
            (26) For applications for standard design approvals which are subject to 10 CFR 50.150(a), the information required by 10 CFR 50.150(b).
            (b) An application for approval of a standard design, which differs significantly from the light-water reactor designs of plants that have been licensed and in commercial operation before April 18, 1989, or uses simplified, inherent, passive, or other innovative means to accomplish its safety functions, must meet the requirements of 10 CFR 50.43(e).
            [72 FR 49517, Aug. 28, 2007, as amended at 74 FR 28147, June 12, 2009]

            § 52.139
            Standards for review of applications.
            Applications filed under this subpart will be reviewed for compliance with the standards set out in 10 CFR parts 20, 50 and its appendices, and 10 CFR parts 73 and 100.

            § 52.141
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            The Commission shall refer a copy of the application to the ACRS. The ACRS shall report on those portions of the application which concern safety.

            § 52.143
            Staff approval of design.

            Upon completion of its review of a submittal under this subpart and receipt of a report by the Advisory Committee on Reactor Safeguards under § 52.141 of this subpart, the NRC staff shall publish a determination in the Federal Register as to whether or not the design is acceptable, subject to appropriate terms and conditions, and make an analysis of the design in the form of a report available at the NRC Web site, http://www.nrc.gov.

            § 52.145
            Finality of standard design approvals; information requests.
            (a) An approved design must be used by and relied upon by the NRC staff and the ACRS in their review of any individual facility license application that incorporates by reference a standard design approved in accordance with this paragraph unless there exists significant new information that substantially affects the earlier determination or other good cause.
            (b) The determination and report by the NRC staff do not constitute a commitment to issue a permit or license, or in any way affect the authority of the Commission, Atomic Safety and Licensing Board Panel, or presiding officers in any proceeding under part 2 of this chapter.
            (c) Except for information requests seeking to verify compliance with the current licensing basis of the standard design approval, information requests to the holder of a standard design approval must be evaluated before issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information. Each evaluation performed by the NRC staff must be in accordance with 10 CFR 50.54(f) and must be approved by the Executive Director for Operations or his or her designee before issuance of the request.

            § 52.147
            Duration of design approval.
            A standard design approval issued under this subpart is valid for 15 years from the date of issuance and may not be renewed. A design approval continues to be valid beyond the date of expiration in any proceeding on an application for a construction permit or an operating license under part 50 or a combined license or manufacturing license under part 52 that references the final design approval and is docketed before the date of expiration of the design approval.

          Subpart F—Manufacturing Licenses

            § 52.151
            Scope of subpart.
            This subpart sets out the requirements and procedures applicable to Commission issuance of a license authorizing manufacture of nuclear power reactors to be installed at sites not identified in the manufacturing license application.

            § 52.153
            Relationship to other subparts.
            (a) A nuclear power reactor manufactured under a manufacturing license issued under this subpart may only be transported to and installed at a site for which either a construction permit under part 50 of this chapter or a combined license under subpart C of this part has been issued.
            (b) Subpart B of this part governs the certification by rulemaking of the design of standard nuclear power facilities. Subpart E of this part governs the NRC staff review and approval of standard designs for a nuclear power facility. A manufacturing license applicant may reference a standard design certification or a standard design approval in its application. These subparts may also be used independently of the provisions in this subpart.

            § 52.155
            Filing of applications.
            (a) Any person, except one excluded by 10 CFR 50.38, may file an application for a manufacturing license under this subpart with the Director, Office of Nuclear Reactor Regulation.
            (b) The application must comply with the applicable filing requirements of §§ 52.3 and 50.30 of this chapter.
            (c) The fees associated with the filing and review of the application are set forth in 10 CFR part 170.
            [72 FR 49517, Aug. 28, 2007, as amended at 84 FR 65645, Nov. 29, 2019]

            § 52.156
            Contents of applications; general information.
            The application must contain all of the information required by 10 CFR 50.33(a) through (d), and (j).

            § 52.157
            Contents of applications; technical information in final safety analysis report.
            The application must contain a final safety analysis report containing the information set forth below, with a level of design information sufficient to enable the Commission to judge the applicant's proposed means of assuring that the manufacturing conforms to the design and to reach a final conclusion on all safety questions associated with the design, permit the preparation of construction and installation specifications by an applicant who seeks to use the manufactured reactor, and permit the preparation of acceptance and inspection requirements by the NRC:
            (a) The principal design criteria for the reactor to be manufactured. Appendix A of 10 CFR part 50, “General Design Criteria for Nuclear Power Plants,” establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units;
            (b) The design bases and the relation of the design bases to the principal design criteria;

            (c) A description and analysis of the structures, systems, and components of the reactor to be manufactured, with emphasis upon the materials of manufacture, performance requirements, the bases, with technical justification therefor, upon which the performance requirements have been established, and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Items such as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems shall be discussed insofar as they are pertinent. The following power reactor design characteristics will be taken into consideration by the Commission:
            (1) Intended use of the manufactured reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials;
            (2) The extent to which generally accepted engineering standards are applied to the design of the reactor; and
            (3) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials;
            (d) The safety features that are engineered into the reactor and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. Special attention must be directed to reactor design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release 11
               from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable postulated site parameters, including site meteorology, to evaluate the offsite radiological consequences. The evaluation must determine that:

                11 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. These accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

            (1) An individual located at any point on the boundary of the exclusion area for any 2 hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem 12
               total effective dose equivalent (TEDE);

                12 A whole body dose of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations at the time could be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, its use is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, to assure that these designs provide assurance of low risk of public exposure to radiation, in the event of an accident.

            (2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE; and
            (e) The kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in part 20 of this chapter.
            (f) Information necessary to establish that the design of the reactor to be manufactured complies with the technical requirements in 10 CFR Chapter I, including:

            (1) An analysis and evaluation of the design and performance of structures, systems, and components with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. Analysis and evaluation of ECCS cooling performance and the need for high-point vents following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of §§ 50.46 and 50.46a of this chapter;
            (2) A description and analysis of the fire protection design features for the reactor necessary to comply with 10 CFR part 50, appendix A, GDC 3 and § 50.48 of this chapter;
            (3) A description of protection provided against pressurized thermal shock events, including projected values of the reference temperature for reactor vessel beltline materials as defined in §§ 50.60 and 50.61 of this chapter;
            (4) An analysis and description of the equipment and systems for combustible gas control as required by § 50.44 of this chapter;
            (5) The coping analysis, and any design features necessary to address station blackout, as described in § 50.63 of this chapter;
            (6) The list of electric equipment important to safety that is required by 10 CFR 50.49(d);
            (7) Information demonstrating how the applicant will comply with requirements for reduction of risk from anticipated transients without scram (ATWS) events in § 50.62;
            (8) Information demonstrating how the applicant will comply with requirements for criticality accidents in § 50.68(b)(2)-(b)(4);
            (9) The information required by § 20.1406 of this chapter;
            (10) [Reserved]
            (11) The information with respect to the design of equipment to maintain control over radioactive materials in gaseous and liquid effluents produced during normal reactor operations, as described in § 50.34a(e) of this chapter;
            (12) The information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in § 50.34(f) of this chapter, except paragraphs (f)(1)(xii), (f)(2)(ix), and (f)(3)(v);
            (13) If the applicant seeks to use risk-informed treatment of SSCs in accordance with § 50.69 of this chapter, the information required by § 50.69(b)(2) of this chapter;
            (14) The information necessary to demonstrate that the manufactured reactor complies with the earthquake engineering criteria in appendix S to 10 CFR part 50;
            (15) Information sufficient to demonstrate compliance with the applicable requirements regarding testing, analysis, and prototypes as set forth in § 50.43(e) of this chapter;
            (16) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter;
            (17) A description of the quality assurance program applied to the design, and to be applied to the manufacture of, the structures, systems, and components of the reactor. Appendix B to 10 CFR part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” sets forth the requirements for quality assurance programs for nuclear power plants. The description of the quality assurance program must include a discussion of how the applicable requirements of appendix B to 10 CFR part 50 have been and will be satisfied; and
            (18) Proposed technical specifications applicable to the reactor being manufactured, prepared in accordance with the requirements of §§ 50.36 and 50.36a of this chapter;
            (19) The site parameters postulated for the design, and an analysis and evaluation of the reactor design in terms of those site parameters;
            (20) The interface requirements between the manufactured reactor and the remaining portions of the nuclear power plant. These requirements must be sufficiently detailed to allow for completion of the final safety analysis;
            (21) Justification that compliance with the interface requirements of paragraph (f)(20) of this section is verifiable through inspections, testing, or analysis. The method to be used for verification of interface requirements must be included as part of the proposed ITAAC required by § 52.158(a);

            (22) A representative conceptual design for a nuclear power facility using the manufactured reactor, to aid the NRC in its review of the final safety analysis required by this section and to permit assessment of the adequacy of the interface requirements in paragraph (f)(20) of this section;

            (23) For light-water reactor designs, a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass;
            (24) [Reserved]
            (25) If the reactor is to be used in modular plant design, a description of the possible operating configurations of the reactor modules with common systems, interface requirements, and system interactions. The final safety analysis must also account for differences among the configurations, including any restrictions that will be necessary during the construction and startup of a given module to ensure the safe operation of any module already operating;
            (26) A description of the management plan for design and manufacturing activities, including:
            (i) The organizational and management structure singularly responsible for direction of design and manufacture of the reactor;
            (ii) Technical resources directed by the applicant, and the qualifications requirements;
            (iii) Details of the interaction of design and manufacture within the applicant's organization and the manner by which the applicant will ensure close integration of the architect engineer and the nuclear steam supply vendor, as applicable;
            (iv) Proposed procedures governing the preparation of the manufactured reactor for shipping to the site where it is to be operated, the conduct of shipping, and verifying the condition of the manufactured reactor upon receipt at the site; and
            (v) The degree of top level management oversight and technical control to be exercised by the applicant during design and manufacture, including the preparation and implementation of procedures necessary to guide the effort;
            (27) Necessary parameters to be used in developing plans for preoperational testing and initial operation;
            (28) Proposed technical resolutions of those Unresolved Safety Issues and medium- and high-priority generic safety issues which are identified in the version of NUREG-0933 current on the date up to 6 months before the docket date of the application and which are technically relevant to the design;
            (29) The information necessary to demonstrate how operating experience insights have been incorporated into the manufactured reactor design;
            (30) For applications for light-water-cooled nuclear power plants, an evaluation of the design to be manufactured against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. The evaluation required by this section shall include an identification and description of all differences in design features, analytical techniques, and procedural measures proposed for the design and those corresponding features, techniques, and measures given in the SRP acceptance criteria. Where a difference exists, the evaluation shall discuss how the proposed alternative provides an acceptable method of complying with the Commission's regulations, or portions thereof, that underlie the corresponding SRP acceptance criteria. The SRP is not a substitute for the regulations, and compliance is not a requirement; and
            (31) A description of the design-specific probabilistic risk assessment and its results.
            (32) For applications for manufacturing licenses which are subject to 10 CFR 50.150(a), the information required by 10 CFR 50.150(b).
            [72 FR 49517, Aug. 28, 2007, as amended at 74 FR 28147, June 12, 2009]

            § 52.158
            Contents of application; additional technical information.
            The application must contain:
            (a)(1) Inspections, tests, analyses, and acceptance criteria (ITAAC). The proposed inspections, tests, and analyses that the licensee who will be operating the reactor shall perform, and the acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met:

            (i) The reactor has been manufactured in conformity with the manufacturing license; the provisions of the Act, and the Commission's rules and regulations; and
            (ii) The manufactured reactor will be operated in conformity with the approved design and any license authorizing operation of the manufactured reactor.
            (2) If the application references a standard design certification, the ITAAC contained in the certified design must apply to those portions of the facility design which are covered by the design certification.

            (3) If the application references a standard design certification, the application may include a notification that a required inspection, test, or analysis in the design certification ITAAC has been successfully completed and that the corresponding acceptance criterion has been met. The Federal Register notification required by § 52.163 must indicate that the application includes this notification.
            (b)(1) An environmental report as required by 10 CFR 51.54.
            (2) If the manufacturing license application references a standard design certification, the environmental report need not contain a discussion of severe accident mitigation design alternatives for the reactor.

            § 52.159
            Standards for review of application.
            Applications filed under this subpart will be reviewed according to the applicable standards set out in 10 CFR parts 20, 50 and its appendices, 51, 73, and 100 and its appendices.

            § 52.161
            [Reserved]

            § 52.163
            Administrative review of applications; hearings.

            A proceeding on a manufacturing license is subject to all applicable procedural requirements contained in 10 CFR part 2, including the requirements for docketing in § 2.101(a)(1) through (4) of this chapter, and the requirements for issuance of a notice of proposed action in § 2.105 of this chapter, provided, however, that the designated sections may not be construed to require that the environmental report or draft or final environmental impact statement include an assessment of the benefits of constructing and/or operating the manufactured reactor or an evaluation of alternative energy sources. All hearings on manufacturing licenses are governed by the hearing procedures contained in 10 CFR part 2, subparts C, E, G, L, and N.
            [72 FR 49517, Aug. 28, 2007, as amended at 78 FR 34249, June 7, 2013]

            § 52.165
            Referral to the Advisory Committee on Reactor Safeguards (ACRS).
            The Commission shall refer a copy of the application to the ACRS. The ACRS shall report on those portions of the application which concern safety.

            § 52.167
            Issuance of manufacturing license.
            (a) After completing any hearing under § 52.163, and receiving the report submitted by the ACRS, the Commission may issue a manufacturing license if the Commission finds that:
            (1) Applicable standards and requirements of the Act and the Commission's regulations have been met;
            (2) There is reasonable assurance that the reactor(s) will be manufactured, and can be transported, incorporated into a nuclear power plant, and operated in conformity with the manufacturing license, the provision of the Act, and the Commission's regulations;
            (3) The proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having characteristics that fall within the site parameters postulated for the design of the manufactured reactor(s) without undue risk to the health and safety of the public;
            (4) The applicant is technically qualified to design and manufacture the proposed nuclear power reactor(s);

            (5) The proposed inspections, tests, analyses and acceptance criteria are necessary and sufficient, within the scope of the manufacturing license, to provide reasonable assurance that the manufactured reactor has been manufactured and will be operated in conformity with the license, the provisions of the Act, and the Commission's regulations;

            (6) The issuance of a license to the applicant will not be inimical to the common defense and security or to the health and safety of the public; and
            (7) The findings required by subpart A of part 51 of this chapter have been made.
            (b) Each manufacturing license issued under this subpart shall specify:
            (1) Terms and conditions as the Commission deems necessary and appropriate;
            (2) Technical specifications for operation of the manufactured reactor, as the Commission deems necessary and appropriate;
            (3) Site parameters and design characteristics for the manufactured reactor; and
            (4) The interface requirements to be met by the site-specific elements of the facility, such as the service water intake structure and the ultimate heat sink, not within the scope of the manufactured reactor.
            (c)(1) A holder of a manufacturing license may not transport or allow to be removed from the place of manufacture the manufactured reactor except to the site of a licensee with either a construction permit under part 50 of this chapter or a combined license under subpart C of this part. The construction permit or combined license must authorize the construction of a nuclear power facility using the manufactured reactor(s).
            (2) A holder of a manufacturing license shall include, in any contract governing the transport of a manufactured reactor from the place of manufacture to any other location, a provision requiring that the person or entity transporting the manufactured reactor to comply with all NRC-approved shipping requirements in the manufacturing license.

            § 52.169
            [Reserved]

            § 52.171
            Finality of manufacturing licenses; information requests.
            (a)(1) Notwithstanding any provision in 10 CFR 50.109, during the term of a manufacturing license the Commission may not modify, rescind, or impose new requirements on the design of the nuclear power reactor being manufactured, or the requirements for the manufacture of the nuclear power reactor, unless the Commission determines that a modification is necessary to bring the design of the reactor or its manufacture into compliance with the Commission's requirements applicable and in effect at the time the manufacturing license was issued, or to provide reasonable assurance of adequate protection to public health and safety or common defense and security.
            (2) Any modification to the design of a manufactured nuclear power reactor which is imposed by the Commission under paragraph (a)(1) of this section will be applied to all reactors manufactured under the license, including those that have already been transported and sited, except those reactors to which the modification has been rendered technically irrelevant by action taken under paragraph (b) of this section.
            (3) In making the findings required for issuance of a construction permit, operating license, combined license, in any hearing under § 52.103, or in any enforcement hearing other than one initiated by the Commission under paragraph (a)(1) of this section, for which a nuclear power reactor manufactured under this subpart is referenced or used, the Commission shall treat as resolved those matters resolved in the proceeding on the application for issuance or renewal of the manufacturing license, including the adequacy of design of the manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives into the design of the reactor to be manufactured.
            (b)(1) The holder of a manufacturing license may not make changes to the design of the nuclear power reactor authorized to be manufactured without prior Commission approval. The request for a change to the design must be in the form of an application for a license amendment, and must meet the requirements of 10 CFR 50.90 and 50.92.

            (2) An applicant or licensee who references or uses a nuclear power reactor manufactured under a manufacturing license under this subpart may request a departure from the design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor. The Commission may grant a request only if it determines that the departure will comply with the requirements of 10 CFR 52.7, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure. The granting of a departure on request of an applicant is subject to litigation in the same manner as other issues in the construction permit or combined license hearing.
            (c) Except for information requests seeking to verify compliance with the current licensing basis of either the manufacturing license or the manufactured reactor, information requests to the holder of a manufacturing license or an applicant or licensee using a manufactured reactor must be evaluated before issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information. Each evaluation performed by the NRC staff must be in accordance with 10 CFR 50.54(f) and must be approved by the Executive Director for Operations or his or her designee before issuance of the request.

            § 52.173
            Duration of manufacturing license.
            A manufacturing license issued under this subpart may be valid for not less than 5, nor more than 15 years from the date of issuance. A holder of a manufacturing license may not initiate the manufacture of a reactor less than 3 years before the expiration of the license even though a timely application for renewal has been docketed with the NRC. Upon expiration of the manufacturing license, the manufacture of any uncompleted reactors must cease unless a timely application for renewal has been docketed with the NRC.

            § 52.175
            Transfer of manufacturing license.
            A manufacturing license may be transferred in accordance with § 50.80 of this chapter.

            § 52.177
            Application for renewal.
            (a) Not less than 12 months, nor more than 5 years before the expiration of the manufacturing license, or any later renewal period, the holder of the manufacturing license may apply for a renewal of the license. An application for renewal must contain all information necessary to bring up to date the information and data contained in the previous application.
            (b) The filing of an application for a renewed license must be in accordance with subpart A of 10 CFR part 2 and 10 CFR 52.3 and 50.30.

            (c) A manufacturing license, either original or renewed, for which a timely application for renewal has been filed, remains in effect until the Commission has made a final determination on the renewal application, provided, however, that in accordance with § 52.173, the holder of a manufacturing license may not begin manufacture of a reactor less than 3 years before the expiration of the license.
            (d) Any person whose interest may be affected by renewal of the permit may request a hearing on the application for renewal. The request for a hearing must comply with 10 CFR 2.309. If a hearing is granted, notice of the hearing will be published in accordance with 10 CFR 2.104.
            (e) The Commission shall refer a copy of the application for renewal to the Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall report on those portions of the application which concern safety and shall apply the criteria set forth in § 52.159.

            § 52.179
            Criteria for renewal.
            The Commission may grant the renewal if the Commission determines:
            (a) The manufacturing license complies with the Atomic Energy Act and the Commission's regulations and orders applicable and in effect at the time the manufacturing license was originally issued; and
            (b) Any new requirements the Commission may wish to impose are:
            (1) Necessary for adequate protection to public health and safety or common defense and security;

            (2) Necessary for compliance with the Commission's regulations and orders applicable and in effect at the time the manufacturing license was originally issued; or
            (3) A substantial increase in overall protection of the public health and safety or the common defense and security to be derived from the new requirements, and the direct and indirect costs of implementation of those requirements are justified in view of this increased protection.

            § 52.181
            Duration of renewal.
            A renewed manufacturing license may be issued for a term of not less than 5, nor more than 15 years, plus any remaining years on the manufacturing license then in effect before renewal. The renewed license shall be subject to the requirements of §§ 52.171 and 52.175.

          Subpart G [Reserved]

          Subpart H—Enforcement

            § 52.301
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of—
            (1) The Atomic Energy Act of 1954, as amended;
            (2) Title II of the Energy Reorganization Act of 1974, as amended; or
            (3) A regulation or order issued under those Acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under Section 234 of the Atomic Energy Act:
            (1) For violations of—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any regulation, or order issued under the sections specified in paragraph (b)(1)(i) of this section;
            (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under Section 186 of the Atomic Energy Act of 1954, as amended.

            § 52.303
            Criminal penalties.
            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under Sections 161b, 161i, or 161o of the Act. For purposes of Section 223, all the regulations in part 52 are issued under one or more of Sections 161b, 161i, or 160o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in part 52 that are not issued under Sections 161b, 161i, or 161o for the purposes of Section 223 are as follows: §§ 52.0, 52.1, 52.2, 52.3, 52.7, 52.8, 52.9, 52.10, 52.11, 52.12, 52.13, 52.15, 52.16, 52.17, 52.18, 52.21, 52.23, 52.24, 52.27, 52.28, 52.29, 52.31, 52.33, 52.39, 52.41, 52.43, 52.45, 52.46, 52.47, 52.48, 52.51, 52.53, 52.54, 52.55, 52.57, 52.59, 52.61, 52.63, 52.71, 52.73, 52.75, 52.77, 52.79, 52.80, 52.81, 52.83, 52.85, 52.87, 52.93, 52.97, 52.98, 52.103, 52.104, 52.105, 52.107, 52.109, 52.131, 52.133, 52.135, 52.136, 52.137, 52.139, 52.141, 52.143, 52.145, 52.147, 52.151, 52.153, 52.155, 52.156, 52.157, 52.158, 52.159, 52.161, 52.163, 52.165, 52.167, 52.171, 52.173, 52.175, 52.177, 52.179, 52.181, 52.301, and 52.303.

            Pt. 52, App. A
            Appendix A to Part 52—Design Certification Rule for the U.S. Advanced Boiling Water Reactor
            I. Introduction
            A. Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (U.S. ABWR) design, in accordance with 10 CFR part 52, subpart B. The applicant for the original certification of the U.S. ABWR design was GE Nuclear Energy (GE).
            B. The applicant for the amendment to the U.S. ABWR design to address the requirements in 10 CFR 50.150, “Aircraft impact assessment,” (AIA rule) is the STP Nuclear Operating Company (STPNOC).
            II. Definitions
            A. Generic design control document (generic DCD) means either or both of the documents containing the Tier 1 and Tier 2 information and generic technical specifications that are incorporated by reference into this appendix.
            B. Generic technical specifications means the information, required by 10 CFR 50.36 and 50.36a, for the portion of the plant that is within the scope of this appendix.

            C. Plant-specific DCD means the document, maintained by an applicant or licensee who references this appendix, consisting of the information in the generic DCD, as modified and supplemented by the plant-specific departures and exemptions made under Section VIII of this appendix.
            D. Tier 1 means the portion of the design-related information contained in the generic DCD that is approved and certified by this appendix (hereinafter Tier 1 information). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes:
            1. Definitions and general provisions;
            2. Design descriptions;
            3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
            4. Significant site parameters; and
            5. Significant interface requirements.
            E. Tier 2 means the portion of the design-related information contained in the generic DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these differences, an applicant or licensee must meet the requirement in Section III.B of this appendix to reference Tier 2 when referencing Tier 1. Tier 2 information includes:
            1. Information required by §§ 52.47(a) and 52.47(c), with the exception of generic technical specifications and conceptual design information;
            2. Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met; and
            3. Combined license (COL) action items (COL license information), which identify certain matters that must be addressed in the site-specific portion of the final safety analysis report (FSAR) by an applicant who references this appendix. These items constitute information requirements but are not the only acceptable set of information in the FSAR. An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the FSAR. After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the FSAR.
            F. Tier 2* means the portion of the Tier 2 information, designated as such in the generic DCD, which is subject to the change process in Section VIII.B.6 of this appendix. This designation expires for some Tier 2* information under Section VIII.B.6.
            G. Departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses means:
            (1) Changing any of the elements of the method described in the plant-specific DCD unless the results of the analysis are conservative or essentially the same; or
            (2) Changing from a method described in the plant-specific DCD to another method unless that method has been approved by NRC for the intended application.
            H. All other terms in this appendix have the meaning set out in 10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable.
            III. Scope and Contents
            A. Design Control Documents
            1. Incorporation by reference approval. Certain documents identified in paragraphs III.A.2 and III.A.3 of this section are approved for incorporation by reference into this appendix by the Director of the Office of the Federal Register in accordance with 5 U.S.C. 552(a) and 1 CFR part 51. Documents approved for incorporation by reference and created or received at the NRC are available online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain entry into ADAMS, which provides text and image files of the NRC's public documents. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, then contact the NRC's Public Document Room (PDR) reference staff at (800) 397-4209, (301) 415-3747, or by email at PDR.Resource@nrc.gov. A copy of these DCDs approved for incorporation by reference are available for examination and copying at the NRC's PDR located at Room O-1F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. Copies are also available for examination at the NRC Library located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852, telephone: (301) 415-5610, email: Library.Resource@nrc.gov. All approved material is available for inspection at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call (202) 741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.

            2. GE DCD: All Tier 1, Tier 2, and the generic technical specifications in the GE Nuclear Energy (GE) “ABWR Design Control Document, Revision 4, March 1997” (GE DCD). You may obtain copies of the GE DCD from the National Technical Information Service, 5285 Port Royal Road, Springfield, Virginia 22161, (703) 605-6515. To view the GE DCD in ADAMS, search under ADAMS Accession No. ML11126A129. The GE DCD can also be viewed at the Federal Rulemaking Web site, http://www.regulations.gov, by searching for documents filed under Docket ID NRC-2010-0134.
            3. STPNOC DCD: All Tier 1 and Tier 2 information in the STP Nuclear Operating Company “Design Control Document ABWR STP Aircraft Impact Assessment Amendment Revision 3, Copyright @ 2010” (STPNOC DCD). You may obtain copies of the STPNOC DCD from the Regulatory Affairs Manager for STP Units 3 and 4, STP Nuclear Operating Company, P.O. Box 289, Wadsworth, Texas 77483, telephone: (361) 972-8440. To view the STPNOC DCD in ADAMS, search under ADAMS Accession No. ML102870017. The STPNOC DCD can also be viewed at the Federal Rulemaking Web site, http://www.regulations.gov, by searching for documents filed under Docket ID NRC-2010-0134.
            B.1. An applicant or licensee referencing this appendix, in accordance with section IV of this appendix, shall incorporate by reference and comply with the requirements of this appendix, including Tier 1, Tier 2, and the generic technical specifications except as otherwise provided in this appendix. An applicant or licensee referencing this appendix may reference either the GE DCD, or both the GE DCD and the STPNOC DCD. An applicant referencing this appendix shall indicate in its application and in all necessary supporting documentation whether it is implementing the GE DCD, or both the GE DCD and the STPNOC DCD.
            2. Conceptual design information, as set forth in the generic DCD, and the “Technical Support Document for the ABWR” are not part of this appendix. Tier 2 references to the probabilistic risk assessment (PRA) in the ABWR standard safety analysis report do not incorporate the PRA into Tier 2.
            C. If there is a conflict between Tier 1 and Tier 2 of a DCD, then Tier 1 controls.
            D. If there is a conflict between the generic DCD and the application for design certification of the U.S. ABWR design, NUREG-1503, “Final Safety Evaluation Report related to the Certification of the Advanced Boiling Water Reactor Design” (ABWR FSER), and Supplement No. 1, or NUREG-1948 “Safety Evaluation Report—The STP Nuclear Operating Company Amendment to the Advanced Boiling Water Reactor (ABWR) Design Certification” (AIA FSER), then the generic DCD controls.
            E. Design activities for structures, systems, and components that are wholly outside the scope of this appendix may be performed using site characteristics, provided the design activities do not affect the DCD or conflict with the interface requirements.
            IV. Additional Requirements and Restrictions
            A. An applicant for a combined license that wishes to reference this appendix shall, in addition to complying with the requirements of 10 CFR 52.77, 52.79, and 52.80, comply with the following requirements:
            1. Incorporate by reference, as part of its application, this appendix;
            2. Include, as part of its application:
            a. A plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the U.S. ABWR design, as modified and supplemented by the applicant's exemptions and departures;
            b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
            c. Plant-specific technical specifications, consisting of the generic and site-specific technical specifications, that are required by 10 CFR 50.36 and 50.36a;
            d. Information demonstrating compliance with the site parameters and interface requirements;
            e. Information that addresses the COL action items; and
            f. Information required by 10 CFR 52.47 that is not within the scope of this appendix.
            3. Include, in the plant-specific DCD, the sensitive unclassified non-safeguards information (including proprietary information) and safeguards information referenced in the GE DCD and the STPNOC DCD, as applicable.
            4.a. Include, as part of its application, a demonstration that an entity other than GE Nuclear Energy is qualified to supply the U.S. ABWR-certified design unless GE Nuclear Energy supplies the design for the applicant's use.
            b. For an applicant referencing the STPNOC-certified design option, include, as part of its application, a demonstration that an entity other than the STPNOC and Toshiba America Nuclear Energy (TANE) acting together is qualified to supply the STPNOC-certified design option, unless the STPNOC and TANE acting together supply the design option for the applicant's use.
            B. The Commission reserves the right to determine in what manner this appendix may be referenced by an applicant for a construction permit or operating license under 10 CFR part 50.
            V. Applicable Regulations
            A.1. Except as indicated in paragraph B of this section, the regulations that apply to the U.S. ABWR design as contained in the GE DCD are in 10 CFR parts 20, 50, 73, and 100, codified as of May 2, 1997, that are applicable and technically relevant, as described in the FSER (NUREG-1503) and Supplement No. 1.

            2. Except as indicated in paragraph B of this section, the regulations that apply to the U.S. ABWR design as contained in the STPNOC DCD are those described in paragraph A.1 of this section and 10 CFR 50.150, codified as of December 7, 2011, as described in the FSER on the STPNOC amendment addressing the AIA rule (NUREG-1948).
            B. The U.S. ABWR design is exempt from portions of the following regulations:
            1. Paragraph (f)(2)(iv) of 10 CFR 50.34—Separate Plant Safety Parameter Display Console;
            2. Paragraph (f)(2)(viii) of 10 CFR 50.34—Post-Accident Sampling for Boron, Chloride, and Dissolved Gases; and
            3. Paragraph (f)(3)(iv) of 10 CFR 50.34—Dedicated Containment Penetration.
            VI. Issue Resolution
            A.1. GE DCD. The Commission has determined that the structures, systems, components, and design features of the U.S. ABWR design, as contained in the GE DCD, comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in section V.A.1 of this appendix; and, therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for the U.S. ABWR design. This conclusion does not include a finding with respect to compliance with the requirements of 10 CFR 50.150.
            2. STPNOC DCD. The Commission has determined that the structures, systems, components, and design features of the STPNOC amendment to the U.S. ABWR design, as contained in the STPNOC DCD, comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in section V.A.2 of this appendix, including 10 CFR 50.150; and, therefore, provide enhanced protection to the health and safety of the public afforded by compliance with 10 CFR 50.150. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, components, design features, design criteria, testing, analyses, acceptance criteria, or justifications to meet the requirements of 10 CFR 50.150 are not necessary for the STPNOC amendment to the U.S. ABWR design.
            3. GE and STPNOC DCD referenced together. The Commission has determined that the structures, systems, components, and design features of the U.S. ABWR, as contained in both the GE DCD and the STPNOC DCD, when referenced together, comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in section V.A. of this appendix; and, therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for the U.S. ABWR design, when the GE DCD and the STPNOC DCD are referenced together.
            B. The Commission considers the following matters resolved within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance of a combined license, amendment of a combined license, or renewal of a combined license, proceedings held under 10 CFR 52.103, and enforcement proceedings involving plants referencing this appendix:
            1. All nuclear safety issues, except for the generic technical specifications and other operational requirements, associated with the information in the ABWR FSER and Supplement No. 1, Tier 1, Tier 2 (including referenced information which the context indicates is intended as requirements), and the rulemaking record for the original certification of the U.S. ABWR design and all nuclear safety issues, except for operational requirements, associated with the information in the AIA FSER, Tier 1, Tier 2 (including referenced information which the context indicates is intended as requirements), and the rulemaking record for certification of the AIA amendment to the U.S. ABWR design;
            2. All nuclear safety and safeguards issues associated with the referenced sensitive unclassified non-safeguards information (including proprietary information) and safeguards information which, in context, are intended as requirements in the GE DCD and the STPNOC DCD;
            3. All generic changes to the DCD under and in compliance with the change processes in sections VIII.A.1 and VIII.B.1 of this appendix;
            4. All exemptions from the DCD under and in compliance with the change processes in sections VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;
            5. All departures from the DCD that are approved by license amendment, but only for that plant;
            6. Except as provided in paragraph VIII.B.5.g of this appendix, all departures from Tier 2 pursuant to and in compliance with the change processes in paragraph VIII.B.5 of this appendix that do not require prior NRC approval, but only for that plant;

            7. All environmental issues concerning severe accident mitigation design alternatives associated with the information in the NRC's final environmental assessment for the U.S. ABWR design and Revision 1 of the technical support document for the U.S. ABWR, dated December 1994, and for the NRC's final environmental assessment and Revision 0 of ABWR-LIC-09-621, “Applicant's Supplemental Environmental Report-Amendment to ABWR Standard Design Certification,” for the AIA amendment to the U.S. ABWR design for plants referencing this appendix whose site parameters are within those specified in the technical support document.

            C. The Commission does not consider operational requirements for an applicant or licensee who references this appendix to be matters resolved within the meaning of 10 CFR 52.63(a)(5). The Commission reserves the right to require operational requirements for an applicant or licensee who references this appendix by rule, regulation, order, or license condition.
            D. Except in accordance with the change processes in Section VIII of this appendix, the Commission may not require an applicant or licensee who references this appendix to:
            1. Modify structures, systems, components, or design features as described in the generic DCD;
            2. Provide additional or alternative structures, systems, components, or design features not discussed in the generic DCD; or
            3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria, or justification for structures, systems, components, or design features discussed in the generic DCD.
            E. The NRC will specify at an appropriate time the procedures to be used by an interested person who wishes to review portions of the design certification or references containing safeguards information or sensitive unclassified non-safeguards information (including proprietary information, such as trade secrets and commercial or financial information obtained from a person that are privileged or confidential (10 CFR 2.390 and 10 CFR part 9)), for the purpose of participating in the hearing required by 10 CFR 52.85, the hearing provided under 10 CFR 52.103, or in any other proceeding relating to this appendix in which interested persons have a right to request an adjudicatory hearing.
            VII. Duration of This Appendix
            This appendix may be referenced for a period of 15 years from June 11, 1997, except as provided for in 10 CFR 52.55(b) and 52.57(b). This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.
            VIII. Processes for Changes and Departures
            A. Tier 1 information.
            1. Generic changes to Tier 1 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 1 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs A.3 or A.4 of this section.
            3. Departures from Tier 1 information that are required by the Commission through plant-specific orders are governed by the requirements in 10 CFR 52.63(a)(4).
            4. Exemptions from Tier 1 information are governed by the requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will deny a request for an exemption from Tier 1, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.
            B. Tier 2 information.
            1. Generic changes to Tier 2 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
            3. The Commission may not require new requirements on Tier 2 information by plant-specific order while this appendix is in effect under §§ 52.55 or 52.61, unless:
            a. A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time this appendix was approved, as set forth in Section V of this appendix, or to assure adequate protection of the public health and safety or the common defense and security; and
            b. Special circumstances as defined in 10 CFR 52.7 are present.
            4. An applicant or licensee who references this appendix may request an exemption from Tier 2 information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7. The Commission will deny a request for an exemption from Tier 2, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design. The grant of an exemption to an applicant must be subject to litigation in the same manner as other issues material to the license hearing. The grant of an exemption to a licensee must be subject to an opportunity for a hearing in the same manner as license amendments.

            5.a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2* information, or the technical specifications, or requires a license amendment under paragraphs B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD.

            b. A proposed departure from Tier 2, other than one affecting resolution of a severe accident issue identified in the plant-specific DCD or one affecting information required by 10 CFR 52.47(a)(28) to address 10 CFR 50.150, requires a license amendment if it would:
            (1) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the plant-specific DCD;
            (2) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the plant-specific DCD;
            (3) Result in more than a minimal increase in the consequences of an accident previously evaluated in the plant-specific DCD;
            (4) Result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety previously evaluated in the plant-specific DCD;
            (5) Create a possibility for an accident of a different type than any evaluated previously in the plant-specific DCD;
            (6) Create a possibility for a malfunction of an SSC important to safety with a different result than any evaluated previously in the plant-specific DCD;
            (7) Result in a design basis limit for a fission product barrier as described in the plant-specific DCD being exceeded or altered; or
            (8) Result in a departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses.
            c. A proposed departure from Tier 2 affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if:
            (1) There is a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously reviewed and determined to be not credible could become credible; or
            (2) There is a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously reviewed.
            d. If an applicant or licensee proposes to depart from the information required by 10 CFR 52.47(a)(28) to be included in the FSAR for the standard design certification, then the applicant or licensee shall consider the effect of the changed feature or capability on the original assessment required by 10 CFR 50.150(a). The applicant or licensee must also document how the modified design features and functional capabilities continue to meet the assessment requirements in 10 CFR 50.150(a)(1) in accordance with section X of this appendix.
            e. If a departure requires a license amendment pursuant to paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
            f. A departure from Tier 2 information that is made under paragraph B.5 of this section does not require an exemption from this appendix.
            g. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under 10 CFR 52.103(a), who believes that an applicant or licensee who references this appendix has not complied with paragraph VIII.B.5 of this appendix when departing from Tier 2 information, may petition the NRC to admit into the proceeding such a contention. In addition to compliance with the general requirements of 10 CFR 2.309, the petition must demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a 10 CFR 52.103 preoperational hearing, or that the change bears directly on the amendment request in the case of a hearing on a license amendment. Any other party may file a response. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. The Commission may admit such a contention if it determines the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5 of this appendix.
            6.a. An applicant who references this appendix may not depart from Tier 2* information, which is designated with italicized text or brackets and an asterisk in the generic DCD, without NRC approval. The departure will not be considered a resolved issue, within the meaning of Section VI of this appendix and 10 CFR 52.63(a)(5).
            b. A licensee who references this appendix may not depart from the following Tier 2* matters without prior NRC approval. A request for a departure will be treated as a request for a license amendment under 10 CFR 50.90.
            (1) Fuel burnup limit (4.2).
            (2) Fuel design evaluation (4.2.3).
            (3) Fuel licensing acceptance criteria (appendix 4B).
            c. A licensee who references this appendix may not, before the plant first achieves full power following the finding required by 10 CFR 52.103(g), depart from the following Tier 2* matters except in accordance with paragraph B.6.b of this section. After the plant first achieves full power, the following Tier 2* matters revert to Tier 2 status and are thereafter subject to the departure provisions in paragraph B.5 of this section.
            (1) ASME Boiler & Pressure Vessel Code, Section III.
            (2) ACI 349 and ANSI/AISC N-690.
            (3) Motor-operated valves.

            (4) Equipment seismic qualification methods.
            (5) Piping design acceptance criteria.
            (6) Fuel system and assembly design (4.2), except burnup limit.
            (7) Nuclear design (4.3).
            (8) Equilibrium cycle and control rod patterns (App. 4A).
            (9) Control rod licensing acceptance criteria (App. 4C).
            (10) Instrument setpoint methodology.
            (11) EMS performance specifications and architecture.
            (12) SSLC hardware and software qualification.
            (13) Self-test system design testing features and commitments.
            (14) Human factors engineering design and implementation process.
            d. Departures from Tier 2* information that are made under paragraph B.6 of this section do not require an exemption from this appendix.
            C. Operational requirements.
            1. Generic changes to generic technical specifications and other operational requirements that were completely reviewed and approved in the design certification rulemaking and do not require a change to a design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Generic changes that do require a change to a design feature in the generic DCD are governed by the requirements in paragraphs A or B of this section.
            2. Generic changes to generic TS and other operational requirements are applicable to all applicants who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this section.
            3. The Commission may require plant-specific departures on generic technical specifications and other operational requirements that were completely reviewed and approved, provided a change to a design feature in the generic DCD is not required and special circumstances as defined in 10 CFR 2.335 are present. The Commission may modify or supplement generic technical specifications and other operational requirements that were not completely reviewed and approved or require additional technical specifications and other operational requirements on a plant-specific basis, provided a change to a design feature in the generic DCD is not required.
            4. An applicant who references this appendix may request an exemption from the generic technical specifications or other operational requirements. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7. The grant of an exemption must be subject to litigation in the same manner as other issues material to the license hearing.
            5. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under 10 CFR 52.103(a), who believes that an operational requirement approved in the DCD or a technical specification derived from the generic technical specifications must be changed may petition to admit into the proceeding such a contention. Such petition must comply with the general requirements of 10 CFR 2.309 and must demonstrate why special circumstances as defined in 10 CFR 2.335 are present, or for compliance with the Commission's regulations in effect at the time this appendix was approved, as set forth in Section V of this appendix. Any other party may file a response thereto. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. All other issues with respect to the plant-specific technical specifications or other operational requirements are subject to a hearing as part of the license proceeding.
            6. After issuance of a license, the generic technical specifications have no further effect on the plant-specific technical specifications and changes to the plant-specific technical specifications will be treated as license amendments under 10 CFR 50.90.
            IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
            A.1. An applicant or licensee who references this appendix shall perform and demonstrate conformance with the ITAAC before fuel load. With respect to activities subject to an ITAAC, an applicant for a license may proceed at its own risk with design and procurement activities, and a licensee may proceed at its own risk with design, procurement, construction, and preoperational activities, even though the NRC may not have found that any particular ITAAC has been met.
            2. The licensee who references this appendix shall notify the NRC that the required inspections, tests, and analyses in the ITAAC have been successfully completed and that the corresponding acceptance criteria have been met.

            3. In the event that an activity is subject to an ITAAC, and the applicant or licensee who references this appendix has not demonstrated that the ITAAC has been met, the applicant or licensee may either take corrective actions to successfully complete that ITAAC, request an exemption from the ITAAC in accordance with Section VIII of this appendix and 10 CFR 52.97(b), or petition for rulemaking to amend this appendix by changing the requirements of the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes to the ITAAC must meet the requirements of paragraph VIII.A.1 of this appendix.

            B.1. The NRC shall ensure that the required inspections, tests, and analyses in the ITAAC are performed. The NRC shall verify that the inspections, tests, and analyses referenced by the licensee have been successfully completed and, based solely thereon, find the prescribed acceptance criteria have been met. At appropriate intervals during construction, the NRC shall publish notices of the successful completion of ITAAC in the Federal Register.

            2. In accordance with 10 CFR 52.103(g), the Commission shall find that the acceptance criteria in the ITAAC for the license are met before fuel load.
            3. After the Commission has made the finding required by 10 CFR 52.103(g), the ITAAC do not, by virtue of their inclusion within the DCD, constitute regulatory requirements either for licensees or for renewal of the license; except for specific ITAAC, which are the subject of a § 52.103(a) hearing, their expiration will occur upon final Commission action in such proceeding. However, subsequent modifications must comply with the Tier 1 and Tier 2 design descriptions in the plant-specific DCD unless the licensee has complied with the applicable requirements of 10 CFR 52.98 and Section VIII of this appendix.
            X. Records and Reporting
            A. Records.
            1. The applicants for this appendix shall maintain a copy of the applicable generic DCD that includes all generic changes to Tier 1, Tier 2, and the generic technical specifications and other operational requirements. The applicants shall maintain the sensitive unclassified non-safeguards information (including proprietary information) and safeguards information referenced in the applicable generic DCD for the period that this appendix may be referenced, as specified in Section VII of this appendix.
            2. An applicant or licensee who references this appendix shall maintain the plant-specific DCD to accurately reflect both generic changes to the generic DCD and plant-specific departures made under Section VIII of this appendix throughout the period of application and for the term of the license (including any period of renewal).
            3. An applicant or licensee who references this appendix shall prepare and maintain written evaluations which provide the bases for the determinations required by Section VIII of this appendix. These evaluations must be retained throughout the period of application and for the term of the license (including any period of renewal).
            4.a. The applicant for the amendment to the U.S. ABWR design to address the requirements in 10 CFR 50.150, “Aircraft impact assessment,” shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) for the term of the certification (including any period of renewal).
            b. An applicant or licensee who references this appendix to include both the GE DCD and the STPNOC DCD shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) throughout the pendency of the application and for the term of the license (including any period of renewal).
            B. Reporting.
            1. An applicant or licensee who references this appendix shall submit a report to the NRC containing a brief description of any plant-specific departures from the DCD, including a summary of the evaluation of each. This report must be filed in accordance with the filing requirements applicable to reports in 10 CFR 52.3.
            2. An applicant or licensee who references this appendix shall submit updates to its DCD, which reflect the generic changes and the plant-specific departures from the generic DCD made under Section VIII of this appendix. These updates must be filed under the filing requirements applicable to final safety analysis report updates in 10 CFR 52.3 and 50.71(e).
            3. The reports and updates required by paragraphs X.B.1 and X.B.2 must be submitted as follows:
            a. On the date that an application for a license referencing this appendix is submitted, the application must include the report and any updates to the generic DCD.
            b. During the interval from the date of application for a license to the date the Commission makes the finding required by 10 CFR 52.103(g), the report must be submitted semiannually. Updates to the plant-specific DCD must be submitted annually and may be submitted along with amendments to the application.
            c. After the Commission makes the finding required by 10 CFR 52.103(g), reports and updates to the plant-specific DCD must be submitted, along with updates to the site-specific portion of the final safety analysis report for the facility, at the intervals required by 10 CFR 50.59(d)(2) and 10 CFR 50.71(e)(4), respectively, or at shorter intervals as specified in the license.
            [72 FR 49517, Aug. 28, 2007, as amended at 76 FR 72085, Nov. 22, 2011; 76 FR 78119, Dec. 16, 2011; 84 FR 63568, Nov. 18, 2019]

            Pt. 52, App. B
            Appendix B to Part 52—Design Certification Rule for the System 80 + Design
            I. Introduction
            Appendix B constitutes design certification for the System 80 + 1
               standard plant design, in accordance with 10 CFR part 52, subpart B. The applicant for certification of the System 80 + design was Combustion Engineering, Inc. (ABB-CE), which is now Westinghouse Electric Company LLC.

                1 “System 80 + ” is a trademark of Westinghouse Electric Company LLC.

            II. Definitions
            A. Generic design control document (generic DCD) means the document containing the Tier 1 and Tier 2 information and generic technical specifications that is incorporated by reference into this appendix.
            B. Generic technical specifications means the information, required by 10 CFR 50.36 and 50.36a, for the portion of the plant that is within the scope of this appendix.
            C. Plant-specific DCD means the document, maintained by an applicant or licensee who references this appendix, consisting of the information in the generic DCD, as modified and supplemented by the plant-specific departures and exemptions made under Section VIII of this appendix.
            D. Tier 1 means the portion of the design-related information contained in the generic DCD that is approved and certified by this appendix (hereinafter Tier 1 information). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes:
            1. Definitions and general provisions;
            2. Design descriptions;
            3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
            4. Significant site parameters; and
            5. Significant interface requirements.
            E. Tier 2 means the portion of the design-related information contained in the generic DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these differences, an applicant or licensee must meet the requirement in Section III.B of this appendix to reference Tier 2 when referencing Tier 1. Tier 2 information includes:
            1. Information required by §§ 52.47(a) and 52.47(c), with the exception of generic technical specifications and conceptual design information;
            2. Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met; and
            3. Combined license (COL) action items (COL license information), which identify certain matters that must be addressed in the site-specific portion of the final safety analysis report (FSAR) by an applicant who references this appendix. These items constitute information requirements but are not the only acceptable set of information in the FSAR. An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the FSAR. After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the FSAR.
            F. Tier 2* means the portion of the Tier 2 information, designated as such in the generic DCD, which is subject to the change process in Section VIII.B.6 of this appendix. This designation expires for some Tier 2* information under Section VIII.B.6 of this appendix.
            G. Departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses means:
            (1) Changing any of the elements of the method described in the plant-specific DCD unless the results of the analysis are conservative or essentially the same; or
            (2) Changing from a method described in the plant-specific DCD to another method unless that method has been approved by NRC for the intended application.
            H. All other terms in this appendix have the meaning set out in 10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable.
            III. Scope and Contents

            A. Tier 1, Tier 2, and the generic technical specifications in the System 80 + Design Control Document, ABB-CE, with revisions dated January 1997, are approved for incorporation by reference by the Director of the Office of the Federal Register in accordance with 5 U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD may be obtained from the National Technical Information Service, 5285 Port Royal Road, Springfield, Virginia 22161. A copy is available for examination and copying at the NRC Public Document Room located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Copies are also available for examination at the NRC Library located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852 and the Office of the Federal Register, 800 North Capitol Street, NW., Suite 700, Washington, DC.

            B. An applicant or licensee referencing this appendix, in accordance with Section IV of this appendix, shall incorporate by reference and comply with the requirements of this appendix, including Tier 1, Tier 2, and the generic technical specifications except as otherwise provided in this appendix. Conceptual design information, as set forth in the generic DCD, and the Technical Support Document for the System 80 + design are not part of this appendix.
            C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then Tier 1 controls.
            D. If there is a conflict between the generic DCD and either the application for design certification of the System 80 + design or NUREG-1462, “Final Safety Evaluation Report Related to the Certification of the System 80 + Design,” (FSER) and Supplement No. 1, then the generic DCD controls.
            E. Design activities for structures, systems, and components that are wholly outside the scope of this appendix may be performed using site characteristics, provided the design activities do not affect the DCD or conflict with the interface requirements.
            IV. Additional Requirements and Restrictions
            A. An applicant for a combined license that wishes to reference this appendix shall, in addition to complying with the requirements of 10 CFR 52.77, 52.79, and 52.80, comply with the following requirements:
            1. Incorporate by reference, as part of its application, this appendix;
            2. Include, as part of its application:
            a. A plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the System 80 + design, as modified and supplemented by the applicant's exemptions and departures;
            b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
            c. Plant-specific technical specifications, consisting of the generic and site-specific technical specifications, that are required by 10 CFR 50.36 and 50.36a;
            d. Information demonstrating compliance with the site parameters and interface requirements;
            e. Information that addresses the COL action items; and
            f. Information required by 10 CFR 52.47 that is not within the scope of this appendix.
            3. Include, in the plant-specific DCD, the proprietary information referenced in the System 80 + DCD.
            B. The Commission reserves the right to determine in what manner this appendix may be referenced by an applicant for a construction permit or operating license under 10 CFR part 50.
            V. Applicable Regulations
            A. Except as indicated in paragraph B of this section, the regulations that apply to the System 80 + design are in 10 CFR parts 20, 50, 73, and 100, codified as of May 9, 1997, that are applicable and technically relevant, as described in the FSER (NUREG-1462) and Supplement No. 1.
            B. The System 80 + design is exempt from portions of the following regulations:
            1. Paragraph (f)(2)(iv) of 10 CFR 50.34—Separate Plant Safety Parameter Display Console;
            2. Paragraphs (f)(2) (vii), (viii), (xxvi), and (xxviii) of 10 CFR 50.34—Accident Source Terms;
            3. Paragraph (f)(2)(viii) of 10 CFR 50.34—Post-Accident Sampling for Hydrogen, Boron, Chloride, and Dissolved Gases;
            4. Paragraph (f)(3)(iv) of 10 CFR 50.34—Dedicated Containment Penetration; and
            5. Paragraphs III.A.1(a) and III.C.3(b) of Appendix J to 10 CFR 50—Containment Leakage Testing.
            VI. Issue Resolution
            A. The Commission has determined that the structures, systems, components, and design features of the System 80 + design comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in Section V of this appendix; and therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for the System 80 + design.
            B. The Commission considers the following matters resolved within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance of a combined license, amendment of a combined license, or renewal of a combined license, proceedings held under 10 CFR 52.103, and enforcement proceedings involving plants referencing this appendix:
            1. All nuclear safety issues, except for the generic technical specifications and other operational requirements, associated with the information in the FSER and Supplement No. 1, Tier 1, Tier 2 (including referenced information which the context indicates is intended as requirements), and the rulemaking record for certification of the System 80 + design;
            2. All nuclear safety and safeguards issues associated with the information in proprietary and safeguards documents, referenced and in context, are intended as requirements in the generic DCD for the System 80 + design;

            3. All generic changes to the DCD under and in compliance with the change processes in Sections VIII.A.1 and VIII.B.1 of this appendix;
            4. All exemptions from the DCD under and in compliance with the change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;
            5. All departures from the DCD that are approved by license amendment, but only for that plant;
            6. Except as provided in paragraph VIII.B.5.f of this appendix, all departures from Tier 2 under and in compliance with the change processes in paragraph VIII.B.5 of this appendix that do not require prior NRC approval, but only for that plant;
            7. All environmental issues concerning severe accident mitigation design alternatives associated with the information in the NRC's final environmental assessment for the System 80 + design and the technical support document for the System 80 + design, dated January 1995, for plants referencing this appendix whose site parameters are within those specified in the technical support document.
            C. The Commission does not consider operational requirements for an applicant or licensee who references this appendix to be matters resolved within the meaning of 10 CFR 52.63(a)(5). The Commission reserves the right to require operational requirements for an applicant or licensee who references this appendix by rule, regulation, order, or license condition.
            D. Except in accordance with the change processes in Section VIII of this appendix, the Commission may not require an applicant or licensee who references this appendix to:
            1. Modify structures, systems, components, or design features as described in the generic DCD;
            2. Provide additional or alternative structures, systems, components, or design features not discussed in the generic DCD; or
            3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria, or justification for structures, systems, components, or design features discussed in the generic DCD.
            E.1. Persons who wish to review proprietary information or other secondary references in the DCD for the System 80 + design, in order to request or participate in the hearing required by 10 CFR 52.85 or the hearing provided under 10 CFR 52.103, or to request or participate in any other hearing relating to this appendix in which interested persons have adjudicatory hearing rights, shall first request access to such information from Westinghouse. The request must state with particularity:
            a. The nature of the proprietary or other information sought;

            b. The reason why the information currently available to the public at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public Document Room, is insufficient;
            c. The relevance of the requested information to the hearing issue(s) which the person proposes to raise; and
            d. A showing that the requesting person has the capability to understand and utilize the requested information.

            2. If a person claims that the information is necessary to prepare a request for hearing, the request must be filed no later than 15 days after publication in the Federal Register of the notice required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse declines to provide the information sought, Westinghouse shall send a written response within ten (10) days of receiving the request to the requesting person setting forth with particularity the reasons for its refusal. The person may then request the Commission (or presiding officer, if a proceeding has been established) to order disclosure. The person shall include copies of the original request (and any subsequent clarifying information provided by the requesting party to the applicant) and the applicant's response. The Commission and presiding officer shall base their decisions solely on the person's original request (including any clarifying information provided by the requesting person to Westinghouse), and Westinghouse's response. The Commission and presiding officer may order Westinghouse to provide access to some or all of the requested information, subject to an appropriate non-disclosure agreement.
            VII. Duration of This Appendix
            This appendix may be referenced for a period of 15 years from June 20, 1997, except as provided for in 10 CFR 52.55(b) and 52.57(b). This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.
            VIII. Processes for Changes and Departures
            A. Tier 1 information.
            1. Generic changes to Tier 1 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 1 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs A.3 or A.4 of this section.
            3. Departures from Tier 1 information that are required by the Commission through plant-specific orders are governed by the requirements in 10 CFR 52.63(a)(4).

            4. Exemptions from Tier 1 information are governed by the requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will deny a request for an exemption from Tier 1, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.
            B. Tier 2 Information
            1. Generic changes to Tier 2 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
            3. The Commission may not require new requirements on Tier 2 information by plant-specific order while this appendix is in effect under §§ 52.55 or 52.61, unless:
            a. A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time this appendix was approved, as set forth in Section V of this appendix, or to assure adequate protection of the public health and safety or the common defense and security; and
            b. Special circumstances as defined in 10 CFR 52.7 are present.
            4. An applicant or licensee who references this appendix may request an exemption from Tier 2 information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7. The Commission will deny a request for an exemption from Tier 2, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design. The grant of an exemption to an applicant must be subject to litigation in the same manner as other issues material to the license hearing. The grant of an exemption to a licensee must be subject to an opportunity for a hearing in the same manner as license amendments.
            5.a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2* information, or the technical specifications, or requires a license amendment under paragraphs B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD.
            b. A proposed departure from Tier 2, other than one affecting resolution of a severe accident issue identified in the plant-specific DCD, requires a license amendment if it would—
            (1) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the plant-specific DCD;
            (2) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the plant-specific DCD;
            (3) Result in more than a minimal increase in the consequences of an accident previously evaluated in the plant-specific DCD;
            (4) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the plant-specific DCD;
            (5) Create a possibility for an accident of a different type than any evaluated previously in the plant-specific DCD;
            (6) Create a possibility for a malfunction of an SSC important to safety with a different result than any evaluated previously in the plant-specific DCD;
            (7) Result in a design basis limit for a fission product barrier as described in the plant-specific DCD being exceeded or altered; or
            (8) Result in a departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses.
            c. A proposed departure from Tier 2 affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if:
            (1) There is a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously reviewed and determined to be not credible could become credible; or
            (2) There is a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously reviewed.
            d. If a departure requires a license amendment under paragraph B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
            e. A departure from Tier 2 information that is made under paragraph B.5 of this section does not require an exemption from this appendix.

            f. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under 10 CFR 52.103(a), who believes that an applicant or licensee who references this appendix has not complied with paragraph VIII.B.5 of this appendix when departing from Tier 2 information, may petition the NRC to admit into the proceeding such a contention. In addition to compliance with the general requirements of 10 CFR 2.309, the petition must demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a 10 CFR 52.103 preoperational hearing, or that the change bears directly on the amendment request in the case of a hearing on a license amendment. Any other party may file a response. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. The Commission may admit such a contention if it determines the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5 of this appendix.
            6.a. An applicant who references this appendix may not depart from Tier 2* information, which is designated with italicized text or brackets and an asterisk in the generic DCD, without NRC approval. The departure will not be considered a resolved issue, within the meaning of Section VI of this appendix and 10 CFR 52.63(a)(5).
            b. A licensee who references this appendix may not depart from the following Tier 2* matters without prior NRC approval. A request for a departure will be treated as a request for a license amendment under 10 CFR 50.90.
            (1) Maximum fuel rod average burnup.
            (2) Control room human factors engineering.
            c. A licensee who references this appendix may not, before the plant first achieves full power following the finding required by 10 CFR 52.103(g), depart from the following Tier 2* matters except in accordance with paragraph B.6.b of this section. After the plant first achieves full power, the following Tier 2* matters revert to Tier 2 status and are thereafter subject to the departure provisions in paragraph B.5 of this section.
            (1) ASME Boiler & Pressure Vessel Code, Section III.
            (2) ACI 349 and ANSI/AISC N-690.
            (3) Motor-operated valves.
            (4) Equipment seismic qualification methods.
            (5) Piping design acceptance criteria.
            (6) Fuel and control rod design, except burnup limit.
            (7) Instrumentation and controls setpoint methodology.
            (8) Instrumentation and controls hardware and software changes.
            (9) Instrumentation and controls environmental qualification.
            (10) Seismic design criteria for non-seismic Category I structures.
            d. Departures from Tier 2* information that are made under paragraph B.6 of this section do not require an exemption from this appendix.
            C. Operational requirements.
            1. Generic changes to generic technical specifications and other operational requirements that were completely reviewed and approved in the design certification rulemaking and do not require a change to a design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Generic changes that do require a change to a design feature in the generic DCD are governed by the requirements in paragraphs A or B of this section.
            2. Generic changes to generic TS and other operational requirements are applicable to all applicants who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this section.
            3. The Commission may require plant-specific departures on generic technical specifications and other operational requirements that were completely reviewed and approved, provided a change to a design feature in the generic DCD is not required and special circumstances as defined in 10 CFR 2.335 are present. The Commission may modify or supplement generic technical specifications and other operational requirements that were not completely reviewed and approved or require additional technical specifications and other operational requirements on a plant-specific basis, provided a change to a design feature in the generic DCD is not required.
            4. An applicant who references this appendix may request an exemption from the generic technical specifications or other operational requirements. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7. The grant of an exemption must be subject to litigation in the same manner as other issues material to the license hearing.

            5. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under 10 CFR 52.103(a), who believes that an operational requirement approved in the DCD or a technical specification derived from the generic technical specifications must be changed may petition to admit into the proceeding such a contention. Such a petition must comply with the general requirements of 10 CFR 2.309 and must demonstrate why special circumstances as defined in 10 CFR 2.335 are present, or for compliance with the Commission's regulations in effect at the time this appendix was approved, as set forth in Section V of this appendix. Any other party may file a response thereto. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. All other issues with respect to the plant-specific technical specifications or other operational requirements are subject to a hearing as part of the license proceeding.

            6. After issuance of a license, the generic technical specifications have no further effect on the plant-specific technical specifications and changes to the plant-specific technical specifications will be treated as license amendments under 10 CFR 50.90.
            IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
            A.1 An applicant or licensee who references this appendix shall perform and demonstrate conformance with the ITAAC before fuel load. With respect to activities subject to an ITAAC, an applicant for a license may proceed at its own risk with design and procurement activities, and a licensee may proceed at its own risk with design, procurement, construction, and preoperational activities, even though the NRC may not have found that any particular ITAAC has been met.
            2. The licensee who references this appendix shall notify the NRC that the required inspections, tests, and analyses in the ITAAC have been successfully completed and that the corresponding acceptance criteria have been met.
            3. In the event that an activity is subject to an ITAAC, and the applicant or licensee who references this appendix has not demonstrated that the ITAAC has been met, the applicant or licensee may either take corrective actions to successfully complete that ITAAC, request an exemption from the ITAAC in accordance with Section VIII of this appendix and 10 CFR 52.97(b), or petition for rulemaking to amend this appendix by changing the requirements of the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes to the ITAAC must meet the requirements of Section VIII.A.1 of this appendix.

            B.1 The NRC shall ensure that the required inspections, tests, and analyses in the ITAAC are performed. The NRC shall verify that the inspections, tests, and analyses referenced by the licensee have been successfully completed and, based solely thereon, find the prescribed acceptance criteria have been met. At appropriate intervals during construction, the NRC shall publish notices of the successful completion of ITAAC in the Federal Register.

            2. In accordance with 10 CFR 52.103(g), the Commission shall find that the acceptance criteria in the ITAAC for the license are met before fuel load.
            3. After the Commission has made the finding required by 10 CFR 52.103(g), the ITAAC do not, by virtue of their inclusion within the DCD, constitute regulatory requirements either for licensees or for renewal of the license; except for specific ITAAC, which are the subject of a § 52.103(a) hearing, their expiration will occur upon final Commission action in such proceeding. However, subsequent modifications must comply with the Tier 1 and Tier 2 design descriptions in the plant-specific DCD unless the licensee has complied with the applicable requirements of 10 CFR 52.98 and Section VIII of this appendix.
            X. Records and Reporting
            A. Records.
            1. The applicant for this appendix shall maintain a copy of the generic DCD that includes all generic changes to Tier 1, Tier 2, and the generic TS and other operational requirements. The applicant shall maintain the proprietary and safeguards information referenced in the generic DCD for the period that this appendix may be referenced, as specified in Section VII of this appendix.
            2. An applicant or licensee who references this appendix shall maintain the plant-specific DCD to accurately reflect both generic changes to the generic DCD and plant-specific departures made under Section VIII of this appendix throughout the period of application and for the term of the license (including any period of renewal).
            3. An applicant or licensee who references this appendix shall prepare and maintain written evaluations which provide the bases for the determinations required by Section VIII of this appendix. These evaluations must be retained throughout the period of application and for the term of the license (including any period of renewal).
            B. Reporting.
            1. An applicant or licensee who references this appendix shall submit a report to the NRC containing a brief description of any plant-specific departures from the DCD, including a summary of the evaluation of each. This report must be filed in accordance with the filing requirements applicable to reports in 10 CFR 52.3.
            2. An applicant or licensee who references this appendix shall submit updates to its DCD, which reflect the generic changes to and plant-specific departures from the generic DCD made under Section VIII of this appendix. These updates must be filed under the filing requirements applicable to final safety analysis report updates in 10 CFR 52.3 and 50.71(e).
            3. The reports and updates required by paragraphs X.B.1 and X.B.2 must be submitted as follows:
            a. On the date that an application for a license referencing this appendix is submitted, the application must include the report and any updates to the generic DCD.
            b. During the interval from the date of application for a license to the date the Commission makes the finding required by 10 CFR 52.103(g), the report must be submitted semi-annually. Updates to the plant-specific DCD must be submitted annually and may be submitted along with amendments to the application.

            c. After the Commission makes the finding required by 10 CFR 52.103(g), the reports and updates to the plant-specific DCD must be submitted, along with updates to the site-specific portion of the final safety analysis report for the facility, at the intervals required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at shorter intervals as specified in the license.
            [72 FR 49517, Aug. 28, 2007, as amended at 76 FR 72085, Nov. 22, 2011; 84 FR 63568, Nov. 18, 2019]

            Pt. 52, App. C
            Appendix C to Part 52—Design Certification Rule for the AP600 Design
            I. Introduction
            Appendix C constitutes the standard design certification for the AP600 1
               design, in accordance with 10 CFR part 52, subpart B. The applicant for certification of the AP600 design is Westinghouse Electric Company LLC.

                1 AP600 is a trademark of Westinghouse Electric Company LLC.

            II. Definitions
            A. Generic design control document (generic DCD) means the document containing the Tier 1 and Tier 2 information and generic technical specifications that is incorporated by reference into this appendix.
            B. Generic technical specifications means the information, required by 10 CFR 50.36 and 50.36a, for the portion of the plant that is within the scope of this appendix.
            C. Plant-specific DCD means the document, maintained by an applicant or licensee who references this appendix, consisting of the information in the generic DCD, as modified and supplemented by the plant-specific departures and exemptions made under Section VIII of this appendix.
            D. Tier 1 means the portion of the design-related information contained in the generic DCD that is approved and certified by this appendix (hereinafter Tier 1 information). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes:
            1. Definitions and general provisions;
            2. Design descriptions;
            3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
            4. Significant site parameters; and
            5. Significant interface requirements.
            E. Tier 2 means the portion of the design-related information contained in the generic DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these differences, an applicant or licensee must meet the requirement in Section III.B of this appendix to reference Tier 2 when referencing Tier 1. Tier 2 information includes:
            1. Information required by §§ 52.47(a) and 52.47(c), with the exception of generic technical specifications and conceptual design information;
            2. Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met; and
            3. Combined license (COL) action items (COL license information), which identify certain matters that must be addressed in the site-specific portion of the final safety analysis report (FSAR) by an applicant who references this appendix. These items constitute information requirements but are not the only acceptable set of information in the FSAR. An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the FSAR. After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the FSAR.
            4. The investment protection short-term availability controls in Section 16.3 of the DCD.
            F. Tier 2* means the portion of the Tier 2 information, designated as such in the generic DCD, which is subject to the change process in Section VIII.B.6 of this appendix. This designation expires for some Tier 2* information under Section VIII.B.6.
            G. Departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses means:
            (1) Changing any of the elements of the method described in the plant-specific DCD unless the results of the analysis are conservative or essentially the same; or
            (2) Changing from a method described in the plant-specific DCD to another method unless that method has been approved by NRC for the intended application.
            H. All other terms in this appendix have the meaning set out in 10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable.
            III. Scope and Contents

            A. Tier 1, Tier 2 (including the investment protection short-term availability controls in Section 16.3), and the generic technical specifications in the AP600 DCD (12/99 revision) are approved for incorporation by reference by the Director of the Office of the Federal Register on January 24, 2000, in accordance with 5 U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD may be obtained from Ronald P. Vijuk, Manager, Passive Plant Engineering, Westinghouse Electric Company, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355. A copy of the generic DCD is available for examination and copying at the NRC Public Document Room located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Copies are also available for examination at the NRC Library located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852; and the Office of the Federal Register, 800 North Capitol Street, NW., Suite 700, Washington, DC.
            B. An applicant or licensee referencing this appendix, in accordance with Section IV of this appendix, shall incorporate by reference and comply with the requirements of this appendix, including Tier 1, Tier 2 (including the investment protection short-term availability controls in Section 16.3), and the generic technical specifications except as otherwise provided in this appendix. Conceptual design information in the generic DCD and the evaluation of severe accident mitigation design alternatives in Appendix 1B of the generic DCD are not part of this appendix.
            C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then Tier 1 controls.
            D. If there is a conflict between the generic DCD and either the application for design certification of the AP600 design or NUREG-1512, “Final Safety Evaluation Report Related to Certification of the AP600 Standard Design,” (FSER), then the generic DCD controls.
            E. Design activities for structures, systems, and components that are wholly outside the scope of this appendix may be performed using site characteristics, provided the design activities do not affect the DCD or conflict with the interface requirements.
            IV. Additional Requirements and Restrictions
            A. An applicant for a combined license that wishes to reference this appendix shall, in addition to complying with the requirements of 10 CFR 52.77, 52.79, and 52.80, comply with the following requirements:
            1. Incorporate by reference, as part of its application, this appendix;
            2. Include, as part of its application:
            a. A plant-specific DCD containing the same type of information and utilizing the same organization and numbering as the generic DCD for the AP600 design, as modified and supplemented by the applicant's exemptions and departures;
            b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
            c. Plant-specific technical specifications, consisting of the generic and site-specific technical specifications, that are required by 10 CFR 50.36 and 50.36a;
            d. Information demonstrating compliance with the site parameters and interface requirements;
            e. Information that addresses the COL action items; and
            f. Information required by 10 CFR 52.47 that is not within the scope of this appendix.
            3. Include, in the plant-specific DCD, the proprietary information and safeguards information referenced in the AP600 DCD.
            B. The Commission reserves the right to determine in what manner this appendix may be referenced by an applicant for a construction permit or operating license under 10 CFR part 50.
            V. Applicable Regulations
            A. Except as indicated in paragraph B of this section, the regulations that apply to the AP600 design are in 10 CFR parts 20, 50, 73, and 100, codified as of December 16, 1999, that are applicable and technically relevant, as described in the FSER (NUREG-1512) and the supplementary information for this section.
            B. The AP600 design is exempt from portions of the following regulations:
            1. Paragraph (a)(1) of 10 CFR 50.34—whole body dose criterion;
            2. Paragraph (f)(2)(iv) of 10 CFR 50.34—Plant Safety Parameter Display Console;
            3. Paragraphs (f)(2)(vii), (viii), (xxvi), and (xxviii) of 10 CFR 50.34—Accident Source Term in TID 14844;
            4. Paragraph (a)(2) of 10 CFR 50.55a—ASME Boiler and Pressure Vessel Code;
            5. Paragraph (c)(1) of 10 CFR 50.62—Auxiliary (or emergency) feedwater system;
            6. Appendix A to 10 CFR part 50, GDC 17—Offsite Power Sources; and
            7. Appendix A to 10 CFR part 50, GDC 19—whole body dose criterion.
            VI. Issue Resolution
            A. The Commission has determined that the structures, systems, components, and design features of the AP600 design comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in Section V of this appendix; and therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for the AP600 design.

            B. The Commission considers the following matters resolved within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance of a combined license, amendment of a combined license, or renewal of a combined license, proceedings held under 10 CFR 52.103, and enforcement proceedings involving plants referencing this appendix:

            1. All nuclear safety issues, except for the generic technical specifications and other operational requirements, associated with the information in the FSER and Supplement No. 1, Tier 1, Tier 2 (including referenced information which the context indicates is intended as requirements and the investment protection short-term availability controls in Section 16.3), and the rulemaking record for certification of the AP600 design;
            2. All nuclear safety and safeguards issues associated with the information in proprietary and safeguards documents, referenced and in context, are intended as requirements in the generic DCD for the AP600 design;
            3. All generic changes to the DCD under and in compliance with the change processes in Sections VIII.A.1 and VIII.B.1 of this appendix;
            4. All exemptions from the DCD under and in compliance with the change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;
            5. All departures from the DCD that are approved by license amendment, but only for that plant;
            6. Except as provided in paragraph VIII.B.5.f of this appendix, all departures from Tier 2 under and in compliance with the change processes in paragraph VIII.B.5 of this appendix that do not require prior NRC approval, but only for that plant;
            7. All environmental issues concerning severe accident mitigation design alternatives associated with the information in the NRC's environmental assessment for the AP600 design and appendix 1B of the generic DCD, for plants referencing this appendix whose site parameters are within those specified in the severe accident mitigation design alternatives evaluation.
            C. The Commission does not consider operational requirements for an applicant or licensee who references this appendix to be matters resolved within the meaning of 10 CFR 52.63(a)(5). The Commission reserves the right to require operational requirements for an applicant or licensee who references this appendix by rule, regulation, order, or license condition.
            D. Except in accordance with the change processes in Section VIII of this appendix, the Commission may not require an applicant or licensee who references this appendix to:
            1. Modify structures, systems, components, or design features as described in the generic DCD;
            2. Provide additional or alternative structures, systems, components, or design features not discussed in the generic DCD; or
            3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria, or justification for structures, systems, components, or design features discussed in the generic DCD.
            E.1. Persons who wish to review proprietary and safeguards information or other secondary references in the AP600 DCD, in order to request or participate in the hearing required by 10 CFR 52.85 or the hearing provided under 10 CFR 52.103, or to request or participate in any other hearing relating to this appendix in which interested persons have adjudicatory hearing rights, shall first request access to such information from Westinghouse. The request must state with particularity:
            a. The nature of the proprietary or other information sought;

            b. The reason why the information currently available to the public at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public Document Room, is insufficient;
            c. The relevance of the requested information to the hearing issue(s) which the person proposes to raise; and
            d. A showing that the requesting person has the capability to understand and utilize the requested information.

            2. If a person claims that the information is necessary to prepare a request for hearing, the request must be filed no later than 15 days after publication in the Federal Register of the notice required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse declines to provide the information sought, Westinghouse shall send a written response within 10 days of receiving the request to the requesting person setting forth with particularity the reasons for its refusal. The person may then request the Commission (or presiding officer, if a proceeding has been established) to order disclosure. The person shall include copies of the original request (and any subsequent clarifying information provided by the requesting party to the applicant) and the applicant's response. The Commission and presiding officer shall base their decisions solely on the person's original request (including any clarifying information provided by the requesting person to Westinghouse), and Westinghouse's response. The Commission and presiding officer may order Westinghouse to provide access to some or all of the requested information, subject to an appropriate non-disclosure agreement.
            VII. Duration of This Appendix
            This appendix may be referenced for a period of 15 years from January 24, 2000, except as provided for in 10 CFR 52.55(b) and 52.57(b). This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.
            VIII. Processes for Changes and Departures
            A. Tier 1 information.

            1. Generic changes to Tier 1 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 1 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs A.3 or A.4 of this section.
            3. Departures from Tier 1 information that are required by the Commission through plant-specific orders are governed by the requirements in 10 CFR 52.63(a)(4).
            4. Exemptions from Tier 1 information are governed by the requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will deny a request for an exemption from Tier 1, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.
            B. Tier 2 information.
            1. Generic changes to Tier 2 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
            3. The Commission may not require new requirements on Tier 2 information by plant-specific order while this appendix is in effect under §§ 52.55 or 52.61, unless:
            a. A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time this appendix was approved, as set forth in Section V of this appendix, or to assure adequate protection of the public health and safety or the common defense and security; and
            b. Special circumstances as defined in 10 CFR 52.7 are present.
            4. An applicant or licensee who references this appendix may request an exemption from Tier 2 information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7. The Commission will deny a request for an exemption from Tier 2, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design. The grant of an exemption to an applicant must be subject to litigation in the same manner as other issues material to the license hearing. The grant of an exemption to a licensee must be subject to an opportunity for a hearing in the same manner as license amendments.
            5.a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2* information, or the technical specifications, or requires a license amendment under paragraphs B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD.
            b. A proposed departure from Tier 2, other than one affecting resolution of a severe accident issue identified in the plant-specific DCD, requires a license amendment if it would:
            (1) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the plant-specific DCD;
            (2) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the plant-specific DCD;
            (3) Result in more than a minimal increase in the consequences of an accident previously evaluated in the plant-specific DCD;
            (4) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the plant-specific DCD;
            (5) Create a possibility for an accident of a different type than any evaluated previously in the plant-specific DCD;
            (6) Create a possibility for a malfunction of an SSC important to safety with a different result than any evaluated previously in the plant-specific DCD;
            (7) Result in a design basis limit for a fission product barrier as described in the plant-specific DCD being exceeded or altered; or
            (8) Result in a departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses.
            c. A proposed departure from Tier 2 affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if:
            (1) There is a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously reviewed and determined to be not credible could become credible; or
            (2) There is a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously reviewed.
            d. If a departure requires a license amendment under paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
            e. A departure from Tier 2 information that is made under paragraph B.5 of this section does not require an exemption from this appendix.

            f. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under 10 CFR 52.103(a), who believes that an applicant or licensee who references this appendix has not complied with paragraph VIII.B.5 of this appendix when departing from Tier 2 information, may petition the NRC to admit into the proceeding such a contention. In addition to compliance with the general requirements of 10 CFR 2.309, the petition must demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a 10 CFR 52.103 preoperational hearing, or that the change bears directly on the amendment request in the case of a hearing on a license amendment. Any other party may file a response. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. The Commission may admit such a contention if it determines the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5 of this appendix.
            6a. An applicant who references this appendix may not depart from Tier 2* information, which is designated with italicized text or brackets and an asterisk in the generic DCD, without NRC approval. The departure will not be considered a resolved issue, within the meaning of Section VI of this appendix and 10 CFR 52.63(a)(5).
            b. A licensee who references this appendix may not depart from the following Tier 2* matters without prior NRC approval. A request for a departure will be treated as a request for a license amendment under 10 CFR 50.90.
            (1) Maximum fuel rod average burn-up.
            (2) Fuel principal design requirements.
            (3) Fuel criteria evaluation process.
            (4) Fire areas.
            (5) Human factors engineering.
            c. A licensee who references this appendix may not, before the plant first achieves full power following the finding required by 10 CFR 52.103(g), depart from the following Tier 2* matters except in accordance with paragraph B.6.b of this section. After the plant first achieves full power, the following Tier 2* matters revert to Tier 2 status and are thereafter subject to the departure provisions in paragraph B.5 of this section.
            (1) Nuclear Island structural dimensions.
            (2) ASME Boiler and Pressure Vessel Code, Section III, and Code Case—284.
            (3) Design Summary of Critical Sections.
            (4) ACI 318, ACI 349, and ANSI/AISC N—690.
            (5) Definition of critical locations and thicknesses.
            (6) Seismic qualification methods and standards.
            (7) Nuclear design of fuel and reactivity control system, except burn-up limit.
            (8) Motor-operated and power-operated valves.
            (9) Instrumentation and control system design processes, methods, and standards.
            (10) PRHR natural circulation test (first plant only).
            (11) ADS and CMT verification tests (first three plants only).
            d. Departures from Tier 2* information that are made under paragraph B.6 of this section do not require an exemption from this appendix.
            C. Operational requirements.
            1. Generic changes to generic technical specifications and other operational requirements that were completely reviewed and approved in the design certification rulemaking and do not require a change to a design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Generic changes that do require a change to a design feature in the generic DCD are governed by the requirements in paragraphs A or B of this section.
            2. Generic changes to generic TS and other operational requirements are applicable to all applicants who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this section.
            3. The Commission may require plant-specific departures on generic technical specifications and other operational requirements that were completely reviewed and approved, provided a change to a design feature in the generic DCD is not required and special circumstances as defined in 10 CFR 2.335 are present. The Commission may modify or supplement generic technical specifications and other operational requirements that were not completely reviewed and approved or require additional technical specifications and other operational requirements on a plant-specific basis, provided a change to a design feature in the generic DCD is not required.
            4. An applicant who references this appendix may request an exemption from the generic technical specifications or other operational requirements. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7. The grant of an exemption must be subject to litigation in the same manner as other issues material to the license hearing.

            5. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under 10 CFR 52.103(a), who believes that an operational requirement approved in the DCD or a technical specification derived from the generic technical specifications must be changed may petition to admit into the proceeding such a contention. Such petition must comply with the general requirements of 10 CFR 2.309 and must demonstrate why special circumstances as defined in 10 CFR 2.335 are present, or for compliance with the Commission's regulations in effect at the time this appendix was approved, as set forth in Section V of this appendix. Any other party may file a response thereto. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. All other issues with respect to the plant-specific technical specifications or other operational requirements are subject to a hearing as part of the license proceeding.
            6. After issuance of a license, the generic technical specifications have no further effect on the plant-specific technical specifications and changes to the plant-specific technical specifications will be treated as license amendments under 10 CFR 50.90.
            IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
            A.1 An applicant or licensee who references this appendix shall perform and demonstrate conformance with the ITAAC before fuel load. With respect to activities subject to an ITAAC, an applicant for a license may proceed at its own risk with design and procurement activities, and a licensee may proceed at its own risk with design, procurement, construction, and preoperational activities, even though the NRC may not have found that any particular ITAAC has been met.
            2. The licensee who references this appendix shall notify the NRC that the required inspections, tests, and analyses in the ITAAC have been successfully completed and that the corresponding acceptance criteria have been met.
            3. In the event that an activity is subject to an ITAAC, and the applicant or licensee who references this appendix has not demonstrated that the ITAAC has been met, the applicant or licensee may either take corrective actions to successfully complete that ITAAC, request an exemption from the ITAAC in accordance with Section VIII of this appendix and 10 CFR 52.97(b), or petition for rulemaking to amend this appendix by changing the requirements of the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes to the ITAAC must meet the requirements of paragraph VIII.A.1 of this appendix.

            B.1. The NRC shall ensure that the required inspections, tests, and analyses in the ITAAC are performed. The NRC shall verify that the inspections, tests, and analyses referenced by the licensee have been successfully completed and, based solely thereon, find the prescribed acceptance criteria have been met. At appropriate intervals during construction, the NRC shall publish notices of the successful completion of ITAAC in the Federal Register.

            2. In accordance with 10 CFR 52.103(g), the Commission shall find that the acceptance criteria in the ITAAC for the license are met before fuel load.
            3. After the Commission has made the finding required by 10 CFR 52.103(g), the ITAAC do not, by virtue of their inclusion within the DCD, constitute regulatory requirements either for licensees or for renewal of the license; except for specific ITAAC, which are the subject of a § 52.103(a) hearing, their expiration will occur upon final Commission action in such proceeding. However, subsequent modifications must comply with the Tier 1 and Tier 2 design descriptions in the plant-specific DCD unless the licensee has complied with the applicable requirements of 10 CFR 52.98 and Section VIII of this appendix.
            X. Records and Reporting
            A. Records.
            1. The applicant for this appendix shall maintain a copy of the generic DCD that includes all generic changes to Tier 1, Tier 2, and the generic TS and other operational requirements. The applicant shall maintain the proprietary and safeguards information referenced in the generic DCD for the period that this appendix may be referenced, as specified in Section VII of this appendix.
            2. An applicant or licensee who references this appendix shall maintain the plant-specific DCD to accurately reflect both generic changes to the generic DCD and plant-specific departures made under Section VIII of this appendix throughout the period of application and for the term of the license (including any period of renewal).
            3. An applicant or licensee who references this appendix shall prepare and maintain written evaluations which provide the bases for the determinations required by Section VIII of this appendix. These evaluations must be retained throughout the period of application and for the term of the license (including any period of renewal).
            B. Reporting.
            1. An applicant or licensee who references this appendix shall submit a report to the NRC containing a brief description of any plant-specific departures from the DCD, including a summary of the evaluation of each. This report must be filed in accordance with the filing requirements applicable to reports in 10 CFR 52.3.

            2. An applicant or licensee who references this appendix shall submit updates to its DCD, which reflect the generic changes to and plant-specific departures from the generic DCD made under Section VIII of this appendix. These updates must be filed under the filing requirements applicable to final safety analysis report updates in 10 CFR 52.3 and 50.71(e).

            3. The reports and updates required by paragraphs X.B.1 and X.B.2 must be submitted as follows:
            a. On the date that an application for a license referencing this appendix is submitted, the application must include the report and any updates to the generic DCD.
            b. During the interval from the date of application for a license to the date the Commission makes the finding required by 10 CFR 52.103(g), the report must be submitted semi-annually. Updates to the plant-specific DCD must be submitted annually and may be submitted along with amendments to the application.
            c. After the Commission makes the finding required by 10 CFR 52.103(g), the reports and updates to the plant-specific DCD must be submitted, along with updates to the site-specific portion of the final safety analysis report for the facility, at the intervals required by 10 CFR 50.59(d)(2) and 50.71(e), respectively, or at shorter intervals as specified in the license.
            [72 FR 49517, Aug. 28, 2007, as amended at 76 FR 72085, Nov. 22, 2011; 84 FR 63568, Nov. 18, 2019]

            Pt. 52, App. D
            Appendix D to Part 52—Design Certification Rule for the AP1000 Design
            I. Introduction
            Appendix D constitutes the standard design certification for the AP1000 1
               design, in accordance with 10 CFR part 52, subpart B. The applicant for certification of the AP1000 design is Westinghouse Electric Company LLC.

                1 AP1000 is a trademark of Westinghouse Electric Company LLC.

            II. Definitions
            A. Generic design control document (generic DCD) means the document containing the Tier 1 and Tier 2 information and generic technical specifications that is incorporated by reference into this appendix.
            B. Generic technical specifications means the information required by 10 CFR 50.36 and 50.36a for the portion of the plant that is within the scope of this appendix.
            C. Plant-specific DCD means the document maintained by an applicant or licensee who references this appendix consisting of the information in the generic DCD as modified and supplemented by the plant-specific departures and exemptions made under Section VIII of this appendix.
            D. Tier 1 means the portion of the design-related information contained in the generic DCD that is approved and certified by this appendix (Tier 1 information). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes:
            1. Definitions and general provisions;
            2. Design descriptions;
            3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
            4. Significant site parameters; and
            5. Significant interface requirements.
            E. Tier 2 means the portion of the design-related information contained in the generic DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these differences, an applicant or licensee must meet the requirement in Section III.B of this appendix to reference Tier 2 when referencing Tier 1. Tier 2 information includes:
            1. Information required by §§ 52.47(a) and 52.47(c), with the exception of generic technical specifications and conceptual design information;
            2. Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met; and
            3. Combined license (COL) action items (COL license information), which identify certain matters that must be addressed in the site-specific portion of the final safety analysis report (FSAR) by an applicant who references this appendix. These items constitute information requirements but are not the only acceptable set of information in the FSAR. An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the FSAR. After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the FSAR.
            4. The investment protection short-term availability controls in Section 16.3 of the DCD.
            F. Tier 2* means the portion of the Tier 2 information, designated as such in the generic DCD, which is subject to the change process in Section VIII.B.6 of this appendix. This designation expires for some Tier 2* information under paragraph VIII.B.6.
            G. Departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses means:
            1. Changing any of the elements of the method described in the plant-specific DCD unless the results of the analysis are conservative or essentially the same; or

            2. Changing from a method described in the plant-specific DCD to another method unless that method has been approved by the NRC for the intended application.
            H. All other terms in this appendix have the meaning set out in 10 CFR 50.2, or 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable.
            III. Scope and Contents

            A. Tier 1, Tier 2 (including the investment protection short-term availability controls in Section 16.3), and the generic TSs in the AP1000 Design Control Document, Revision 19, (Public Version) (AP1000 DCD), APP-GW-GL-702, dated June 13, 2011, are approved for incorporation by reference by the Director of the Office of the Federal Register under 5 U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD may be obtained from Stanley E. Ritterbusch, Manager, AP1000 Design Certification, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066, telephone (412) 374-3037. A copy of the generic DCD is also available for examination and copying at the NRC's PDR, Room O-1F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. Copies are available for examination at the NRC Library, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852, telephone (301) 415-5610, email LIBRARY.RESOURCE@NRC.GOV. The DCD can also be viewed online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html by searching under ADAMS Accession No. ML11171A500. All approved material is available for inspection at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call (202) 741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.

            B. An applicant or licensee referencing this appendix, in accordance with Section IV of this appendix, shall incorporate by reference and comply with the requirements of this appendix, including Tier 1, Tier 2 (including the investment protection short-term availability controls in Section 16.3 of the DCD), and the generic TS except as otherwise provided in this appendix. Conceptual design information in the generic DCD and the evaluation of severe accident mitigation design alternatives in appendix 1B of the generic DCD are not part of this appendix.
            C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then Tier 1 controls.
            D.1. If there is a conflict between the generic DCD and either the application for the initial design certification of the AP1000 design or NUREG-1793, “Final Safety Evaluation Report Related to Certification of the Westinghouse Standard Design,” and Supplement No. 1, then the generic DCD controls.
            2. If there is a conflict between the generic DCD and either the application for Amendment 1 to the design certification of the AP1000 design or NUREG-1793, “Final Safety Evaluation Report Related to Certification of the Westinghouse Standard Design,” Supplement No. 2, then the generic DCD controls.
            E. Design activities for structures, systems, and components that are wholly outside the scope of this appendix may be performed using site characteristics, provided the design activities do not affect the DCD or conflict with the interface requirements.
            IV. Additional Requirements and Restrictions
            A. An applicant for a combined license that wishes to reference this appendix shall, in addition to complying with the requirements of 10 CFR 52.77, 52.79, and 52.80, comply with the following requirements:
            1. Incorporate by reference, as part of its application, this appendix.
            2. Include, as part of its application:
            a. A plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the AP1000 design, as modified and supplemented by the applicant's exemptions and departures;
            b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
            c. Plant-specific TS, consisting of the generic and site-specific TS that are required by 10 CFR 50.36 and 50.36a;
            d. Information demonstrating compliance with the site parameters and interface requirements;
            e. Information that addresses the COL action items; and
            f. Information required by 10 CFR 52.47(a) that is not within the scope of this appendix.
            3. Include, in the plant-specific DCD, the sensitive unclassified non-safeguards information (including proprietary information) and safeguards information referenced in the AP1000 DCD.
            4. Include, as part of its application, a demonstration that an entity other than Westinghouse is qualified to supply the AP1000 design, unless Westinghouse supplies the design for the applicant's use.
            B. The Commission reserves the right to determine in what manner this appendix may be referenced by an applicant for a construction permit or operating license under 10 CFR part 50.
            V. Applicable Regulations
            A.1. Except as indicated in paragraph B of this section, the regulations that apply to the AP1000 design are in 10 CFR parts 20, 50, 73, and 100, codified as of January 23, 2006, that are applicable and technically relevant, as described in the FSER (NUREG-1793) and Supplement No. 1.

            2. The regulations that apply to those portions of the AP1000 design approved by Amendment 1 are in 10 CFR parts 20, 50, 73, and 100, codified as of December 30, 2011, that are applicable and technically relevant, as described in the Supplement No. 2 of the FSER (NUREG-1793).
            B. The AP1000 design is exempt from portions of the following regulations:
            1. Paragraph (f)(2)(iv) of 10 CFR 50.34—Plant Safety Parameter Display Console;
            2. Paragraph (c)(1) of 10 CFR 50.62—Auxiliary (or emergency) feedwater system; and
            3. Appendix A to 10 CFR part 50, GDC 17—Second offsite power supply circuit.
            VI. Issue Resolution
            A. The Commission has determined that the structures, systems, components, and design features of the AP1000 design comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in Section V of this appendix; and therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for the AP1000 design.
            B. The Commission considers the following matters resolved within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance of a COL, amendment of a COL, or renewal of a COL, proceedings held under 10 CFR 52.103, and enforcement proceedings involving plants referencing this appendix:
            1. All nuclear safety issues, except for the generic TS and other operational requirements, associated with the information in the FSER and Supplement Nos. 1 and 2, Tier 1, Tier 2 (including referenced information, which the context indicates is intended as requirements, and the investment protection short-term availability controls in Section 16.3 of the DCD), and the rulemaking records for initial certification and Amendment 1 of the AP1000 design;
            2. All nuclear safety and safeguards issues associated with the referenced sensitive unclassified non-safeguards information (including proprietary information) and safeguards information which, in context, are intended as requirements in the generic DCD for the AP1000 design;
            3. All generic changes to the DCD under and in compliance with the change processes in Sections VIII.A.1 and VIII.B.1 of this appendix;
            4. All exemptions from the DCD under and in compliance with the change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;
            5. All departures from the DCD that are approved by license amendment, but only for that plant;
            6. Except as provided in paragraph VIII.B.5.g of this appendix, all departures from Tier 2 under and in compliance with the change processes in paragraph VIII.B.5 of this appendix that do not require prior NRC approval, but only for that plant;
            7. All environmental issues concerning severe accident mitigation design alternatives associated with the information in the NRC's EA for the AP1000 design, Appendix 1B of Revision 15 of the generic DCD, the NRC's final EA for Amendment 1 to the AP1000 design, and Appendix 1B of Revision 19 of the generic DCD, for plants referencing this appendix whose site parameters are within those specified in the severe accident mitigation design alternatives evaluation.
            C. The Commission does not consider operational requirements for an applicant or licensee who references this appendix to be matters resolved within the meaning of 10 CFR 52.63(a)(5). The Commission reserves the right to require operational requirements for an applicant or licensee who references this appendix by rule, regulation, order, or license condition.
            D. Except under the change processes in Section VIII of this appendix, the Commission may not require an applicant or licensee who references this appendix to:
            1. Modify structures, systems, components, or design features as described in the generic DCD;
            2. Provide additional or alternative structures, systems, components, or design features not discussed in the generic DCD; or
            3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria, or justification for structures, systems, components, or design features discussed in the generic DCD.
            E. The NRC will specify at an appropriate time the procedures to be used by an interested person who wishes to review portions of the design certification or references containing safeguards information or sensitive unclassified non-safeguards information (including proprietary information, such as trade secrets or financial information obtained from a person that are privileged or confidential (10 CFR 2.390 and 10 CFR part 9)), for the purpose of participating in the hearing required by 10 CFR 52.85, the hearing provided under 10 CFR 52.103, or in any other proceeding relating to this appendix in which interested persons have a right to request an adjudicatory hearing.
            VII. Duration of This Appendix

            This appendix may be referenced for a period of 15 years from February 27, 2006, except as provided for in 10 CFR 52.55(b) and 52.57(b). This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.

            VIII. Processes for Changes and Departures
            A. Tier 1 information.
            1. Generic changes to Tier 1 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 1 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs A.3 or A.4 of this section.
            3. Departures from Tier 1 information that are required by the Commission through plant-specific orders are governed by the requirements in 10 CFR 52.63(a)(4).
            4. Exemptions from Tier 1 information are governed by the requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will deny a request for an exemption from Tier 1, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.
            B. Tier 2 information.
            1. Generic changes to Tier 2 information are governed by the requirements in 10 CFR 52.63(a)(1).
            2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
            3. The Commission may not require new requirements on Tier 2 information by plant-specific order while this appendix is in effect under 10 CFR 52.55 or 52.61, unless:
            a. A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time this appendix was approved, as set forth in Section V of this appendix, or to ensure adequate protection of the public health and safety or the common defense and security; and
            b. Special circumstances as defined in 10 CFR 50.12(a) are present.
            4. An applicant or licensee who references this appendix may request an exemption from Tier 2 information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 50.12(a). The Commission will deny a request for an exemption from Tier 2, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design. The grant of an exemption to an applicant must be subject to litigation in the same manner as other issues material to the license hearing. The grant of an exemption to a licensee must be subject to an opportunity for a hearing in the same manner as license amendments.
            5.a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2* information, or the TS, or requires a license amendment under paragraphs B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD.
            b. A proposed departure from Tier 2, other than one affecting resolution of a severe accident issue identified in the plant-specific DCD or one affecting information required by 10 CFR52.47(a)(28) to address 10 CFR 50.150, requires a license amendment if it would:
            (1) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the plant-specific DCD;
            (2) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety and previously evaluated in the plant-specific DCD;
            (3) Result in more than a minimal increase in the consequences of an accident previously evaluated in the plant-specific DCD;
            (4) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the plant-specific DCD;
            (5) Create a possibility for an accident of a different type than any evaluated previously in the plant-specific DCD;
            (6) Create a possibility for a malfunction of an SSC important to safety with a different result than any evaluated previously in the plant-specific DCD;
            (7) Result in a design basis limit for a fission product barrier as described in the plant-specific DCD being exceeded or altered; or
            (8) Result in a departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses.
            c. A proposed departure from Tier 2 affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if:
            (1) There is a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously reviewed and determined to be not credible could become credible; or
            (2) There is a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously reviewed.

            d. If an applicant or licensee proposes to depart from the information required by 10 CFR 52.47(a)(28) to be included in the FSAR for the standard design certification, then the applicant or licensee shall consider the effect of the changed feature or capability on the original assessment required by 10 CFR 50.150(a). The applicant or licensee must also document how the modified design features and functional capabilities continue to meet the assessment requirements in 10 CFR 50.150(a)(1) in accordance with Section X of this appendix.
            e. If a departure requires a license amendment under paragraph B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
            f. A departure from Tier 2 information that is made under paragraph B.5 of this section does not require an exemption from this appendix.
            g. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under 10 CFR 52.103(a), who believes that an applicant or licensee who references this appendix has not complied with paragraph VIII.B.5 of this appendix when departing from Tier 2 information, may petition to admit into the proceeding such a contention. In addition to compliance with the general requirements of 10 CFR 2.309, the petition must demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a 10 CFR 52.103 preoperational hearing, or that the change bears directly on the amendment request in the case of a hearing on a license amendment. Any other party may file a response. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. The Commission may admit such a contention if it determines the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5 of this appendix.
            6.a. An applicant who references this appendix may not depart from Tier 2* information, which is designated with italicized text or brackets and an asterisk in the generic DCD, without NRC approval. The departure will not be considered a resolved issue, within the meaning of Section VI of this appendix and 10 CFR 52.63(a)(5).
            b. A licensee who references this appendix may not depart from the following Tier 2* matters without prior NRC approval. A request for a departure will be treated as a request for a license amendment under 10 CFR 50.90.
            (1) Maximum fuel rod average burn-up.
            (2) Fuel principal design requirements.
            (3) Fuel criteria evaluation process.
            (4) Fire areas.
            (5) Reactor coolant pump type.
            (6) Small-break loss-of-coolant accident (LOCA) analysis methodology.
            (7) Screen design criteria.
            (8) Heat sink data for containment pressure analysis.
            c. A licensee who references this appendix may not, before the plant first achieves full power following the finding required by 10 CFR 52.103(g), depart from the following Tier 2* matters except under paragraph B.6.b of this section. After the plant first achieves full power, the following Tier 2* matters revert to Tier 2 status and are subject to the departure provisions in paragraph B.5 of this section.
            (1) Nuclear Island structural dimensions.
            (2) American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME Code) piping design and welding restrictions, and ASME Code Cases.
            (3) Design Summary of Critical Sections.
            (4) American Concrete Institute (ACI) 318, ACI 349, American National Standards Institute/American Institute of Steel Construction (ANSI/AISC)N-690, and American Iron and Steel Institute (AISI), “Specification for the Design of Cold Formed Steel Structural Members, Part 1 and 2,” 1996 Edition and 2000 Supplement.
            (5) Definition of critical locations and thicknesses.
            (6) Seismic qualification methods and standards.
            (7) Nuclear design of fuel and reactivity control system, except burn-up limit.
            (8) Motor-operated and power-operated valves.
            (9) Instrumentation and control system design processes, methods, and standards.
            (10) Passive residual heat removal (PRHR) natural circulation test (first plant only).
            (11) Automatic depressurization system (ADS) and core make-up tank (CMT) verification tests (first three plants only).
            (12) Polar crane parked orientation.
            (13) Piping design acceptance criteria.
            (14) Containment vessel design parameters, including ASME Code, Section III, Subsection NE.
            (15) Human factors engineering.
            (16) Steel composite structural module details.
            d. Departures from Tier 2* information that are made under paragraph B.6 of this section do not require an exemption from this appendix.
            C. Operational requirements.
            1. Generic changes to generic TS and other operational requirements that were completely reviewed and approved in the design certification rulemaking and do not require a change to a design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Generic changes that require a change to a design feature in the generic DCD are governed by the requirements in paragraphs A or B of this section.

            2. Generic changes to generic TS and other operational requirements are applicable to all applicants who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this section.
            3. The Commission may require plant-specific departures on generic TS and other operational requirements that were completely reviewed and approved, provided a change to a design feature in the generic DCD is not required and special circumstances as defined in 10 CFR 2.335 are present. The Commission may modify or supplement generic TS and other operational requirements that were not completely reviewed and approved or require additional TS and other operational requirements on a plant-specific basis, provided a change to a design feature in the generic DCD is not required.
            4. An applicant who references this appendix may request an exemption from the generic technical specifications or other operational requirements. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7. The grant of an exemption must be subject to litigation in the same manner as other issues material to the license hearing.
            5. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license, or for operation under 10 CFR 52.103(a), who believes that an operational requirement approved in the DCD or a TS derived from the generic TS must be changed may petition to admit such a contention into the proceeding. The petition must comply with the general requirements of 10 CFR 2.309 and must demonstrate why special circumstances as defined in 10 CFR 2.335 are present, or demonstrate compliance with the Commission's regulations in effect at the time this appendix was approved, as set forth in Section V of this appendix. Any other party may file a response to the petition. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. All other issues with respect to the plant-specific TS or other operational requirements are subject to a hearing as part of the license proceeding.
            6. After issuance of a license, the generic TS have no further effect on the plant-specific TS. Changes to the plant-specific TS will be treated as license amendments under 10 CFR 50.90.
            IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
            A.1. An applicant or licensee who references this appendix shall perform and demonstrate conformance with the ITAAC before fuel load. With respect to activities subject to an ITAAC, an applicant for a license may proceed at its own risk with design and procurement activities. A licensee may also proceed at its own risk with design, procurement, construction, and preoperational activities, even though the NRC may not have found that any particular ITAAC has been met.
            2. The licensee who references this appendix shall notify the NRC that the required inspections, tests, and analyses in the ITAAC have been successfully completed and that the corresponding acceptance criteria have been met.
            3. If an activity is subject to an ITAAC and the applicant or licensee who references this appendix has not demonstrated that the ITAAC has been met, the applicant or licensee may either take corrective actions to successfully complete that ITAAC, request an exemption from the ITAAC under Section VIII of this appendix and 10 CFR 52.97(b), or petition for rulemaking to amend this appendix by changing the requirements of the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes to the ITAAC must meet the requirements of paragraph VIII.A.1 of this appendix.

            B.1. The NRC shall ensure that the required inspections, tests, and analyses in the ITAAC are performed. The NRC shall verify that the inspections, tests, and analyses referenced by the licensee have been successfully completed and, based solely thereon, find that the prescribed acceptance criteria have been met. At appropriate intervals during construction, the NRC shall publish notices of the successful completion of ITAAC in the Federal Register.

            2. In accordance with 10 CFR 52.103(g), the Commission shall find that the acceptance criteria in the ITAAC for the license are met before fuel load.
            3. After the Commission has made the finding required by 10 CFR 52.103(g), the ITAAC do not, by virtue of their inclusion within the DCD, constitute regulatory requirements either for licensees or for renewal of the license; except for specific ITAAC, which are the subject of a § 52.103(a) hearing, their expiration will occur upon final Commission action in such a proceeding. However, subsequent modifications must comply with the Tier 1 and Tier 2 design descriptions in the plant-specific DCD unless the licensee has complied with the applicable requirements of 10 CFR 52.98 and Section VIII of this appendix.
            X. Records and Reporting
            A. Records

            1. The applicant for this appendix shall maintain a copy of the generic DCD that includes all generic changes it makes to Tier 1 and Tier 2, and the generic TS and other operational requirements. The applicant shall maintain sensitive unclassified non-safeguards information (including proprietary information) and safeguards information referenced in the generic DCD for the period that this appendix may be referenced, as specified in Section VII of this appendix.
            2. An applicant or licensee who references this appendix shall maintain the plant-specific DCD to accurately reflect both generic changes to the generic DCD and plant-specific departures made under Section VIII of this appendix throughout the period of application and for the term of the license (including any period of renewal).
            3. An applicant or licensee who references this appendix shall prepare and maintain written evaluations which provide the bases for the determinations required by Section VIII of this appendix. These evaluations must be retained throughout the period of application and for the term of the license (including any period of renewal).
            4.a. The applicant for the AP1000 design shall maintain a copy of the AIA performed to comply with the requirements of 10 CFR 50.150(a) for the term of the certification (including any period of renewal).
            b. An applicant or licensee who references this appendix shall maintain a copy of the AIA performed to comply with the requirements of 10 CFR 50.150(a) throughout the pendency of the application and for the term of the license (including any period of renewal).
            B. Reporting
            1. An applicant or licensee who references this appendix shall submit a report to the NRC containing a brief description of any plant-specific departures from the DCD, including a summary of the evaluation of each. This report must be filed in accordance with the filing requirements applicable to reports in 10 CFR 52.3.
            2. An applicant or licensee who references this appendix shall submit updates to its DCD, which reflect the generic changes to and plant-specific departures from the generic DCD made under Section VIII of this appendix. These updates must be filed under the filing requirements applicable to final safety analysis report updates in 10 CFR 52.3 and 50.71(e).
            3. The reports and updates required by paragraphs X.B.1 and X.B.2 must be submitted as follows:
            a. On the date that an application for a license referencing this appendix is submitted, the application must include the report and any updates to the generic DCD.
            b. During the interval from the date of application for a license to the date the Commission makes its findings required by 10 CFR 52.103(g), the report must be submitted semi-annually. Updates to the plant-specific DCD must be submitted annually and may be submitted along with amendments to the application.
            c. After the Commission makes the finding required by 10 CFR 52.103(g), the reports and updates to the plant-specific DCD must be submitted, along with updates to the site-specific portion of the final safety analysis report for the facility, at the intervals required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at shorter intervals as specified in the license.
            [72 FR 49517, Aug. 28, 2007, as amended at 76 FR 82102, Dec. 30, 2011; 84 FR 63568, Nov. 18, 2019]

            Pt. 52, App. E
            Appendix E to Part 52—Design Certification Rule for the ESBWR Design
            I. Introduction
            Appendix E constitutes the standard design certification for the Economic Simplified Boiling-Water Reactor (ESBWR) design, in accordance with 10 CFR part 52, subpart B. The applicant for certification of the ESBWR design is GE-Hitachi Nuclear Energy.
            II. Definitions
            A. Generic design control document (generic DCD) means the document containing the Tier 1 and Tier 2 information and generic technical specifications that is incorporated by reference into this appendix.
            B. Generic technical specifications (generic TS) means the information required by 10 CFR 50.36 and 50.36a for the portion of the plant that is within the scope of this appendix.
            C. Plant-specific DCD means that portion of the combined license (COL) final safety analysis report (FSAR) that sets forth both the generic DCD information and any plant-specific changes to generic DCD information.
            D. Tier 1 means the portion of the design-related information contained in the generic DCD that is approved and certified by this appendix (Tier 1 information). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes:
            1. Definitions and general provisions;
            2. Design descriptions;
            3. Inspections, tests, analyses, and acceptance criteria (ITAACs);
            4. Significant site parameters; and
            5. Significant interface requirements.
            E. Tier 2 means the portion of the design-related information contained in the generic DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these differences, an applicant or licensee must meet the requirement in paragraph III.B of this appendix to reference Tier 2 when referencing Tier 1. Tier 2 information includes:
            1. Information required by §§ 52.47(a) and 52.47(c), with the exception of generic TS and conceptual design information;
            2. Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAACs have been met;
            3. COL action items (COL license information), which identify certain matters that must be addressed in the site-specific portion of the FSAR by an applicant who references this appendix. These items constitute information requirements but are not the only acceptable set of information in the FSAR. An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the FSAR. After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the FSAR; and
            4. The availability controls in Appendix 19ACM of the DCD.
            F. Tier 2* means the portion of the Tier 2 information, designated as such in the generic DCD, which is subject to the change process in paragraph VIII.B.6 of this appendix. This designation expires for some Tier 2* information under paragraph VIII.B.6 of this appendix.
            G. Departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses means:
            1. Changing any of the elements of the method described in the plant-specific DCD unless the results of the analysis are conservative or essentially the same; or
            2. Changing from a method described in the plant-specific DCD to another method unless that method has been approved by the NRC for the intended application.
            H. All other terms in this appendix have the meaning set out in 10 CFR 50.2, 10 CFR 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable.
            III. Scope and Contents

            A. Incorporation by reference approval. The documents in Table 1 are approved for incorporation by reference by the Director of the Office of the Federal Register under 5 U.S.C. 552(a) and 1 CFR part 51. You may obtain copies of the generic DCD from Jerald G. Head, Senior Vice President, Regulatory Affairs, GE-Hitachi Nuclear Energy, 3901 Castle Hayne Road, MC A-18, Wilmington, NC 28401, telephone: 1-910-819-5692. You can view the generic DCD online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. In ADAMS, search under the ADAMS Accession No. listed in Table 1. If you do not have access to ADAMS or if you have problems accessing documents located in ADAMS, contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 1-301-415-3747, or by email at PDR.Resource@nrc.gov. These documents can also be viewed at the Federal rulemaking Web site, http://www.regulations.gov, by searching for documents filed under Docket ID NRC-2010-0135. Copies of these documents are available for examination and copying at the NRC's PDR located at Room O-1F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. Copies are also available for examination at the NRC Library located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852, telephone: 301-415-5610, email: Library.Resource@nrc.gov. All approved material is available for inspection at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 1-202-741-6030 or go to http://www.archives.gov/federal-register/cfr/ibrlocations.html.

              Table 1—Documents Approved for Incorporation by Reference

                Document No.
                Document title
                ADAMS Accession No.

                GE Hitachi:

                26A6642AB Rev. 10
                ESBWR Design Control Document, Revision 10, Tier 1, dated April 2014
                ML14104A929 (package)

                26A6642AB Rev. 10
                ESBWR Design Control Document, Revision 10, Tier 2, dated April 2014
                ML14104A929 (package)

                Bechtel Power Corporation:

                BC-TOP-3-A
                “Tornado and Extreme Wind Design Criteria for Nuclear Power Plants,” Topical Report, Revision 3, August 1974
                ML14093A218

                BC-TOP-9A
                “Design of Structures for Missile Impact,” Topical Report, Revision 2, September 1974
                ML14093A217

                General Electric:

                GEZ-4982A
                General Electric Large Steam Turbine Generator Quality Control Program, The STG Global Supply Chain Quality Management System (MFGGLO-GEZ-0010) Revision 1.2, February 7, 2006
                ML14093A215

                GE Nuclear Energy:

                NEDO-11209-04A
                “GE Nuclear Energy Quality Assurance Program Description,” Class 1, Revision 8, March 31, 1989
                ML14093A209

                NEDO-31960-A
                “BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,” Class I, November 1995
                ML14093A212

                NEDO-31960-A—Supplement 1
                “BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,” Class I, November 1995
                ML14093A211

                NEDO-32465-A
                GE Nuclear Energy and BWR Owners' Group, “Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications,” Class I, August 1996
                ML14093A210

                GE-Hitachi Nuclear Energy:

                NEDO-33181
                “NP-2010 COL Demonstration Project Quality Assurance Plan,” Revision 6, August 2009
                ML14248A297

                NEDO-33219
                “ESBWR Human Factors Engineering Functional Requirements Analysis Implementation Plan,” Revision 4, Class I, February 2010
                ML100350104

                NEDO-33260
                “Quality Assurance Requirements for Suppliers of Equipment and Services to the GEH ESBWR Project,” Revision 5, Class I, April 2008
                ML14248A648

                NEDO-33262
                “ESBWR Human Factors Engineering Operating Experience Review Implementation Plan,” Revision 3, Class I, January 2010
                ML100340030

                NEDO-33266
                “ESBWR Human Factors Engineering Staffing and Qualifications Implementation Plan,” Revision 3, Class I, January 2010
                ML100350167

                NEDO-33267
                “ESBWR Human Factors Engineering Human Reliability Analysis Implementation Plan,” Revision 4, Class I, January 2010
                ML100330609

                NEDO-33277
                “ESBWR Human Factors Engineering Human Performance Monitoring Implementation Plan,” Revision 4, Class I, January 2010
                ML100270770

                NEDO-33278
                “ESBWR Human Factors Engineering Design Implementation Plan,” Revision 4, Class I, January 2010
                ML100270468

                NEDO-33289
                “ESBWR Reliability Assurance Program,” Revision 2, Class II, September 2008
                ML14248A662

                NEDO-33337
                “ESBWR Initial Core Transient Analyses,” Revision 1, Class I, April 2009
                ML091130628

                NEDO-33338
                “ESBWR Feedwater Temperature Operating Domain Transient and Accident Analysis,” Revision 1, Class I, May 2009
                ML091380173

                NEDO-33373-A
                “Dynamic, Load-Drop, and Thermal-Hydraulic Analyses for ESBWR Fuel Racks,” Revision 5, Class I, October 2010
                ML102990226 (part 1)ML102990228 (part 2)

                NEDO-33411
                “Risk Significance of Structures, Systems and Components for the Design Phase of the ESBWR,” Revision 2, Class I, February 2010
                ML100610417

            B. An applicant or licensee referencing this appendix, in accordance with Section IV of this appendix, shall incorporate by reference and comply with the requirements of this appendix, including Tier 1, Tier 2 (including the availability controls in Appendix 19ACM of the DCD), and the generic TS except as otherwise provided in this appendix. Conceptual design information in the generic DCD and the evaluation of severe accident mitigation design alternatives in NEDO-33306, Revision 4, “ESBWR Severe Accident Mitigation Design Alternatives,” are not part of this appendix.
            C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then Tier 1 controls.
            D. If there is a conflict between the generic DCD and either the application for design certification of the ESBWR design or NUREG-1966, “Final Safety Evaluation Report Related to Certification of the ESBWR Standard Design,” (FSER) and Supplement No. 1 to NUREG-1966, then the generic DCD controls.
            E. Design activities for structures, systems, and components that are wholly outside the scope of this appendix may be performed using site characteristics, provided the design activities do not affect the DCD or conflict with the interface requirements.
            IV. Additional Requirements and Restrictions
            A. An applicant for a COL who references this appendix shall, in addition to complying with the requirements of §§ 52.77, 52.79, and 52.80, comply with the following requirements:

            1. Incorporate by reference, as part of its application, this appendix.

            2. Include, as part of its application:
            a. A plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the ESBWR design, either by including or incorporating by reference the generic DCD information, and as modified and supplemented by the applicant's exemptions and departures;
            b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
            c. Plant-specific TS, consisting of the generic and site-specific TS that are required by 10 CFR 50.36 and 50.36a;
            d. Information demonstrating that the site characteristics fall within the site parameters and that the interface requirements have been met;
            e. Information that addresses the COL action items;
            f. Information required by § 52.47(a) that is not within the scope of this appendix;
            g. Information demonstrating that hurricane loads on those structures, systems, and components described in Section 3.3.2 of the generic DCD are either bounded by the total tornado loads analyzed in Section 3.3.2 of the generic DCD or will meet applicable NRC requirements with consideration of hurricane loads in excess of the total tornado loads; and hurricane-generated missile loads on those structures, systems, and components described in Section 3.5.2 of the generic DCD are either bounded by tornado-generated missile loads analyzed in Section 3.5.1.4 of the generic DCD or will meet applicable NRC requirements with consideration of hurricane-generated missile loads in excess of the tornado-generated missile loads; and
            h. Information demonstrating that the spent fuel pool level instrumentation is designed to allow the connection of an independent power source, and that the instrumentation will maintain its design accuracy following a power interruption or change in power source without requiring recalibration.
            3. Include, in the plant-specific DCD, the sensitive, unclassified, non-safeguards information (including proprietary information and security-related information) and safeguards information referenced in the ESBWR generic DCD.
            4. Include, as part of its application, a demonstration that an entity other than GE-Hitachi Nuclear Energy is qualified to supply the ESBWR design unless GE-Hitachi Nuclear Energy supplies the design for the applicant's use.
            B. The Commission reserves the right to determine in what manner this appendix may be referenced by an applicant for a construction permit or operating license under 10 CFR part 50.
            V. Applicable Regulations
            A. Except as indicated in paragraph B of this section, the regulations that apply to the ESBWR design are in 10 CFR parts 20, 50, 73, and 100, codified as of October 6, 2014, that are applicable and technically relevant, as described in the FSER (NUREG-1966) and Supplement No. 1.
            B. The ESBWR design is exempt from portions of the following regulations:
            1. Paragraph (f)(2)(iv) of 10 CFR 50.34—Contents of Applications: Technical Information—codified as of October 6, 2014.
            VI. Issue Resolution
            A. The Commission has determined that the structures, systems, components, and design features of the ESBWR design comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in Section V of this appendix; and therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for the ESBWR design.
            B. The Commission considers the following matters resolved within the meaning of § 52.63(a)(5) in subsequent proceedings for issuance of a COL, amendment of a COL, or renewal of a COL, proceedings held under § 52.103, and enforcement proceedings involving plants referencing this appendix:

            1. All nuclear safety issues associated with the information in the FSER and Supplement No. 1; Tier 1, Tier 2 (including referenced information, which the context indicates is intended as requirements, and the availability controls in Appendix 19ACM of the DCD), the 20 documents referenced in Table 1 of paragraph III.A, and the rulemaking record for certification of the ESBWR design, with the exception of: generic TS and other operational requirements such as human factors engineering procedure development and training program development in Sections 18.9 and 18.10 of the generic DCD; hurricane loads on those structures, systems, and components described in Section 3.3.2 of the generic DCD that are not bounded by the total tornado loads analyzed in Section 3.3.2 of the generic DCD; hurricane-generated missile loads on those structures, systems, and components described in Section 3.5.2 of the generic DCD that are not bounded by tornado-generated missile loads analyzed in Section 3.5.1.4 of the generic DCD; and spent fuel pool level instrumentation design in regard to the connection of an independent power source, and how the instrumentation will maintain its design accuracy following a power interruption or change in power source without recalibration;
            2. All nuclear safety and safeguards issues associated with the referenced information in the 50 non-public documents in Tables 1.6-1 and 1.6-2 of Tier 2 of the DCD which contain sensitive unclassified non-safeguards information (including proprietary information and security-related information) and safeguards information and which, in context, are intended as requirements in the generic DCD for the ESBWR design, with the exception of human factors engineering procedure development and training program development in Chapters 18.9 and 18.10 of the generic DCD;
            3. All generic changes to the DCD under and in compliance with the change processes in paragraphs VIII.A.1 and VIII.B.1 of this appendix;
            4. All exemptions from the DCD under and in compliance with the change processes in paragraphs VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;
            5. All departures from the DCD that are approved by license amendment, but only for that plant;
            6. Except as provided in paragraph VIII.B.5.g of this appendix, all departures from Tier 2 under and in compliance with the change processes in paragraph VIII.B.5 of this appendix that do not require prior NRC approval, but only for that plant;
            7. All environmental issues concerning severe accident mitigation design alternatives associated with the information in the NRC's Environmental Assessment for the ESBWR design (ADAMS Accession No. ML111730382) and NEDO-33306, Revision 4, “ESBWR Severe Accident Mitigation Design Alternatives,” (ADAMS Accession No. ML102990433) for plants referencing this appendix whose site characteristics fall within those site parameters specified in NEDO-33306.
            C. The Commission does not consider operational requirements for an applicant or licensee who references this appendix to be matters resolved within the meaning of § 52.63(a)(5). The Commission reserves the right to require operational requirements for an applicant or licensee who references this appendix by rule, regulation, order, or license condition.
            D. Except under the change processes in Section VIII of this appendix, the Commission may not require an applicant or licensee who references this appendix to:
            1. Modify structures, systems, components, or design features as described in the generic DCD;
            2. Provide additional or alternative structures, systems, components, or design features not discussed in the generic DCD; or
            3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria, or justification for structures, systems, components, or design features discussed in the generic DCD.
            E. The NRC will specify at an appropriate time the procedures to be used by an interested person who seeks to review portions of the design certification or references containing safeguards information or sensitive unclassified non-safeguards information (including proprietary information, such as trade secrets and commercial or financial information obtained from a person that are privileged or confidential (10 CFR 2.390 and 10 CFR part 9), and security-related information), for the purpose of participating in the hearing required by § 52.85, the hearing provided under § 52.103, or in any other proceeding relating to this appendix in which interested persons have a right to request an adjudicatory hearing.
            VII. Duration of This Appendix
            This appendix may be referenced for a period of 15 years from November 14, 2014, except as provided for in §§ 52.55(b) and 52.57(b). This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.
            VIII. Processes for Changes and Departures
            A. Tier 1 information
            1. Generic changes to Tier 1 information are governed by the requirements in § 52.63(a)(1).
            2. Generic changes to Tier 1 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs A.3 or A.4 of this section.
            3. Departures from Tier 1 information that are required by the Commission through plant-specific orders are governed by the requirements in § 52.63(a)(4).
            4. Exemptions from Tier 1 information are governed by the requirements in §§ 52.63(b)(1) and 52.98(f). The Commission will deny a request for an exemption from Tier 1, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.
            B. Tier 2 information
            1. Generic changes to Tier 2 information are governed by the requirements in 10 CFR 52.63(a)(1).

            2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.

            3. The Commission may not require new requirements on Tier 2 information by plant-specific order while this appendix is in effect under 10 CFR 52.55 or 52.61, unless:
            a. A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time this appendix was approved, as set forth in Section V of this appendix, or to ensure adequate protection of the public health and safety or the common defense and security; and
            b. Special circumstances as defined in 10 CFR 50.12(a) are present.
            4. An applicant or licensee who references this appendix may request an exemption from Tier 2 information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 50.12(a). The Commission will deny a request for an exemption from Tier 2, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design. The grant of an exemption to an applicant must be subject to litigation in the same manner as other issues material to the license hearing. The grant of an exemption to a licensee must be subject to an opportunity for a hearing in the same manner as license amendments.
            5.a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2* information, or the TS, or requires a license amendment under paragraph B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD.
            b. A proposed departure from Tier 2, other than one affecting resolution of a severe accident issue identified in the plant-specific DCD or one affecting information required by § 52.47(a)(28) to address aircraft impacts, requires a license amendment if it would:
            (1) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the plant-specific DCD;
            (2) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety and previously evaluated in the plant-specific DCD;
            (3) Result in more than a minimal increase in the consequences of an accident previously evaluated in the plant-specific DCD;
            (4) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the plant-specific DCD;
            (5) Create a possibility for an accident of a different type than any evaluated previously in the plant-specific DCD;
            (6) Create a possibility for a malfunction of an SSC important to safety with a different result than any evaluated previously in the plant-specific DCD;
            (7) Result in a design-basis limit for a fission product barrier as described in the plant-specific DCD being exceeded or altered; or
            (8) Result in a departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses.
            c. A proposed departure from Tier 2 affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if:
            (1) There is a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously reviewed and determined to be not credible could become credible; or
            (2) There is a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously reviewed.
            d. A proposed departure from Tier 2 information required by § 52.47(a)(28) to address aircraft impacts shall consider the effect of the changed design feature or functional capability on the original aircraft impact assessment required by 10 CFR 50.150(a). The applicant or licensee shall describe in the plant-specific DCD how the modified design features and functional capabilities continue to meet the aircraft impact assessment requirements in 10 CFR 50.150(a)(1).
            e. If a departure requires a license amendment under paragraph B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
            f. A departure from Tier 2 information that is made under paragraph B.5 of this section does not require an exemption from this appendix.

            g. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under § 52.103(a), who believes that an applicant or licensee who references this appendix has not complied with paragraph VIII.B.5 of this appendix when departing from Tier 2 information, may petition to admit into the proceeding such a contention. In addition to compliance with the general requirements of 10 CFR 2.309, the petition must demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a § 52.103 preoperational hearing, or that the change bears directly on the amendment request in the case of a hearing on a license amendment. Any other party may file a response. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. The Commission may admit such a contention if it determines the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5 of this appendix.
            6.a. An applicant who references this appendix may not depart from Tier 2* information, which is designated with italicized text or brackets and an asterisk in the generic DCD, without NRC approval. The departure will not be considered a resolved issue, within the meaning of Section VI of this appendix and § 52.63(a)(5).
            b. A licensee who references this appendix may not depart from the following Tier 2* matters without prior NRC approval. A request for a departure will be treated as a request for a license amendment under 10 CFR 50.90.
            (1) Fuel mechanical and thermal-mechanical design evaluation reports, including fuel burnup limits.
            (2) Control rod mechanical and nuclear design reports.
            (3) Fuel nuclear design report.
            (4) Critical power correlation.
            (5) Fuel licensing acceptance criteria.
            (6) Control rod licensing acceptance criteria.
            (7) Mechanical and structural design of spent fuel storage racks.
            (8) Steam dryer pressure load analysis methodology.
            c. A licensee who references this appendix may not, before the plant first achieves full power following the finding required by § 52.103(g), depart from the following Tier 2* matters except under paragraph B.6.b of this section. After the plant first achieves full power, the following Tier 2* matters revert to Tier 2 status and are subject to the departure provisions in paragraph B.5 of this section.
            (1) ASME Boiler and Pressure Vessel Code, Section III, Subsections NE (Division 1) and CC (Division 2) for containment vessel design.
            (2) American Concrete Institute 349 and American National Standards Institute/American Institute of Steel Construction—N690.
            (3) Power-operated valves.
            (4) Equipment seismic qualification methods.
            (5) Piping design acceptance criteria.
            (6) Instrument setpoint methodology.
            (7) Safety-Related Distribution Control and Information System performance specification and architecture.
            (8) Safety System Logic and Control hardware and software.
            (9) Human factors engineering design and implementation.
            (10) First of a kind testing for reactor stability (first plant only).
            (11) Reactor precritical heatup with reactor water cleanup/shutdown cooling (first plant only).
            (12) Isolation condenser system heatup and steady state operation (first plant only).
            (13) Power maneuvering in the feedwater temperature operating domain (first plant only).
            (14) Load maneuvering capability (first plant only).
            (15) Defense-in-depth stability solution evaluation test (first plant only).
            d. Departures from Tier 2* information that are made under paragraph B.6 of this section do not require an exemption from this appendix.
            C. Operational requirements.
            1. Generic changes to generic TS and other operational requirements that were completely reviewed and approved in the design certification rulemaking and do not require a change to a design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Generic changes that require a change to a design feature in the generic DCD are governed by the requirements in paragraphs A or B of this section.
            2. Generic changes to generic TS and other operational requirements are applicable to all applicants who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this section.
            3. The Commission may require plant-specific departures on generic TS and other operational requirements that were completely reviewed and approved, provided a change to a design feature in the generic DCD is not required and special circumstances as defined in 10 CFR 2.335 are present. The Commission may modify or supplement generic TS and other operational requirements that were not completely reviewed and approved or require additional TS and other operational requirements on a plant-specific basis, provided a change to a design feature in the generic DCD is not required.
            4. An applicant who references this appendix may request an exemption from the generic TS or other operational requirements. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of § 52.7. The grant of an exemption must be subject to litigation in the same manner as other issues material to the license hearing.

            5. A party to an adjudicatory proceeding for the issuance, amendment, or renewal of a license, or for operation under § 52.103(a), who believes that an operational requirement approved in the DCD or a TS derived from the generic TS must be changed may petition to admit such a contention into the proceeding. The petition must comply with the general requirements of 10 CFR 2.309 and must demonstrate why special circumstances as defined in 10 CFR 2.335 are present, or demonstrate compliance with the Commission's regulations in effect at the time this appendix was approved, as set forth in Section V of this appendix. Any other party may file a response to the petition. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. All other issues with respect to the plant-specific TS or other operational requirements are subject to a hearing as part of the license proceeding.
            6. After issuance of a license, the generic TS have no further effect on the plant-specific TS. Changes to the plant-specific TS will be treated as license amendments under 10 CFR 50.90.
            IX. [Reserved]
            X. Records and Reporting
            A. Records
            1. The applicant for this appendix shall maintain a copy of the generic DCD that includes all generic changes it makes to Tier 1 and Tier 2, and the generic TS and other operational requirements. The applicant shall maintain the sensitive unclassified non-safeguards information (including proprietary information and security-related information) and safeguards information referenced in the generic DCD for the period that this appendix may be referenced, as specified in Section VII of this appendix.
            2. An applicant or licensee who references this appendix shall maintain the plant-specific DCD to accurately reflect both generic changes to the generic DCD and plant-specific departures made under Section VIII of this appendix throughout the period of application and for the term of the license (including any period of renewal).
            3. An applicant or licensee who references this appendix shall prepare and maintain written evaluations that provide the bases for the determinations required by Section VIII of this appendix. These evaluations must be retained throughout the period of application and for the term of the license (including any period of renewal).
            4.a. The applicant for the ESBWR design shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) for the term of the certification (including any period of renewal).
            b. An applicant or licensee who references this appendix shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) throughout the pendency of the application and for the term of the license (including any period of renewal).
            B. Reporting
            1. An applicant or licensee who references this appendix shall submit a report to the NRC containing a brief description of any plant-specific departures from the DCD, including a summary of the evaluation of each. This report must be filed in accordance with the filing requirements applicable to reports in § 52.3.
            2. An applicant or licensee who references this appendix shall submit updates to its plant-specific DCD that reflect the generic changes to and plant-specific departures from the generic DCD made under Section VIII of this appendix. These updates shall be filed under the filing requirements applicable to final safety analysis report updates in 10 CFR 52.3 and 50.71(e).
            3. The reports and updates required by paragraphs X.B.1 and X.B.2 of this appendix must be submitted as follows:
            a. On the date that an application for a license referencing this appendix is submitted, the application must include the report and any updates to the generic DCD.
            b. During the interval from the date of application for a license to the date the Commission makes its finding required by § 52.103(g), the report must be submitted semi-annually. Updates to the plant-specific DCD must be submitted annually and may be submitted along with amendments to the application.
            c. After the Commission makes the finding required by § 52.103(g), the reports and updates to the plant-specific DCD must be submitted, along with updates to the site-specific portion of the final safety analysis report for the facility, at the intervals required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at shorter intervals as specified in the license.
            [79 FR 61983, Oct. 15, 2014, as amended at 84 FR 63568, Nov. 18, 2019]

            Pt. 52, App. F
            Appendix F to Part 52—Design Certification Rule for the APR1400 Design
            I. Introduction
            Appendix F constitutes the standard design certification for the Advanced Power Reactor 1400 (APR1400) design, in accordance with 10 CFR part 52, subpart B. The applicant for certification of the APR1400 design is Korea Electric Power Corporation and Korea Hydro & Nuclear Power Co., Ltd. (KEPCO/KHNP).
            II. Definitions
            A. Generic design control document (generic DCD) means the document containing the Tier 1 and Tier 2 information (including the technical and topical reports referenced in Chapter 1) and generic technical specifications that is incorporated by reference into this appendix.
            B. Generic technical specifications (generic TS) means the information required by 10 CFR 50.36 and 50.36a for the portion of the plant that is within the scope of this appendix.
            C. Plant-specific DCD means that portion of the combined license (COL) final safety analysis report that sets forth both the generic DCD information and any plant-specific changes to generic DCD information.
            D. Tier 1 means the portion of the design-related information contained in the generic DCD that is approved and certified by this appendix (Tier 1 information). The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes:
            1. Definitions and general provisions;
            2. Design descriptions;
            3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
            4. Significant site parameters; and
            5. Significant interface requirements.
            E. Tier 2 means the portion of the design-related information contained in the generic DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1. Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these differences, an applicant or licensee must meet the requirement in paragraph III.B of this appendix to reference Tier 2 when referencing Tier 1. Tier 2 information includes:
            1. Information required by § 52.47(a) and (c), with the exception of generic TS and conceptual design information;
            2. Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met; and
            3. COL Items (COL license information), which identify certain matters that must be addressed in the site-specific portion of the final safety analysis report by an applicant who references this appendix. These items constitute information requirements but are not the only acceptable set of information in the final safety analysis report. An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the final safety analysis report. After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the final safety analysis report.
            F. Departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses means:
            1. Changing any of the elements of the method described in the plant-specific DCD unless the results of the analysis are conservative or essentially the same; or
            2. Changing from a method described in the plant-specific DCD to another method unless that method has been approved by the NRC for the intended application.
            G. All other terms in this appendix have the meaning set out in 10 CFR 50.2, 10 CFR 52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable.
            III. Scope and Contents

            A. Incorporation by reference approval. The APR1400 material is approved for incorporation by reference by the Director of the Office of the Federal Register under 5 U.S.C. 552(a) and 1 CFR part 51. You may obtain copies of the generic DCD from Yun-Ho Kim, President, KHNP Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon, 34101, Korea. You can view the generic DCD online in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. In ADAMS, search under ADAMS Accession No. ML18228A667. If you do not have access to ADAMS or if you have problems accessing documents located in ADAMS, contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-3747, or by email at PDR.Resource@nrc.gov. Copies of this document are available for examination and copying at the NRC's PDR located at Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852. Copies are also available for examination at the NRC Library located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852, telephone: 301-415-5610, email: Library.Resource@nrc.gov. All approved material is available for inspection at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 202-741-6030 or go to https://www.archives.gov/federal-register/cfr/ibrlocations.html.

            1. Korea Electric Power Corporation and Korea Hydro & Nuclear Power Co, Ltd
            a. APR1400 Design Control Document Tier 1 (APR1400-K-X-IT-14001-NP), Revision 3 (August 2018).
            b. APR1400 Design Control Document Tier 2 (APR1400-K-X-FS-14002-NP), Revision 3 (August 2018), including:
            i. Chapter 1, Introduction and General Description of the Plant.
            ii. Chapter 2, Site Characteristics.
            iii. Chapter 3, Design of Structures, Systems, Components, and Equipment.
            iv. Chapter 4, Reactor.

            v. Chapter 5, Reactor Coolant System and Connecting Systems.
            vi. Chapter 6, Engineered Safety Features.
            vii. Chapter 7, Instrumentation and Controls.
            viii. Chapter 8, Electric Power.
            ix. Chapter 9, Auxiliary Systems.
            x. Chapter 10, Steam and Power Conversion System.
            xi. Chapter 11, Radioactive Waste Management.
            xii. Chapter 12, Radiation Protection.
            xiii. Chapter 13, Conduct of Operations.
            xiv. Chapter 14, Verification Programs.
            xv. Chapter 15, Transient and Accident Analyses.
            xvi. Chapter 16, Technical Specifications.
            xvii. Chapter 17, Quality Assurance and Reliability Assurance.
            xviii. Chapter 18, Human Factors Engineering.
            xix. Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation.
            c. APR1400-E-B-NR-16001-NP, Evaluation of Main Steam and Feedwater Piping Applied to the Graded Approach for the APR1400, Rev. 0 (July 2017).
            d. APR1400-E-B-NR-16002-NP, Evaluation of Safety Injection and Shutdown Cooling Piping Applied to the Graded Approach for the APR1400, Rev. 1 (May 2018).
            e. APR1400-E-I-NR-14001-NP, Human Factors Engineering Program Plan, Rev. 4 (July 2018).
            f. APR1400-E-I-NR-14002-NP, Operating Experience Review Implementation Plan, Rev. 2 (January 2018).
            g. APR1400-E-I-NR-14003-NP, Functional Requirements Analysis and Function Allocation Implementation Plan, Rev. 2 (January 2018).
            h. APR1400-E-I-NR-14004-NP, Task Analysis Implementation Plan, Rev. 3 (May 2018).
            i. APR1400-E-I-NR-14006-NP, Treatment of Important Human Actions Implementation Plan, Rev. 3 (May 2018).
            j. APR1400-E-I-NR-14007-NP, Human-System Interface Design Implementation Plan, Rev. 3 (May 2018).
            k. APR1400-E-I-NR-14008-NP, Human Factors Verification and Validation Implementation Plan, Rev. 3 (May 2018).
            l. APR1400-E-I-NR-14010-NP, Human Factors Verification and Validation Scenarios, Rev. 2 (January 2018).
            m. APR1400-E-I-NR-14011-NP, Basic Human-System Interface, Rev. 3 (May 2018).
            n. APR1400-E-I-NR-14012-NP, Style Guide, Rev. 2 (January 2018).
            o. APR1400-E-J-NR-14001-NP, Component Interface Module, Rev. 1 (March 2017).
            p. APR1400-E-J-NR-17001-NP, Secure Development and Operational Environment for APR1400 Computer-Based I&C Safety Systems, Rev. 0 (September 2017).
            q. APR1400-E-N-NR-14001-NP, Design Features To Address GSI-191, Rev. 3 (February 2018).
            r. APR1400-E-P-NR-14005-NP, Evaluations and Design Enhancements To Incorporate Lessons Learned from Fukushima Dai-Ichi Nuclear Accident, Rev. 2 (July 2017).
            s. APR1400-E-S-NR-14004-NP, Evaluation of Effects of HRHF Response Spectra on SSCs, Rev. 3 (December 2017).
            t. APR1400-E-S-NR-14005-NP, Evaluation of Structure-Soil-Structure Interaction (SSSI) Effects, Rev. 2 (December 2017).
            u. APR1400-E-S-NR-14006-NP, Stability Check for NI Common Basemat, Rev. 5 (May 2018).
            v. APR1400-E-X-NR-14001-NP, Equipment Qualification Program, Rev. 4 (July 2018).
            w. APR1400-F-A-NR-14001-NP, Small Break LOCA Evaluation Model, Rev. 1 (March 2017).
            x. APR1400-F-A-NR-14003-NP, Post-LOCA Long Term Cooling Evaluation Model, Rev. 1 (March 2017).
            y. APR1400-F-A-TR-12004-NP-A, Realistic Evaluation Methodology for Large-Break LOCA of the APR1400 (August 2018).
            z. APR1400-F-C-NR-14001-NP, CPC Setpoint Analysis Methodology for APR1400, Rev. 3 (June 2018).
            aa. APR1400-F-C-NR-14002-NP, Functional Design Requirements for a Core Operating Limit Supervisory System for APR1400, Rev. 1 (February 2017).
            ab. APR1400-F-C-NR-14003-NP, Functional Design Requirements for a Core Protection Calculator System for APR1400, Rev. 1 (March 2017).
            ac. APR1400-F-C-TR-12002-NP-A, KCE-1 Critical Heat Flux Correlation for PLUS7 Thermal Design (April 2017).
            ad. APR1400-F-M-TR-13001-NP-A, PLUS7 Fuel Design for the APR1400 (August 2018).
            ae. APR1400-H-N-NR-14005-NP, Summary Stress Report for Primary Piping, Rev. 2 (September 2016).
            af. APR1400-H-N-NR-14012-NP, Mechanical Analysis for New and Spent Fuel Storage Racks, Rev. 3 (August 2017).
            ag. APR1400-K-I-NR-14005-NP, Staffing and Qualifications Implementation Plan, Rev. 1 (February 2017).
            ah. APR1400-K-I-NR-14009-NP, Design Implementation Plan, Rev. 1 (February 2017).
            ai. APR1400-K-Q-TR-11005-NP-A, KHNP Quality Assurance Program Description (QAPD) for the APR1400 Design Certification, Rev. 2 (October 2016).
            aj. APR1400-Z-A-NR-14006-NP, Non-LOCA Safety Analysis Methodology, Rev. 1 (February 2017).
            ak. APR1400-Z-A-NR-14007-NP, Mass and Energy Release Methodologies for LOCA and MSLB, Rev. 2 (May 2018).

            al. APR1400-Z-A-NR-14011-NP, Criticality Analysis of New and Spent Fuel Storage Racks, Rev. 3 (May 2018).

            am. APR1400-Z-A-NR-14019-NP, CCF Coping Analysis, Rev. 3 (July 2018).
            an. APR1400-Z-J-NR-14001-NP, Safety I&C System, Rev. 3 (May 2018).
            ao. APR1400-Z-J-NR-14002-NP, Diversity and Defense-in-Depth, Rev. 3 (May 2018).
            ap. APR1400-Z-J-NR-14003-NP, Software Program Manual, Rev. 3 (May 2018).
            aq. APR1400-Z-J-NR-14004-NP, Uncertainty Methodology and Application for Instrumentation, Rev. 2 (January 2018).
            ar. APR1400-Z-J-NR-14005-NP, Setpoint Methodology for Safety-Related Instrumentation, Rev. 2 (January 2018).
            as. APR1400-Z-J-NR-14012-NP, Control System CCF Analysis, Rev. 3 (May 2018).
            at. APR1400-Z-J-NR-14013-NP, Response Time Analysis of Safety I&C System, Rev. 2 (January 2018).
            au. APR1400-Z-M-NR-14008-NP, Pressure-Temperature Limits Methodology for RCS Heatup and Cooldown, Rev. 1 (January 2018).
            av. APR1400-Z-M-TR-12003-NP-A, Fluidic Device Design for the APR1400 (April 2017).
            2. Combustion Engineering, Inc.
            a. CEN-310-NP-A, CPC and Methodology Changes for the CPC Improvement Program (April 1986).
            b. CEN-312-NP, Overview Description of the Core Operating Limit Supervisory System (COLSS), Rev. 01-NP (November 1986).
            3. Westinghouse
            a. WCAP-10697-NP-A, Common Qualified Platform Topical Report, Rev. 3 (February 2013).
            b. WCAP-17889-NP (APR1400-A-N-NR-17001-NP), Validation of SCALE 6.1.2 with 238-Group ENDF/B-VII.0 Cross Section Library for APR1400 Design Certification, Rev. 0 (June 2014).
            B. An applicant or licensee referencing this appendix, in accordance with Section IV of this appendix, shall incorporate by reference and comply with the requirements of this appendix except as otherwise provided in this appendix.
            C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then Tier 1 controls.
            D. If there is a conflict between the generic DCD and either the application for the design certification of the APR1400 design or the NUREG, “Final Safety Evaluation Report Related to Certification of the APR1400 Standard Design,” then the generic DCD controls.
            E. Design activities for structures, systems, and components that are entirely outside the scope of this appendix may be performed using site characteristics, provided the design activities do not affect the DCD or conflict with the interface requirements.
            IV. Additional Requirements and Restrictions
            A. An applicant for a COL that wishes to reference this appendix shall, in addition to complying with the requirements of §§ 52.77, 52.79, and 52.80, comply with the following requirements:
            1. Incorporate by reference, as part of its application, this appendix.
            2. Include, as part of its application:
            a. A plant-specific DCD containing the same type of information and using the same organization and numbering as the generic DCD for the APR1400 design, either by including or incorporating by reference the generic DCD information, and as modified and supplemented by the applicant's exemptions and departures;
            b. The reports on departures from and updates to the plant-specific DCD required by paragraph X.B of this appendix;
            c. Plant-specific TS, consisting of the generic and site-specific TS that are required by 10 CFR 50.36 and 50.36a;
            d. Information demonstrating that the site characteristics fall within the site parameters and that the interface requirements have been met;
            e. Information that addresses the COL items; and
            f. Information required by § 52.47(a) that is not within the scope of this appendix.
            3. Include, in the plant-specific DCD, the sensitive, unclassified, non-safeguards information (including proprietary information and security-related information) and safeguards information referenced in the APR1400 generic DCD.
            4. Include, as part of its application, a demonstration that an entity other than KEPCO/KHNP is qualified to supply the APR1400 design, unless KEPCO/KHNP supplies the design for the applicant's use.
            B. The Commission reserves the right to determine in what manner this appendix may be referenced by an applicant for a construction permit or operating license under 10 CFR part 50.
            V. Applicable Regulations
            A. Except as indicated in paragraph B of this section, the regulations that apply to the APR1400 design are in 10 CFR parts 20, 50, 52, 73, and 100, codified as of September 19, 2019, that are applicable and technically relevant, as described in the final safety evaluation report.
            B. The APR1400 design is exempt from portions of the following regulations:
            1. Paragraph (f)(2)(iv) of 10 CFR 50.34—Contents of Applications: Technical Information—codified as of September 19, 2019.
            VI. Issue Resolution

            A. The Commission has determined that the structures, systems, and components and design features of the APR1400 design comply with the provisions of the Atomic Energy Act of 1954, as amended, and the applicable regulations identified in Section V of this appendix; and therefore, provide adequate protection to the health and safety of the public. A conclusion that a matter is resolved includes the finding that additional or alternative structures, systems, and components, design features, design criteria, testing, analyses, acceptance criteria, or justifications are not necessary for the APR1400 design.
            B. The Commission considers the following matters resolved within the meaning of § 52.63(a)(5) in subsequent proceedings for issuance of a COL, amendment of a COL, or renewal of a COL, proceedings held under § 52.103, and enforcement proceedings involving plants referencing this appendix:
            1. All nuclear safety issues associated with the information in the final safety evaluation report, Tier 1, Tier 2, and the rulemaking record for certification of the APR1400 design, with the exception of generic TS and other operational requirements;
            2. All nuclear safety and safeguards issues associated with the referenced information in the 53 non-public documents in Tables 1.6-1 and 1.6-2 of Tier 2 of the DCD, which contain sensitive unclassified non-safeguards information (including proprietary information and security-related information) and safeguards information and which, in context, are intended as requirements in the generic DCD for the APR1400 design;
            3. All generic changes to the DCD under and in compliance with the change processes in paragraphs VIII.A.1 and VIII.B.1 of this appendix;
            4. All exemptions from the DCD under and in compliance with the change processes in paragraphs VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;
            5. All departures from the DCD that are approved by license amendment, but only for that plant;
            6. Except as provided in paragraph VIII.B.5.f of this appendix, all departures from Tier 2 under and in compliance with the change processes in paragraph VIII.B.5 of this appendix that do not require prior NRC approval, but only for that plant; and
            7. All environmental issues concerning severe accident mitigation design alternatives associated with the information in the NRC's environmental assessment for the APR1400 design (ADAMS Accession No. ML18306A607) and APR1400-E-P-NR-14006, Revision 2, “Severe Accident Mitigation Design Alternatives (SAMDAs) for the APR1400” (ML18235A158) for plants referencing this appendix whose site characteristics fall within those site parameters specified in APR1400-E-P-NR-14006.
            C. The Commission does not consider operational requirements for an applicant or licensee who references this appendix to be matters resolved within the meaning of § 52.63(a)(5). The Commission reserves the right to require operational requirements for an applicant or licensee who references this appendix by rule, regulation, order, or license condition.
            D. Except under the change processes in Section VIII of this appendix, the Commission may not require an applicant or licensee who references this appendix to:
            1. Modify structures, systems, components, or design features as described in the generic DCD;
            2. Provide additional or alternative structures, systems, components, or design features not discussed in the generic DCD; or
            3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria, or justification for structures, systems, components, or design features discussed in the generic DCD.
            E. The NRC will specify, at an appropriate time, the procedures to be used by an interested person who wishes to review portions of the design certification or references containing safeguards information or sensitive unclassified non-safeguards information (including proprietary information, such as trade secrets and commercial or financial information obtained from a person that are privileged or confidential (10 CFR 2.390 and 10 CFR part 9), and security-related information), for the purpose of participating in the hearing required by § 52.85, the hearing provided under § 52.103, or in any other proceeding relating to this appendix, in which interested persons have a right to request an adjudicatory hearing.
            VII. Duration of This Appendix
            This appendix may be referenced for a period of 15 years from September 19, 2019, except as provided for in §§ 52.55(b) and 52.57(b). This appendix remains valid for an applicant or licensee who references this appendix until the application is withdrawn or the license expires, including any period of extended operation under a renewed license.
            VIII. Processes for Changes and Departures
            A. Tier 1 Information
            1. Generic changes to Tier 1 information are governed by the requirements in § 52.63(a)(1).
            2. Generic changes to Tier 1 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs A.3 or A.4 of this section.
            3. Departures from Tier 1 information that are required by the Commission through plant-specific orders are governed by the requirements in § 52.63(a)(4).

            4. Exemptions from Tier 1 information are governed by the requirements in §§ 52.63(b)(1) and 52.98(f). The Commission will deny a request for an exemption from Tier 1, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.
            B. Tier 2 Information
            1. Generic changes to Tier 2 information are governed by the requirements in § 52.63(a)(1).
            2. Generic changes to Tier 2 information are applicable to all applicants or licensees who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs B.3, B.4, or B.5, of this section.
            3. The Commission may not require new requirements on Tier 2 information by plant-specific order, while this appendix is in effect under § 52.55 or § 52.61, unless:
            a. A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time this appendix was approved, as set forth in Section V of this appendix, or to ensure adequate protection of the public health and safety or the common defense and security; and
            b. Special circumstances as defined in 10 CFR 50.12(a) are present.
            4. An applicant or licensee who references this appendix may request an exemption from Tier 2 information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 50.12(a). The Commission will deny a request for an exemption from Tier 2, if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design. The granting of an exemption to an applicant must be subject to litigation in the same manner as other issues material to the license hearing. The granting of an exemption to a licensee must be subject to an opportunity for a hearing in the same manner as license amendments.
            5.a. An applicant or licensee who references this appendix may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, or the TS, or requires a license amendment under paragraph B.5.b or B.5.c of this section. When evaluating the proposed departure, an applicant or licensee shall consider all matters described in the plant-specific DCD.
            b. A proposed departure from Tier 2, other than one affecting resolution of a severe accident issue identified in the plant-specific DCD or one affecting information required by § 52.47(a)(28) to address aircraft impacts, requires a license amendment if it would:
            (1) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the plant-specific DCD;
            (2) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety and previously evaluated in the plant-specific DCD;
            (3) Result in more than a minimal increase in the consequences of an accident previously evaluated in the plant-specific DCD;
            (4) Result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the plant-specific DCD;
            (5) Create a possibility for an accident of a different type than any evaluated previously in the plant-specific DCD;
            (6) Create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any evaluated previously in the plant-specific DCD;
            (7) Result in a design-basis limit for a fission product barrier as described in the plant-specific DCD being exceeded or altered; or
            (8) Result in a departure from a method of evaluation described in the plant-specific DCD used in establishing the design bases or in the safety analyses.
            c. A proposed departure from Tier 2, affecting resolution of an ex-vessel severe accident design feature identified in the plant-specific DCD, requires a license amendment if:
            (1) There is a substantial increase in the probability of an ex-vessel severe accident such that a particular ex-vessel severe accident previously reviewed and determined to be not credible could become credible; or
            (2) There is a substantial increase in the consequences to the public of a particular ex-vessel severe accident previously reviewed.
            d. A proposed departure from Tier 2 information required by § 52.47(a)(28) to address aircraft impacts shall consider the effect of the changed design feature or functional capability on the original aircraft impact assessment required by 10 CFR 50.150(a). The applicant or licensee shall describe, in the plant-specific DCD, how the modified design features and functional capabilities continue to meet the aircraft impact assessment requirements in 10 CFR 50.150(a)(1).
            e. If a departure requires a license amendment under paragraph B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
            f. A departure from Tier 2 information that is made under paragraph B.5 of this section does not require an exemption from this appendix.

            g. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal of a license or for operation under § 52.103(a), who believes that an applicant or licensee who references this appendix has not complied with paragraph VIII.B.5 of this appendix when departing from Tier 2 information, may petition to admit into the proceeding such a contention. In addition to complying with the general requirements of 10 CFR 2.309, the petition must demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. Further, the petition must demonstrate that the change bears on an asserted noncompliance with an ITAAC acceptance criterion in the case of a § 52.103 preoperational hearing, or that the change bears directly on the amendment request in the case of a hearing on a license amendment. Any other party may file a response. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. The Commission may admit such a contention if it determines the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5 of this appendix.
            C. Operational Requirements
            1. Changes to APR1400 DC generic TS and other operational requirements that were completely reviewed and approved in the design certification rulemaking and do not require a change to a design feature in the generic DCD are governed by the requirements in 10 CFR 50.109. Changes that require a change to a design feature in the generic DCD are governed by the requirements in paragraphs A or B of this section.
            2. Changes to APR1400 DC generic TS and other operational requirements are applicable to all applicants who reference this appendix, except those for which the change has been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this section.
            3. The Commission may require plant-specific departures on generic TS and other operational requirements that were completely reviewed and approved, provided a change to a design feature in the generic DCD is not required and special circumstances, as defined in 10 CFR 2.335 are present. The Commission may modify or supplement generic TS and other operational requirements that were not completely reviewed and approved or require additional TS and other operational requirements on a plant-specific basis, provided a change to a design feature in the generic DCD is not required.
            4. An applicant who references this appendix may request an exemption from the generic TS or other operational requirements. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of § 52.7. The granting of an exemption must be subject to litigation in the same manner as other issues material to the license hearing.
            5. A party to an adjudicatory proceeding for the issuance, amendment, or renewal of a license, or for operation under § 52.103(a), who believes that an operational requirement approved in the DCD or a TS derived from the generic TS must be changed, may petition to admit such a contention into the proceeding. The petition must comply with the general requirements of 10 CFR 2.309 and must demonstrate why special circumstances as defined in 10 CFR 2.335 are present, or demonstrate compliance with the Commission's regulations in effect at the time this appendix was approved, as set forth in Section V of this appendix. Any other party may file a response to the petition. If, on the basis of the petition and any response, the presiding officer determines that a sufficient showing has been made, the presiding officer shall certify the matter directly to the Commission for determination of the admissibility of the contention. All other issues with respect to the plant-specific TS or other operational requirements are subject to a hearing as part of the licensing proceeding.
            6. After issuance of a license, the generic TS have no further effect on the plant-specific TS. Changes to the plant-specific TS will be treated as license amendments under 10 CFR 50.90.
            IX. [Reserved]
            X. Records and Reporting
            A. Records
            1. The applicant for this appendix shall maintain a copy of the generic DCD that includes all generic changes that are made to Tier 1 and Tier 2, and the generic TS and other operational requirements. The applicant shall maintain the sensitive unclassified non-safeguards information (including proprietary information and security-related information) and safeguards information referenced in the generic DCD for the period that this appendix may be referenced, as specified in Section VII of this appendix.
            2. An applicant or licensee who references this appendix shall maintain the plant-specific DCD to accurately reflect both generic changes to the generic DCD and plant-specific departures made under Section VIII of this appendix throughout the period of application and for the term of the license (including any periods of renewal).

            3. An applicant or licensee who references this appendix shall prepare and maintain written evaluations which provide the bases for the determinations required by Section VIII of this appendix. These evaluations must be retained throughout the period of application and for the term of the license (including any periods of renewal).

            4.a. The applicant for the APR1400 design shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) for the term of the certification (including any period of renewal).
            b. An applicant or licensee who references this appendix shall maintain a copy of the aircraft impact assessment performed to comply with the requirements of 10 CFR 50.150(a) throughout the pendency of the application and for the term of the license (including any periods of renewal).
            B. Reporting
            1. An applicant or licensee who references this appendix shall submit a report to the NRC containing a brief description of any plant-specific departures from the DCD, including a summary of the evaluation of each departure. This report must be filed in accordance with the filing requirements applicable to reports in § 52.3.
            2. An applicant or licensee who references this appendix shall submit updates to its plant-specific DCD, which reflect the generic changes to and plant-specific departures from the generic DCD made under Section VIII of this appendix. These updates shall be filed under the filing requirements applicable to final safety analysis report updates in 10 CFR 50.71(e) and 52.3.
            3. The reports and updates required by paragraphs X.B.1 and X.B.2 of this appendix must be submitted as follows:
            a. On the date that an application for a license referencing this appendix is submitted, the application must include the report and any updates to the generic DCD.
            b. During the interval from the date of application for a license to the date the Commission makes its finding required by § 52.103(g), the report must be submitted semi-annually. Updates to the plant-specific DCD must be submitted annually and may be submitted along with amendments to the application.
            c. After the Commission makes the finding required by § 52.103(g), the reports and updates to the plant-specific DCD must be submitted, along with updates to the site-specific portion of the final safety analysis report for the facility, at the intervals required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at shorter intervals as specified in the license.
            [84 FR 23452, May 22, 2019]

            Appendixes G-M to Part 52 [Reserved]

            Pt. 52, App. N
            Appendix N to Part 52—Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites
            The Commission's regulations in part 2 of this chapter specifically provide for the holding of hearings on particular issues separately from other issues involved in hearings in licensing proceedings, and for the consolidation of adjudicatory proceedings and of the presentations of parties in adjudicatory proceedings such as licensing proceedings (§§ 2.316 and 2.317 of this chapter).
            This appendix sets out the particular requirements and provisions applicable to situations in which applications for combined licenses under subpart C of this part are filed by one or more applicants for licenses to construct and operate nuclear power reactors of identical design (“common design”) to be located at multiple sites. 1

                1 If the design for the power reactor(s) proposed in a particular application is not identical to the others, that application may not be processed under this appendix and subpart D of part 2 of this chapter.

            1. Except as otherwise specified in this appendix or as the context otherwise indicates, the provisions of subpart C of this part and subpart D of part 2 of this chapter apply to combined license applications subject to this appendix.
            2. Each combined license application submitted pursuant to this appendix must be submitted as specified in § 52.75 and 10 CFR 2.101. Each application must state that the applicant wishes to have the application considered under 10 CFR part 52, appendix N, and must list each of the applications to be treated together under this appendix.

            3. Each application must include the information required by §§ 52.77, 52.79, and 52.80(a), provided however, that the application must identify the common design, and, if applicable, reference a standard design certification under subpart B of this part, or the use of a reactor manufactured under subpart F of this part. The final safety analysis report for each application must either incorporate by reference or include the final safety analysis of the common design, including, if applicable, the final safety analysis report for the referenced design certification or the manufactured reactor. 2

                2 As used in this appendix, the design of a nuclear power reactor included in a single referenced safety analysis report means the design of those structures, systems, and components important to radiological health and safety and the common defense and security.

            4. Each combined license application submitted pursuant to this appendix must contain an environmental report as required by § 52.80(b), and which complies with the applicable provisions of 10 CFR part 51, provided, however, that the application may incorporate by reference a single environmental report on the environmental impacts of the common design.

            5. Upon a determination that each application is acceptable for docketing under 10 CFR 2.101, each application will be docketed and a notice of docketing for each application will be published in the Federal Register, in accordance with 10 CFR 2.104, provided, however, that the notice must state that the application will be processed under the provisions of 10 CFR part 52, appendix N, and subpart D of part 2 of this chapter. As the discretion of the Commission, a single notice of docketing for multiple applications may be published in the Federal Register.

            6. The NRC staff shall prepare draft and final environmental impact statements for each of the applications under part 51 of this chapter. Scoping under 10 CFR 51.28 and 51.29 for each of the combined license applications may be conducted simultaneously and joint scoping may be conducted with respect to the environmental issues relevant to the common design.
            If the applications reference a standard design certification, then the environmental impact statement for each of the applications must incorporate by reference the design certification environmental assessment. If the applications do not reference a standard design certification, then the NRC staff shall prepare draft and final supplemental environmental impact statements which address severe accident mitigation design alternatives for the common design, which must be incorporated by reference into the environmental impact statement prepared for each application. Scoping under 10 CFR 51.28 and 51.29 for the supplemental environmental impact statement may be conducted simultaneously, and may be part of the scoping for each of the combined license applications.

            7. The ACRS shall report on each of the applications as required by § 52.87. Each report must be limited to those safety matters for each application which are not relevant to the common design. In addition, the ACRS shall separately report on the safety of the common design, provided, however, that the report need not address the safety of a referenced standard design certification or reactor manufactured under subpart F of this part.
            8. The Commission shall designate a presiding officer to conduct the proceeding with respect to the health and safety, common defense and security, and environmental matters relating to the common design. The hearing will be governed by the applicable provisions of subparts A, C, G, L, N, and O of part 2 of this chapter relating to applications for combined licenses. The presiding officer shall issue a partial initial decision on the common design.

        PART 53 [RESERVED]

        Pt. 54
        PART 54—REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR POWER PLANTS

            General Provisions
            Sec.
            54.1
            Purpose.
            54.3
            Definitions.
            54.4
            Scope.
            54.5
            Interpretations.
            54.7
            Written communications.
            54.9
            Information collection requirements: OMB approval.
            54.11
            Public inspection of applications.
            54.13
            Completeness and accuracy of information.
            54.15
            Specific exemptions.
            54.17
            Filing of application.
            54.19
            Contents of application—general information.
            54.21
            Contents of application—technical information.
            54.22
            Contents of application—technical specifications.
            54.23
            Contents of application—environmental information.
            54.25
            Report of the Advisory Committee on Reactor Safeguards.
            54.27
            Hearings.
            54.29
            Standards for issuance of a renewed license.
            54.30
            Matters not subject to a renewal review.
            54.31
            Issuance of a renewed license.
            54.33
            Continuation of CLB and conditions of renewed license.
            54.35
            Requirements during term of renewed license.
            54.37
            Additional records and recordkeeping requirements.
            54.41
            Violations.
            54.43
            Criminal penalties.

          Authority:

          Atomic Energy Act of 1954, secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 223, 234 (42 U.S.C. 2132, 2133, 2134, 2136, 2137, 2201, 2231, 2232, 2233, 2236, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206 (42 U.S.C. 5841, 5842, 5846); 44 U.S.C. 3504 note.

          Section 54.17 also issued under E.O. 12829, 58 FR 3479, 3 CFR, 1993 Comp., p. 570; E.O. 13526, 75 FR 707, 3 CFR, 2009 Comp., p. 298; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p. 391.

          Source:
          60 FR 22491, May 8, 1995, unless otherwise noted.

          General Provisions

            § 54.1
            Purpose.
            This part governs the issuance of renewed operating licenses and renewed combined licenses for nuclear power plants licensed pursuant to Sections 103 or 104b of the Atomic Energy Act of 1954, as amended, and Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242).
            [72 FR 49560, Aug. 28, 2007]

            § 54.3
            Definitions.
            (a) As used in this part,

              Current licensing basis (CLB) is the set of NRC requirements applicable to a specific plant and a licensee's written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modifications and additions to such commitments over the life of the license) that are docketed and in effect. The CLB includes the NRC regulations contained in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 51, 52, 54, 55, 70, 72, 73, 100 and appendices thereto; orders; license conditions; exemptions; and technical specifications. It also includes the plant-specific design-basis information defined in 10 CFR 50.2 as documented in the most recent final safety analysis report (FSAR) as required by 10 CFR 50.71 and the licensee's commitments remaining in effect that were made in docketed licensing correspondence such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments documented in NRC safety evaluations or licensee event reports.

              Integrated plant assessment (IPA) is a licensee assessment that demonstrates that a nuclear power plant facility's structures and components requiring aging management review in accordance with § 54.21(a) for license renewal have been identified and that the effects of aging on the functionality of such structures and components will be managed to maintain the CLB such that there is an acceptable level of safety during the period of extended operation.

              Nuclear power plant means a nuclear power facility of a type described in 10 CFR 50.21(b) or 50.22.

              Renewed combined license means a combined license originally issued under part 52 of this chapter for which an application for renewal is filed in accordance with 10 CFR 52.107 and issued under this part.

              Time-limited aging analyses, for the purposes of this part, are those licensee calculations and analyses that:
            (1) Involve systems, structures, and components within the scope of license renewal, as delineated in § 54.4(a);
            (2) Consider the effects of aging;
            (3) Involve time-limited assumptions defined by the current operating term, for example, 40 years;
            (4) Were determined to be relevant by the licensee in making a safety determination;
            (5) Involve conclusions or provide the basis for conclusions related to the capability of the system, structure, and component to perform its intended functions, as delineated in § 54.4(b); and
            (6) Are contained or incorporated by reference in the CLB.
            (b) All other terms in this part have the same meanings as set out in 10 CFR 50.2 or Section 11 of the Atomic Energy Act, as applicable.
            [60 FR 22491, May 8, 1995, as amended at 72 FR 49560, Aug. 28, 2007]

            § 54.4
            Scope.
            (a) Plant systems, structures, and components within the scope of this part are—
            (1) Safety-related systems, structures, and components which are those relied upon to remain functional during and following design-basis events (as defined in 10 CFR 50.49 (b)(1)) to ensure the following functions—
            (i) The integrity of the reactor coolant pressure boundary;
            (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition; or

            (iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in § 50.34(a)(1), § 50.67(b)(2), or § 100.11 of this chapter, as applicable.
            (2) All nonsafety-related systems, structures, and components whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraphs (a)(1) (i), (ii), or (iii) of this section.
            (3) All systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for fire protection (10 CFR 50.48), environmental qualification (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61), anticipated transients without scram (10 CFR 50.62), and station blackout (10 CFR 50.63).
            (b) The intended functions that these systems, structures, and components must be shown to fulfill in § 54.21 are those functions that are the bases for including them within the scope of license renewal as specified in paragraphs (a) (1)-(3) of this section.
            [60 FR 22491, May 8, 1995, as amended at 61 FR 65175, Dec. 11, 1996; 64 FR 72002, Dec. 23, 1999]

            § 54.5
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

            § 54.7
            Written communications.
            All applications, correspondence, reports, and other written communications shall be filed in accordance with applicable portions of 10 CFR 50.4.

            § 54.9
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501, et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0155.
            (b) The approved information requirements contained in this part appear in §§ 54.13, 54.15, 54.17, 54.19, 54.21, 54.22, 54.23, 54.33, and 54.37.
            [60 FR 22491, May 8, 1995, as amended at 62 FR 52188, Oct. 6, 1997; 67 FR 67100, Nov. 4, 2002]

            § 54.11
            Public inspection of applications.
            Applications and documents submitted to the Commission in connection with renewal applications may be made available for public inspection in accordance with the provisions of the regulations contained in 10 CFR part 2.

            § 54.13
            Completeness and accuracy of information.
            (a) Information provided to the Commission by an applicant for a renewed license or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant must be complete and accurate in all material respects.
            (b) Each applicant shall notify the Commission of information identified by the applicant as having, for the regulated activity, a significant implication for public health and safety or common defense and security. An applicant violates this paragraph only if the applicant fails to notify the Commission of information that the applicant has identified as having a significant implication for public health and safety or common defense and security. Notification must be provided to the Administrator of the appropriate regional office within 2 working days of identifying the information. This requirement is not applicable to information that is already required to be provided to the Commission by other reporting or updating requirements.

            § 54.15
            Specific exemptions.
            Exemptions from the requirements of this part may be granted by the Commission in accordance with 10 CFR 50.12.

            § 54.17
            Filing of application.
            (a) The filing of an application for a renewed license must be in accordance with subpart A of 10 CFR part 2 and 10 CFR 50.4 and 50.30.
            (b) Any person who is a citizen, national, or agent of a foreign country, or any corporation, or other entity which the Commission knows or has reason to know is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government, is ineligible to apply for and obtain a renewed license.
            (c) An application for a renewed license may not be submitted to the Commission earlier than 20 years before the expiration of the operating license or combined license currently in effect.
            (d) An applicant may combine an application for a renewed license with applications for other kinds of licenses.
            (e) An application may incorporate by reference information contained in previous applications for licenses or license amendments, statements, correspondence, or reports filed with the Commission, provided that the references are clear and specific.
            (f) If the application contains Restricted Data or other defense information, it must be prepared in such a manner that all Restricted Data and other defense information are separated from unclassified information in accordance with 10 CFR 50.33(j).
            (g) As part of its application, and in any event before the receipt of Restricted Data or classified National Security Information or the issuance of a renewed license, the applicant shall agree in writing that it will not permit any individual to have access to or any facility to possess Restricted Data or classified National Security Information until the individual and/or facility has been approved for such access under the provisions of 10 CFR parts 25 and/or 95. The agreement of the applicant in this regard shall be deemed part of the renewed license, whether so stated therein or not.
            [60 FR 22491, May 8, 1995, as amended at 62 FR 17690, Apr. 11, 1997; 72 FR 49560, Aug. 28, 2007]

            § 54.19
            Contents of application—general information.
            (a) Each application must provide the information specified in 10 CFR 50.33 (a) through (e), (h), and (i). Alternatively, the application may incorporate by reference other documents that provide the information required by this section.
            (b) Each application must include conforming changes to the standard indemnity agreement, 10 CFR 140.92, Appendix B, to account for the expiration term of the proposed renewed license.

            § 54.21
            Contents of application—technical information.
            Each application must contain the following information:
            (a) An integrated plant assessment (IPA). The IPA must—
            (1) For those systems, structures, and components within the scope of this part, as delineated in § 54.4, identify and list those structures and components subject to an aging management review. Structures and components subject to an aging management review shall encompass those structures and components—

            (i) That perform an intended function, as described in § 54.4, without moving parts or without a change in configuration or properties. These structures and components include, but are not limited to, the reactor vessel, the reactor coolant system pressure boundary, steam generators, the pressurizer, piping, pump casings, valve bodies, the core shroud, component supports, pressure retaining boundaries, heat exchangers, ventilation ducts, the containment, the containment liner, electrical and mechanical penetrations, equipment hatches, seismic Category I structures, electrical cables and connections, cable trays, and electrical cabinets, excluding, but not limited to, pumps (except casing), valves (except body), motors, diesel generators, air compressors, snubbers, the control rod drive, ventilation dampers, pressure transmitters, pressure indicators, water level indicators, switchgears, cooling fans, transistors, batteries, breakers, relays, switches, power inverters, circuit boards, battery chargers, and power supplies; and

            (ii) That are not subject to replacement based on a qualified life or specified time period.
            (2) Describe and justify the methods used in paragraph (a)(1) of this section.
            (3) For each structure and component identified in paragraph (a)(1) of this section, demonstrate that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the CLB for the period of extended operation.
            (b) CLB changes during NRC review of the application. Each year following submittal of the license renewal application and at least 3 months before scheduled completion of the NRC review, an amendment to the renewal application must be submitted that identifies any change to the CLB of the facility that materially affects the contents of the license renewal application, including the FSAR supplement.
            (c) An evaluation of time-limited aging analyses. (1) A list of time-limited aging analyses, as defined in § 54.3, must be provided. The applicant shall demonstrate that—
            (i) The analyses remain valid for the period of extended operation;
            (ii) The analyses have been projected to the end of the period of extended operation; or
            (iii) The effects of aging on the intended function(s) will be adequately managed for the period of extended operation.
            (2) A list must be provided of plant-specific exemptions granted pursuant to 10 CFR 50.12 and in effect that are based on time-limited aging analyses as defined in § 54.3. The applicant shall provide an evaluation that justifies the continuation of these exemptions for the period of extended operation.
            (d) An FSAR supplement. The FSAR supplement for the facility must contain a summary description of the programs and activities for managing the effects of aging and the evaluation of time-limited aging analyses for the period of extended operation determined by paragraphs (a) and (c) of this section, respectively.

            § 54.22
            Contents of application—technical specifications.
            Each application must include any technical specification changes or additions necessary to manage the effects of aging during the period of extended operation as part of the renewal application. The justification for changes or additions to the technical specifications must be contained in the license renewal application.

            § 54.23
            Contents of application—environmental information.
            Each application must include a supplement to the environmental report that complies with the requirements of subpart A of 10 CFR part 51.

            § 54.25
            Report of the Advisory Committee on Reactor Safeguards.
            Each renewal application will be referred to the Advisory Committee on Reactor Safeguards for a review and report. Any report will be made part of the record of the application and made available to the public, except to the extent that security classification prevents disclosure.

            § 54.27
            Hearings.

            A notice of an opportunity for a hearing will be published in the Federal Register in accordance with 10 CFR 2.105 and 2.309. In the absence of a request for a hearing filed within 60 days by a person whose interest may be affected, the Commission may issue a renewed operating license or renewed combined license without a hearing upon a 30-day notice and publication in the Federal Register of its intent to do so.
            [77 FR 46600, Aug. 3, 2012]

            § 54.29
            Standards for issuance of a renewed license.
            A renewed license may be issued by the Commission up to the full term authorized by § 54.31 if the Commission finds that:

            (a) Actions have been identified and have been or will be taken with respect to the matters identified in paragraphs (a)(1) and (a)(2) of this section, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the CLB, and that any changes made to the plant's CLB in order to comply with this paragraph are in accord with the Act and the Commission's regulations. These matters are:
            (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under § 54.21(a)(1); and
            (2) time-limited aging analyses that have been identified to require review under § 54.21(c).
            (b) Any applicable requirements of subpart A of 10 CFR part 51 have been satisfied.
            (c) Any matters raised under § 2.335 have been addressed.
            [60 FR 22491, May 8, 1995, as amended at 69 FR 2279, Jan. 14, 2004]

            § 54.30
            Matters not subject to a renewal review.
            (a) If the reviews required by § 54.21 (a) or (c) show that there is not reasonable assurance during the current license term that licensed activities will be conducted in accordance with the CLB, then the licensee shall take measures under its current license, as appropriate, to ensure that the intended function of those systems, structures or components will be maintained in accordance with the CLB throughout the term of its current license.
            (b) The licensee's compliance with the obligation under Paragraph (a) of this section to take measures under its current license is not within the scope of the license renewal review.

            § 54.31
            Issuance of a renewed license.
            (a) A renewed license will be of the class for which the operating license or combined license currently in effect was issued.
            (b) A renewed license will be issued for a fixed period of time, which is the sum of the additional amount of time beyond the expiration of the operating license or combined license (not to exceed 20 years) that is requested in a renewal application plus the remaining number of years on the operating license or combined license currently in effect. The term of any renewed license may not exceed 40 years.
            (c) A renewed license will become effective immediately upon its issuance, thereby superseding the operating license or combined license previously in effect. If a renewed license is subsequently set aside upon further administrative or judicial appeal, the operating license or combined license previously in effect will be reinstated unless its term has expired and the renewal application was not filed in a timely manner.
            (d) A renewed license may be subsequently renewed in accordance with all applicable requirements.
            [60 FR 22491, May 8, 1995, as amended at 72 FR 49560, Aug. 28, 2007]

            § 54.33
            Continuation of CLB and conditions of renewed license.
            (a) Whether stated therein or not, each renewed license will contain and otherwise be subject to the conditions set forth in 10 CFR 50.54.
            (b) Each renewed license will be issued in such form and contain such conditions and limitations, including technical specifications, as the Commission deems appropriate and necessary to help ensure that systems, structures, and components subject to review in accordance with § 54.21 will continue to perform their intended functions for the period of extended operation. In addition, the renewed license will be issued in such form and contain such conditions and limitations as the Commission deems appropriate and necessary to help ensure that systems, structures, and components associated with any time-limited aging analyses will continue to perform their intended functions for the period of extended operation.

            (c) Each renewed license will include those conditions to protect the environment that were imposed pursuant to 10 CFR 50.36b and that are part of the CLB for the facility at the time of issuance of the renewed license. These conditions may be supplemented or amended as necessary to protect the environment during the term of the renewed license and will be derived from information contained in the supplement to the environmental report submitted pursuant to 10 CFR part 51, as analyzed and evaluated in the NRC record of decision. The conditions will identify the obligations of the licensee in the environmental area, including, as appropriate, requirements for reporting and recordkeeping of environmental data and any conditions and monitoring requirements for the protection of the nonaquatic environment.
            (d) The licensing basis for the renewed license includes the CLB, as defined in § 54.3(a); the inclusion in the licensing basis of matters such as licensee commitments does not change the legal status of those matters unless specifically so ordered pursuant to paragraphs (b) or (c) of this section.

            § 54.35
            Requirements during term of renewed license.
            During the term of a renewed license, licensees shall be subject to and shall continue to comply with all Commission regulations contained in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 51, 52, 54, 55, 70, 72, 73, and 100, and the appendices to these parts that are applicable to holders of operating licenses or combined licenses, respectively.
            [72 FR 49560, Aug. 28, 2007]

            § 54.37
            Additional records and recordkeeping requirements.
            (a) The licensee shall retain in an auditable and retrievable form for the term of the renewed operating license or renewed combined license all information and documentation required by, or otherwise necessary to document compliance with, the provisions of this part.
            (b) After the renewed license is issued, the FSAR update required by 10 CFR 50.71(e) must include any systems, structures, and components newly identified that would have been subject to an aging management review or evaluation of time-limited aging analyses in accordance with § 54.21. This FSAR update must describe how the effects of aging will be managed such that the intended function(s) in § 54.4(b) will be effectively maintained during the period of extended operation.
            [60 FR 22491, May 8, 1995, as amended at 72 FR 49560, Aug. 28, 2007]

            § 54.41
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of the following acts—
            (1) The Atomic Energy Act of 1954, as amended.
            (2) Title II of the Energy Reorganization Act of 1974, as amended or
            (3) A regulation or order issued pursuant to those acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under Section 234 of the Atomic Energy Act—
            (1) For violations of the following—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section;
            (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under Section 186 of the Atomic Energy Act of 1954, as amended.

            § 54.43
            Criminal penalties.
            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violations of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 54 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in part 54 that are not issued under Sections 161b, 161i, or 161o for the purposes of Section 223 are as follows: §§ 54.1, 54.3, 54.4, 54.5, 54.7, 54.9, 54.11, 54.15, 54.17, 54.19, 54.21, 54.22, 54.23, 54.25, 54.27, 54.29, 54.31, 54.41, and 54.43.

        Pt. 55
        PART 55—OPERATORS' LICENSES

            Subpart A—General Provisions
            Sec.
            55.1
            Purpose.
            55.2
            Scope.
            55.3
            License requirements.
            55.4
            Definitions.
            55.5
            Communications.
            55.6
            Interpretations.
            55.7
            Additional requirements.

            55.8
            Information collection requirements: OMB approval.
            55.9
            Completeness and accuracy of information.

            Subpart B—Exemptions
            55.11
            Specific exemptions.
            55.13
            General exemptions.

            Subpart C—Medical Requirements
            55.21
            Medical examination.
            55.23
            Certification.
            55.25
            Incapacitation because of disability or illness.
            55.27
            Documentation.

            Subpart D—Applications
            55.31
            How to apply.
            55.33
            Disposition of an initial application.
            55.35
            Re-applications.

            Subpart E—Written Examinations and Operating Tests
            55.40
            Implementation.
            55.41
            Written examination: Operators.
            55.43
            Written examination: Senior operators.
            55.45
            Operating tests.
            55.46
            Simulation facilities.
            55.47
            Waiver of examination and test requirements.
            55.49
            Integrity of examinations and tests.

            Subpart F—Licenses
            55.51
            Issuance of licenses.
            55.53
            Conditions of licenses.
            55.55
            Expiration.
            55.57
            Renewal of licenses.
            55.59
            Requalification.

            Subpart G—Modification and Revocation of Licenses
            55.61
            Modification and revocation of licenses.

            Subpart H—Enforcement
            55.71
            Violations.
            55.73
            Criminal penalties.

          Authority:
          Atomic Energy Act of 1954, secs. 107, 161, 181, 182, 183, 186, 187, 223, 234 (42 U.S.C. 2137, 2201, 2231, 2232, 2233, 2236, 2237, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); Nuclear Waste Policy Act of 1982, sec. 306 (42 U.S.C. 10226); 44 U.S.C. 3504 note.
          Section 55.61 also issued under Atomic Energy Act secs. 186, 187 (42 U.S.C. 2236, 2237).

          Source:
          52 FR 9460, Mar. 25, 1987, unless otherwise noted.

          Editorial Note:
          Nomenclature changes to part 55 appear at 80 FR 74980, Dec. 1, 2015.

          Subpart A—General Provisions

            § 55.1
            Purpose.
            The regulations in this part:
            (a) Establish procedures and criteria for the issuance of licenses to operators and senior operators of utilization facilities licensed under the Atomic Energy Act of 1954, as amended, or Section 202 of the Energy Reorganization Act of 1974, as amended, and part 50, part 52, or part 54 of this chapter,
            (b) Provide for the terms and conditions upon which the Commission will issue or modify these licenses, and
            (c) Provide for the terms and conditions to maintain and renew these licenses.
            [52 FR 9460, Mar. 25, 1987, as amended at 72 FR 49560, Aug. 28, 2007]

            § 55.2
            Scope.
            The regulations in this part apply to—
            (a) Any individual who manipulates the controls of any utilization facility licensed under parts 50, 52, or 54 of this chapter,
            (b) Any individual designated by a facility licensee to be responsible for directing any licensed activity of a licensed operator.
            (c) Any facility license.
            [52 FR 9460, Mar. 25, 1987, as amended at 59 FR 5938, Feb. 9, 1994; 72 FR 49560, Aug. 28, 2007]

            § 55.3
            License requirements.
            A person must be authorized by a license issued by the Commission to perform the function of an operator or a senior operator as defined in this part.

            § 55.4
            Definitions.
            As used in this part:

              Act means the Atomic Energy Act of 1954, including any amendments to the Act.

              Actively performing the functions of an operator or senior operator means that an individual has a position on the shift crew that requires the individual to be licensed as defined in the facility's technical specifications, and that the individual carries out and is responsible for the duties covered by that position.

              Commission means the Nuclear Regulatory Commission or its duly authorized representatives.

              Controls when used with respect to a nuclear reactor means apparatus and mechanisms the manipulation of which directly affects the reactivity or power level of the reactor.

              Facility means any utilization facility as defined in part 50 of this chapter. In cases for which a license is issued for operation of two or more facilities, facility means all facilities identified in the license.

              Facility licensee means an applicant for or holder of a license for a facility.

              Licensee means an individual licensed operator or senior operator.

              Operator means any individual licensed under this part to manipulate a control of a facility.

              Performance testing means testing conducted to verify a simulation facility's performance as compared to actual or predicted reference plant performance.

              Physician means an individual licensed by a State or territory of the United States, the District of Columbia or the Commonwealth of Puerto Rico to dispense drugs in the practice of medicine.

              Plant-referenced simulator means a simulator modeling the systems of the reference plant with which the operator interfaces in the control room, including operating consoles, and which permits use of the reference plant's procedures.

              Reference plant means the specific nuclear power plant from which a simulation facility's control room configuration, system control arrangement, and design data are derived.

              Senior operator means any individual licensed under this part to manipulate the controls of a facility and to direct the licensed activities of licensed operators.

              Simulation facility means one or more of the following components, alone or in combination: used for either the partial conduct of operating tests for operators, senior operators, and license applicants, or to establish on-the-job training and experience prerequisites for operator license eligibility:
            (1) A plant-referenced simulator;
            (2) A Commission-approved simulator under § 55.46(b); or
            (3) Another simulation device, including part-task and limited scope simulation devices, approved under § 55.46(b).

              Systems approach to training means a training program that includes the following five elements:
            (1) Systematic analysis of the jobs to be performed.
            (2) Learning objectives derived from the analysis which describe desired performance after training.
            (3) Training design and implementation based on the learning objectives.
            (4) Evaluation of trainee mastery of the objectives during training.
            (5) Evaluation and revision of the training based on the performance of trained personnel in the job setting.

              United States, when used in a geographical sense, includes Puerto Rico and all territories and possessions of the United States.
            [52 FR 9460, Mar. 25, 1987, as amended at 66 FR 52667, Oct. 17, 2001]

            § 55.5
            Communications.
            (a) Except as provided under a regional licensing program identified in paragraph (b) of this section, an applicant or licensee or facility licensee shall submit any communication or report concerning the regulations in this part and shall submit any application filed under these regulations to the Commission as follows:
            (1) By mail addressed to—Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; or
            (2) By delivery in person to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland, or

            (3) Where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.
            (b)(1) Except for test and research reactor facilities, the Director, Office of Nuclear Reactor Regulation, has delegated to the Regional Administrators of Regions I, II, III, and IV authority and responsibility under the regulations in this part for the issuance and renewal of licenses for operators and senior operators of nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter and located in these regions.
            (2) Any application for a license or license renewal filed under the regulations in this part involving a nuclear power reactor licensed under 10 CFR part 50 or part 52 of this chapter and any related inquiry, communication, information, or report must be submitted to the Regional Administrator by an appropriate method listed in paragraph (a) of this section. The Regional Administrator or the Administrator's designee will transmit to the Director, Office of Nuclear Reactor Regulation, any matter that is not within the scope of the Regional Administrator's delegated authority.

            (i) If the nuclear power reactor is located in Region I, submissions must be made to the Regional Administrator of Region I. Submissions by mail or hand delivery must be addressed to the Administrator at U.S. Nuclear Regulatory Commission, 2100 Renaissance Boulevard, Suite 100, King of Prussia, PA 19406-2713; where email is appropriate it should be addressed to RidsRgn1MailCenter.Resource@nrc.gov.

            (ii) If the nuclear power reactor is located in Region II, submissions must be made to the Regional Administrator of Region II. Submissions by mail or hand delivery must be addressed to the Regional Administrator at U.S. Nuclear Regulatory Commission, 245 Peachtree Center Avenue, NE., Suite 1200, Atlanta, Georgia 30303-1257. Where e-mail is appropriate, it should be addressed to RidsRgn2MailCenter@nrc.gov.

            (iii) If the nuclear power reactor is located in Region III, submissions must be made to the Regional Administrator of Region III. Submissions by mail or hand delivery must be addressed to the Administrator at U.S. Nuclear Regulatory Commission, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; where e-mail is appropriate it should be addressed to RidsRgn3MailCenter@nrc.gov.

            (iv) If the nuclear power reactor is located in Region IV, submissions must be made to the Regional Administrator of Region IV. Submission by mail or hand delivery must be addressed to the Administrator at U.S. Nuclear Regulatory Commission, 1600 E. Lamar Blvd., Arlington, TX 76011-4511; where email is appropriate, it should be addressed to RidsRgn4MailCenter@nrc.gov.

            (3) Any application for a license or license renewal filed under the regulations in this part and all other submissions involving a test and research reactor or non-power reactor facility licensed under 10 CFR part 50 and any related inquiry, communication, information, or report must be submitted to the Office of Nuclear Reactor Regulation, Director of the Division of Advanced Reactors and Non-Power Production and Utilization Facilities at the NRC's headquarters, by an appropriate method listed in paragraph (a) of this section.
            [52 FR 9460, Mar. 25, 1987]

              Editorial Note:
              For Federal Register citations affecting § 55.5, see the List of CFR Sections Affected, which appears in the Finding Aids section of the printed volume and at www.govinfo.gov.

            § 55.6
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

            § 55.7
            Additional requirements.

            The Commission may, by rule, regulation, or order, impose upon any licensee such requirements, in addition to those established in the regulations in this part, as it deems appropriate or necessary to protect health and to minimize danger to life or property.

            § 55.8
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0018.
            (b) The approved information collection requirements contained in this part appear in §§ 55.11, 55.25, 55.27, 55.31, 55.35, 55.40, 55.41, 55.43, 55.45, 55.47, 55.53, 55.57, and 55.59.
            (c) This part contains information collection requirements in addition to those approved under the control number specified in paragraph (a) of this section. These information collection requirements and the control numbers under which they are approved are as follows:
            (1) In §§ 55.23, 55.25, 55.27, 55.31, NRC Form 396 is approved under control number 3150-0024.
            (2) In §§ 55.31, 55.35, 55.47, and 55.57, NRC Form 398 is approved under control number 3150-0090.
            [62 FR 52188, Oct. 6, 1997, as amended at 64 FR 19878, Apr. 23, 1999; 66 FR 52667, Oct. 17, 2001; 67 FR 67100, Nov. 4, 2002]

            § 55.9
            Completeness and accuracy of information.
            Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.
            [52 FR 49372, Dec. 31, 1987]

          Subpart B—Exemptions

            § 55.11
            Specific exemptions.
            The Commission may, upon application by an interested person, or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property and are otherwise in the public interest.

            § 55.13
            General exemptions.
            The regulations in this part do not require a license for an individual who—
            (a) Under the direction and in the presence of a licensed operator or senior operator, manipulates the controls of—
            (1) A research or training reactor as part of the individual's training as a student, or
            (2) A facility as a part of the individual's training in a facility licensee's training program as approved by the Commission to qualify for an operator license under this part.
            (b) Under the direction and in the presence of a licensed senior operator, manipulates the controls of a facility to load or unload the fuel into, out of, or within the reactor vessel.

          Subpart C—Medical Requirements

            § 55.21
            Medical examination.
            An applicant for a license shall have a medical examination by a physician. A licensee shall have a medical examination by a physician every two years. The physician shall determine that the applicant or licensee meets the requirements of § 55.33(a)(1).

            § 55.23
            Certification.

            To certify the medical fitness of the applicant, an authorized representative of the facility licensee shall complete and sign NRC Form 396, “Certification of Medical Examination by Facility Licensee,” which can be obtained by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by calling (301) 415-7232, or by visiting the NRC's Web site at http://www.nrc.gov and selecting forms from the index found on the home page.
            (a) Form NRC-396 must certify that a physician has conducted the medical examination of the applicant as required in § 55.21.
            (b) When the certification requests a conditional license based on medical evidence, the medical evidence must be submitted on NRC Form 396 to the Commission and the Commission then makes a determination in accordance with § 55.33.
            [52 FR 9460, Mar. 25, 1987, as amended at 53 FR 43421, Oct. 27, 1988; 68 FR 58813, Oct. 10, 2003; 73 FR 30458, May 28, 2008]

            § 55.25
            Incapacitation because of disability or illness.
            If, during the term of the license, the licensee develops a permanent physical or mental condition that causes the licensee to fail to meet the requirements of § 55.21 of this part, the facility licensee shall notify the Commission, within 30 days of learning of the diagnosis, in accordance with § 50.74(c). For conditions for which a conditional license (as described in § 55.33(b) of this part) is requested, the facility licensee shall provide medical certification on Form NRC 396 to the Commission (as described in § 55.23 of this part).
            [60 FR 13617, Mar. 14, 1995]

            § 55.27
            Documentation.
            The facility licensee shall document and maintain the results of medical qualifications data, test results, and each operator's or senior operator's medical history for the current license period and provide the documentation to the Commission upon request. The facility licensee shall retain this documentation while an individual performs the functions of an operator or senior operator.

          Subpart D—Applications

            § 55.31
            How to apply.
            (a) The applicant shall:

            (1) Complete NRC Form 398, “Personal Qualification Statement—Licensee,” which can be obtained by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by calling (301) 415-7232, or by visiting the NRC's Web site at http://www.nrc.gov and selecting forms from the index found on the home page;
            (2) File an original of NRC Form 398, together with the information required in paragraphs (a)(3), (4), (5) and (6) of this section, with the appropriate Regional Administrator;
            (3) Submit a written request from an authorized representative of the facility licensee by which the applicant will be employed that the written examination and operating test be administered to the applicant;
            (4) Provide evidence that the applicant has successfully completed the facility licensee's requirements to be licensed as an operator or senior operator and of the facility licensee's need for an operator or a senior operator to perform assigned duties. An authorized representative of the facility licensee shall certify this evidence on Form NRC-398. This certification must include details of the applicant's qualifications, and details on courses of instruction administered by the facility licensee, and describe the nature of the training received at the facility, and the startup and shutdown experience received. In lieu of these details, the Commission may accept certification that the applicant has successfully completed a Commission-approved training program that is based on a systems approach to training and that uses a simulation facility acceptable to the Commission under § 55.45(b) of this part;

            (5) Provide evidence that the applicant, as a trainee, has successfully manipulated the controls of either the facility for which a license is sought or a plant-referenced simulator that meets the requirements of § 55.46(c). At a minimum, five significant control manipulations must be performed that affect reactivity or power level. Control manipulations performed on the plant-referenced simulator may be chosen from a representative sampling of the control manipulations and plant evolutions described in § 55.59(c)(3)(i)(A-F), (R), (T), (W), and (X) of this part, as applicable to the design of the plant for which the license application is submitted. For licensed operators applying for a senior operator license, certification that the operator has successfully operated the controls of the facility as a licensed operator shall be accepted; and
            (6) Provide certification by the facility licensee of medical condition and general health on Form NRC-396, to comply with §§ 55.21, 55.23 and 55.33(a)(1).
            (b) The Commission may at any time after the application has been filed, and before the license has expired, require futher information under oath or affirmation in order to enable it to determine whether to grant or deny the application or whether to revoke, modify, or suspend the license.
            (c) An applicant whose application has been denied because of a medical condition or general health may submit a further medical report at any time as a supplement to the application.
            (d) Each application and statement must contain complete and accurate disclosure as to all matters required to be disclosed. The applicant shall sign statements required by paragraphs (a) (1) and (2) of this section.
            [52 FR 9460, Mar. 25, 1987, as amended at 53 FR 43421, Oct. 27, 1988; 66 FR 52667, Oct. 17, 2001; 68 FR 58813, Oct. 10, 2003; 73 FR 30458, May 28, 2008]

            § 55.33
            Disposition of an initial application.
            (a) Requirements for the approval of an initial application. The Commission will approve an initial application for a license pursuant to the regulations in this part, if it finds that—
            (1) Health. The applicants medical condition and general health will not adversely affect the performance of assigned operator job duties or cause operational errors endangering public health and safety. The Commission will base its finding upon the certification by the facility licensee as detailed in § 55.23.
            (2) Written examination and operating test. The applicant has passed the requisite written examination and operating test in accordance with §§ 55.41 and 55.45 or 55.43 and 55.45. These examinations and tests determine whether the applicant for an operator's license has learned to operate a facility competently and safely, and additionally, in the case of a senior operator, whether the applicant has learned to direct the licensed activities of licensed operators competently and safely.
            (b) Conditional license. If an applicant's general medical condition does not meet the minimum standards under § 55.33(a)(1) of this part, the Commission may approve the application and include conditions in the license to accommodate the medical defect. The Commission will consider the recommendations and supporting evidence of the facility licensee and of the examining physician (provided on Form NRC-396) in arriving at its decision.

            § 55.35
            Re-applications.
            (a) An applicant whose application for a license has been denied because of failure to pass the written examination or operating test, or both, may file a new application two months after the date of denial. The application must be submitted on Form NRC-398 and include a statement signed by an authorized representative of the facility licensee by whom the applicant will be employed that states in detail the extent of the applicant's additional training since the denial and certifies that the applicant is ready for re-examination. An applicant may file a third application six months after the date of denial of the second application, and may file further successive applications two years after the date of denial of each prior application. The applicant shall submit each successive application on Form NRC-398 and include a statement of additional training.
            (b) An applicant who has passed either the written examination or operating test and failed the other may request in a new application on Form NRC-398 to be excused from re-examination on the portions of the examination or test which the applicant has passed. The Commission may in its discretion grant the request, if it determines that sufficient justification is presented.

          Subpart E—Written Examinations and Operating Tests

            § 55.40
            Implementation.
            (a) The Commission shall use the criteria in NUREG-1021, “Operator Licensing Examination Standards for Power Reactors,” 1
               in effect six months before the examination date to prepare the written examinations required by §§ 55.41 and 55.43 and the operating tests required by § 55.45. The Commission shall also use the criteria in NUREG-1021 to evaluate the written examinations and operating tests prepared by power reactor facility licensees pursuant to paragraph (b) of this section.

                1 Copies of NUREGs may be purchased from the Superintendent of Documents, U.S. Government Publishing Office, P.O. Box 38082, Washington, DC 20402-9328. Copies are also available from the National Technical Information Service, 5301 Shawnee Road, Alexandria, VA 22312. A copy is available for inspection and/or copying in the NRC Public Document Room, One White Flint North, 11555 Rockville Pike (O-1 F21), Rockville, MD.

            (b) Power reactor facility licensees may prepare, proctor, and grade the written examinations required by §§ 55.41 and 55.43 and may prepare the operating tests required by § 55.45, subject to the following conditions:
            (1) Power reactor facility licensees shall prepare the required examinations and tests in accordance with the criteria in NUREG-1021 as described in paragraph (a) of this section;
            (2) Pursuant to § 55.49, power reactor facility licensees shall establish, implement, and maintain procedures to control examination security and integrity;
            (3) An authorized representative of the power reactor facility licensee shall approve the required examinations and tests before they are submitted to the Commission for review and approval; and
            (4) Power reactor facility licensees must receive Commission approval of their proposed written examinations and operating tests.
            (c) In lieu of paragraph (b) of this section and upon written request from a power reactor facility licensee pursuant to § 55.31(a)(3), the Commission shall, for that facility licensee, prepare, proctor, and grade, the written examinations required by §§ 55.41 and 55.43 and the operating tests required by § 55.45. In addition, the Commission may exercise its discretion and reject a power reactor facility licensee's determination to elect paragraph (b) of this section, in which case the Commission shall prepare, proctor, and grade the required written examinations and operating tests for that facility licensee.
            (d) The Commission shall use the criteria in NUREG-1478, “Operator Licensing Examiner Standards for Research and Test Reactors,” for all test and research reactors to prepare, proctor, and grade the written examinations required by §§ 55.41 and 55.43 and the operating tests required by § 55.45 for non-power reactor facility licensees.
            [64 FR 19878, Apr. 23, 1999, as amended at 69 FR 76600, Dec. 22, 2004; 79 FR 66604, Nov. 10, 2014; 80 FR 45844, Aug. 3, 2015; 80 FR 74980, Dec. 1, 2015]

            § 55.41
            Written examination: Operators.
            (a) Content. The written examination for an operator will contain a representative selection of questions on the knowledge, skills, and abilities needed to perform licensed operator duties. The knowledge, skills, and abilities will be identified, in part, from learning objectives derived from a systematic analysis of licensed operator duties performed by each facility licensee and contained in its training program and from information in the Final Safety Analysis Report, system description manuals and operating procedures, facility license and license amendments, Licensee Event Reports, and other materials requested from the facility licensee by the Commission.
            (b) The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.
            (1) Fundamentals of reactor theory, including fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

            (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

            (3) Mechanical components and design features of the reactor primary system.
            (4) Secondary coolant and auxiliary systems that affect the facility.
            (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.
            (6) Design, components, and functions of reactivity control mechanisms and instrumentation.
            (7) Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
            (8) Components, capacity, and functions of emergency systems.
            (9) Shielding, isolation, and containment design features, including access limitations.
            (10) Administrative, normal, abnormal, and emergency operating procedures for the facility.
            (11) Purpose and operation of radiation monitoring systems, including alarms and survey equipment.
            (12) Radiological safety principles and procedures.
            (13) Procedures and equipment available for handling and disposal of radioactive materials and effluents.
            (14) Principles of heat transfer thermodynamics and fluid mechanics.

            § 55.43
            Written examination: Senior operators.
            (a) Content. The written examination for a senior operator will contain a representative selection of questions on the knowledge, skills, and abilities needed to perform licensed senior operator duties. The knowledge, skills, and abilities will be identified, in part, from learning objectives derived from a systematic analysis of licensed senior operator duties performed by each facility licensee and contained in its training program and from information in the Final Safety Analysis Report, system description manuals and operating procedures, facility license and license amendments, Licensee Event Reports, and other materials requested from the facility licensee by the Commission.
            (b) The written examination for a senior operator for a facility will include a representative sample from among the following seven items and the 14 items specified in § 55.41 of this part, to the extent applicable to the facility:
            (1) Conditions and limitations in the facility license.
            (2) Facility operating limitations in the technical specifications and their bases.
            (3) Facility licensee procedures required to obtain authority for design and operating changes in the facility.
            (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
            (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
            (6) Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.
            (7) Fuel handling facilities and procedures.

            § 55.45
            Operating tests.
            (a) Content. The operating tests administered to applicants for operator and senior operator licenses in accordance with paragraph (b)(1) of this section are generally similar in scope. The content will be identified, in part, from learning objectives derived from a systematic analysis of licensed operator or senior operator duties performed by each facility licensee and contained in its training program and from information in the Final Safety Analysis Report, system description manuals and operating procedures, facility license and license amendments, Licensee Event Reports, and other materials requested from the facility licensee by the Commission. The operating test, to the extent applicable, requires the applicant to demonstrate an understanding of and the ability to perform the actions necessary to accomplish a representative sample from among the following 13 items.

            (1) Perform pre-startup procedures for the facility, including operating of those controls associated with plant equipment that could affect reactivity.
            (2) Manipulate the console controls as required to operate the facility between shutdown and designated power levels.
            (3) Identify annunciators and condition-indicating signals and perform appropriate remedial actions where appropriate.
            (4) Identify the instrumentation systems and the significance of facility instrument readings.
            (5) Observe and safely control the operating behavior characteristics of the facility.
            (6) Perform control manipulations required to obtain desired operating results during normal, abnormal, and emergency situations.
            (7) Safely operate the facility's heat removal systems, including primary coolant, emergency coolant, and decay heat removal systems, and identify the relations of the proper operation of these systems to the operation of the facility.
            (8) Safely operate the facility's auxiliary and emergency systems, including operation of those controls associated with plant equipment that could affect reactivity or the release of radioactive materials to the environment.
            (9) Demonstrate or describe the use and function of the facility's radiation monitoring systems, including fixed radiation monitors and alarms, portable survey instruments, and personnel monitoring equipment.
            (10) Demonstrate knowledge of significant radiation hazards, including permissible levels in excess of those authorized, and ability to perform other procedures to reduce excessive levels of radiation and to guard against personnel exposure.
            (11) Demonstrate knowledge of the emergency plan for the facility, including, as appropriate, the operator's or senior operator's responsibility to decide whether the plan should be executed and the duties under the plan assigned.
            (12) Demonstrate the knowledge and ability as appropriate to the assigned position to assume the responsibilities associated with the safe operation of the facility.
            (13) Demonstrate the applicant's ability to function within the control room team as appropriate to the assigned position, in such a way that the facility licensee's procedures are adhered to and that the limitations in its license and amendments are not violated.
            (b) Implementation—Administration. The operating test will be administered in a plant walkthrough and in either—
            (1) A simulation facility that the Commission has approved for use after application has been made by the facility licensee under § 55.46(b);
            (2) A plant-referenced simulator (§ 55.46(c)); or
            (3) The plant, if approved for use in the administration of the operating test by the Commission under § 55.46(b).
            [52 FR 9460, Mar. 25, 1987, as amended at 53 FR 43421, Oct. 27, 1988; 62 FR 59276, Nov. 3, 1997; 66 FR 52667, Oct. 17, 2001]

            § 55.46
            Simulation facilities.
            (a) General. This section addresses the use of a simulation facility for the administration of the operating test and plant-referenced simulators to meet experience requirements for applicants for operator and senior operator licenses.
            (b) Commission-approved simulation facilities and Commission approval of use of the plant in the administration of the operating test. (1) Facility licensees that propose to use a simulation facility, other than a plant-referenced simulator, or the plant in the administration of the operating test under §§ 55.45(b)(1) or 55.45(b)(3), shall request approval from the Commission. This request must include:
            (i) A description of the components of the simulation facility intended to be used, or the way the plant would be used for each part of the operating test, unless previously approved; and
            (ii) A description of the performance tests for the simulation facility as part of the request, and the results of these tests; and
            (iii) A description of the procedures for maintaining examination and test integrity consistent with the requirements of § 55.49.

            (2) The Commission will approve a simulation facility or use of the plant for administration of operating tests if it finds that the simulation facility and its proposed use, or the proposed use of the plant, are suitable for the conduct of operating tests for the facility licensee's reference plant under § 55.45(a).
            (c) Plant-referenced simulators. (1) A plant-referenced simulator used for the administration of the operating test or to meet experience requirements in § 55.31(a)(5) must demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. The plant-referenced simulator must be designed and implemented so that it:
            (i) Is sufficient in scope and fidelity to allow conduct of the evolutions listed in §§ 55.45(a)(1) through (13), and 55.59(c)(3)(i)(A) through (AA), as applicable to the design of the reference plant.
            (ii) Allows for the completion of control manipulations for operator license applicants.
            (2) Facility licensees that propose to use a plant-referenced simulator to meet the control manipulation requirements in § 55.31(a)(5) must ensure that:
            (i) The plant-referenced simulator utilizes models relating to nuclear and thermal-hydraulic characteristics that replicate the most recent core load in the nuclear power reference plant for which a license is being sought; and
            (ii) Simulator fidelity has been demonstrated so that significant control manipulations are completed without procedural exceptions, simulator performance exceptions, or deviation from the approved training scenario sequence.
            (3) A simulation facility consisting solely of a plant-referenced simulator must meet the requirements of paragraph (c)(1) of this section and the criteria in paragraphs (d)(1) and (4) of this section for the Commission to accept the plant-referenced simulator for conducting operating tests as described in § 55.45(a) of this part, requalification training as described in § 55.59(c)(3) of this part, or for performing control manipulations that affect reactivity to establish eligibility for an operator's license as described in § 55.31(a)(5).
            (d) Continued assurance of simulator fidelity. Facility licensees that maintain a simulation facility shall:
            (1) Conduct performance testing throughout the life of the simulation facility in a manner sufficient to ensure that paragraphs (c)(2)(ii), as applicable, and (d)(3) of this section are met. The results of performance tests must be retained for four years after the completion of each performance test or until superseded by updated test results;
            (2) Correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing;
            (3) Make results of any uncorrected performance test failures that may exist at the time of the operating test or requalification program inspection available for NRC review, prior to or concurrent with preparations for each operating test or requalification program inspection; and
            (4) Maintain the provisions for license application, examination, and test integrity consistent with § 55.49.
            [66 FR 52667, Oct. 17, 2001]

            § 55.47
            Waiver of examination and test requirements.
            (a) On application, the Commission may waive any or all of the requirements for a written examination and operating test, if it finds that the applicant—
            (1) Has had extensive actual operating experience at a comparable facility, as determined by the Commission, within two years before the date of application;
            (2) Has discharged his or her responsibilities competently and safely and is capable of continuing to do so; and
            (3) Has learned the operating procedures for and is qualified to operate competently and safely the facility designated in the application.

            (b) The Commission may accept as proof of the applicant's past performance a certification of an authorized representative of the facility licensee or of a holder of an authorization by which the applicant was previously employed. The certification must contain a description of the applicant's operating experience, including an approximate number of hours the applicant operated the controls of the facility, the duties performed, and the extent of the applicant's responsibility.
            (c) The Commission may accept as proof of the applicant's current qualifications a certification of an authorized representative of the facility licensee or of a holder of an authorization where the applicant's services will be utilized.

            § 55.49
            Integrity of examinations and tests.
            Applicants, licensees, and facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. This includes activities related to the preparation and certification of license applications and all activities related to the preparation, administration, and grading of the tests and examinations required by this part.
            [64 FR 19878, Apr. 23, 1999]

          Subpart F—Licenses

            § 55.51
            Issuance of licenses.

              Operator and senior operator licenses. If the Commission determines that an applicant for an operator license or a senior operator license meets the requirements of the Act and its regulations, it will issue a license in the form and containing any conditions and limitations it considers appropriate and necessary.

            § 55.53
            Conditions of licenses.
            Each license contains and is subject to the following conditions whether stated in the license or not:
            (a) Neither the license nor any right under the license may be assigned or otherwise transferred.
            (b) The license is limited to the facility for which it is issued.
            (c) The license is limited to those controls of the facility specified in the license.
            (d) The license is subject to, and the licensee shall observe, all applicable rules, regulations, and orders of the Commission.
            (e) If a licensee has not been actively performing the functions of an operator or senior operator, the licensee may not resume activities authorized by a license issued under this part except as permitted by paragraph (f) of this section. To maintain active status, the licensee shall actively perform the functions of an operator or senior operator on a minimum of seven 8-hour or five 12-hour shifts per calendar quarter. For test and research reactors, the licensee shall actively perform the functions of an operator or senior operator for a minimum of four hours per calendar quarter.
            (f) If paragraph (e) of this section is not met, before resumption of functions authorized by a license issued under this part, an authorized representative of the facility licensee shall certify the following:
            (1) That the qualifications and status of the licensee are current and valid; and
            (2) That the licensee has completed a minimum of 40 hours of shift functions under the direction of an operator or senior operator as appropriate and in the position to which the individual will be assigned. The 40 hours must have included a complete tour of the plant and all required shift turnover procedures. For senior operators limited to fuel handling under paragraph (c) of this section, one shift must have been completed. For test and research reactors, a minimum of six hours must have been completed.
            (g) The licensee shall notify the Commission within 30 days about a conviction for a felony.
            (h) The licensee shall complete a requalification program as described by § 55.59.
            (i) The licensee shall have a biennial medical examination.

            (j) The licensee shall not consume or ingest alcoholic beverages within the protected area of power reactors, or the controlled access area of non-power reactors. The licensee shall not use, possess, or sell any illegal drugs. The licensee shall not perform activities authorized by a license issued under this part while under the influence of alcohol or any prescription, over-the-counter, or illegal substance that could adversely affect his or her ability to safely and competently perform his or her licensed duties. For the purpose of this paragraph, with respect to alcoholic beverages and drugs, the term “under the influence” means the licensee exceeded, as evidenced by a confirmed test result, the lower of the cutoff levels for drugs or alcohol contained in subparts E, F, and G of part 26 of this chapter, or as established by the facility licensee. The term “under the influence” also means the licensee could be mentally or physically impaired as a result of substance use including prescription and over-the-counter drugs, as determined under the provisions, policies, and procedures established by the facility licensee for its fitness-for-duty program, in such a manner as to adversely affect his or her ability to safely and competently perform licensed duties.
            (k) Each licensee at power reactors shall participate in the drug and alcohol testing programs established pursuant to 10 CFR part 26. Each licensee at non-power reactors shall participate in any drug and alcohol testing program that may be established for that non-power facility.
            (l) The licensee shall comply with any other conditions that the Commission may impose to protect health or to minimize danger to life or property.
            [52 FR 9460, Mar. 25, 1987, as amended at 56 FR 32070, July 15, 1991; 74 FR 45545, Sept. 3, 2009; 79 FR 66604, Nov. 10, 2014]

            § 55.55
            Expiration.
            (a) Each operator license and senior operator license expires six years after the date of issuance, upon termination of employment with the facility licensee, or upon determination by the facility licensee that the licensed individual no longer needs to maintain a license.
            (b) If a licensee files an application for renewal or an upgrade of an existing license on Form NRC-398 at least 30 days before the expiration of the existing license, it does not expire until disposition of the application for renewal or for an upgraded license has been finally determined by the Commission. Filing by mail will be deemed to be complete at the time the application is deposited in the mail.
            [52 FR 9460, Mar. 25, 1987, as amended at 79 FR 66605, Nov. 10, 2014]

            § 55.57
            Renewal of licenses.
            (a) The applicant for renewal of a license shall—
            (1) Complete and sign Form NRC-398 and include the number of the license for which renewal is sought.
            (2) File an original of NRC Form 398 with the appropriate Regional Administrator specified in § 55.5(b).
            (3) Provide written evidence of the applicant's experience under the existing license and the approximate number of hours that the licensee has operated the facility.
            (4) Provide a statement by an authorized representative of the facility licensee that during the effective term of the current license the applicant has satisfactorily completed the requalification program for the facility for which operator or senior operator license renewal is sought.
            (5) Provide evidence that the applicant has discharged the license responsibilities competently and safely. The Commission may accept as evidence of the applicant's having met this requirement a certificate of an authorized representative of the facility licensee or holder of an authorization by which the licensee has been employed.
            (6) Provide certification by the facility licensee of medical condition and general health on Form NRC-396, to comply with §§ 55.21, 55.23 and 55.27.
            (b) The license will be renewed if the Commission finds that—
            (1) The medical condition and the general health of the licensee continue to be such as not to cause operational errors that endanger public health and safety. The Commission will base this finding upon the certification by the facility licensee as described in § 55.23.
            (2) The licensee—

            (i) Is capable of continuing to competently and safely assume licensed duties;

            (ii) Has successfully completed a requalification program that has been approved by the Commission as required by § 55.59; and
            (iii) Has passed the requalification examinations and annual operating tests as required by § 55.59.
            (3) There is a continued need for a licensee to operate or for a senior operator to direct operators at the facility designated in the application.
            (4) The past performance of the licensee has been satisfactory to the Commission. In making its finding, the Commission will include in its evaluation information such as notices of violations or letters of reprimand in the licensee's docket.
            [52 FR 9460, Mar. 25, 1987, as amended at 59 FR 5938, Feb. 9, 1994; 68 FR 58813, Oct. 10, 2003]

            § 55.59
            Requalification.
            (a) Requalification requirements. Each licensee shall—
            (1) Successfully complete a requalification program developed by the facility licensee that has been approved by the Commission. This program shall be conducted for a continuous period not to exceed 24 months in duration.
            (2) Pass a comprehensive requalification written examination and an annual operating test.
            (i) The written examination will sample the items specified in §§ 55.41 and 55.43 of this part, to the extent applicable to the facility, the licensee, and any limitation of the license under § 55.53(c) of this part.
            (ii) The operating test will require the operator or senior operator to demonstrate an understanding of and the ability to perform the actions necessary to accomplish a comprehensive sample of items specified in § 55.45(a) (2) through (13) inclusive to the extent applicable to the facility.
            (iii) In lieu of the Commission accepting a certification by the facility licensee that the licensee has passed written examinations and operating tests administered by the facility licensee within its Commission-approved program developed by using a systems approach to training under paragraph (c) of this section, the Commission may administer a comprehensive requalification written examination and an annual operating test.
            (b) Additional training. If the requirements of paragraphs (a) (1) and (2) of this section are not met, the Commission may require the licensee to complete additional training and to submit evidence to the Commission of successful completion of this training before returning to licensed duties.
            (c) Requalification program requirements. A facility licensee shall have a requalification program reviewed and approved by the Commission and shall, upon request consistent with the Commission's inspection program needs, submit to the Commission a copy of its comprehensive requalification written examinations or annual operating tests. The requalification program must meet the requirements of paragraphs (c) (1) through (7) of this section. In lieu of paragraphs (c) (2), (3), and (4) of this section, the Commission may approve a program developed by using a systems approach to training.
            (1) Schedule. The requalification program must be conducted for a continuous period not to exceed two years, and upon conclusion must be promptly followed, pursuant to a continuous schedule, by successive requalification programs.
            (2) Lectures. The requalification program must include preplanned lectures on a regular and continuing basis throughout the license period in those areas where operator and senior operator written examinations and facility operating experience indicate that emphasis in scope and depth of coverage is needed in the following subjects:
            (i) Theory and principles of operation.
            (ii) General and specific plant operating characteristics.
            (iii) Plant instrumentation and control systems.
            (iv) Plant protection systems.
            (v) Engineered safety systems.
            (vi) Normal, abnormal, and emergency operating procedures.
            (vii) Radiation control and safety.
            (viii) Technical specifications.
            (ix) Applicable portions of title 10, chapter I, Code of Federal Regulations.
            (3) On-the-job training. The requalification program must include on-the-job training so that—

            (i) Each licensed operator of a utilization facility manipulates the plant controls and each licensed senior operator either manipulates the controls or directs the activities of individuals during plant control manipulations during the term of the licensed operator's or senior operator's license. For reactor operators and senior operators, these manipulations must consist of the following control manipulations and plant evolutions if they are applicable to the plant design. Items described in paragraphs (c)(3)(i) (A) through (L) of this section must be performed annually; all other items must be performed on a two-year cycle. However, the requalification programs must contain a commitment that each individual shall perform or participate in a combination of reactivity control manipulations based on the availability of plant equipment and systems. Those control manipulations which are not performed at the plant may be performed on a simulator. The use of the Technical Specifications should be maximized during the simulator control manipulations. Senior operator licensees are credited with these activities if they direct control manipulations as they are performed.
            (A) Plant or reactor startups to include a range that reactivity feedback from nuclear heat addition is noticeable and heatup rate is established.
            (B) Plant shutdown.
            (C) Manual control of steam generators or feedwater or both during startup and shutdown.
            (D) Boration or dilution during power operation.
            (E) Significant (≥10 percent) power changes in manual rod control or recirculation flow.
            (F) Reactor power change of 10 percent or greater where load change is performed with load limit control or where flux, temperature, or speed control is on manual (for HTGR).
            (G) Loss of coolant, including—
            (1) Significant PWR steam generator leaks
            (2) Inside and outside primary containment
            (3) Large and small, including leak-rate determination
            (4) Saturated reactor coolant response (PWR).
            (H) Loss of instrument air (if simulated plant specific).
            (I) Loss of electrical power (or degraded power sources).
            (J) Loss of core coolant flow/natural circulation.
            (K) Loss of feedwater (normal and emergency).
            (L) Loss of service water, if required for safety.
            (M) Loss of shutdown cooling.
            (N) Loss of component cooling system or cooling to an individual component.
            (O) Loss of normal feedwater or normal feedwater system failure.
            (P) Loss of condenser vacuum.
            (Q) Loss of protective system channel.
            (R) Mispositioned control rod or rods (or rod drops).
            (S) Inability to drive control rods.
            (T) Conditions requiring use of emergency boration or standby liquid control system.
            (U) Fuel cladding failure or high activity in reactor coolant or offgas.
            (V) Turbine or generator trip.
            (W) Malfunction of an automatic control system that affects reactivity.
            (X) Malfunction of reactor coolant pressure/volume control system.
            (Y) Reactor trip.
            (Z) Main steam line break (inside or outside containment).
            (AA) A nuclear instrumentation failure.
            (ii) Each licensed operator and senior operator has demonstrated satisfactory understanding of the operation of the apparatus and mechanisms associated with the control manipulations in paragraph (c)(3)(i) of this section, and knows the operating procedures in each area for which the operator or senior operator is licensed.
            (iii) Each licensed operator and senior operator is cognizant of facility design changes, procedure changes, and facility license changes.
            (iv) Each licensed operator and senior operator reviews the contents of all abnormal and emergency procedures on a regularly scheduled basis.

            (v) A simulator may be used in meeting the requirements of paragraphs (c) (3)(i) and (3)(ii) of this section, if it reproduces the general operating characteristics of the facility involved and the arrangement of the instrumentation and controls of the simulator is similar to that of the facility involved. If the simulator or simulation device is used to administer operating tests for a facility, as provided in § 55.45(b)(1), the device approved to meet the requirements of § 55.45(b)(1) must be used for credit to be given for meeting the requirements of paragraphs (c)(3)(i) (G through AA) of this section.
            (4) Evaluation. The requalification program must include—
            (i) Comprehensive requalification written examinations and annual operating tests which determine areas in which retraining is needed to upgrade licensed operator and senior operator knowledge.
            (ii) Written examinations which determine licensed operators' and senior operators' knowledge of subjects covered in the requalification program and provide a basis for evaluating their knowledge of abnormal and emergency procedures.
            (iii) Systematic observation and evaluation of the performance and competency of licensed operators and senior operators by supervisors and/or training staff members, including evaluation of actions taken or to be taken during actual or simulated abnormal and emergency procedures.
            (iv) Simulation of emergency or abnormal conditions that may be accomplished by using the control panel of the facility involved or by using a simulator. When the control panel of the facility is used for simulation, the actions taken or to be taken for the emergency or abnormal condition shall be discussed; actual manipulation of the plant controls is not required. If a simulator is used in meeting the requirements of paragraph (c)(4)(iii) of this section, it must accurately reproduce the operating characteristics of the facility involved and the arrangement of the instrumentation and controls of the simulator must closely parallel that of the facility involved. After the provisions of § 55.46 have been implemented at a facility, the Commission approved or plant-referenced simulator must be used to comply with this paragraph.
            (v) Provisions for each licensed operator and senior operator to participate in an accelerated requalification program where performance evaluations conducted pursuant to paragraphs (c)(4) (i) through (iv) of this section clearly indicated the need.
            (5) Records. The requalification program documentation must include the following:
            (i) The facility licensee shall maintain records documenting the participation of each licensed operator and senior operator in the requalification program. The records must contain copies of written examinations administered, the answers given by the licensee, and the results of evaluations and documentation of operating tests and of any additional training administered in areas in which an operator or senior operator has exhibited deficiencies. The facility licensee shall retain these records until the operator's or senior operator's license is renewed.
            (ii) Each record required by this part must be legible throughout the retention period specified by each Commission regulation. The record may be the original or a reproduced copy or a microform provided that the copy or microform is authenticated by authorized personnel and that the microform is capable of producing a clear copy throughout the required retention period.
            (iii) If there is a conflict between the Commission's regulations in this part, and any license condition, or other written Commission approval or authorization pertaining to the retention period for the same type of record, the retention period specified for these records by the regulations in this part apply unless the Commission, pursuant to § 55.11, grants a specific exemption from this record retention requirement.
            (6) Alternative training programs. The requirements of this section may be met by requalification programs conducted by persons other than the facility licensee if the requalification programs are similar to the program described in paragraphs (c) (1) through (5) of this section and the alternative program has been approved by the Commission.

            (7) Applicability to research and test reactor facilities. To accommodate specialized modes of operation and differences in control, equipment, and operator skills and knowledge, the requalification program for each licensed operator and senior operator of a research reactor or test reactor facility must conform generally but need not be identical to the requalification program outlined in paragraphs (c) (1) through (6) of this section. Significant deviations from the requirements of paragraphs (c) (1) through (6) of this section will be permitted only if supported by written justification and approved by the Commission.
            [52 FR 9460, Mar. 25, 1987, as amended at 59 FR 5938, Feb. 9, 1994; 66 FR 52668, Oct. 17, 2001, 81 FR 86909, Dec. 2, 2016]

          Subpart G—Modification and Revocation of Licenses

            § 55.61
            Modification and revocation of licenses.
            (a) The terms and conditions of all licenses are subject to amendment, revision, or modification by reason of rules, regulations, or orders issued in accordance with the Act or any amendments thereto.
            (b) Any license may be revoked, suspended, or modified, in whole or in part:
            (1) For any material false statement in the application or in any statement of fact required under section 182 of the Act,
            (2) Because of conditions revealed by the application or statement of fact or any report, record, inspection or other means that would warrant the Commission to refuse to grant a license on an original application,
            (3) For willful violation of, or failure to observe any of the terms and conditions of the Act, or the license, or of any rule, regulation, or order of the Commission, or
            (4) For any conduct determined by the Commission to be a hazard to safe operation of the facility.
            (5) For the sale, use or possession of illegal drugs, or refusal to participate in the facility drug and alcohol testing program, or a confirmed positive test for drugs, drug metabolites, or alcohol in violation of the conditions and cutoff levels established by § 55.53(j) or the consumption of alcoholic beverages within the protected area of power reactors or the controlled access area of non-power reactors, or a determination of unfitness for scheduled work as a result of the consumption of alcoholic beverages.
            [52 FR 9460, Mar. 25, 1987, as amended at 56 FR 32070, July 15, 1991]

          Subpart H—Enforcement

            § 55.71
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of—
            (1) The Atomic Energy Act of 1954, as amended;
            (2) Title II of the Energy Reorganization Act of 1974, as amended; or
            (3) A regulation or order issued pursuant to those Acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:
            (1) For violations of—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section;
            (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.
            [57 FR 55076, Nov. 24, 1992]

            § 55.73
            Criminal penalties.

            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy of violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 55 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in part 55 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 55.1, 55.2, 55.4, 55.5, 55.6, 55.7, 55.8, 55.11. 55.13, 55.31, 55.33, 55.35, 55.41, 55.43, 55.47, 55.51, 55.55, 55.57, 55.61, 55.71, and 55.73.
            [57 FR 55076, Nov. 24, 1992]

        Pt. 60
        PART 60—DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES

            Subpart A—General Provisions
            Sec.
            60.1
            Purpose and scope.
            60.2
            Definitions.
            60.3
            License required.
            60.4
            Communications and records.
            60.5
            Interpretations.
            60.6
            Exemptions.
            60.7
            License not required for certain preliminary activities.
            60.8
            Information collection requirements: Approval.
            60.9
            Employee protection.
            60.10
            Completeness and accuracy of information.
            60.11
            Deliberate misconduct.

            Subpart B—Licenses

              Preapplication Review
              60.15
              Site characterization.
              60.16
              Site characterization plan required.
              60.17
              Contents of site characterization plan.
              60.18
              Review of site characterization activities.

              License Applications
              60.21
              Content of application.
              60.22
              Filing and distribution of application.
              60.23
              Elimination of repetition.
              60.24
              Updating of application and environmental impact statement.

              Construction Authorization
              60.31
              Construction authorization.
              60.32
              Conditions of construction authorization.
              60.33
              Amendment of construction authorization.

              License Issuance and Amendment
              60.41
              Standards for issuance of a license.
              60.42
              Conditions of license.
              60.43
              License specification.
              60.44
              Changes, tests, and experiments.
              60.45
              Amendment of license.
              60.46
              Particular activities requiring license amendment.

              US/IAEA Safeguards Agreement
              60.47
              Facility information and verification.

              Permanent Closure
              60.51
              License amendment for permanent closure.
              60.52
              Termination of license.

            Subpart C—Participation by State Governments and Affected Indian Tribes
            60.61
            Provision of information.
            60.62
            Site review.
            60.63
            Participation in license reviews.
            60.64
            Notice to States.
            60.65
            Representation.

            Subpart D—Records, Reports, Tests, and Inspections
            60.71
            Records and reports.
            60.72
            Construction records.
            60.73
            Reports of deficiencies.
            60.74
            Tests.
            60.75
            Inspections.
            60.78
            Material control and accounting records and reports.

            Subpart E—Technical Criteria
            60.101
            Purpose and nature of findings.
            60.102
            Concepts.

              Performance Objectives
              60.111
              Performance of the geologic repository operations area through permanent closure.
              60.112
              Overall system performance objective for the geologic repository after permanent closure.
              60.113
              Performance of particular barriers after permanent closure.

              Land Ownership and Control
              60.121
              Requirements for ownership and control interests in land.

              Siting Criteria
              60.122
              Siting criteria.

              Design Criteria for the Geologic Repository Operations Area
              60.130
              General considerations.
              60.131
              General design criteria for the geologic repository operations area.
              60.132
              Additional design criteria for surface facilities in the geologic repository operations area.
              60.133
              Additional design criteria for the underground facility.
              60.134
              Design of seals for shafts and boreholes.

              Design Criteria for the Waste Package
              60.135
              Criteria for the waste package and its components.

              Preclosure Controlled Area
              60.136
              Preclosure controlled area.

              Performance Confirmation Requirements
              60.137
              General requirements for performance confirmation.

            Subpart F—Performance Confirmation Program
            60.140
            General requirements.
            60.141
            Confirmation of geotechnical and design parameters.
            60.142
            Design testing.
            60.143
            Monitoring and testing waste packages.

            Subpart G—Quality Assurance
            60.150
            Scope.
            60.151
            Applicability.
            60.152
            Implementation.

            Subpart H—Training and Certification of Personnel
            60.160
            General requirements.
            60.161
            Training and certification program.
            60.162
            Physical requirements.

            Subpart I—Emergency Planning Criteria [Reserved]

            Subpart J—Violations
            60.181
            Violations.
            60.183
            Criminal penalties.

          Authority:
          Atomic Energy Act of 1954, secs. 51, 53, 62, 63, 65, 81, 161, 182, 183, 223, 234 (42 U.S.C. 2071, 2073, 2092, 2093, 2095, 2111, 2201, 2232, 2233, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 42 U.S.C. 2021a; National Environmental Policy Act of 1969 (42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 114, 117, 121 (42 U.S.C. 10134, 10137, 10141), 44 U.S.C. 3504 note.

          Source:
          46 FR 13980, Feb. 25, 1981, unless otherwise noted.

          Subpart A—General Provisions

            § 60.1
            Purpose and scope.
            This part prescribes rules governing the licensing (including issuance of a construction authorization) of the U.S. Department of Energy to receive and possess source, special nuclear, and byproduct material at a geologic repository operations area sited, constructed, or operated in accordance with the Nuclear Waste Policy Act of 1982, as amended. This part does not apply to any activity licensed under another part of this chapter. This part does not apply to the licensing of the U.S. Department of Energy to receive and possess source, special nuclear, and byproduct material at a geologic repository operations area sited, constructed, or operated at Yucca Mountain, Nevada, in accordance with the Nuclear Waste Policy Act of 1992, as amended, and the Energy Policy Act of 1992, subject to part 63 of this chapter. This part also gives notice to all persons who knowingly provide to any licensee, applicant, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licensee's or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 60.11.
            [69 FR 2279, Jan. 14, 2004]

            § 60.2
            Definitions.
            As used in this part:

              Accessible environment means:
            (1) The atmosphere;
            (2) The land surface;
            (3) Surface water;
            (4) Oceans; and
            (5) The portion of the lithosphere that is outside the postclosure controlled area.

              Affected Indian Tribe means any Indian Tribe (1) within whose reservation boundaries a repository for high-level radioactive waste or spent fuel is proposed to be located; or (2) whose Federally defined possessory or usage rights to other lands outside of the reservation's boundaries arising out of Congressionally ratified treaties or other Federal law may be substantially and adversely affected by the locating of such a facility; Provided, That the Secretary of the Interior finds, upon the petition of the appropriate governmental officials of the Tribe, that such effects are both substantial and adverse to the Tribe.

              Anticipated processes and events means those natural processes and events that are reasonably likely to occur during the period the intended performance objective must be achieved. To the extent reasonable in the light of the geologic record, it shall be assumed that those processes operating in the geologic setting during the Quaternary Period continue to operate but with the perturbations caused by the presence of emplaced radioactive waste superimposed thereon.

              Barrier means any material or structure that prevents or substantially delays movement of water or radionuclides.

              Candidate area means a geologic and hydrologic system within which a geologic repository may be located.

              Commencement of construction means clearing of land, surface or subsurface excavation, or other substantial action that would adversely affect the environment of a site, but does not include changes desirable for the temporary use of the land for public recreational uses, site characterization activities, other preconstruction monitoring and investigation necessary to establish background information related to the suitability of a site or to the protection of environmental values, or procurement or manufacture of components of the geologic repository operations area.

              Commission means the Nuclear Regulatory Commission or its duly authorized representatives.

              Containment means the confinement of radioactive waste within a designated boundary.

              Controlled area means a surface location, to be marked by suitable monuments, extending horizontally no more than 10 kilometers in any direction from the outer boundary of the underground facility, and the underlying subsurface, which area has been committed to use as a geologic repository and from which incompatible activities would be restricted following permanent closure.

              Design bases means that information that identifies the specific functions to be performed by a structure, system, or component of a facility and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be restraints derived from generally accepted “state-of-the-art” practices for achieving functional goals or requirements derived from analysis (based on calculation or experiments) of the effects of a postulated event under which a structure, system, or component must meet its functional goals. The values for controlling parameters for external events include:
            (1) Estimates of severe natural events to be used for deriving design bases that will be based on consideration of historical data on the associated parameters, physical data, or analysis of upper limits of the physical processes involved; and
            (2) Estimates of severe external man-induced events, to be used for deriving design bases, that will be based on analysis of human activity in the region, taking into account the site characteristics and the risks associated with the event.

              Design basis events means:
            (1)(i) Those natural and human-induced events that are reasonably likely to occur regularly, moderately frequently, or one or more times before permanent closure of the geologic repository operations area; and
            (ii) Other natural and man-induced events that are considered unlikely, but sufficiently credible to warrant consideration, taking into account the potential for significant radiological impacts on public health and safety.
            (2) The events described in paragraph (1)(i) of this definition are referred to as “Category 1” design basis events. The events described in paragraph (1)(ii) of this definition are referred to as “Category 2” design basis events.

              Director means the Director of the Nuclear Regulatory Commission's Office of Nuclear Material Safety and Safeguards.

              Disposal means the isolation of radioactive wastes from the accessible environment.

              Disturbed zone means that portion of the postclosure controlled area, the physical or chemical properties of which have changed as a result of underground facility construction or as a result of heat generated by the emplaced radioactive wastes, such that the resultant change of properties may have a significant effect on the performance of the geologic repository.

              DOE means the U.S. Department of Energy or its duly authorized representatives.

              Engineered barrier system means the waste packages and the underground facility.

              Geologic repository means a system which is intended to be used for, or may be used for, the disposal of radioactive wastes in excavated geologic media. A geologic repository includes: (1) The geologic repository operations area, and (2) the portion of the geologic setting that provides isolation of the radioactive waste.

              Geologic repository operations area means a high-level radioactive waste facility that is part of a geologic repository, including both surface and subsurface areas, where waste handling activities are conducted.

              Geologic setting means the geologic, hydrologic, and geochemical systems of the region in which a geologic repository operations area is or may be located.

              Groundwater means all water which occurs below the land surface.

              High-level radioactive waste or HLW means: (1) Irradiated reactor fuel, (2) liquid wastes resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated wastes from subsequent extraction cycles, or equivalent, in a facility for reprocessing irradiated reactor fuel, and (3) solids into which such liquid wastes have been converted.

              HLW facility means a facility subject to the licensing and related regulatory authority of the Commission pursuant to Sections 202(3) and 202(4) of the Energy Reorganization Act of 1974 (88 Stat. 1244). 1

                1 These are DOE “facilities used primarily for the receipt and storage of high-level radioactive wastes resulting from activities licensed under such Act [the Atomic Energy Act]” and “Retrievable Surface Storage Facilities and other facilities authorized for the express purpose of subsequent long-term storage of high-level radioactive wastes generated by [DOE], which are not used for, or are part of, research and development activities.”

              Host rock means the geologic medium in which the waste is emplaced.

              Important to safety, with reference to structures, systems, and components, means those engineered features of the repository whose function is:
            (1) To provide reasonable assurance that high-level waste can be received, handled, packaged, stored, emplaced, and retrieved without exceeding the requirements of § 60.111(a) for Category 1 design basis events; or
            (2) To prevent or mitigate Category 2 design basis events that could result in doses equal to or greater than the values specified in § 60.136 to any individual located on or beyond any point on the boundary of the preclosure controlled area.

              Isolation means inhibiting the transport of radioactive material so that amounts and concentrations of this material entering the accessible environment will be kept within prescribed limits.

              NRC Public Document Room means the facility at One White Flint North, 11555 Rockville Pike, Room 0-1F23, Rockville, Maryland 20852, where certain public records of the NRC that were made available for public inspection in paper or microfiche prior to the implementation of the NRC Agency wide Documents Access and Management System, commonly referred to as ADAMS, will remain available for public inspection. It is also the place where computer terminals are available to access the NRC Library components of ADAMS on the NRC Website, http://www.nrc.gov, where copies can be made or ordered as set forth in § 9.35 of this chapter. The facility is staffed with reference librarians to assist the public in identifying and locating documents and in using the NRC Web site and ADAMS. The NRC Public Document Room is open from 7:30 am to 4:15 pm, Monday through Friday, except on Federal holidays, Reference service and access to documents may also be requested by telephone (1-800-397-4209) between 8:30 am and 4:15 pm, or by e-mail (PDR.Resource@nrc.gov), fax (301-415-3548), or letter (NRC Public Document Room, One White Flint North, 11555 Rockville Pike, Room 0-1F23, Rockville, Maryland 20852).
            NRC Web site, http://www.nrc.gov is the Internet uniform resource locator name for the Internet address of the Web site where NRC will ordinarily make available its public records for inspection.

              Permanent closure means final backfilling of the underground facility and the sealing of shafts and boreholes.

              Performance confirmation means the program of tests, experiments, and analyses which is conducted to evaluate the accuracy and adequacy of the information used to determine with reasonable assurance that the performance objectives for the period after permanent closure will be met.

              Postclosure controlled area means a surface location, to be marked by suitable monuments, extending horizontally no more than 10 kilometers in any direction from the outer boundary of the underground facility, and the underlying subsurface, which area has been committed to use as a geologic repository and from which incompatible activities would be restricted following permanent closure.

              Preclosure controlled area means that surface area surrounding the geologic repository operations area for which the licensee exercises authority over its use, in accordance with the provisions of this part, until permanent closure has been completed.

              Radioactive waste or waste means HLW and other radioactive materials other than HLW that are received for emplacement in a geologic repository.

              Restricted area means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set aside as a restricted area.

              Retrieval means the act of intentionally removing radioactive waste from the underground location at which the waste had been previously emplaced for disposal.

              Saturated zone means that part of the earth's crust beneath the regional water table in which all voids, large and small, are ideally filled with water under pressure greater than atmospheric.

              Site means the location of the preclosure controlled area, or of the postclosure controlled area, or both.

              Site characterization means the program of exploration and research, both in the laboratory and in the field, undertaken to establish the geologic conditions and the ranges of those parameters of a particular site relevant to the procedures under this part. Site characterization includes borings, surface excavations, excavation of exploratory shafts, limited subsurface lateral excavations and borings, and in situ testing at depth needed to determine the suitability of the site for a geologic repository, but does not include preliminary borings and geophysical testing needed to decide whether site characterization should be undertaken.

              Unanticipated processes and events means those processes and events affecting the geologic setting that are judged not to be reasonably likely to occur during the period the intended performance objective must be achieved, but which are nevertheless sufficiently credible to warrant consideration. Unanticipated processes and events may be either natural processes or events or processes and events initiated by human activities other than those activities licensed under this part. Processes and events initiated by human activities may only be found to be sufficiently credible to warrant consideration if it is assumed that: (1) The monuments provided for by this part are sufficiently permanent to serve their intended purpose; (2) the value to future generations of potential resources within the site can be assessed adequately under the applicable provisions of this part; (3) an understanding of the nature of radioactivity, and an appreciation of its hazards, have been retained in some functioning institutions; (4) institutions are able to assess risk and to take remedial action at a level of social organization and technological competence equivalent to, or superior to, that which was applied in initiating the processes or events concerned; and (5) relevant records are preserved, and remain accessible, for several hundred years after permanent closure.

              Underground facility means the underground structure, including openings and backfill materials, but excluding shafts, boreholes, and their seals.

              Unrestricted area means an area, access to which is neither limited nor controlled by the licensee.

              Unsaturated zone means the zone between the land surface and the regional water table. Generally, fluid pressure in this zone is less than atmospheric pressure, and some of the voids may contain air or other gases at atmospheric pressure. Beneath flooded areas or in perched water bodies the fluid pressure locally may be greater than atmospheric.

              Waste form means the radioactive waste materials and any encapsulating or stabilizing matrix.

              Waste package means the waste form and any containers, shielding, packing and other absorbent materials immediately surrounding an individual waste container.

              Water table means that surface in a groundwater body at which the water pressure is atmospheric.
            [48 FR 28217, June 21, 1983, as amended at 50 FR 29647, July 22, 1985; 51 FR 27162, July 30, 1986; 53 FR 43421, Oct. 27, 1988; 61 FR 64267, Dec. 4, 1996; 64 FR 48953, Sept. 9, 1999; 69 FR 76601, Dec. 22, 2004; 76 FR 72086, Nov. 22, 2011]

            § 60.3
            License required.
            (a) DOE shall not receive or possess source, special nuclear, or byproduct material at a geologic repository operations area except as authorized by a license issued by the Commission pursuant to this part.
            (b) DOE shall not commence construction of a geologic repository operations area unless it has filed an application with the Commission and has obtained construction authorization as provided in this part. Failure to comply with this requirement shall be grounds for denial of a license.

            § 60.4
            Communications and records.

            (a) Except where otherwise specified, all communications and reports concerning the regulations in this part and applications filed under them should be sent by mail addressed: ATTN: Document Control Desk: Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.
            (b) Each record required by this part must be legible throughout the retention period specified by each Commission regulation. The record may be the original or a reproduced copy or a microform provided that the copy or microform is authenticated by authorized personnel and that the microform is capable of producing a clear copy throughout the required retention period. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, specifications, must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.
            [53 FR 19251, May 27, 1988, as amended at 53 FR 43421, Oct. 27, 1988; 68 FR 58813, Oct. 10, 2003; 74 FR 62682, Dec. 1, 2009; 80 FR 74980, Dec. 1, 2015]

            § 60.5
            Interpretations.
            Except as specifically authorized by the Commission, in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be considered binding upon the Commission.

            § 60.6
            Exemptions.
            The Commission may, upon application by DOE, any interested person, or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law, will not endanger life or property or the common defense and security, and are otherwise in the public interest.

            § 60.7
            License not required for certain preliminary activities.
            The requirement for a license set forth in § 60.3(a) of this part is not applicable to the extent that DOE receives and possesses source, special nuclear, and byproduct material at a geologic repository:
            (a) For purposes of site characterization; or
            (b) For use, during site characterization or construction, as components of radiographic, radiation monitoring, or similar equipment or instrumentation.

            § 60.8
            Information collection requirements: Approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0127.
            (b) The approved information collection requirements contained in this part appear in §§ 60.47, 60.62, 60.63, 60.65.
            (c) In § 60.47, Form N-71 and associated forms are approved under control number 3150-0056, and DOC/NRC Forms AP-1, AP-A, and associated forms are approved under control number 0694-0135.
            [61 FR 64268, Dec. 4, 1996, as amended at 62 FR 52188, Oct. 6, 1997; 73 FR 78605, Dec. 23, 2008]

            § 60.9
            Employee protection.
            (a) Discrimination by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act or the Energy Reorganization Act.
            (1) The protected activities include but are not limited to:
            (i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) introductory text of this section or possible violations of requirements imposed under either of those statutes;
            (ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) introductory text or under these requirements if the employee has identified the alleged illegality to the employer;
            (iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;
            (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) introductory text.
            (v) Assisting or participating in, or is about to assist or participate in, these activities.
            (2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.

            (3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.

            (b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.
            (c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant may be grounds for—
            (1) Denial, revocation, or suspension of the license.
            (2) Imposition of a civil penalty on the licensee, applicant, or a contractor or subcontractor of the licensee or applicant.
            (3) Other enforcement action.
            (d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.
            (e)(1) Each licensee and each applicant for a license shall prominently post the revision of NRC Form 3, “Notice to Employees,” referenced in 10 CFR 19.11(c). This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, and for 30 days following license termination.

            (2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, via email to Forms.Resource@nrc.gov, or by visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.

            (f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.
            [58 FR 52411, Oct. 8, 1993, as amended at 60 FR 24552, May 9, 1995; 61 FR 6765, Feb. 22, 1996; 68 FR 58813, Oct. 10, 2003; 72 FR 63974, Nov. 14, 2007; 73 FR 30459, May 28, 2008; 79 FR 66605, Nov. 10, 2014]

            § 60.10
            Completeness and accuracy of information.
            (a) Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.

            (b) Each applicant or licensee shall notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant or licensee violates this paragraph only if the applicant or licensee fails to notify the Commission of information that the applicant or licensee has identified as having a significant implication for public health and safety or common defense and security. Notification shall be provided to the Administrator of the appropriate Regional Office within two working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.
            [52 FR 49372, Dec. 31, 1987]

            § 60.11
            Deliberate misconduct.
            (a) Any licensee, applicant for a license, employee of a licensee or applicant; or any contractor (including a supplier or consultant), subcontractor, employee of a contractor or subcontractor of any licensee or applicant for a license who knowingly provides to any licensee, applicant, contractor, or subcontractor, any components, equipment, materials, or other goods or services that relate to a licensee's or applicant's activities in this part, may not:
            (1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission; or
            (2) Deliberately submit to the NRC, a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.
            (b) A person who violates paragraph (a)(1) or (a)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.
            (c) For the purposes of paragraph (a)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:
            (1) Would cause a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license issued by the Commission; or
            (2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor.
            [63 FR 1898, Jan. 13, 1998]

          Subpart B—Licenses

            Preapplication Review

              § 60.15
              Site characterization.
              (a) Prior to submittal of an application for a license to be issued under this part DOE shall conduct a program of site characterization with respect to the site to be described in such application.
              (b) Unless the Commission determines with respect to the site described in the application that it is not necessary, site characterization shall include a program of in situ exploration and testing at the depths that wastes would be emplaced.
              (c) The program of site characterization shall be conducted in accordance with the following:
              (1) Investigations to obtain the required information shall be conducted in such a manner as to limit adverse effects on the long-term performance of the geologic repository to the extent practical.
              (2) The number of exploratory boreholes and shafts shall be limited to the extent practical consistent with obtaining the information needed for site characterization.
              (3) To the extent practical, exploratory boreholes and shafts in the geologic repository operations area shall be located where shafts are planned for underground facility construction and operation or where large unexcavated pillars are planned.
              (4) Subsurface exploratory drilling, excavation, and in situ testing before and during construction shall be planned and coordinated with geologic repository operations area design and construction.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28219, June 21, 1983. Redesignated and amended at 51 FR 27162, July 30, 1986; 54 FR 27871, July 3, 1989]

              § 60.16
              Site characterization plan required.

              Before proceeding to sink shafts at any area which has been approved by the President for site characterization, DOE shall submit to the Director, for review and comment, a site characterization plan for such area. DOE shall defer the sinking of such shafts until such time as there has been an opportunity for Commission comments thereon to have been solicited and considered by DOE.
              [51 FR 27162, July 30, 1986]

              § 60.17
              Contents of site characterization plan.
              The site characterization plan shall contain—
              (a) A general plan for site characterization activities to be conducted at the area to be characterized, which general plan shall include:
              (1) A description of such area, including information on quality assurance programs that have been applied to the collection, recording, and retention of information used in preparing such description.
              (2) A description of such site characterization activities, including the following—
              (i) The extent of planned excavations;
              (ii) Plans for any onsite testing with radioactive material, including radioactive tracers, or nonradioactive material;
              (iii) Plans for any investigation activities that may affect the capability of such area to isolate high-level radioactive waste;
              (iv) Plans to control any adverse impacts from such site characterization activities that are important to safety or that are important to waste isolation; and
              (v) Plans to apply quality assurance to data collection, recording, and retention.
              (3) Plans for the decontamination and decommissioning of such area, and for the mitigation of any significant adverse environmental impacts caused by site characterization activities, if such area is determined unsuitable for application for a construction authorization for a geologic repository operations area;
              (4) Criteria, developed pursuant to section 112(a) of the Nuclear Waste Policy Act of 1982, to be used to determine the suitability of such area for the location of a geologic repository; and
              (5) Any other information which the Commission, by rule or order, requires.
              (b) A description of the possible waste form or waste package for the high-level radioactive waste to be emplaced in such geologic repository, a description (to the extent practicable) of the relationship between such waste form or waste package and the host rock at such area, and a description of the activities being conducted by DOE with respect to such possible waste form or waste package or their relationship; and
              (c) A conceptual design for the geologic repository operations area that takes into account likely site-specific requirements.
              [51 FR 27163, July 30, 1986]

              § 60.18
              Review of site characterization activities. 2

                  2 In addition to the review of site characterization activities specified in this section, the Commission contemplates an ongoing review of other information on site investigation and site characterization, in order to allow early identification of potential licensing issues for timely resolution. This activity will include, for example, a review of the environmental assessments prepared by DOE at the time of site nomination, and review of issues related to long lead time exploratory shaft planning and procurement actions by DOE prior to issuance of site characterization plans.

              (a) The Director shall cause to be published in the Federal Register a notice that a site characterization plan has been received from DOE and that a staff review of such plan has begun. The notice shall identify the area to be characterized and the NRC staff members to be consulted for further information.

              (b) The Director shall make a copy of the site characterization plan available at the Public Document Room. The Director shall also transmit copies of the published notice of receipt to the Governor and legislature of the State in which the area to be characterized is located and to the governing body of any affected Indian Tribe. The Director shall provide an opportunity, with respect to any area to be characterized, for the State in which such area is located and for affected Indian Tribes to present their views on the site characterization plan and their suggestions with respect to comments thereon which may be made by NRC. In addition, the Director shall make NRC staff available to consult with States and affected Indian Tribes as provided in Subpart C of this part.
              (c) The Director shall review the site characterization plan and prepare a site characterization analysis with respect to such plan. In the preparation of such site characterization analysis, the Director may invite and consider the views of interested persons on DOE's site characterization plan and may review and consider comments made in connection with public hearings held by DOE.
              (d) The Director shall provide to DOE the site characterization analysis together with such additional comments as may be warranted. These comments shall include either a statement that the Director has no objection to the DOE's site characterization program, if such a statement is appropriate, or specific objections with respect to DOE's program for characterization of the area concerned. In addition, the Director may make specific recommendations pertinent to DOE's site characterization program.
              (e) If DOE's planned site characterization activities include onsite testing with radioactive material, including radioactive tracers, the Director's comments shall include a determination regarding whether or not the Commission concurs that the proposed use of such radioactive material is necessary to provide data for the preparation of the environmental reports required by law and for an application to be submitted under § 60.22 of this part.
              (f) The Director shall publish in the Federal Register a notice of availability of the site characterization analysis and a request for public comment within a reasonable period, as specified (not less than 90 days). The notice along with copies of the site characterization analysis shall be available at the NRC Web site, http://www.nrc.gov, and copies of any comments received will also be made available there.
              (g) During the conduct of site characterization activities, DOE shall report not less than once every six months to the Commission on the nature and extent of such activities and the information that has been developed, and on the progress of waste form and waste package research and development. The semiannual reports shall include the results of site characterization studies, the identification of new issues, plans for additional studies to resolve new issues, elimination of planned studies no longer necessary, identification of decision points reached and modifications to schedules where appropriate. DOE shall also report its progress in developing the design of a geologic repository operations area appropriate for the area being characterized, noting when key design parameters or features which depend upon the results of site characterization will be established. Other topics related to site characterization shall also be covered if requested by the Director.
              (h) During the conduct of site characterization activities, NRC staff shall be permitted to visit and inspect the locations at which such activities are carried out and to observe excavations, borings, and in situ tests as they are done.
              (i) The Director may comment at any time in writing to DOE, expressing current views on any aspect of site characterization. In particular, such comments shall be made whenever the Director, upon review of comments invited on the site characterization analysis or upon review of DOE's semiannual reports, determines that there are substantial new grounds for making recommendations or stating objections to DOE's site characterization program. The Director shall invite public comment on any comments which the Director makes to DOE upon review of the DOE semiannual reports or on any other comments which the Director makes to DOE on site characterization.

              (j) The Director shall transmit copies of the site characterization analysis and all comments to DOE made by the Director under this section to the Governor and legislature of the State in which the area to be characterized is located and to the governing body of any affected Indian Tribe. When transmitting the site characterization analysis under this paragraph, the Director shall invite the addressees to review and comment thereon.

              (k) All correspondence between DOE and the NRC under this section, including the reports described in paragraph (g), shall be placed in the Public Document Room.
              (l) The activities described in paragraphs (a) through (k) of this section constitute informal conference between a prospective applicant and the staff, as described in § 2.101(a)(1) of this chapter, and are not part of a proceeding under the Atomic Energy Act of 1954, as amended. Accordingly, neither the issuance of a site characterization analysis nor any other comments of the Director made under this section constitutes a commitment to issue any authorization or license or in any way affect the authority of the Commission, the Atomic Safety and Licensing Appeal Board, Atomic Safety and Licensing Boards, other presiding officers, or the Director, in any such proceeding.
              [51 FR 27163, July 30, 1986, as amended at 64 FR 48954, Sept. 9, 1999]

            License Applications

              § 60.21
              Content of application.
              (a) An application shall consist of general information and a Safety Analysis Report. An environmental impact statement shall be prepared in accordance with the Nuclear Waste Policy Act of 1982, as amended, and shall accompany the application. Any Restricted Data or National Security Information shall be separated from unclassified information.
              (b) The general information shall include:
              (1) A general description of the proposed geologic repository identifying the location of the geologic repository operations area, the general character of the proposed activities, and the basis for the exercise of licensing authority by the Commission.
              (2) Proposed schedules for construction, receipt of waste, and emplacement of wastes at the proposed geologic repository operations area.
              (3) A detailed plan to provide physical protection of high-level radioactive waste in accordance with § 73.51 of this chapter. This plan must include the design for physical protection, the licensee's safeguards contingency plan, and security organization personnel training and qualification plan. The plan must list tests, inspections, audits, and other means to be used to demonstrate compliance with such requirements.
              (4) A description of the program to meet the requirements of § 60.78.
              (5) A description of site characterization work actually conducted by DOE at all sites considered in the application and, as appropriate, explanations of why such work differed from the description of the site characterization program described in the Site Characterization Report for each site.
              (c) The Safety Analysis Report shall include:
              (1) A description and assessment of the site at which the proposed geologic repository operations area is to be located with appropriate attention to those features of the site that might affect geologic repository operations area design and performance. The description of the site shall identify the location of the geologic repository operations area with respect to the boundary of the accessible environment.
              (i) The description of the site shall also include the following information regarding subsurface conditions. This description shall, in all cases, include this information with respect to the postclosure controlled area. In addition, where subsurface conditions outside the postclosure controlled area may affect isolation within the postclosure controlled area, the description shall include information with respect to subsurface conditions outside the postclosure controlled area to the extent the information is relevant and material. The detailed information referred to in this paragraph shall include:
              (A) The orientation, distribution, aperture in-filling and origin of fractures, discontinuities, and heterogeneities;
              (B) The presence and characteristics of other potential pathways such as solution features, breccia pipes, or other potentially permeable features;
              (C) The geomechanical properties and conditions, including pore pressure and ambient stress conditions;

              (D) The hydrogeologic properties and conditions;

              (E) The geochemical properties; and
              (F) The anticipated response of the geomechanical, hydrogeologic, and geochemical systems to the maximum design thermal loading, given the pattern of fractures and other discontinuities and the heat transfer properties of the rock mass and groundwater.
              (ii) The assessment shall contain:
              (A) An analysis of the geology, geophysics, hydrogeology, geochemistry, climatology, and meteorology of the site,
              (B) Analyses to determine the degree to which each of the favorable and potentially adverse conditions, if present, has been characterized, and the extent to which it contributes to or detracts from isolation. For the purpose of determining the presence of the potentially adverse conditions, investigations shall extend from the surface to a depth sufficient to determine critical pathways for radionuclide migration from the underground facility to the accessible environment. Potentially adverse conditions shall be investigated outside of the postclosure controlled area if they affect isolation within the postclosure controlled area.
              (C) An evaluation of the performance of the proposed geologic repository for the period after permanent closure, assuming anticipated processes and events, giving the rates and quantities of releases of radionuclides to the accessible environment as a function of time; and a similar evaluation which assumes the occurrence of unanticipated processes and events.
              (D) The effectiveness of engineered and natural barriers, including barriers that may not be themselves a part of the geologic repository operations area, against the release of radioactive material to the environment. The analysis shall also include a comparative evaluation of alternatives to the major design features that are important to waste isolation, with particular attention to the alternatives that would provide longer radionuclide containment and isolation.
              (E) An analysis of the performance of the major design structures, systems, and components, both surface and subsurface, to identify those that are important to safety. For the purposes of this analysis, it shall be assumed that operations at the geologic repository operations area will be carried out at the maximum capacity and rate of receipt of radioactive waste stated in the application.
              (F) An explanation of measures used to support the models used to perform the assessments required in paragraphs (A) through (D). Analyses and models that will be used to predict future conditions and changes in the geologic setting shall be supported by using an appropriate combination of such methods as field tests, in situ tests, laboratory tests which are representative of field conditions, monitoring data, and natural analog studies.
              (2) A description and discussion of the design, both surface and subsurface, of the geologic repository operations area including:
              (i) The principal design criteria and their relationship to any general performance objectives promulgated by the Commission,
              (ii) The design bases and the relation of the design bases to the principal design criteria,
              (iii) Information relative to materials of construction (including geologic media, general arrangement, and approximate dimensions), and
              (iv) Codes and standards that DOE proposes to apply to the design and construction of the geologic repository operations area.
              (3) A description and analysis of the design and performance requirements for structures, systems, and components of the geologic repository that are important to safety. The analysis must include a demonstration that—
              (i) The requirements of § 60.111(a) will be met, assuming occurrence of Category 1 design basis events; and
              (ii) The requirements of § 60.136 will be met, assuming occurrence of Category 2 design basis events.
              (4) A description of the quality assurance program to be applied to the structures, systems, and components important to safety and to the engineered and natural barriers important to waste isolation.

              (5) A description of the kind, amount, and specifications of the radioactive material proposed to be received and possessed at the geologic repository operations area.

              (6) An identification and justification for the selection of those variables, conditions, or other items which are determined to be probable subjects of license specifications. Special attention shall be given to those items that may significantly influence the final design.
              (7) A description of the program for control and monitoring of radioactive effluents and occupational radiation exposures to maintain such effluents and exposures in accordance with the requirements of part 20 of this chapter.
              (8) A description of the controls that the applicant will apply to restrict access and to regulate land use at the site and adjacent areas, including a conceptual design of monuments which would be used to identify the postclosure controlled area after permanent closure.
              (9) Plans for coping with radiological emergencies at any time prior to permanent closure and decontamination or dismantlement of surface facilities.
              (10) A description of the program to be used to maintain the records described in §§ 60.71 and 60.72.
              (11) A description of design considerations that are intended to facilitate permanent closure and decontamination or dismantlement of surface facilities.
              (12) A description of plans for retrieval and alternate storage of the radioactive wastes should the geologic repository prove to be unsuitable for disposal of radioactive wastes.
              (13) An identification and evaluation of the natural resources of the geologic setting, including estimates as to undiscovered deposits, the exploitation of which could affect the ability of the geologic repository to isolate radioactive wastes. Undiscovered deposits of resources characteristic of the area shall be estimated by reasonable inference based on geological and geophysical evidence. This evaluation of resources, including undiscoverd deposits, shall be conducted for the site and for areas of similar size that are representative of and are within the geologic setting. For natural resources with current markets the resources shall be assessed, with estimates provided of both gross and net value. The estimate of net value shall take into account current development, extraction and marketing costs. For natural resources without current markets, but which would be marketable given credible projected changes in economic or technological factors, the resources shall be described by physical factors such as tonnage or other amount, grade, and quality.
              (14) An identification of those structures, systems, and components of the geologic repository, both surface and subsurface, which require research and development to confirm the adequacy of design. For structures, systems, and components important to safety and for the engineered and natural barriers important to waste isolation, DOE shall provide a detailed description of the programs designed to resolve safety questions, including a schedule indicating when these questions would be resolved.
              (15) The following information concerning activities at the geologic repository operations area:
              (i) The organizational structure of DOE as it pertains to construction and operation of the geologic repository operations area including a description of any delegations of authority and assignments of responsibilities, whether in the form of regulations, administrative directives, contract provisions, or otherwise.
              (ii) Identification of key positions which are assigned responsibility for safety at and operation of the geologic repository operations area.
              (iii) Personnel qualifications and training requirements.
              (iv) Plans for startup activities and startup testing.
              (v) Plans for conduct of normal activities, including maintenance, surveillance, and periodic testing of structures, systems, and components of the geologic repository operation area.
              (vi) Plans for permanent closure and plans for the decontamination or dismantlement of surface facilities.

              (vii) Plans for any uses of the geologic repository operations area for purposes other than disposal of radioactive wastes, with an analysis of the effects, if any, that such uses may have upon the operation of the structures, systems, and components important to safety and the engineered and natural barriers important to waste isolation.
              (d) The applicant for a license to receive and possess source, special nuclear, and byproduct material at a geologic repository operations area sited, constructed, or operated in accordance with the Nuclear Waste Policy Act of 1982 shall protect Safeguards Information in accordance with the requirements in § 73.21 and the requirements in § 73.22 or § 73.23 of this chapter, as applicable, and shall protect classified information in accordance with the requirements of parts 25 and 95 of this chapter, as applicable.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28219, June 21, 1983; 54 FR 27871, July 3, 1989; 61 FR 64268, Dec. 4, 1996; 63 FR 26961, May 15, 1998; 73 FR 63571, Oct. 24, 2008]

              § 60.22
              Filing and distribution of application.
              (a) An application for a construction authorization for a high-level radioactive waste repository at a geologic repository operations area, and an application for a license to receive and possess source, special nuclear, or byproduct material at a geologic repository operations area at a site which has been characterized, and any amendments thereto, and an accompanying environmental impact statement and any supplements, shall be signed by the Secretary of Energy or the Secretary's authorized representative and must be filed with the Director.
              (b) DOE shall maintain the capability to generate additional copies for distribution in accordance with written instructions from the Director or the Director's designee.
              (c) DOE shall, upon notification of the appointment of an Atomic Safety and Licensing Board, update the application, eliminating all superseded information, and supplement the environmental impact statement if necessary, and serve the updated application and environmental impact statement (as it may have been supplemented) as directed by the Board. At that time DOE shall also serve one such copy of the application and environmental impact statement on the Atomic Safety and Licensing Appeal Panel. Any subsequent amendments to the application or supplements to the environmental impact statement shall be served in the same manner.
              (d) At the time of filing of an application and any amendments thereto, one copy shall be made available in an appropriate location near the proposed geologic repository operations area (which shall be a public document room, if one has been established) for inspection by the public and updated as amendments to the application are made. The environmental impact statement and any supplements thereto shall be made available in the same manner. An updated copy of the application, and the environmental impact statement and supplements, shall be produced at any public hearing held by the Commission on the application, for use by any party to the proceeding.
              (e) The DOE shall certify that the updated copies of the application, and the environmental impact statement as it may have been supplemented, as referred to in paragraphs (c) and (d) of this section, contain the current contents of such documents submitted in accordance with the requirements of this part.
              [54 FR 27871, July 3, 1989, as amended at 68 FR 58814, Oct. 10, 2003; 69 FR 2279, Jan. 14, 2004]

              § 60.23
              Elimination of repetition.

              In its application, environmental report, or Site Characterization Report, the DOE may incorporate by reference information contained in previous applications, statements, or reports filed with the Commission: Provided, That such references are clear and specific and that copies of the information so incorporated are available in the public document room located near the site of the proposed geologic repository.

              § 60.24
              Updating of application and environmental impact statement.
              (a) The application shall be as complete as possible in the light of information that is reasonably available at the time of docketing.
              (b) The DOE shall update its application in a timely manner so as to permit the Commission to review, prior to issuance of a license:

              (1) Additional geologic, geophysical, geochemical, hydrologic, meteorologic and other data obtained during construction.
              (2) Conformance of construction of structures, systems, and components with the design.
              (3) Results of research programs carried out to confirm the adequacy of designs.
              (4) Other information bearing on the Commission's issuance of a license that was not available at the time a construction authorization was issued.
              (c) The DOE shall supplement its environmental impact statement in a timely manner so as to take into account the environmental impacts of any substantial changes in its proposed actions or any significant new circumstances or information relevant to environmental concerns and bearing on the proposed action or its impacts.
              [46 FR 13980, Feb. 25, 1981, as amended at 54 FR 27872, July 3, 1989]

            Construction Authorization

              § 60.31
              Construction authorization.
              Upon review and consideration of an application and environmental impact statement submitted under this part, the Commission may authorize construction if it determines:
              (a) Safety. That there is reasonable assurance that the types and amounts of radioactive materials described in the application can be received, possessed, and disposed of in a geologic repository operations area of the design proposed without unreasonable risk to the health and safety of the public. In arriving at this determination, the Commission shall consider whether:
              (1) DOE has described the proposed geologic repository including but not limited to: (i) The geologic, geophysical, geochemical and hydrologic characteristics of the site; (ii) the kinds and quantities of radioactive waste to be received, possessed, stored, and disposed of in the geologic repository operations area; (iii) the principal architectural and engineering criteria for the design of the geologic repository operations area; (iv) construction procedures which may affect the capability of the geologic repository to serve its intended function; and (v) features or components incorporated in the design for the protection of the health and safety of the public.
              (2) The site and design comply with the performance objectives and criteria contained in Subpart E of this part.
              (3) The DOE's quality assurance program complies with the requirements of Subpart G of this part.
              (4) The DOE's personnel training program complies with the criteria contained in Subpart H of this part.
              (5) The DOE's emergency plan complies with the criteria contained in Subpart I of this part.
              (6) The DOE's proposed operating procedures to protect health and to minimize danger to life or property are adequate.
              (b) Common defense and security. That there is reasonable assurance that the activities proposed in the application will not be inimical to the common defense and security.
              (c) Environmental. That, after weighing the environmental, economic, technical and other benefits against environmental costs and considering available alternatives, the action called for is issuance of the construction authorization, with any appropriate conditions to protect environmental values.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28220, June 21, 1983; 54 FR 27872, July 3, 1989; 63 FR 26961, May 15, 1998]

              § 60.32
              Conditions of construction authorization.
              (a) A construction authorization shall include such conditions as the Commission finds to be necessary to protect the health and safety of the public, the common defense and security, or environmental values.

              (b) The Commission will incorporate in the construction authorization provisions requiring DOE to furnish periodic or special reports regarding: (1) Progress of construction, (2) any data about the site obtained during construction which are not within the predicted limits upon which the facility design was based, (3) any deficiencies in design and construction which, if uncorrected, could adversely affect safety at any future time, and (4) results of research and development programs being conducted to resolve safety questions.

              (c) The construction authorization will include restrictions on subsequent changes to the features of the geologic repository and the procedures authorized. The restrictions that may be imposed under this paragraph can include measures to prevent adverse effects on the geologic setting as well as measures related to the design and construction of the geologic repository operations area. These restrictions will fall into three categories of descending importance to public health and safety as follows: (1) Those features and procedures which may not be changed without: (i) 60 days prior notice to the Commission (ii) 30 days notice of opportunity for a prior hearing, and (iii) prior Commission approval; (2) those features and procedures which may not be changed without (i) 60 days prior notice to the Commission, and (ii) prior Commission approval; and (3) those features and procedures which may not be changed without 60 days notice to the Commission. Features and procedures falling in paragraph (c)(3) of this section may not be changed without prior Commission approval if the Commission, after having received the required notice, so orders.
              (d) A construction authorization shall be subject to the limitation that a license to receive and possess source, special nuclear, or byproduct material at the geologic repository operations area shall not be issued by the Commission until (1) the DOE has updated its application as specified in § 60.24, and (2) the Commission has made the findings stated in § 60.41.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28221, June 21, 1983]

              § 60.33
              Amendment of construction authorization.
              (a) An application for amendment of a construction authorization shall be filed with the Commission fully describing any changes desired and following as far as applicable the format prescribed in § 60.21.
              (b) In determining whether an amendment of a construction authorization will be approved, the Commission will be guided by the considerations which govern the issuance of the initial construction authorization, to the extent applicable.

            License Issuance and Amendment

              § 60.41
              Standards for issuance of a license.
              A license to receive and possess source, special nuclear, or byproduct material at a geologic repository operations area may be issued by the Commission upon finding that:
              (a) Construction of the geologic repository operations area has been substantially completed in conformity with the application as amended, the provisions of the Atomic Energy Act, and the rules and regulations of the Commission. Construction may be deemed to be substantially complete for the purposes of this paragraph if the construction of (1) surface and interconnecting structures, systems, and components, and (2) any underground storage space required for initial operation are substantially complete.
              (b) The activities to be conducted at the geologic repository operations area will be in conformity with the application as amended, the provisions of the Atomic Energy Act and the Energy Reorganization Act, and the rules and regulations of the Commission.
              (c) The issuance of the license will not be inimical to the common defense and security and will not constitute an unreasonable risk to the health and safety of the public.
              (d) All applicable requirements of part 51 have been satisfied.
              [46 FR 13980, Feb. 25, 1981, as amended at 63 FR 26961, May 15, 1998]

              § 60.42
              Conditions of license.
              (a) A license issued pursuant to this part shall include such conditions, including license specifications, as the Commission finds to be necessary to protect the health and safety of the public, the common defense and security, and environmental values.
              (b) Whether stated therein or not, the following shall be deemed conditions in every license issued:

              (1) The license shall be subject to revocation, suspension, modification, or amendment for cause as provided by the Atomic Energy Act and the Commission's regulations.

              (2) The DOE shall at any time while the license is in effect, upon written request of the Commission, submit written statements to enable the Commission to determine whether or not the license should be modified, suspended or revoked.
              (3) The license shall be subject to the provisions of the Atomic Energy Act now or hereafter in effect and to all rules, regulations, and orders of the Commission. The terms and conditions of the license shall be subject to amendment, revision, or modification, by reason of amendments to or by reason of rules, regulations, and orders issued in accordance with the terms of the Atomic Energy Act.
              (c) Each license shall be deemed to contain the provisions set forth in Section 183 b-d, inclusive, of the Atomic Energy Act, whether or not these provisions are expressly set forth in the license.
              (d) The licensee (Department of Energy) shall ensure that Safeguards Information is protected against unauthorized disclosure in accordance with the requirements in § 73.21 and the requirements in § 73.22 or § 73.23 of this chapter, as applicable. The licensee (Department of Energy) shall ensure that classified information is protected in accordance with the requirements of parts 25 and 95 of this chapter, as applicable.
              [46 FR 13980, Feb. 25, 1981, as amended at 73 FR 63571, Oct. 24, 2008]

              § 60.43
              License specification.
              (a) A license issued under this part shall include license conditions derived from the analyses and evaluations included in the application, including amendments made before a license is issued, together with such additional conditions as the Commission finds appropriate.
              (b) License conditions shall include items in the following categories:
              (1) Restrictions as to the physical and chemical form and radioisotopic content of radioactive waste.
              (2) Restrictions as to size, shape, and materials and methods of construction of radioactive waste packaging.
              (3) Restrictions as to the amount of waste permitted per unit volume of storage space considering the physical characteristics of both the waste and the host rock.
              (4) Requirements relating to test, calibration, or inspection to assure that the foregoing restrictions are observed.
              (5) Controls to be applied to restricted access and to avoid disturbance to the postclosure controlled area and to areas outside the controlled area where conditions may affect isolation within the controlled area.
              (6) Administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure that activities at the facility are conducted in a safe manner and in conformity with the other license specifications.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28221, June 21, 1983; 61 FR 64268, Dec. 4, 1996]

              § 60.44
              Changes, tests, and experiments.
              (a)(1) Following authorization to receive and possess source, special nuclear, or byproduct material at a geologic repository operations area, the DOE may (i) make changes in the geologic repository operations area as described in the application, (ii) make changes in the procedures as described in the application, and (iii) conduct tests or experiments not described in the application, without prior Commission approval, provided the change, test, or experiment involves neither a change in the license conditions incorporated in the license nor an unreviewed safety question.
              (2) A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question if (i) the likelihood of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the application is increased, (ii) the possibility of an accident or malfunction of a different type than any previously evaluated in the application is created, or (iii) the margin of safety as defined in the basis for any license condition is reduced.

              (b) The DOE shall maintain records of changes in the geologic repository operations area and of changes in procedures made pursuant to this section, to the extent that such changes constitute changes in the geologic repository operations area or procedures as described in the application. Records of tests and experiments carried out pursuant to paragraph (a) of this section shall also be maintained. These records shall include a written safety evaluation which provides the basis for the determination that the change, test, or experiment does not involve an unreviewed safety question. The DOE shall prepare annually, or at such shorter intervals as may be specified in the license, a report containing a brief description of such changes, tests, and experiments, including a summary of the safety evaluation of each. The DOE shall furnish the report to the appropriate NRC Regional Office shown in appendix D to part 20 of this chapter, by an appropriate method listed in § 60.4(a), with a copy to the Director of the NRC's Office of Nuclear Material Safety and Safeguards. Any report submitted pursuant to this paragraph shall be made a part of the public record of the licensing proceedings.
              [46 FR 13980, Feb. 25, 1981, as amended at 52 FR 31612, Aug. 21, 1987; 68 FR 58814, Oct. 10, 2003]

              § 60.45
              Amendment of license.
              (a) An application for amendment of a license may be filed with the Commission fully describing the changes desired and following as far as applicable the format prescribed for license applications.
              (b) In determining whether an amendment of a license will be approved, the Commission will be guided by the considerations that govern the issuance of the initial license, to the extent applicable.

              § 60.46
              Particular activities requiring license amendment.
              (a) Unless expressly authorized in the license, an amendment of the license shall be required with respect to any of the following activities:
              (1) Any action which would make emplaced high-level radioactive waste irretrievable or which would substantially increase the difficulty of retrieving such emplaced waste.
              (2) Dismantling of structures.
              (3) Removal or reduction of controls applied to restrict access to or avoid disturbance of the controlled area and to areas outside the postclosure controlled area where conditions may affect isolation within the controlled area.
              (4) Destruction or disposal of records required to be maintained under the provisions of this part.
              (5) Any substantial change to the design or operating procedures from that specified in the license.
              (6) Permanent closure.
              (7) Any other activity involving an unreviewed safety question.
              (b) An application for such an amendment shall be filed, and shall be reviewed, in accordance with the provisions of § 60.45.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28221, June 21, 1983; 61 FR 64268, Dec. 4, 1996]

            US/IAEA Safeguards Agreement

              § 60.47
              Facility information and verification.
              (a) In response to a written request by the Commission, each applicant for a construction authorization or license and each recipient of a construction authorization or a license shall submit facility information, as described in § 75.10 of this chapter, on Form N-71 and associated forms, and site information on DOC/NRC Form AP-A and associated forms;
              (b) As required by the Additional Protocol, applicants and licensees specified in paragraph (a) of this section shall submit location information described in § 75.11 of this chapter on DOC/NRC Form AP-1 and associated forms;
              (c) Shall permit verification thereof by the International Atomic Energy Agency (IAEA) and take other action as necessary to implement the US/IAEA Safeguards Agreement, as described in part 75 of this chapter.
              [73 FR 78605, Dec. 23, 2008]

            Permanent Closure

              § 60.51
              License amendment for permanent closure.
              (a) DOE shall submit an application to amend the license prior to permanent closure. The submission shall consist of an update of the license application submitted under §§ 60.21 and 60.22, including:
              (1) A description of the program for post-permanent closure monitoring of the geologic repository.
              (2) A detailed description of the measures to be employed—such as land use controls, construction of monuments, and preservation of records—to regulate or prevent activities that could impair the long-term isolation of emplaced waste within the geologic repository and to assure that relevant information will be preserved for the use of future generations. As a minimum, such measures shall include:
              (i) Identification of the postclosure controlled area and geologic repository operations area by monuments that have been designed, fabricated, and emplaced to be as permanent as is practicable; and
              (ii) Placement of records in the archives and land record systems of local State, and Federal government agencies, and archives elsewhere in the world, that would be likely to be consulted by potential human intruders—such records to identify the location of the geologic repository operations area, including the underground facility, boreholes and shafts, and the boundaries of the postclosure controlled area, and the nature and hazard of the waste.
              (3) Geologic, geophysical, geochemical, hydrologic, and other site data that are obtained during the operational period pertinent to the long-term isolation of emplaced radioactive wastes.
              (4) The results of tests, experiments, and any other analyses relating to backfill of excavated areas, shaft sealing, waste interaction with the host rock, and any other tests, experiments, or analyses pertinent to the long-term isolation of emplaced wastes within the geologic repository.
              (5) Any substantial revision of plans for permanent closure.
              (6) Other information bearing upon permanent closure that was not available at the time a license was issued.
              (b) If necessary, so as to take into account the environmental impact of any substantial changes in the permanent closure activities proposed to be carried out or any significant new information regarding the environmental impacts of such closure, DOE shall also supplement its environmental impact statement and submit such statement, as supplemented, with the application for license amendment.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28221, June 21, 1983; 54 FR 27872, July 3, 1989; 61 FR 64268, Dec. 4, 1996]

              § 60.52
              Termination of license.
              (a) Following permanent closure and the decontamination or dismantlement of surface facilities, DOE may apply for an amendment to terminate the license.
              (b) Such application shall be filed, and will be reviewed, in accordance with the provisions of § 60.45 and this section.
              (c) A license shall be terminated only when the Commission finds with respect to the geologic repository:
              (1) That the final disposition of radioactive wastes has been made in conformance with the DOE's plan, as amended and approved as part of the license.
              (2) That the final state of the geologic repository operations area conforms to DOE's plans for permanent closure and DOE's plans for the decontamination or dismantlement of surface facilities, as amended and approved as part of the license.
              (3) That the termination of the license is authorized by law, including sections 57, 62, and 81 of the Atomic Energy Act, as amended.
              [46 FR 13980, Feb. 25, 1981, as amended at 48 FR 28222, June 21, 1983]

          Subpart C—Participation by State Governments and Affected Indian Tribes

            Source:
            51 FR 27164, July 30, 1986, unless otherwise noted.

            § 60.61
            Provision of information.
            (a) The Director shall provide to the Governor and legislature of any State in which a geologic repository operations area is or may be located, and to the governing body of any affected Indian Tribe, timely and complete information regarding determinations or plans made by the Commission with respect to the site characterization, siting, development, design, licensing, construction, operation, regulation, permanent closure, or decontamination and dismantlement of surface facilities, of such geologic repository operations area.
            (b) For purposes of this section, a geologic repository operations area shall be considered to be one which “may be located” in a State if the location thereof in such State has been described in a site characterization plan submitted to the Commission under this part.
            (c) Notwithstanding paragraph (a) of this section, the Director is not required to distribute any document to any entity if, with respect to such document, that entity or its counsel is included on a service list prepared pursuant to part 2 of this chapter.

            (d) Copies of all communications by the Director under this section are available at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public Document Room, and copies are furnished to DOE.
            [51 FR 27164, July 30, 1986, as amended at 64 FR 48954, Sept. 9, 1999]

            § 60.62
            Site review.
            (a) Whenever an area has been approved by the President for site characterization, and upon request of a State or an affected Indian Tribe, the Director shall make NRC staff available to consult with representatives of such States and Tribes.
            (b) Requests for consultation shall be made in writing to the Director.
            (c) Consultation under this section may include:
            (1) Keeping the parties informed of the Director's views on the progress of site characterization.
            (2) Review of applicable NRC regulations, licensing procedures, schedules, and opportunities for State and Tribe participation in the Commission's regulatory activities.
            (3) Cooperation in development of proposals for State and Tribe participation in license reviews.

            § 60.63
            Participation in license reviews.
            (a) State, local governmental bodies, and affected, Federally-recognized Indian Tribes may participate in license reviews as provided in subpart J of part 2 of this chapter. A State in which a repository for high-level radioactive waste is proposed to be located and any affected, Federally-recognized Indian Tribe shall have an unquestionable legal right to participate as a party in such proceedings.
            (b) In addition, whenever an area has been approved by the President for site characterization, a State or an affected Indian Tribe may submit to the Director a proposal to facilitate its participation in the review of a site characterization plan and/or license application. The proposal may be submitted at any time and must contain a description and schedule of how the State or affected Indian Tribe wishes to participate in the review, or what services or activities the State or affected Indian Tribe wishes NRC to carry out, and how the services or activities proposed to be carried out by NRC would contribute to such participation. The proposal may include educational or information services (seminars, public meetings) or other actions on the part of NRC, such as employment or exchange of State personnel under the Intergovernmental Personnel Act.
            (c) The Director shall arrange for a meeting between the representatives of the State or affected Indian Tribe and the NRC staff to discuss any proposal submitted under paragraph (b) of this section, with a view to identifying any modifications that may contribute to the effective participation by such State or Tribe.
            (d) Subject to the availability of funds, the Director shall approve all or any part of a proposal, as it may be modified through the meeting described above, if it is determined that:

            (1) The proposed activities are suitable in light of the type and magnitude of impacts which the State or affected Indian Tribe may bear;
            (2) The proposed activities:
            (i) Will enhance communications between NRC and the State or affected Indian Tribe;
            (ii) Will make a productive and timely contribution to the review; and
            (iii) Are authorized by law.
            (e) The Director will advise the State or affected Indian Tribe whether its proposal has been accepted or denied, and if all or any part of proposal is denied, the Director shall state the reason for the denial.

            (f) Proposals submitted under this section, and responses thereto, shall be made available at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public Document Room.
            [51 FR 27164, July 30, 1986, as amended at 64 FR 48954, Sept. 9, 1999; 69 FR 2279, Jan. 14, 2004]

            § 60.64
            Notice to States.
            If the Governor and legislature of a State have jointly designated on their behalf a single person or entity to receive notice and information from the Commission under this part, the Commission will provide such notice and information to the jointly designated person or entity instead of the Governor and legislature separately.

            § 60.65
            Representation.
            Any person who acts under this subpart as a representative for a State (or for the Governor or legislature thereof) or for an affected Indian Tribe shall include in the request or other submission, or at the request of the Commission, a statement of the basis of his or her authority to act in such representative capacity.

          Subpart D—Records, Reports, Tests, and Inspections

            § 60.71
            Records and reports.
            (a) DOE shall maintain such records and make such reports in connection with the licensed activity as may be required by the conditions of the license or by rules, regulations, and orders of the Commission as authorized by the Atomic Energy Act and the Energy Reorganization Act.
            (b) Records of the receipt, handling, and disposition of radioactive waste at a geologic repository operations area shall contain sufficient information to provide a complete history of the movement of the waste from the shipper through all phases of storage and disposal. DOE shall retain these records in a manner that ensures their useability for future generations in accordance with § 60.51(a)(2).
            [48 FR 28222, June 21, 1983, as amended at 53 FR 19251, May 27, 1988]

            § 60.72
            Construction records.
            (a) DOE shall maintain records of construction of the geologic repository operations area in a manner that ensures their useability for future generations in accordance with § 60.51(a)(2).
            (b) The records required under paragraph (a) shall include at least the following:
            (1) Surveys of the underground facility excavations, shafts, and boreholes referenced to readily identifiable surface features or monuments;
            (2) A description of the materials encountered;
            (3) Geologic maps and geologic cross sections;
            (4) Locations and amount of seepage;
            (5) Details of equipment, methods, progress, and sequence of work;
            (6) Construction problems;
            (7) Anomalous conditions encountered;
            (8) Instrument locations, readings, and analysis;
            (9) Location and description of structural support systems;
            (10) Location and description of dewatering systems; and
            (11) Details, methods of emplacement, and location of seals used.
            [48 FR 28222, June 21, 1983, as amended at 53 FR 19251, May 27, 1988]

            § 60.73
            Reports of deficiencies.

            DOE shall promptly notify the Commission of each deficiency found in the characteristics of the site, and design and construction of the geologic repository operations area which, were it to remain uncorrected, could: (a) Be a substantial safety hazard, (b) represent a significant deviation from the design criteria and design bases stated in the application, or (c) represent a deviation from the conditions stated in the terms of a construction authorization or the license, including license specifications. The notification shall be in the form of a written report, copies of which shall be sent to the Director and to the appropriate Nuclear Regulatory Commission Regional Office listed in appendix D of part 20 of this chapter.
            [48 FR 28222, June 21, 1983]

            § 60.74
            Tests.
            (a) DOE shall perform, or permit the Commission to perform, such tests as the Commission deems appropriate or necessary for the administration of the regulations in this part. These may include tests of:
            (1) Radioactive waste,
            (2) The geologic repository including its structures, systems, and components,
            (3) Radiation detection and monitoring instruments, and
            (4) Other equipment and devices used in connection with the receipt, handling, or storage of radioactive waste.
            (b) The tests required under this section shall include a performance confirmation program carried out in accordance with subpart F of this part.
            [48 FR 28222, June 21, 1983]

            § 60.75
            Inspections.
            (a) DOE shall allow the Commission to inspect the premises of the geologic repository operations area and adjacent areas to which DOE has rights of access.
            (b) DOE shall make available to the Commission for inspection, upon reasonable notice, records kept by DOE pertaining to activities under this part.
            (c)(1) DOE shall upon requests by the Director, Office of Nuclear Material Safety and Safeguards, provide rent-free office space for the exclusive use of the Commission inspection personnel. Heat, air-conditioning, light, electrical outlets and janitorial services shall be furnished by DOE. The office shall be convenient to and have full access to the facility and shall provide the inspector both visual and acoustic privacy.
            (2) The space provided shall be adequate to accommodate a full-time inspector, a part-time secretary and transient NRC personnel and will be generally commensurate with other office facilities at the geologic repository operations area. A space of 250 square feet either within the geologic repository operations area's office complex or in an office trailer or other onsite space at the geologic repository operations area is suggested as a guide. For locations at which activities are carried out under licenses issued under other parts of this chapter, additional space may be requested to accommodate additional full-time inspectors. The Office space that is provided shall be subject to the approval of the Director, Office of Nuclear Material Safety and Safeguards. All furniture, supplies and communication equipment will be furnished by the Commission.
            (3) DOE shall afford any NRC resident inspector assigned to that location, or other NRC inspectors identified by the Regional Administrator as likely to inspect the facility, immediate unfettered access, equivalent to access provided regular employees, following proper identification and compliance with applicable access control measures for security, radiological protection and personal safety.
            [48 FR 28222, June 21, 1983, as amended at 52 FR 31612, Aug. 21, 1987; 76 FR 72086, Nov. 22, 2011]

            § 60.78
            Material control and accounting records and reports.
            DOE shall implement a program of material control and accounting (and accidental criticality reporting) that is the same as that specified in §§ 72.72, 72.74, 72.76, and 72.78 of this chapter.
            [63 FR 26961, May 15, 1998]

          Subpart E—Technical Criteria

            Source:
            48 FR 28222, June 21, 1983, unless otherwise noted.

            § 60.101
            Purpose and nature of findings.

            (a)(1) Subpart B of this part prescribes the standards for issuance of a license to receive and possess source, special nuclear, or byproduct material at a geologic repository operations area. In particular, § 60.41(c) requires a finding that the issuance of a license will not constitute an unreasonable risk to the health and safety of the public. The purpose of this subpart is to set out performance objectives and site and design criteria which, if satisfied, will support such a finding of no unreasonable risk.
            (2) While these performance objectives and criteria are generally stated in unqualified terms, it is not expected that complete assurance that they will be met can be presented. A reasonable assurance, on the basis of the record before the Commission, that the objectives and criteria will be met is the general standard that is required. For § 60.112, and other portions of this subpart that impose objectives and criteria for repository performance over long times into the future, there will inevitably be greater uncertainties. Proof of the future performance of engineered barrier systems and the geologic setting over time periods of many hundreds or many thousands of years is not to be had in the ordinary sense of the word. For such long-term objectives and criteria, what is required is reasonable assurance, making allowance for the time period, hazards, and uncertainties involved, that the outcome will be in conformance with those objectives and criteria. Demonstration of compliance with such objectives and criteria will involve the use of data from accelerated tests and predictive models that are supported by such measures as field and laboratory tests, monitoring data and natural analog studies.
            (b) Subpart B of this part also lists findings that must be made in support of an authorization to construct a geologic repository operations area. In particular, § 60.31(a) requires a finding that there is reasonable assurance that the types and amounts of radioactive materials described in the application can be received, possessed, and disposed of in a geologic repository operations area of the design proposed without unreasonable risk to the health and safety of the public. As stated in that paragraph, in arriving at this determination, the Commission will consider whether the site and design comply with the criteria contained in this subpart. Once again, while the criteria may be written in unqualified terms, the demonstration of compliance may take uncertainties and gaps in knowledge into account, provided that the Commission can make the specified finding of reasonable assurance as specified in paragraph (a) of this section.

            § 60.102
            Concepts.
            This section provides a functional overview of subpart E. In the event of any inconsistency with definitions found in § 60.2, those definitions shall prevail.
            (a) The HLW facility. NRC exercises licensing and related regulatory authority over those facilities described in section 202 (3) and (4) of the Energy Reorganization Act of 1974. Any of these facilities is designated a HLW facility.

            (b) The geologic repository operations area. (1) This part deals with the exercise of authority with respect to a particular class of HLW facility—namely a geologic repository operations area.

            (2) A geologic repository operations area consists of those surface and subsurface areas that are part of a geologic repository where radioactive waste handling activities are conducted. The underground structure, including openings and backfill materials, but excluding shafts, boreholes, and their seals, is designated the underground facility.

            (3) The exercise of Commission authority requires that the geologic repository operations area be used for storage (which includes disposal) of high-level radioactive wastes (HLW).

            (4) HLW includes irradiated reactor fuel as well as reprocessing wastes. However, if DOE proposes to use the geologic repository operations area for storage of radioactive waste other than HLW, the storage of this radioactive waste is subject to the requirements of this part.
            (c) Areas related to isolation. Although the activities subject to regulation under this part are those to be carried out at the geologic repository operations area, the licensing process also considers characteristics of adjacent areas that are defined in other ways. There is to be an area surrounding the underground facility referred to above, which is designated the postclosure controlled area, within which DOE is to exercise specified controls to prevent adverse human actions following permanent closure. The location of the controlled area is the site. The accessible environment is the atmosphere, land surface, surface water, oceans, and the portion of the lithosphere that is outside the controlled area. There is an area, designated the geologic setting, which includes the geologic, hydrologic, and geochemical systems of the region in which a geologic repository operations area is or may be located. The geologic repository operations area plus the portion of the geologic setting that provides isolation of the radioactive waste make up the geologic repository.

            (d) Stages in the licensing process. There are several stages in the licensing process. The site characterization stage, though begun before submission of a license application, may result in consequences requiring evaluation in the license review. The construction stage would follow, after issuance of a construction authorization. A period of operations follows the issuance of a license by the Commission. The period of operations includes the time during which emplacement of wastes occurs; any subsequent period before permanent closure during which the emplaced wastes are retrievable; and permanent closure, which includes sealing of shafts. Permanent closure represents the end of active human intervention with respect to the engineered barrier system.
            (e) Isolation of waste. (1) During the first several hundred years following permanent closure of a geologic repository, when radiation and thermal levels are high and the uncertainties in assessing repository performance are large, special emphasis is placed upon the ability to contain the wastes by waste packages within an engineered barrier system. This is known as the containment period. The engineered barrier system includes the waste packages and the underground facility. A waste package is composed of the waste form and any containers, shielding, packing, and absorbent materials immediately surrounding an individual waste container. The underground facility means the underground structure, including openings and backfill materials, but excluding, shafts, boreholes, and their seals.

            (2) Following the containment period special emphasis is placed upon the ability to achieve isolation of the wastes by virtue of the characteristics of the geologic repository. The engineered barrier system works to control the release of radioactive material to the geologic setting and the geologic setting works to control the release of radioactive material to the accessible environment. Isolation means inhibiting the transport of radioactive material so that amounts and concentrations of the materials entering the accessible environment will be kept within prescribed limits.
            [48 FR 28222, June 21, 1983, as amended at 61 FR 64268, Dec. 4, 1996]

            Performance Objectives

              § 60.111
              Performance of the geologic repository operations area through permanent closure.
              (a) Protection against radiation exposures and releases of radioactive material. The geologic repository operations area shall be designed so that until permanent closure has been completed, radiation exposures and radiation levels, and releases of radioactive materials to unrestricted areas, will be maintained within the limits specified in part 20 of this chapter and such generally applicable environmental standards for radioactivity as may have been established by the Environmental Protection Agency.
              (b) Retrievability of waste. (1) The geologic repository operations area shall be designed to preserve the option of waste retrieval throughout the period during which wastes are being emplaced and, thereafter, until the completion of a preformance confirmation program and Commission review of the information obtained from such a program. To satisfy this objective, the geologic repository operations area shall be designed so that any or all of the emplaced waste could be retrieved on a reasonable schedule starting at any time up to 50 years after waste emplacement operations are initiated, unless a different time period is approved or specified by the Commission. This different time period may be established on a case-by-case basis consistent with the emplacement schedule and the planned performance confirmation program.
              (2) This requirement shall not preclude decisions by the Commission to allow backfilling part or all of, or permanent closure of, the geologic repository operations area prior to the end of the period of design for retrievability.
              (3) For purposes of this paragraph, a reasonable schedule for retrieval is one that would permit retrieval in about the same time as that devoted to construction of the geologic repository operations area and the emplacement of wastes.
              [48 FR 28222, June 21, 1983, as amended at 61 FR 64268, Dec. 4, 1996; 62 FR 59276, Nov. 3, 1997]

              § 60.112
              Overall system performance objective for the geologic repository after permanent closure.
              The geologic setting shall be selected and the engineered barrier system and the shafts, boreholes and their seals shall be designed to assure that releases of radioactive materials to the accessible environment following permanent closure conform to such generally applicable environmental standards for radioactivity as may have been established by the Environmental Protection Agency with respect to both anticipated processes and events and unanticipated processes and events.

              § 60.113
              Performance of particular barriers after permanent closure.
              (a) General provisions—(1) Engineered barrier system. (i) The engineered barrier system shall be designed so that assuming anticipated processes and events: (A) Containment of HLW will be substantially complete during the period when radiation and thermal conditions in the engineered barrier system are dominated by fission product decay; and (B) any release of radionuclides from the engineered barrier system shall be a gradual process which results in small fractional releases to the geologic setting over long times. For disposal in the saturated zone, both the partial and complete filling with groundwater of available void spaces in the underground facility shall be appropriately considered and analysed among the anticipated processes and events in designing the engineered barrier system.
              (ii) In satisfying the preceding requirement, the engineered barrier system shall be designed, assuming anticipated processes and events, so that:
              (A) Containment of HLW within the waste packages will be substantially complete for a period to be determined by the Commission taking into account the factors specified in § 60.113(b) provided, that such period shall be not less than 300 years nor more than 1,000 years after permanent closure of the geologic repository; and
              (B) The release rate of any radionuclide from the engineered barrier system following the containment period shall not exceed one part in 100,000 per year of the inventory of that radionuclide calculated to be present at 1,000 years following permanent closure, or such other fraction of the inventory as may be approved or specified by the Commission; provided, that this requirement does not apply to any radionuclide which is released at a rate less than 0.1% of the calculated total release rate limit. The calculated total release rate limit shall be taken to be one part in 100,000 per year of the inventory of radioactive waste, originally emplaced in the underground facility, that remains after 1,000 years of radioactive decay.
              (2) Geologic setting. The geologic repository shall be located so that pre-waste-emplacement groundwater travel time along the fastest path of likely radionuclide travel from the disturbed zone to the accessible environment shall be at least 1,000 years or such other travel time as may be approved or specified by the Commission.

              (b) On a case-by-case basis, the Commission may approve or specify some other radionuclide release rate, designed containment period or pre-waste-emplacement groundwater travel time, provided that the overall system performance objective, as it relates to anticipated processes and events, is satisfied. Among the factors that the Commission may take into account are:

              (1) Any generally applicable environmental standard for radioactivity established by the Environmental Protection Agency;
              (2) The age and nature of the waste, and the design of the underground facility, particularly as these factors bear upon the time during which the thermal pulse is dominated by the decay heat from the fission products;
              (3) The geochemical characteristics of the host rock, surrounding strata and groundwater; and
              (4) Particular sources of uncertainty in predicting the performance of the geologic repository.
              (c) Additional requirements may be found to be necessary to satisfy the overall system performance objective as it relates to unanticipated processes and events.

            Land Ownership and Control

              § 60.121
              Requirements for ownership and control of interests in land.
              (a) Ownership of land. (1) Both the geologic repository operations area and the postclosure controlled area shall be located in and on lands that are either acquired lands under the jurisdiction and control of DOE, or lands permanently withdrawn and reserved for its use.
              (2) These lands shall be held free and clear of all encumbrances, if significant, such as: (i) Rights arising under the general mining laws; (ii) easements for right-of-way; and (iii) all other rights arising under lease, rights of entry, deed, patent, mortgage, appropriation, prescription, or otherwise.
              (b) Additional controls. Appropriate controls shall be established outside of the postclosure controlled area. DOE shall exercise any jurisdiction and control over surface and subsurface estates necessary to prevent adverse human actions that could significantly reduce the geologic repository's ability to achieve isolation. The rights of DOE may take the form of appropriate possessory interests, servitudes, or withdrawals from location or patent under the general mining laws.
              (c) Water rights. (1) DOE shall also have obtained such water rights as may be needed to accomplish the purpose of the geologic repository operations area.
              (2) Water rights are included in the additional controls to be established under paragraph (b) of this section.
              [48 FR 28222, June 21, 1983, as amended at 61 FR 64268, Dec. 4, 1996]

            Siting Criteria

              § 60.122
              Siting criteria.
              (a)(1) A geologic setting shall exhibit an appropriate combination of the conditions specified in paragraph (b) of this section so that, together with the engineered barriers system, the favorable conditions present are sufficient to provide reasonable assurance that the performance objectives relating to isolation of the waste will be met.
              (2) If any of the potentially adverse conditions specified in paragraph (c) of this section is present, it may compromise the ability of the geologic repository to meet the performance objectives relating to isolation of the waste. In order to show that a potentially adverse condition does not so compromise the performance of the geologic repository the following must be demonstrated:
              (i) The potentially adverse human activity or natural condition has been adequately investigated, including the extent to which the condition may be present and still be undetected taking into account the degree of resolution achieved by the investigations; and
              (ii) The effect of the potentially adverse human activity or natural condition on the site has been adequately evaluated using analyses which are sensitive to the potentially adverse human activity or natural condition and assumptions which are not likely to underestimate its effect; and
              (iii)(A) The potentially adverse human activity or natural condition is shown by analysis pursuant to paragraph (a)(2)(ii) of this section not to affect significantly the ability of the geologic repository to meet the performance objectives relating to isolation of the waste, or

              (B) The effect of the potentially adverse human activity or natural condition is compensated by the presence of a combination of the favorable characteristics so that the performance objectives relating to isolation of the waste are met, or
              (C) The potentially adverse human activity or natural condition can be remedied.
              (b) Favorable conditions. (1) The nature and rates of tectonic, hydrogeologic, geochemical, and geomorphic processes (or any of such processes) operating within the geologic setting during the Quaternary Period, when projected, would not affect or would favorably affect the ability of the geologic repository to isolate the waste.
              (2) For disposal in the saturated zone, hydrogeologic conditions that provide:
              (i) A host rock with low horizontal and vertical permeability;
              (ii) Downward or dominantly horizontal hydraulic gradient in the host rock and immediately surrounding hydrogeologic units; and
              (iii) Low vertical permeability and low hydraulic gradient between the host rock and the surrounding hydrogeologic units.
              (3) Geochemical conditions that:
              (i) Promote precipitation or sorption of radionuclides;
              (ii) Inhibit the formation of particulates, colloids, and inorganic and organic complexes that increase the mobility of radionuclides; or
              (iii) Inhibit the transport of radionuclides by particulates, colloids, and complexes.
              (4) Mineral assemblages that, when subjected to anticipated thermal loading, will remain unaltered or alter to mineral assemblages having equal or increased capacity to inhibit radionuclide migration.
              (5) Conditions that permit the emplacement of waste at a minimum depth of 300 meters from the ground surface. (The ground surface shall be deemed to be the elevation of the lowest point on the surface above the disturbed zone.)
              (6) A low population density within the geologic setting and a postclosure controlled area that is remote from population centers.
              (7) Pre-waste-emplacement groundwater travel time along the fastest path of likely radionuclide travel from the disturbed zone to the accessible environment that substantially exceeds 1,000 years.
              (8) For disposal in the unsaturated zone, hydrogeologic conditions that provide—
              (i) Low moisture flux in the host rock and in the overlying and underlying hydrogeologic units;
              (ii) A water table sufficiently below the underground facility such that fully saturated voids contiguous with the water table do not encounter the underground facility;
              (iii) A laterally extensive low-permeability hydrogeologic unit above the host rock that would inhibit the downward movement of water or divert downward moving water to a location beyond the limits of the underground facility;
              (iv) A host rock that provides for free drainage; or
              (v) A climatic regime in which the average annual historic precipitation is a small percentage of the average annual potential evapotranspiration.
              (c) Potentially adverse conditions. The following conditions are potentially adverse conditions if they are characteristic of the postclosure controlled area or may affect isolation within the controlled area.
              (1) Potential for flooding of the underground facility, whether resulting from the occupancy and modification of floodplains or from the failure of existing or planned man-made surface water impoundments.
              (2) Potential for foreseeable human activity to adversely affect the groundwater flow system, such as groundwater withdrawal, extensive irrigation, subsurface injection of fluids, underground pumped storage, military activity or construction of large scale surface water impoundments.
              (3) Potential for natural phenomena such as landslides, subsidence, or volcanic activity of such a magnitude that large-scale surface water impoundments could be created that could change the regional groundwater flow system and thereby adversely affect the performance of the geologic repository.

              (4) Structural deformation, such as uplift, subsidence, folding, or faulting that may adversely affect the regional groundwater flow system.
              (5) Potential for changes in hydrologic conditions that would affect the migration of radionuclides to the accessible environment, such as changes in hydraulic gradient, average interstitial velocity, storage coefficient, hydraulic conductivity, natural recharge, potentiometric levels, and discharge points.
              (6) Potential for changes in hydrologic conditions resulting from reasonably foreseeable climatic changes.
              (7) Groundwater conditions in the host rock, including chemical composition, high ionic strength or ranges of Eh-pH, that could increase the solubility or chemical reactivity of the engineered barrier system.
              (8) Geochemical processes that would reduce sorption of radionuclides, result in degradation of the rock strength, or adversely affect the performance of the engineered barrier system.
              (9) Groundwater conditions in the host rock that are not reducing.
              (10) Evidence of dissolutioning such as breccia pipes, dissolution cavities, or brine pockets.
              (11) Structural deformation such as uplift, subsidence, folding, and faulting during the Quaternary Period.
              (12) Earthquakes which have occurred historically that if they were to be repeated could affect the site significantly.
              (13) Indications, based on correlations of earthquakes with tectonic processes and features, that either the frequency of occurrence or magnitude of earthquakes may increase.
              (14) More frequent occurrence of earthquakes or earthquakes of higher magnitude than is typical of the area in which the geologic setting is located.
              (15) Evidence of igneous activity since the start of the Quaternary Period.
              (16) Evidence of extreme erosion during the Quaternary Period.
              (17) The presence of naturally occurring materials, whether identified or undiscovered, within the site, in such form that:
              (i) Economic extraction is currently feasible or potentially feasible during the foreseeable future; or
              (ii) Such materials have greater gross value or net value than the average for other areas of similar size that are representative of and located within the geologic setting.
              (18) Evidence of subsurface mining for resources within the site.
              (19) Evidence of drilling for any purpose within the site.
              (20) Rock or groundwater conditions that would require complex engineering measures in the design and construction of the underground facility or in the sealing of boreholes and shafts.
              (21) Geomechanical properties that do not permit design of underground opening that will remain stable through permanent closure.
              (22) Potential for the water table to rise sufficiently so as to cause saturation of an underground facility located in the unsaturated zone.
              (23) Potential for existing or future perched water bodies that may saturate portions of the underground facility or provide a faster flow path from an underground facility located in the unsaturated zone to the accessible environment.
              (24) Potential for the movement of radionuclides in a gaseous state through air-filled pore spaces of an unsaturated geologic medium to the accessible environment.
              [48 FR 28222, June 21, 1983, as amended at 50 FR 29647, July 22, 1985; 61 FR 64269, Dec. 4, 1996]

            Design Criteria for the Geologic Repository Operations Area

              § 60.130
              General considerations.

              (a) Pursuant to the provisions of § 60.21(c)(2)(i), an application for construction authorization for a high-level radioactive waste repository at a geologic repository operations area, and an application for a license to receive, possess, store, and dispose of high-level radioactive waste in the geologic repository operations area, must include the principal design criteria for a proposed facility. The principal design criteria establish the necessary design, fabrication, construction, testing, maintenance, and performance requirements for structures, systems, and components important to safety and/or important to waste isolation. Sections 60.131 through 60.134 specify minimum requirements for the principal design criteria for the geologic repository operations area.
              (b) These design criteria are not intended to be exhaustive. However, omissions in §§ 60.131 through 60.134 do not relieve DOE from any obligation to provide such features in a specific facility needed to achieve the performance objectives.
              [69 FR 2280, Jan. 14, 2004]

              § 60.131
              General design criteria for the geologic repository operations area.
              (a) Radiological protection. The geologic repository operations area shall be designed to maintain radiation doses, levels, and concentrations of radioactive material in air in restricted areas within the limits specified in part 20 of this chapter. Design shall include:
              (1) Means to limit concentrations of radioactive material in air;
              (2) Means to limit the time required to perform work in the vicinity of radioactive materials, including, as appropriate, designing equipment for ease of repair and replacement and providing adequate space for ease of operation;
              (3) Suitable shielding;
              (4) Means to monitor and control the dispersal of radioactive contamination;
              (5) Means to control access to high radiation areas or airborne radioactivity areas; and
              (6) A radiation alarm system to warn of significant increases in radiation levels, concentrations of radioactive material in air, and of increased radioactivity released in effluents. The alarm system shall be designed with provisions for calibration and for testing its operability.
              (b) Protection against design basis events. The structures, systems, and components important to safety shall be designed so that they will perform their necessary safety functions, assuming occurrence of design basis events.
              (c) Protection against dynamic effects of equipment failure and similar events. The structures, systems, and components important to safety shall be designed to withstand dynamic effects such as missile impacts, that could result from equipment failure, and similar events and conditions that could lead to loss of their safety functions.
              (d) Protection against fires and explosions. (1) The structures, systems, and components important to safety shall be designed to perform their safety functions during and after credible fires or explosions in the geologic repository operations area.
              (2) To the extent practicable, the geologic repository operations area shall be designed to incorporate the use of noncombustible and heat resistant materials.
              (3) The geologic repository operations area shall be designed to include explosion and fire detection alarm systems and appropriate suppression systems with sufficient capacity and capability to reduce the adverse effects of fires and explosions on structures, systems, and components important to safety.
              (4) The geologic repository operations area shall be designed to include means to protect systems, structures, and components important to safety against the adverse effects of either the operation or failure of the fire suppression systems.
              (e) Emergency capability. (1) The structures, systems, and components important to safety shall be designed to maintain control of radioactive waste and radioactive effluents, and permit prompt termination of operations and evacuation of personnel during an emergency.
              (2) The geologic repository operations area shall be designed to include onsite facilities and services that ensure a safe and timely response to emergency conditions and that facilitate the use of available offsite services (such as fire, police, medical, and ambulance service) that may aid in recovery from emergencies.
              (f) Utility services. (1) Each utility service system that is important to safety shall be designed so that essential safety functions can be performed, assuming occurrence of the design basis events.

              (2) The utility services important to safety shall include redundant systems to the extent necessary to maintain, with adequate capacity, the ability to perform their safety functions.
              (3) Provisions shall be made so that, if there is a loss of the primary electric power source or circuit, reliable and timely emergency power can be provided to instruments, utility service systems, and operating systems, including alarm systems, important to safety.
              (g) Inspection, testing, and maintenance. The structures, systems, and components important to safety shall be designed to permit periodic inspection, testing, and maintenance, as necessary, to ensure their continued functioning and readiness.
              (h) Criticality control. All systems for processing, transporting, handling, storage, retrieval, emplacement, and isolation of radioactive waste shall be designed to ensure that nuclear criticality is not possible unless at least two unlikely, independent, and concurrent or sequential changes have occurred in the conditions essential to nuclear criticality safety. Each system must be designed for criticality safety assuming occurrence of design basis events. The calculated effective multiplication factor (keff) must be sufficiently below unity to show at least a 5 percent margin, after allowance for the bias in the method of calculation and the uncertainty in the experiments used to validate the method of calculation.
              (i) Instrumentation and control systems. The design shall include provisions for instrumentation and control systems to monitor and control the behavior of systems important to safety, assuming occurrence of design basis events.
              (j) Compliance with mining regulations. To the extent that DOE is not subject to the Federal Mine Safety and Health Act of 1977, as to the construction and operation of the geologic repository operations area, the design of the geologic repository operations area shall nevertheless include provisions for worker protection necessary to provide reasonable assurance that all structures, systems, and components important to safety can perform their intended functions. Any deviation from relevant design requirements in 30 CFR, chapter I, subchapters D, E, and N will give rise to a rebuttable presumption that this requirement has not been met.
              (k) Shaft conveyances used in radioactive waste handling. (1) Hoists important to safety shall be designed to preclude cage free fall.
              (2) Hoists important to safety shall be designed with a reliable cage location system.
              (3) Loading and unloading systems for hoists important to safety shall be designed with a reliable system of interlocks that will fail safely upon malfunction.
              (4) Hoists important to safety shall be designed to include two independent indicators to indicate when waste packages are in place and ready for transfer.
              [48 FR 28222, June 21, 1983, as amended at 61 FR 64269, Dec. 4, 1996]

              § 60.132
              Additional design criteria for surface facilities in the geologic repository operations area.
              (a) Facilities for receipt and retrieval of waste. Surface facilities in the geologic repository operations area shall be designed to allow safe handling and storage of wastes at the geologic repository operations area, whether these wastes are on the surface before emplacement or as a result of retrieval from the underground facility.
              (b) Surface facility ventilation. Surface facility ventilation systems supporting waste transfer, inspection, decontamination, processing, or packaging shall be designed to provide protection against radiation exposures and offsite releases as provided in § 60.111(a).
              (c) Radiation control and monitoring—(1) Effluent control. The surface facilities shall be designed to control the release of radioactive materials in effluents during Category 1 design basis events so as to meet the performance objectives of § 60.111(a).
              (2) Effluent monitoring. The effluent monitoring systems shall be designed to measure the amount and concentration of radionuclides in any effluent with sufficient precision to determine whether releases conform to the design requirement for effluent control. The monitoring systems shall be designed to include alarms that can be periodically tested.

              (d) Waste treatment. Radioactive waste treatment facilities shall be designed to process any radioactive wastes generated at the geologic repository operations area into a form suitable to permit safe disposal at the geologic repository operations area or to permit safe transportation and conversion to a form suitable for disposal at an alternative site in accordance with any regulations that are applicable.
              (e) Consideration of decommissioning. The surface facility shall be designed to facilitate decontamination or dismantlement to the same extent as would be required, under other parts of this chapter, with respect to equivalent activities licensed thereunder.
              [48 FR 28222, June 21, 1983, as amended at 61 FR 64270, Dec. 4, 1996]

              § 60.133
              Additional design criteria for the underground facility.
              (a) General criteria for the underground facility. (1) The orientation, geometry, layout, and depth of the underground facility, and the design of any engineered barriers that are part of the underground facility shall contribute to the containment and isolation of radionuclides.
              (2) The underground facility shall be designed so that the effects of credible disruptive events during the period of operations, such as flooding, fires and explosions, will not spread through the facility.
              (b) Flexibility of design. The underground facility shall be designed with sufficient flexibility to allow adjustments where necessary to accommodate specific site conditions identified through in situ monitoring, testing, or excavation.
              (c) Retrieval of waste. The underground facility shall be designed to permit retrieval of waste in accordance with the performance objectives of § 60.111.
              (d) Control of water and gas. The design of the underground facility shall provide for control of water or gas intrusion.
              (e) Underground openings. (1) Openings in the underground facility shall be designed so that operations can be carried out safely and the retrievability option maintained.
              (2) Openings in the underground facility shall be designed to reduce the potential for deleterious rock movement or fracturing of overlying or surrounding rock.
              (f) Rock excavation. The design of the underground facility shall incorporate excavation methods that will limit the potential for creating a preferential pathway for groundwater to contact the waste packages or radionuclide migration to the accessible environment.
              (g) Underground facility ventilation. The ventilation system shall be designed to:
              (1) Control the transport of radioactive particulates and gases within and releases from the underground facility in accordance with the performance objectives of § 60.111(a),
              (2) Assure the ability to perform essential safety functions assuming occurrence of design basis events.
              (3) Separate the ventilation of excavation and waste emplacement areas.
              (h) Engineered barriers. Engineered barriers shall be designed to assist the geologic setting in meeting the performance objectives for the period following permanent closure.
              (i) Thermal loads. The underground facility shall be designed so that the performance objectives will be met taking into account the predicted thermal and thermomechanical response of the host rock, and surrounding strata, groundwater system.
              [48 FR 28222, June 21, 1983, as amended at 50 FR 29648, July 22, 1985; 61 FR 64270, Dec. 4, 1996]

              § 60.134
              Design of seals for shafts and boreholes.
              (a) General design criterion. Seals for shafts and boreholes shall be designed so that following permanent closure they do not become pathways that compromise the geologic repository's ability to meet the performance objectives or the period following permanent closure.
              (b) Selection of materials and placement methods. Materials and placement methods for seals shall be selected to reduce, to the extent practicable:

              (1) The potential for creating a preferential pathway for groundwater to contact the waste packages or

              (2) For radionuclide migration through existing pathways.
              [48 FR 28222, June 21, 1983, as amended at 50 FR 29648, July 22, 1985]

            Design Criteria for the Waste Package

              § 60.135
              Criteria for the waste package and its components.
              (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste package and its interactions with the emplacement environment do not compromise the function of the waste packages or the performance of the underground facility or the geologic setting.
              (2) The design shall include but not be limited to consideration of the following factors: solubility, oxidation/reduction reactions, corrosion, hydriding, gas generation, thermal effects, mechanical strength, mechanical stress, radiolysis, radiation damage, radionuclide retardation, leaching, fire and explosion hazards, thermal loads, and synergistic interactions.
              (b) Specific criteria for HLW package design—(1) Explosive, pyrophoric, and chemically reactive materials. The waste package shall not contain explosive or pyrophoric materials or chemically reactive materials in an amount that could compromise the ability of the underground facility to contribute to waste isolation or the ability of the geologic repository to satisfy the performance objectives.
              (2) Free liquids. The waste package shall not contain free liquids in an amount that could compromise the ability of the waste packages to achieve the performance objectives relating to containment of HLW (because of chemical interactions or formation of pressurized vapor) or result in spillage and spread of contamination in the event of waste package perforation during the period through permanent closure.
              (3) Handling. Waste packages shall be designed to maintain waste containment during transportation, emplacement, and retrieval.
              (4) Unique identification. A label or other means of identification shall be provided for each waste package. The identification shall not impair the integrity of the waste package and shall be applied in such a way that the information shall be legible at least to the end of the period of retrievability. Each waste package identification shall be consistent with the waste package's permanent written records.
              (c) Waste form criteria for HLW. High-level radioactive waste that is emplaced in the underground facility shall be designed to meet the following criteria:
              (1) Solidification. All such radioactive wastes shall be in solid form and placed in sealed containers.
              (2) Consolidation. Particulate waste forms shall be consolidated (for example, by incorporation into an encapsulating matrix) to limit the availability and generation of particulates.
              (3) Combustibles. All combustible radioactive wastes shall be reduced to a noncombustible form unless it can be demonstrated that a fire involving the waste packages containing combustibles will not compromise the integrity of other waste packages, adversely affect any structures, systems, or components important to safety, or compromise the ability of the underground facility to contribute to waste isolation.
              (d) Design criteria for other radioactive wastes. Design criteria for waste types other than HLW will be addressed on an individual basis if and when they are proposed for disposal in a geologic repository.

            Preclosure Controlled Area

              § 60.136
              Preclosure controlled area.
              (a) A preclosure controlled area must be established for the geologic repository operations area.

              (b) The geologic repository operations area shall be designed so that, for Category 2 design basis events, no individual located on or beyond any point on the boundary of the preclosure controlled area will receive the more limiting of a total effective dose equivalent of 0.05 Sv (5 rem), or the sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue (other than the lens of the eye) of 0.5 Sv (50 rem). The eye dose equivalent shall not exceed 0.15 Sv (15 rem), and the shallow dose equivalent to skin shall not exceed 0.5 Sv (50 rem). The minimum distance from the surface facilities in the geologic repository operations area to the boundary of the preclosure controlled area must be at least 100 meters.
              (c) The preclosure controlled area may be traversed by a highway, railroad, or waterway, so long as appropriate and effective arrangements are made to control traffic and to protect public health and safety.
              [61 FR 64270, Dec. 4, 1996]

            Performance Confirmation Requirements

              § 60.137
              General requirements for performance confirmation.
              The geologic repository operations area shall be designed so as to permit implementation of a performance confirmation program that meets the requirements of subpart F of this part.

          Subpart F—Performance Confirmation Program

            Source:
            48 FR 28228, June 21, 1983, unless otherwise noted.

            § 60.140
            General requirements.
            (a) The performance confirmation program shall provide data which indicates, where practicable, whether:
            (1) Actual subsurface conditions encountered and changes in those conditions during construction and waste emplacement operations are within the limits assumed in the licensing review; and
            (2) Natural and engineered systems and components required for repository operation, or which are designed or assumed to operate as barriers after permanent closure, are functioning as intended and anticipated.
            (b) The program shall have been started during site characterization and it will continue until permanent closure.
            (c) The program shall include in situ monitoring, laboratory and field testing, and in situ experiments, as may be appropriate to accomplish the objective as stated above.
            (d) The program shall be implemented so that:
            (1) It does not adversely affect the ability of the natural and engineered elements of the geologic repository to meet the performance objectives.
            (2) It provides baseline information and analysis of that information on those parameters and natural processes pertaining to the geologic setting that may be changed by site characterization, construction, and operational activities.
            (3) It monitors and analyzes changes from the baseline condition of parameters that could affect the performance of a geologic repository.
            (4) It provides an established plan for feedback and analysis of data, and implementation of appropriate action.

            § 60.141
            Confirmation of geotechnical and design parameters.
            (a) During repository construction and operation, a continuing program of surveillance, measurement, testing, and geologic mapping shall be conducted to ensure that geotechnical and design parameters are confirmed and to ensure that appropriate action is taken to inform the Commission of changes needed in design to accommodate actual field conditions encountered.
            (b) Subsurface conditions shall be monitored and evaluated against design assumptions.
            (c) As a minimum, measurements shall be made of rock deformations and displacement, changes in rock stress and strain, rate and location of water inflow into subsurface areas, changes in groundwater conditions, rock pore water pressures including those along fractures and joints, and the thermal and thermomechanical response of the rock mass as a result of development and operations of the geologic repository.

            (d) These measurements and observations shall be compared with the original design bases and assumptions. If significant differences exist between the measurements and observations and the original design bases and assumptions, the need for modifications to the design or in construction methods shall be determined and these differences and the recommended changes reported to the Commission.

            (e) In situ monitoring of the thermomechanical response of the underground facility shall be conducted until permanent closure to ensure that the performance of the natural and engineering features are within design limits.

            § 60.142
            Design testing.
            (a) During the early or developmental stages of construction, a program for in situ testing of such features as borehole and shaft seals, backfill, and the thermal interaction effects of the waste packages, backfill, rock, and groundwater shall be conducted.
            (b) The testing shall be initiated as early as is practicable.
            (c) A backfill test section shall be constructed to test the effectiveness of backfill placement and compaction procedures against design requirements before permanent backfill placement is begun.
            (d) Test sections shall be established to test the effectiveness of borehole and shaft seals before full-scale operation proceeds to seal boreholes and shafts.

            § 60.143
            Monitoring and testing waste packages.
            (a) A program shall be established at the geologic repository operations area for monitoring the condition of the waste packages. Waste packages chosen for the program shall be representative of those to be emplaced in the underground facility.
            (b) Consistent with safe operation at the geologic repository operations area, the environment of the waste packages selected for the waste package monitoring program shall be representative of the environment in which the wastes are to be emplaced.
            (c) The waste package monitoring program shall include laboratory experiments which focus on the internal condition of the waste packages. To the extent practical, the environment experienced by the emplaced waste packages within the underground facility during the waste package monitoring program shall be duplicated in the laboratory experiments.
            (d) The waste package monitoring program shall continue as long as practical up to the time of permanent closure.

          Subpart G—Quality Assurance

            Source:
            48 FR 28228, June 21, 1983, unless otherwise noted.

            § 60.150
            Scope.
            As used in this part, quality assurance comprises all those planned and systematic actions necessary to provide adequate confidence that the geologic repository and its subsystems or components will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to the physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system to predetermined requirements.

            § 60.151
            Applicability.
            The quality assurance program applies to all systems, structures and components important to safety, to design and characterization of barriers important to waste isolation and to activities related thereto. These activities include: site characterization, facility and equipment construction, facility operation, performance confirmation, permanent closure, and decontamination and dismantling of surface facilities.

            § 60.152
            Implementation.
            DOE shall implement a quality assurance program based on the criteria of appendix B of 10 CFR part 50 as applicable, and appropriately supplemented by additional criteria as required by § 60.151.

          Subpart H—Training and Certification of Personnel

            Source:
            48 FR 28229, June 21, 1983, unless otherwise noted.

            § 60.160
            General requirements.

            Operations of systems and components that have been identified as important to safety in the Safety Analysis Report and in the license shall be performed only by trained and certified personnel or by personnel under the direct visual supervision of an individual with training and certification in such operation. Supervisory personnel who direct operations that are important to safety must also be certified in such operations.

            § 60.161
            Training and certification program.
            DOE shall establish a program for training, proficiency testing, certification and requalification of operating and supervisory personnel.

            § 60.162
            Physical requirements.
            The physical condition and the general health of personnel certified for operations that are important to safety shall not be such as might cause operational errors that could endanger the public health and safety. Any condition which might cause impaired judgment or motor coordination must be considered in the selection of personnel for activities that are important to safety. These conditions need not categorically disqualify a person, so long as appropriate provisions are made to accommodate such conditions.

          Subpart I—Emergency Planning Criteria [Reserved]

          Subpart J—Violations

            § 60.181
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of—
            (1) The Atomic Energy Act of 1954, as amended;
            (2) Title II of the Energy Reorganization Act of 1974, as amended; or
            (3) A regulation or order issued pursuant to those Acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:
            (1) For violations of—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section;
            (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.
            [57 FR 55076, Nov. 24, 1992]

            § 60.183
            Criminal penalties.
            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 60 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in part 60 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 60.1, 60.2, 60.3, 60.5, 60.6, 60.7, 60.8, 60.15, 60.16, 60.17, 60.18, 60.21, 60.22, 60.23, 60.24, 60.31, 60.32, 60.33, 60.41, 60.42, 60.43, 60.44, 60.45, 60.46, 60.51, 60.52, 60.61, 60.62, 60.63, 60.64, 60.65, 60.101, 60.102, 60.111, 60.112, 60.113, 60.121, 60.122, 60.130, 60.131, 60.132, 60.133, 60.134, 60.135, 60.137, 60.140, 60.141, 60.142, 60.143, 60.150, 60.151, 60.152, 60.162, 60.181, and 60.183.
            [57 FR 55076, Nov. 24, 1992]

        Pt. 61
        PART 61—LICENSING REQUIREMENTS FOR LAND DISPOSAL OF RADIOACTIVE WASTE

            Subpart A—General Provisions
            Sec.
            61.1
            Purpose and scope.
            61.2
            Definitions.
            61.3
            License required.
            61.4
            Communications.
            61.5
            Interpretations.
            61.6
            Exemptions.
            61.7
            Concepts.
            61.8
            Information collection requirements: OMB approval.
            61.9
            Employee protection.
            61.9a
            Completeness and accuracy of information.

            61.9b
            Deliberate misconduct.

            Subpart B—Licenses
            61.10
            Content of application.
            61.11
            General information.
            61.12
            Specific technical information.
            61.13
            Technical analyses.
            61.14
            Institutional information.
            61.15
            Financial information.
            61.16
            Other information.
            61.20
            Filing and distribution of application.
            61.21
            Elimination of repetition.
            61.22
            Updating of application.
            61.23
            Standards for issuance of a license.
            61.24
            Conditions of licenses.
            61.25
            Changes.
            61.26
            Amendment of license.
            61.27
            Application for renewal or closure.
            61.28
            Contents of application for closure.
            61.29
            Post-closure observation and maintenance.
            61.30
            Transfer of license.
            61.31
            Termination of license.

              US/IAEA Safeguards Agreement
              61.32
              Facility information and verification.

            Subpart C—Performance Objectives
            61.40
            General requirement.
            61.41
            Protection of the general population from releases of radioactivity.
            61.42
            Protection of individuals from inadvertent intrusion.
            61.43
            Protection of individuals during operations.
            61.44
            Stability of the disposal site after closure.

            Subpart D—Technical Requirements for Land Disposal Facilities
            61.50
            Disposal site suitability requirements for land disposal.
            61.51
            Disposal site design for land disposal.
            61.52
            Land disposal facility operation and disposal site closure.
            61.53
            Environmental monitoring.
            61.54
            Alternative requirements for design and operations.
            61.55
            Waste classification.
            61.56
            Waste characteristics.
            61.57
            Labeling.
            61.58
            Alternative requirements for waste classification and characteristics.
            61.59
            Institutional requirements.

            Subpart E—Financial Assurances
            61.61
            Applicant qualifications and assurances.
            61.62
            Funding for disposal site closure and stabilization.
            61.63
            Financial assurances for institutional controls.

            Subpart F—Participation by State Governments and Indian Tribes
            61.70
            Scope.
            61.71
            State and Tribal government consultation.
            61.72
            Filing of proposals for State and Tribal participation.
            61.73
            Commission approval of proposals.

            Subpart G—Records, Reports, Tests, and Inspections
            61.80
            Maintenance of records, reports, and transfers.
            61.81
            Tests at land disposal facilities.
            61.82
            Commission inspections of land disposal facilities.
            61.83
            Violations.
            61.84
            Criminal penalties.

          Authority:
          Atomic Energy Act of 1954, secs. 53, 57, 62, 63, 65, 81, 161, 181, 182, 183, 223, 234 (42 U.S.C. 2073, 2077, 2092, 2093, 2095, 2111, 2201, 2231, 2232, 2233, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 206, 211 (42 U.S.C. 5841, 5846, 5851); Low-Level Radioactive Waste Policy Amendments Act of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 3504 note.

          Source:
          47 FR 57463, Dec. 27, 1982, unless otherwise noted.

          Editorial Note:
          Nomenclature changes to part 61 appear at 80 FR 74980, Dec. 1, 2015, and at 81 FR 86909, Dec. 2, 2016.

          Subpart A—General Provisions

            § 61.1
            Purpose and scope.
            (a) The regulations in this part establish, for land disposal of radioactive waste, the procedures, criteria, and terms and conditions upon which the Commission issues licenses for the disposal of radioactive wastes containing byproduct, source and special nuclear material received from other persons. Disposal of waste by an individual licensee is set forth in part 20 of this chapter. Applicability of the requirements in this part to Commission licenses for waste disposal facilities in effect on the effective date of this rule will be determined on a case-by-case basis and implemented through terms and conditions of the license or by orders issued by the Commission.

            (b) Except as provided in part 150 of this chapter, which addresses assumption of certain regulatory authority by Agreement States, and § 61.6 “Exemptions,” the regulations in this part apply to all persons in the United States. The regulations in this part do not apply to—
            (1) Disposal of high-level waste as provided for in part 60 or 63 of this chapter;
            (2) Disposal of uranium or thorium tailings or wastes (byproduct material as defined in § 40.4 (a-1) as provided for in part 40 of this chapter in quantities greater than 10,000 kilograms and containing more than 5 millicuries of radium-226; or
            (3) Disposal of licensed material as provided for in part 20 of this chapter.
            (c) This part also gives notice to all persons who knowingly provide to any licensee, applicant, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licensee's or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 61.9b.
            [47 FR 57463, Dec. 27, 1982, as amended at 56 FR 40690, Aug. 15, 1991; 63 FR 1898, Jan. 13, 1998; 66 FR 55791, Nov. 2, 2001]

            § 61.2
            Definitions.
            As used in this part:

              Active maintenance means any significant remedial activity needed during the period of institutional control to maintain a reasonable assurance that the performance objectives in §§ 61.41 and 61.42 are met. Such active maintenance includes ongoing activities such as the pumping and treatment of water from a disposal unit or one-time measures such as replacement of a disposal unit cover. Active maintenance does not include custodial activities such as repair of fencing, repair or replacement of monitoring equipment, revegetation, minor additions to soil cover, minor repair of disposal unit covers, and general disposal site upkeep such as mowing grass.

              Buffer zone is a portion of the disposal site that is controlled by the licensee and that lies under the disposal units and between the disposal units and the boundary of the site.

              Chelating agent means amine polycarboxylic acids (e.g., EDTA, DTPA), hydroxy-carboxylic acids, and polycarboxylic acids (e.g., citric acid, carbolic acid, and glucinic acid).

              Commencement of construction means any clearing of land, excavation, or other substantial action that would adversely affect the environment of a land disposal facility. The term does not mean disposal site exploration, necessary roads for disposal site exploration, borings to determine foundation conditions, or other preconstruction monitoring or testing to establish background information related to the suitability of the disposal site or the protection of environmental values.

              Commission means the Nuclear Regulatory Commission or its duly authorized representatives.

              Custodial Agency means an agency of the government designated to act on behalf of the government owner of the disposal site.

              Director means the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission.

              Disposal means the isolation of radioactive wastes from the biosphere inhabited by man and containing his food chains by emplacement in a land disposal facility.

              Disposal site means that portion of a land disposal facility which is used for disposal of waste. It consists of disposal units and a buffer zone.

              Disposal unit means a discrete portion of the disposal site into which waste is placed for disposal. For near-surface disposal the unit is usually a trench.

              Engineered barrier means a man-made structure or device that is intended to improve the land disposal facility's ability to meet the performance objectives in subpart C.

              Explosive material means any chemical compound, mixture, or device, which produces a substantial instantaneous release of gas and heat spontaneously or by contact with sparks or flame.

              Government agency means any executive department, commission, independent establishment, or corporation, wholly or partly owned by the United States of America which is an instrumentality of the United States; or any board, bureau, division, service, office, officer, authority, administration, or other establishment in the executive branch of the government.

              Hazardous waste means those wastes designated as hazardous by Environmental Protection Agency regulations in 40 CFR part 261.

              Hydrogeologic unit means any soil or rock unit or zone which by virtue of its porosity or permeability, or lack thereof, has a distinct influence on the storage or movement of groundwater.

              Inadvertent intruder means a person who might occupy the disposal site after closure and engage in normal activities, such as agriculture, dwelling construction, or other pursuits in which the person might be unknowingly exposed to radiation from the waste.

              Indian Tribe means an Indian Tribe as defined in the Indian Self-Determination and Education Assistance Act (25 U.S.C. 5301).

              Intruder barrier means a sufficient depth of cover over the waste that inhibits contact with waste and helps to ensure that radiation exposures to an inadvertent intruder will meet the performance objectives set forth in this part, or engineered structures that provide equivalent protection to the inadvertent intruder.

              Land disposal facility means the land, building, and structures, and equipment which are intended to be used for the disposal of radioactive wastes. For purposes of this chapter, a “geologic repository” as defined in part 60 or 63 is not considered a land disposal facility.

              License means a license issued under the regulations in part 61 of this chapter. Licensee means the holder of such a license.

              Monitoring means observing and making measurements to provide data to evaluate the performance and characteristics of the disposal site.

              Near-surface disposal facility means a land disposal facility in which radioactive waste is disposed of in or within the upper 30 meters of the earth's surface.

              Person means (1) any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, government agency other than the Commission or the Department of Energy (except that the Department of Energy is considered a person within the meaning of the regulations in this part to the extent that its facilities and activities are subject to the licensing and related regulatory authority of the Commission pursuant to law), any State or any political subdivision of or any political entity within a State, any foreign government or nation or any political subdivision of any such government or nation, or other entity; and (2) any legal successor, representative, agent, or agency of the foregoing.

              Pyrophoric liquid means any liquid that ignites spontaneously in dry or moist air at or below 130 °F (54.5 °C). A pyrophoric solid is any solid material, other than one classed as an explosive, which under normal conditions is liable to cause fires through friction, retained heat from manufacturing or processing, or which can be ignited readily and when ignited burns so vigorously and persistently as to create a serious transportation, handling, or disposal hazard. Included are spontaneously combustible and water-reactive materials.

              Site closure and stablization means those actions that are taken upon completion of operations that prepare the disposal site for custodial care and that assure that the disposal site will remain stable and will not need ongoing active maintenance.

              State means any State, Territory, or possession of the United States, Puerto Rico, and the District of Columbia.

              Stability means structural stabillity.

              Surveillance means observation of the disposal site for purposes of visual detection of need for maintenance, custodial care, evidence of intrusion, and compliance with other license and regulatory requirements.

              Tribal Governing Body means a Tribal organization as defined in the Indian Self-Determination and Education Assistance Act (25 U.S.C. 5301).

              Waste means those low-level radioactive wastes containing source, special nuclear, or byproduct material that are acceptable for disposal in a land disposal facility. For the purposes of this definition, low-level radioactive waste means radioactive waste not classified as high-level radioactive waste, transuranic waste, spent nuclear fuel, or byproduct material as defined in paragraphs (2), (3), and (4) of the definition of Byproduct material set forth in § 20.1003 of this chapter.
            [47 FR 57463, Dec. 27, 1982, as amended at 54 FR 22583, May 25, 1989; 58 FR 33891, June 22, 1993; 66 FR 55792, Nov. 2, 2001; 72 FR 55933, Oct. 1, 2007; 73 FR 5725, Jan. 31, 2008; 79 FR 75740, Dec. 19, 2014; 82 FR 52825, Nov. 15, 2017]

            § 61.3
            License required.
            (a) No person may receive, possess, and dispose of radioactive waste containing source, special nuclear, or byproduct material at a land disposal facility unless authorized by a license issued by the Commission pursuant to this part, or unless exemption has been granted by the Commission under § 61.6 of this part.
            (b) Each person shall file an application with the Commission and obtain a license as provided in this part before commencing construction of a land disposal facility. Failure to comply with this requirement may be grounds for denial of a license.

            § 61.4
            Communications.

            Except where otherwise specified, all communications and reports concerning the regulations in this part and applications filed under them should be sent by mail addressed: ATTN: Document Control Desk; Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand delivery to the NRC's Offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.
            [73 FR 5725, Jan. 31, 2008, as amended at 74 FR 62683, Dec. 1, 2009; 79 FR 75740, Dec. 19, 2014; 80 FR 74980, Dec. 1, 2015]

            § 61.5
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be considered binding upon the Commission.

            § 61.6
            Exemptions.
            The Commission may, upon application by any interested person, or upon its own initiative, grant any exemption from the requirements of the regulations in this part as it determines is authorized by law, will not endanger life or property or the common defense and security, and is otherwise in the public interest.

            § 61.7
            Concepts.
            (a) The disposal facility. (1) Part 61 is intended to apply to land disposal of radioactive waste and not to other methods such as sea or extraterrestrial disposal. Part 61 contains procedural requirements and performance objectives applicable to any method of land disposal. It contains specific technical requirements for near-surface disposal of radioactive waste, a subset of land disposal, which involves disposal in the uppermost portion of the earth, approximately 30 meters. Near-surface disposal includes disposal in engineered facilities which may be built totally or partially above-grade provided that such facilities have protective earthen covers. Near-surface disposal does not include disposal facilities which are partially or fully above-grade with no protective earthen cover, which are referred to as “above-ground disposal.” Burial deeper than 30 meters may also be satisfactory. Technical requirements for alternative methods may be added in the future.

            (2) Near-surface disposal of radioactive waste takes place at a near-surface disposal facility, which includes all of the land and buildings necessary to carry out the disposal. The disposal site is that portion of the facility which is used for disposal of waste and consists of disposal units and a buffer zone. A disposal unit is a discrete portion of the disposal site into which waste is placed for disposal. For near-surface disposal, the disposal unit is usually a trench. A buffer zone is a portion of the disposal site that is controlled by the licensee and that lies under the site and between the boundary of the disposal site and any disposal unit. It provides controlled space to establish monitoring locations which are intended to provide an early warning of radionuclide movement, and to take mitigative measures if needed. In choosing a disposal site, site characteristics should be considered in terms of the indefinite future and evaluated for at least a 500-year timeframe.
            (b) Waste classification and near-surface disposal. (1) Disposal of radioactive waste in near-surface disposal facilities has the following safety objectives: protection of the general population from releases of radioactivity, protection of individuals from inadvertent intrusion, and protection of individuals during operations. A fourth objective is to ensure stability of the site after closure.

            (2) A cornerstone of the system is stability—stability of the waste and the disposal site so that once emplaced and covered, the access of water to the waste can be minimized. Migration of radionuclides is thus minimized, long-term active maintenance can be avoided, and potential exposures to intruders reduced. While stability is a desirable characteristic for all waste much radioactive waste does not contain sufficient amounts of radionuclides to be of great concern from these standpoints; this waste, however, tends to be unstable, such as ordinary trash type wastes. If mixed with the higher activity waste, their deterioration could lead to failure of the system and permit water to penetrate the disposal unit and cause problems with the higher activity waste. Therefore, in order to avoid placing requirements for a stable waste form on relatively innocuous waste, these wastes have been classed as Class A waste. The Class A waste will be disposed of in separate disposal units at the disposal site. However, Class A waste that is stable may be mixed with other classes of waste. Those higher activity wastes that should be stable for proper disposal are classed as Class B and C waste. To the extent that it is practicable, Class B and C waste forms or containers should be designed to be stable, i.e., maintain gross physical properties and identity, over 300 years. For certain radionuclides prone to migration, a maximum disposal site inventory based on the characteristics of the disposal site may be established to limit potential exposure.
            (3) It is possible but unlikely that persons might occupy the site in the future and engage in normal pursuits without knowing that they were receiving radiation exposure. These persons are referred to as inadvertent intruders. Protection of such intruders can involve two principal controls: institutional control over the site after operations by the site owner to ensure that no such occupation or improper use of the site occurs; or, designating which waste could present an unacceptable risk to an intruder, and disposing of this waste in a manner that provides some form of intruder barrier that is intended to prevent contact with the waste. This regulation incorporates both types of protective controls.
            (4) Institutional control of access to the site is required for up to 100 years. This permits the disposal of Class A and Class B waste without special provisions for intrusion protection, since these classes of waste contain types and quantities of radioisotopes that will decay during the 100-year period and will present an acceptable hazard to an intruder. The government landowner administering the active institutional control program has flexibility in controlling site access which may include allowing productive uses of the land provided the integrity and long-term performance of the site are not affected.

            (5) Waste that will not decay to levels which present an acceptable hazard to an intruder within 100 years is designated as Class C waste. This waste is disposed of at a greater depth than the other classes of waste so that subsequent surface activities by an intruder will not disturb the waste. Where site conditions prevent deeper disposal, intruder barriers such as concrete covers may be used. The effective life of these intruder barriers should be 500 years. A maximum concentration of radionuclides is specified for all wastes so that at the end of the 500 year period, remaining radioactivity will be at a level that does not pose an unacceptable hazard to an intruder or public health and safety. Waste with concentrations above these limits is generally unacceptable for near-surface disposal. There may be some instances where waste with concentrations greater than permitted for Class C would be acceptable for near-surface disposal with special processing or design. These will be evaluated on a case-by-case basis. Class C waste must also be stable.
            (c) The licensing process. (1) During the preoperational phase, the potential applicant goes through a process of disposal site selection by selecting a region of interest, examining a number of possible disposal sites within the area of interest and narrowing the choice to the proposed site. Through a detailed investigation of the disposal site characteristics the potential applicant obtains data on which to base an analysis of the disposal site's suitability. Along with these data and analyses, the applicant submits other more general information to the Commission in the form of an application for a license for land disposal. The Commission's review of the application is in accordance with administrative procedures established by rule and may involve participation by affected State governments or Indian Tribes. While the proposed disposal site must be owned by a State or the Federal government before the Commission will issue a license, it may be privately owned during the preoperational phase if suitable arrangements have been made with a State or the Federal government to take ownership in fee of the land before the license is issued.
            (2) During the operational phase, the licensee carries out disposal activities in accordance with the requirements of this regulation and any conditions on the license. Periodically, the authority to conduct the above ground operations and dispose of waste will be subject to a license renewal, at which time the operating history will be reviewed and a decision made to permit or deny continued operation. When disposal operations are to cease, the licensee applies for an amendment to his license to permit site closure. After final review of the licensee's site closure and stabilization plan, the Commission may approve the final activities necessary to prepare the disposal site so that ongoing active maintenance of the site is not required during the period of institutional control.
            (3) During the period when the final site closure and stabilization activities are being carried out, the licensee is in a disposal site closure phase. Following that, for a period of 5 years, the licensee must remain at the disposal site for a period of post-closure observation and maintenance to assure that the disposal site is stable and ready for institutional control. The Commission may approve shorter or require longer periods if conditions warrant. At the end of this period, the licensee applies for a license transfer to the disposal site owner.
            (4) After a finding of satisfactory disposal site closure, the Commission will transfer the license to the State or Federal government that owns the disposal site. If the Department of Energy is the Federal agency administering the land on bahalf of the Federal government the license will be terminated because the Commission lacks regulatory authority over the Department for this activity. Under the conditions of the transferred license, the owner will carry out a program of monitoring to assure continued satisfactory disposal site performance, physical surveillance to restrict access to the site and carry out minor custodial activities. During this period, productive uses of the land might be permitted if those uses do not affect the stability of the site and its ability to meet the performance objectives. At the end of the prescribed period of institutional control, the license will be terminated by the Commission.
            [47 FR 57463, Dec. 27, 1982, as amended at 58 FR 33891, June 22, 1993]

            § 61.8
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0135.
            (b) The approved information collection requirements contained in this Part appear in §§ 61.3, 61.6, 61.9, 61.10, 61.11, 61.12, 61.13, 61.14, 61.15, 61.16, 61.20, 61.22, 61.24, 61.26, 61.27, 61.28, 61.30, 61.31, 61.32, 61.53, 61.55, 61.57, 61.58, 61.61, 61.62, 61.63, 61.72, and 61.80.
            (c) In § 61.32, Form N-71 and associated forms are approved under control number 3150-0056, and DOC/NRC Forms AP-1, AP-A, and associated forms are approved under control numbers 0694-0135.
            [58 FR 33891, June 22, 1993, as amended at 62 FR 52188, Oct. 6, 1997; 73 FR 78606, Dec. 23, 2008]

            § 61.9
            Employee protection.
            (a) Discrimination by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act or the Energy Reorganization Act.
            (1) The protected activities include but are not limited to:
            (i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) introductory text of the section or possible violations of requirements imposed under either of those statutes;
            (ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) introductory text or under these requirements if the employee has identified the alleged illegality to the employer;
            (iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;
            (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) introductory text.
            (v) Assisting or participating in, or is about to assist or participate in, these activities.
            (2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.
            (3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.
            (b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.

            (c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant may be grounds for—

            (1) Denial, revocation, or suspension of the license.
            (2) Imposition of a civil penalty on the licensee, applicant, or a contractor or subcontractor of the licensee or applicant.
            (3) Other enforcement action.
            (d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.
            (e)(1) Each licensee and each applicant for a license shall prominently post the revision of NRC Form 3, “Notice to Employees,” referenced in 10 CFR 19.11(c). This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, and for 30 days following license termination.

            (2) Copies of NRC Form 3 can be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, via email to Forms.Resource@nrc.gov, or by visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.

            (f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.
            [58 FR 52412, Oct. 8, 1993, as amended at 60 FR 24552, May 9, 1995; 61 FR 6765, Feb. 22, 1996; 68 FR 58814, Oct. 10, 2003; 72 FR 63974, Nov. 14, 2007; 73 FR 30459, May 28, 2008; 79 FR 66605, Nov. 10, 2014]

            § 61.9a
            Completeness and accuracy of information.
            (a) Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.
            (b) Each applicant or licensee shall notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant or licensee violates this paragraph only if the applicant or licensee fails to notify the Commission of information that the applicant or licensee has identified as having a significant implication for public health and safety or common defense and security. Notification shall be provided to the Administrator of the appropriate Regional Office within two working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.
            [52 FR 49372, Dec. 31, 1987]

            § 61.9b
            Deliberate misconduct.
            (a) Any licensee, applicant for a license, employee of a licensee or applicant; or any contractor (including a supplier or consultant), subcontractor, employee of a contractor or subcontractor of any licensee or applicant for a license, who knowingly provides to any licensee, applicant, contractor, or subcontractor, any components, equipment, materials, or other goods or services that relate to a licensee's or applicant's activities in this part, may not:

            (1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission; or
            (2) Deliberately submit to the NRC, a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.
            (b) A person who violates paragraph (a)(1) or (a)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.
            (c) For the purposes of paragraph (a)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:
            (1) Would cause a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license issued by the Commission; or
            (2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor.
            [63 FR 1898, Jan. 13, 1998]

          Subpart B—Licenses

            § 61.10
            Content of application.
            An application to receive from others, possess and dispose of wastes containing or contaminated with source, byproduct or special nuclear material by land disposal must consist of general information, specific technical information, institutional information, and financial information as set forth in §§ 61.11 through 61.16. An environmental report prepared in accordance with subpart A of part 51 of this chapter must accompany the application.
            [49 FR 9405, Mar. 12, 1984]

            § 61.11
            General information.
            The general information must include each of the following:
            (a) Identity of the applicant including:
            (1) The full name, address, telephone number and description of the business or occupation of the applicant;
            (2) If the applicant is a partnership, the name, and address of each partner and the principal location where the partnership does business;
            (3) If the applicant is a corporation or an unincorporated association, (i) the state where it is incorporated or organized and the principal location where it does business, and (ii) the names and addresses of its directors and principal officers; and
            (4) If the applicant is acting as an agent or representative of another person in filing the application, all information required under this paragraph must be supplied with respect to the other person.
            (b) Qualifications of the applicant:
            (1) The organizational structure of the applicant, both offsite and onsite, including a description of lines of authority and assignments of responsibilities, whether in the form of administrative directives, contract provisions, or otherwise;
            (2) The technical qualifications, including training and experience, of the applicant and members of the applicant's staff to engage in the proposed activities. Minimum training and experience requirements for personnel filling key positions described in paragraph (b)(1) of this section must be provided;
            (3) A description of the applicant's personnel training program; and
            (4) The plan to maintain an adequate complement of trained personnel to carry out waste receipt, handling, and disposal operations in a safe manner.
            (c) A description of:
            (1) The location of the proposed disposal site;
            (2) The general character of the proposed activities;
            (3) The types and quantities of radioactive waste to be received, possessed, and disposed of;
            (4) Plans for use of the land disposal facility for purposes other than disposal of radioactive wastes; and
            (5) The proposed facilities and equipment.

            (d) Proposed schedules for construction, receipt of waste, and first emplacement of waste at the proposed land disposal facility.

            § 61.12
            Specific technical information.
            The specific technical information must include the following information needed for demonstration that the performance objectives of subpart C of this part and the applicable technical requirements of subpart D of this part will be met:
            (a) A description of the natural and demographic disposal site characteristics as determined by disposal site selection and characterization activities. The description must include geologic, geotechnical, hydrologic, meteorologic, climatologic, and biotic features of the disposal site and vicinity.
            (b) A description of the design features of the land disposal facility and the disposal units. For near-surface disposal, the description must include those design features related to infiltration of water; integrity of covers for disposal units; structural stability of backfill, wastes, and covers; contact of wastes with standing water; disposal site drainage; disposal site closure and stabilization; elimination to the extent practicable of long-term disposal site maintenance; inadvertent intrusion; occupational exposures; disposal site monitoring; and adequacy of the size of the buffer zone for monitoring and potential mitigative measures.
            (c) A description of the principal design criteria and their relationship to the performance objectives.
            (d) A description of the design basis natural events or phenomena and their relationship to the principal design criteria.
            (e) A description of codes and standards which the applicant has applied to the design and which will apply to construction of the land disposal facilities.
            (f) A description of the construction and operation of the land disposal facility. The description must include as a minimum the methods of construction of disposal units; waste emplacement; the procedures for and areas of waste segregation; types of intruder barriers; onsite traffic and drainage systems; survey control program; methods and areas of waste storage; and methods to control surface water and groundwater access to the wastes. The description must also include a description of the methods to be employed in the handling and disposal of wastes containing chelating agents or other non-radiological substances that might affect meeting the performance objectives in subpart C of this part.
            (g) A description of the disposal site closure plan, including those design features which are intended to facilitate disposal site closure and to eliminate the need for ongoing active maintenance.
            (h) An identification of the known natural resources at the disposal site, the exploitation of which could result in inadvertent intrusion into the low-level wastes after removal of active institutional control.
            (i) A description of the kind, amount, classification and specifications of the radioactive material proposed to be received, possessed, and disposed of at the land disposal facility.
            (j) A description of the quality assurance program, tailored to LLW disposal, developed and applied by the applicant for the determination of natural disposal site characteristics and for quality assurance during the design, construction, operation, and closure of the land disposal facility and the receipt, handling, and emplacement of waste.
            (k) A description of the radiation safety program for control and monitoring of radioactive effluents to ensure compliance with the performance objective in § 61.41 of this part and occupational radiation exposure to ensure compliance with the requirements of part 20 of this chapter and to control contamination of personnel, vehicles, equipment, buildings, and the disposal site. Both routine operations and accidents must be addressed. The program description must include procedures, instrumentation, facilities, and equipment.
            (l) A description of the environmental monitoring program to provide data to evaluate potential health and environmental impacts and the plan for taking corrective measures if migration of radionuclides is indicated.

            (m) A description of the administrative procedures that the applicant will apply to control activities at the land disposal facility.
            (n) A description of the facility electronic recordkeeping system as required in § 61.80.
            [47 FR 57463, Dec. 27, 1982, as amended at 58 FR 33891, June 22, 1993; 60 FR 15666, Mar. 27, 1995]

            § 61.13
            Technical analyses.
            The specific technical information must also include the following analyses needed to demonstrate that the performance objectives of subpart C of this part will be met:
            (a) Pathways analyzed in demonstrating protection of the general population from releases of radioactivity must include air, soil, groundwater, surface water, plant uptake, and exhumation by burrowing animals. The analyses must clearly identify and differentiate between the roles performed by the natural disposal site characteristics and design features in isolating and segregating the wastes. The analyses must clearly demonstrate that there is reasonable assurance that the exposure to humans from the release of radioactivity will not exceed the limits set forth in § 61.41.
            (b) Analyses of the protection of individuals from inadvertent intrusion must include demonstration that there is reasonable assurance the waste classification and segregation requirements will be met and that adequate barriers to inadvertent intrusion will be provided.
            (c) Analyses of the protection of individuals during operations must include assessments of expected exposures due to routine operations and likely accidents during handling, storage, and disposal of waste. The analyses must provide reasonable assurance that exposures will be controlled to meet the requirements of part 20 of this chapter.
            (d) Analyses of the long-term stability of the disposal site and the need for ongoing active maintenance after closure must be based upon analyses of active natural processes such as erosion, mass wasting, slope failure, settlement of wastes and backfill, infiltration through covers over disposal areas and adjacent soils, and surface drainage of the disposal site. The analyses must provide reasonable assurance that there will not be a need for ongoing active maintenance of the disposal site following closure.

            § 61.14
            Institutional information.
            The institutional information must include:
            (a) A certification by the Federal or State government which owns the disposal site that the Federal or State government is prepared to accept transfer of the license when the provisions of § 61.30 are met, and will assume responsibility for custodial care after site closure and postclosure observation and maintenance.
            (b) Where the proposed disposal site is on land not owned by the Federal or a State government, the applicant must submit evidence that arrangements have been made for assumption of ownership in fee by the Federal or a State government before the Commission issues a license.

            § 61.15
            Financial information.
            The financial information must be sufficient to demonstrate that the financial qualifications of the applicant are adequate to carry out the activities for which the license is sought and meet other financial assurance requirements as specified in subpart E of this part.

            § 61.16
            Other information.
            Depending upon the nature of the wastes to be disposed of, and the design and proposed operation of the land disposal facility, additional information may be requested by the Commission including the following:
            (a) Physical security measures, if appropriate. Any application to receive and possess special nuclear material in quantities subject to the requirements of part 73 of this chapter shall demonstrate how the physical security requirements of part 73 will be met. In determining whether receipt and possession will be subject to the requirements of part 73, the applicant shall not consider the quantity of special nuclear material that has been disposed of.
            (b) Safety information concerning criticality, if appropriate. (1) Any application to receive and possess special nuclear material in quantities that would be subject to the requirements of § 70.24, “Criticality accident requirements” of part 70 of this chapter shall demonstrate how the requirements of that section will be met, unless the applicant requests an exemption pursuant to § 70.24(d). In determining whether receipt and possession would be subject to the requirements of § 70.24, the applicant shall not consider the quantity of special nuclear material that has been disposed of.
            (2) Any application to receive and possess special nuclear material shall describe proposed procedures for avoiding accidental criticality, which address both storage of special nuclear material prior to disposal and waste emplacement for disposal.

            § 61.20
            Filing and distribution of application.
            (a) An application for a license under this part, and any amendments thereto, must be filed with the Director, must be signed by the applicant or the applicant's authorized representative under oath or affirmation, and, if the document is in paper form, must be the signed original.
            (b) The applicant shall maintain the capability to generate additional copies of the application for distribution in accordance with written instructions from the Director or the Director's designee.
            (c) Fees. Application, amendment, and inspection fees applicable to a license covering the receipt and disposal of radioactive wastes in a land disposal facility are required by part 170 of this chapter.
            [47 FR 57463, Dec. 27, 1982, as amended at 49 FR 9405, Mar. 12, 1984; 68 FR 58814, Oct. 10, 2003]

            § 61.21
            Elimination of repetition.
            In its application, the applicant may incorporate by reference information contained in previous applications, statements, or reports filed with the Commission if these references are clear and specific.
            [49 FR 9405, Mar. 12, 1984]

            § 61.22
            Updating of application.
            (a) The application must be as complete as possible in the light of information that is available at the time of submittal.
            (b) The applicant shall supplement its application in a timely manner, as necessary, to permit the Commission to review, prior to issuance of a license, any changes in the activities proposed to be carried out or new information regarding the proposed activities.
            [49 FR 9405, Mar. 12, 1984]

            § 61.23
            Standards for issuance of a license.
            A license for the receipt, possession, and disposal of waste containing or contaminated with source, special nuclear, or byproduct material will be issued by the Commission upon finding that the issuance of the license will not be inimical to the common defense and security and will not constitute an unreasonable risk to the health and safety of the public, and:
            (a) The applicant is qualified by reason of training and experience to carry out the disposal operations requested in a manner that protects health and minimizes danger to life or property.
            (b) The applicant's proposed disposal site, disposal design, land disposal facility operations (including equipment, facilities, and procedures), disposal site closure, and postclosure institutional control are adequate to protect the public health and safety in that they provide reasonable assurance that the general population will be protected from releases of radioactivity as specified in the performance objective in § 61.41, Protection of the general population from releases of radioactivity.
            (c) The applicant's proposed disposal site, disposal site design, land disposal facility operations (including equipment, facilities, and procedures), disposal site closure, and postclosure institutional control are adequate to protect the public health and safety in that they will provide reasonable assurance that individual inadvertent intruders are protected in accordance with the performance objective in § 61.42, Protection of individuals from inadvertent intrusion.

            (d) The applicant's proposed land disposal facility operations, including equipment, facilities, and procedures, are adequate to protect the public health and safety in that they will provide reasonable assurance that the standards for radiation protection set out in part 20 of this chapter will be met.
            (e) The applicant's proposed disposal site, disposal site design, land disposal facility operations, disposal site closure, and postclosure institutional control are adequate to protect the public health and safety in that they will provide reasonable assurance that long-term stability of the disposed waste and the disposal site will be achieved and will eliminate to the extent practicable the need for ongoing active maintenance of the disposal site following closure.
            (f) The applicant's demonstration provides reasonable assurance that the applicable technical requirements of subpart D of this part will be met.
            (g) The applicant's proposal for institutional control provides reasonable assurance that institutional control will be provided for the length of time found necessary to ensure the findings in paragraphs (b) through (e) of this section and that the institutional control meets the requirements of § 61.59, Institutional requirements.
            (h) The information on financial assurances meets the requirements of subpart E of this part.
            (i) The applicant's physical security information provides reasonable assurance that the requirements of part 73 of this chapter will be met, insofar as they are applicable to special nuclear material to be possessed before disposal under the license.
            (j) The applicant's criticality safety procedures are adequate to protect the public health and safety and provide reasonable assurance that the requirements of § 70.24, Criticality accident requirements, of part 70 of this chapter will be met, insofar as they are applicable to special nuclear material to be possessed before disposal under the license.
            (k) Any additional information submitted as requested by the Commission pursuant to § 61.16, Other information, is adequate.
            (l) The requirements of subpart A of part 51 of this chapter have been met.
            [47 FR 57463, Dec. 27, 1982, as amended at 49 FR 9405, Mar. 12, 1984]

            § 61.24
            Conditions of licenses.
            (a) A license issued under this part, or any right thereunder, may be transferred, assigned, or in any manner disposed of, either voluntarily or involuntarily, directly or indirectly, through transfer of control of the license to any person, only if the Commission finds, after securing full information, that the transfer is in accordance with the provisions of the Atomic Energy Act and gives its consent in writing in the form of a license amendment.
            (b) The licensee shall submit written statements under oath upon request of the Commission, at any time before termination of the license, to enable the Commission to determine whether or not the license should be modified, suspended, or revoked.
            (c) The license will be transferred to the site owner only on the full implementation of the final closure plan as approved by the Commission, including post-closure observation and maintenance.
            (d) The licensee shall be subject to the provisions of the Atomic Energy Act now or hereafter in effect, and to all rules, regulations, and orders of the Commission. The terms and conditions of the license are subject to amendment, revision, or modification, by reason of amendments to, or by reason of rules, regulations, and orders issued in accordance with the terms of the Atomic Energy Act.
            (e) Any license may be revoked, suspended or modified in whole or in part for any material false statement in the application or any statement of fact required under Section 182 of the Act, or because of conditions revealed by any application or statement of fact or any report, record, or inspection or other means which would warrant the Commission to refuse to grant a license to the original application, or for failure to operate the facility in accordance with the terms of the license, or for any violation of, or failure to observe any of the terms and conditions of the Act, or any rule, regulation, license or order of the Commission.

            (f) Each person licensed by the Commission pursuant to the regulations in this part shall confine possession and use of materials to the locations and purposes authorized in the license.

            (g) No radioactive waste may be disposed of until the Commission has inspected the land disposal facility and has found it to be in conformance with the description, design, and construction described in the application for a license.
            (h) The Commission may incorporate in any license at the time of issuance, or thereafter, by appropriate rule, regulation or order, additional requirements and conditions with respect to the licensee's receipt, possession, and disposal of source, special nuclear or byproduct material as it deems appropriate or necessary in order to:
            (1) Promote the common defense and security;
            (2) Protect health or to minimize danger to life or property;
            (3) Require reports and the keeping of records, and to provide for inspections of activities under the license that may be necessary or appropriate to effectuate the purposes of the Act and regulations thereunder.
            (i) Any licensee who receives and possesses special nuclear material under this part in quantities that would be subject to the requirements of § 70.24 of part 70 of this chapter shall comply with the requirements of that section. The licensee shall not consider the quantity of special nuclear material that has been disposed of.
            (j) The authority to dispose of wastes expires on the date stated in the license except as provided in § 61.27(a) of this part.
            (k)(1) Each licensee shall notify the appropriate NRC Regional Administrator, in writing, immediately following the filing of a voluntary or involuntary petition for bankruptcy under any Chapter of Title 11 (Bankruptcy) of the United States Code by or against:
            (i) The licensee;
            (ii) An entity (as that term is defined in 11 U.S.C. 101(14)) controlling the licensee or listing the license or licensee as property of the estate; or
            (iii) An affiliate (as that term is defined in 11 U.S.C. 101(2)) of the licensee.
            (2) This notification must indicate:
            (i) The bankruptcy court in which the petition for bankruptcy was filed; and
            (ii) The date of the filing of the petition.
            [47 FR 57463, Dec. 27, 1982, as amended at 52 FR 1295, Jan. 12, 1987]

            § 61.25
            Changes.
            (a) Except as provided for in specific license conditions, the licensee shall not make changes in the land disposal facility or procedures described in the license application. The license will include conditions restricting subsequent changes to the facility and the procedures authorized which are important to public health and safety. These license restrictions will fall into three categories of descending importance to public health and safety as follows: (1) those features and procedures which may not be changed without (i) 60 days prior notice to the Commission, (ii) 30 days notice of opportunity for a prior hearing, and (iii) prior Commission approval; (2) those features and procedures which may not be changed without (i) 60 days prior notice to the Commisson, and (ii) prior Commission approval; and (3) those features and procedures which may not be changed without 60 days prior notice to the Commission. Features and procedures falling in paragraph (a)(3) of this section may not be changed without prior Commission approval if the Commission, after having received the required notice, so orders.
            (b) Amendments authorizing site closure, license transfer, or license termination shall be included in paragraph (a)(1) of this section.
            (c) The Commission shall provide a copy of the notices of opportunity for hearing provided in paragraph (a)(1) of this section to State and local officials or tribal governing bodies specified in § 2.104(c) of this chapter.
            [47 FR 57463, Dec. 27, 1982, as amended at 77 FR 46600, Aug. 3, 2012]

            § 61.26
            Amendment of license.
            (a) An application for amendment of a license must be filed in accordance with § 61.20 and shall fully describe the changes desired.
            (b) In determining whether an amendment to a license will be approved, the Commission will apply the criteria set forth in § 61.23.

            § 61.27
            Application for renewal or closure.
            (a) Any expiration date on a license applies only to the above ground activities and to the authority to dispose of waste. Failure to renew the license shall not relieve the licensee of responsibility for carrying out site closure, postclosure observation and transfer of the license to the site owner. An application for renewal or an application for closure under § 61.28 must be filed at least 30 days prior to license expiration.
            (b) Applications for renewal of a license must be filed in accordance with §§ 61.10 through 61.16 and § 61.20. Applications for closure must be filed in accordance with §§ 61.20 and 61.28. Information contained in previous applications, statements or reports filed with the Commission under the license may be incorporated by reference if the references are clear and specific.
            (c) In any case in which a licensee has timely filed an application for renewal of a license, the license for continued receipt and disposal of licensed materials does not expire until the Commission has taken final action on the application for renewal.
            (d) In determining whether a license will be renewed, the Commission will apply the criteria set forth in § 61.23.

            § 61.28
            Contents of application for closure.
            (a) Prior to final closure of the disposal site, or as otherwise directed by the Commission, the applicant shall submit an application to amend the license for closure. This closure application must include a final revision and specific details of the disposal site closure plan included as part of the license application submitted under § 61.12(g) that includes each of the following:
            (1) Any additional geologic, hydrologic, or other disposal site data pertinent to the long-term containment of emplaced radioactive wastes obtained during the operational period.
            (2) The results of tests, experiments, or any other analyses relating to backfill of excavated areas, closure and sealing, waste migration and interaction with emplacement media, or any other tests, experiments, or analysis pertinent to the long-term containment of emplaced waste within the disposal site.
            (3) Any proposed revision of plans for:
            (i) Decontamination and/or dismantlement of surface facilities;
            (ii) Backfilling of excavated areas; or
            (iii) Stabilization of the disposal site for post-closure care.
            (b) An environmental report or a supplement to an environmental report prepared in accordance with subpart A of part 51 of this chapter must accompany the application.
            (c) Upon review and consideration of an application to amend the license for closure submitted in accordance with paragraph (a) of this section, the Commission shall issue an amendment authorizing closure if there is reasonable assurance that the long-term performance objectives of subpart C of this part will be met.
            [47 FR 57463, Dec. 27, 1982, as amended at 49 FR 9406, Mar. 12, 1984]

            § 61.29
            Post-closure observation and maintenance.
            Following completion of closure authorized in § 61.28, the licensee shall observe, monitor, and carry out necessary maintenance and repairs at the disposal site until the license is transferred by the Commission in accordance with § 61.30. Responsibility for the disposal site must be maintained by the licensee for 5 years. A shorter or longer time period for post-closure observation and maintenance may be established and approved as part of the site closure plan, based on site-specific conditions.

            § 61.30
            Transfer of license.
            (a) Following closure and the period of post-closure observation and maintenance, the licensee may apply for an amendment to transfer the license to the disposal site owner. The license shall be transferred when the Commission finds:
            (1) That the closure of the disposal site has been made in conformance with the licensee's disposal site closure plan, as amended and approved as part of the license;

            (2) That reasonable assurance has been provided by the licensee that the performance objectives of subpart C of this part are met;
            (3) That any funds for care and records required by § 61.80 (e) and (f) have been transferred to the disposal site owner;
            (4) That the post-closure monitoring program is operational for implementation by the disposal site owner; and
            (5) That the Federal or State government agency which will assume responsibility for institutional control of the disposal site is prepared to assume responsibility and ensure that the institutional requirements found necessary under § 61.23(g) will be met.
            [47 FR 57463, Dec. 27, 1982, as amended at 61 FR 24674, May 16, 1996]

            § 61.31
            Termination of license.
            (a) Following any period of institutional control needed to meet the requirements found necessary under § 61.23, the licensee may apply for an amendment to terminate the license.
            (b) This application must be filed, and will be reviewed, in accordance with the provision of § 61.20 and of this section.
            (c) A license is terminated only when the Commission finds:
            (1) That the institutional control requirements found necessary under § 61.23(g) have been met; and
            (2) That any additional requirements resulting from new information developed during the institutional control period have been met, and that permanent monuments or markers warning against intrusion have been installed.
            (3) That the records required by § 61.80 (e) and (f) have been sent to the party responsible for institutional control of the disposal site and a copy has been sent to the Commission immediately prior to license termination.
            [47 FR 57463, Dec. 27, 1982, as amended at 61 FR 24674, May 16, 1996]

            US/IAEA Safeguards Agreement

              § 61.32
              Facility information and verification.
              (a) In response to a written request by the Commission, each applicant for a license and each recipient of a license shall submit facility information, as described in § 75.10 of this chapter, on Form N-71 and associated forms and site information on DOC/NRC Form AP-A, and associated forms;
              (b) As required by the Additional Protocol, applicants and licensees specified in paragraph (a) of this section shall submit location information described in § 75.11 of this chapter on DOC/NRC Form AP-1 and associated forms; and
              (c) Shall permit verification thereof by the International Atomic Energy Agency (IAEA) and take other action as necessary to implement the US/IAEA Safeguards Agreement, as described in Part 75 of this chapter.
              [73 FR 78606, Dec. 23, 2008]

          Subpart C—Performance Objectives

            § 61.40
            General requirement.
            Land disposal facilities must be sited, designed, operated, closed, and controlled after closure so that reasonable assurance exists that exposures to humans are within the limits established in the performance objectives in §§ 61.41 through 61.44.

            § 61.41
            Protection of the general population from releases of radioactivity.
            Concentrations of radioactive material which may be released to the general environment in groundwater, surface water, air, soil, plants, or animals must not result in an annual dose exceeding an equivalent of 25 millirems to the whole body, 75 millirems to the thyroid, and 25 millirems to any other organ of any member of the public. Reasonable effort should be made to maintain releases of radioactivity in effluents to the general environment as low as is reasonably achievable.

            § 61.42
            Protection of individuals from inadvertent intrusion.
            Design, operation, and closure of the land disposal facility must ensure protection of any individual inadvertently intruding into the disposal site and occupying the site or contacting the waste at any time after active institutional controls over the disposal site are removed.

            § 61.43
            Protection of individuals during operations.
            Operations at the land disposal facility must be conducted in compliance with the standards for radiation protection set out in part 20 of this chapter, except for releases of radioactivity in effluents from the land disposal facility, which shall be governed by § 61.41 of this part. Every reasonable effort shall be made to maintain radiation exposures as low as is reasonably achievable.

            § 61.44
            Stability of the disposal site after closure.
            The disposal facility must be sited, designed, used, operated, and closed to achieve long-term stability of the disposal site and to eliminate to the extent practicable the need for ongoing active maintenance of the disposal site following closure so that only surveillance, monitoring, or minor custodial care are required.

          Subpart D—Technical Requirements for Land Disposal Facilities

            § 61.50
            Disposal site suitability requirements for land disposal.
            (a) Disposal site suitability for near-surface disposal. (1) The purpose of this section is to specify the minimum characteristics a disposal site must have to be acceptable for use as a near-surface disposal facility. The primary emphasis in disposal site suitability is given to isolation of wastes, a matter having long-term impacts, and to disposal site features that ensure that the long-term performance objectives of subpart C of this part are met, as opposed to short-term convenience or benefits.
            (2) The disposal site shall be capable of being characterized, modeled, analyzed and monitored.
            (3) Within the region or state where the facility is to be located, a disposal site should be selected so that projected population growth and future developments are not likely to affect the ability of the disposal facility to meet the performance objectives of subpart C of this part.
            (4) Areas must be avoided having known natural resources which, if exploited, would result in failure to meet the performance objectives of subpart C of this part.
            (5) The disposal site must be generally well drained and free of areas of flooding or frequent ponding. Waste disposal shall not take place in a 100-year flood plain, coastal high-hazard area or wetland, as defined in Executive Order 11988, “Floodplain Management Guidelines.”
            (6) Upstream drainage areas must be minimized to decrease the amount of runoff which could erode or inundate waste disposal units.
            (7) The disposal site must provide sufficient depth to the water table that groundwater intrusion, perennial or otherwise, into the waste will not occur. The Commission will consider an exception to this requirement to allow disposal below the water table if it can be conclusively shown that disposal site characteristics will result in molecular diffusion being the predominant means of radionuclide movement and the rate of movement will result in the performance objectives of subpart C of this part being met. In no case will waste disposal be permitted in the zone of fluctuation of the water table.
            (8) The hydrogeologic unit used for disposal shall not discharge groundwater to the surface within the disposal site.
            (9) Areas must be avoided where tectonic processes such as faulting, folding, seismic activity, or vulcanism may occur with such frequency and extent to significantly affect the ability of the disposal site to meet the performance objectives of subpart C of this part, or may preclude defensible modeling and prediction of long-term impacts.

            (10) Areas must be avoided where surface geologic processes such as mass wasting, erosion, slumping, landsliding, or weathering occur with such frequency and extent to significantly affect the ability of the disposal site to meet the performance objectives of subpart C of this part, or may preclude defensible modeling and prediction of long-term impacts.

            (11) The disposal site must not be located where nearby facilities or activities could adversely impact the ability of the site to meet the performance objectives of subpart C of this part or significantly mask the environmental monitoring program.
            (b) Disposal site suitability requirements for land disposal other than near-surface. [Reserved]

            § 61.51
            Disposal site design for land disposal.
            (a) Disposal site design for near-surface disposal. (1) Site design features must be directed toward long-term isolation and avoidance of the need for continuing active maintenance after site closure.
            (2) The disposal site design and operation must be compatible with the disposal site closure and stabilization plan and lead to disposal site closure that provides reasonable assurance that the performance objectives of subpart C of this part will be met.
            (3) The disposal site must be designed to complement and improve, where appropriate, the ability of the disposal site's natural characteristics to assure that the performance objectives of subpart C of this part will be met.
            (4) Covers must be designed to minimize to the extent practicable water infiltration, to direct percolating or surface water away from the disposed waste, and to resist degradation by surface geologic processes and biotic activity.
            (5) Surface features must direct surface water drainage away from disposal units at velocities and gradients which will not result in erosion that will require ongoing active maintenance in the future.
            (6) The disposal site must be designed to minimize to the extent practicable the contact of water with waste during storage, the contact of standing water with waste during disposal, and the contact of percolating or standing water with wastes after disposal.
            (b) Disposal site design for other than near-surface disposal. [Reserved]

            § 61.52
            Land disposal facility operation and disposal site closure.
            (a) Near-surface disposal facility operation and disposal site closure. (1) Wastes designated as Class A pursuant to § 61.55, must be segregated from other wastes by placing in disposal units which are sufficiently separated from disposal units for the other waste classes so that any interaction between Class A wastes and other wastes will not result in the failure to meet the performance objectives in subpart C of this Part. This segregation is not necessary for Class A wastes if they meet the stability requirements in § 61.56(b) of this part.
            (2) Wastes designated as Class C pursuant to § 61.55, must be disposed of so that the top of the waste is a minimum of 5 meters below the top surface of the cover or must be disposed of with intruder barriers that are designed to protect against an inadvertent intrusion for a least 500 years.
            (3) All wastes shall be disposed of in accordance with the requirements of paragraphs (a) (4) through (11) of this section.
            (4) Wastes must be emplaced in a manner that maintains the package integrity during emplacement, minimizes the void spaces between packages, and permits the void spaces to be filled.
            (5) Void spaces between waste packages must be filled with earth or other material to reduce future subsidence within the fill.
            (6) Waste must be placed and covered in a manner that limits the radiation dose rate at the surface of the cover to levels that at a minimum will permit the licensee to comply with all provisions of §§ 20.1301 and 20.1302 of this chapter at the time the license is transferred pursuant to § 61.30 of this part.

            (7) The boundaries and locations of each disposal unit (e.g., trenches) must be accurately located and mapped by means of a land survey. Near-surface disposal units must be marked in such a way that the boundaries of each unit can be easily defined. Three permanent survey marker control points, referenced to United States Geological Survey (USGS) or National Geodetic Survey (NGS) survey control stations, must be established on the site to facilitate surveys. The USGS or NGS control stations must provide horizontal and vertical controls as checked against USGS or NGS record files.
            (8) A buffer zone of land must be maintained between any buried waste and the disposal site boundary and beneath the disposed waste. The buffer zone shall be of adequate dimensions to carry out environmental monitoring activities specified in § 61.53(d) of this part and take mitigative measures if needed.
            (9) Closure and stabilization measures as set forth in the approved site closure plan must be carried out as each disposal unit (e.g., each trench) is filled and covered.
            (10) Active waste disposal operations must not have an adverse effect on completed closure and stabilization measures.
            (11) Only wastes containing or contaminated with radioactive materials shall be disposed of at the disposal site.
            (b) Facility operation and disposal site closure for land disposal facilities other than near-surface. [Reserved]
            [47 FR 57463, Dec. 27, 1982, as amended at 56 FR 23474, May 21, 1991; 56 FR 61352, Dec. 3, 1991; 58 FR 67662, Dec. 22, 1993]

            § 61.53
            Environmental monitoring.
            (a) At the time a license application is submitted, the applicant shall have conducted a preoperational monitoring program to provide basic environmental data on the disposal site characteristics. The applicant shall obtain information about the ecology, meteorology, climate, hydrology, geology, geochemistry, and seismology of the disposal site. For those characteristics that are subject to seasonal variation, data must cover at least a twelve month period.
            (b) The licensee must have plans for taking corrective measures if migration of radionuclides would indicate that the performance objectives of subpart C may not be met.
            (c) During the land disposal facility site construction and operation, the licensee shall maintain a monitoring program. Measurements and observations must be made and recorded to provide data to evaluate the potential health and environmental impacts during both the construction and the operation of the facility and to enable the evaluation of long-term effects and the need for mitigative measures. The monitoring system must be capable of providing early warning of releases of radionuclides from the disposal site before they leave the site boundary.
            (d) After the disposal site is closed, the licensee responsible for post-operational surveillance of the disposal site shall maintain a monitoring system based on the operating history and the closure and stabilization of the disposal site. The monitoring system must be capable of providing early warning of releases of radionuclides from the disposal site before they leave the site boundary.

            § 61.54
            Alternative requirements for design and operations.
            The Commission may, upon request or on its own initiative, authorize provisions other than those set forth in §§ 61.51 through 61.53 for the segregation and disposal of waste and for the design and operation of a land disposal facility on a specific basis, if it finds reasonable assurance of compliance with the performance objectives of subpart C of this part.

            § 61.55
            Waste classification.
            (a) Classification of waste for near surface disposal—(1) Considerations. Determination of the classification of radioactive waste involves two considerations. First, consideration must be given to the concentration of long-lived radionuclides (and their shorter-lived precursors) whose potential hazard will persist long after such precautions as institutional controls, improved waste form, and deeper disposal have ceased to be effective. These precautions delay the time when long-lived radionuclides could cause exposures. In addition, the magnitude of the potential dose is limited by the concentration and availability of the radionuclide at the time of exposure. Second, consideration must be given to the concentration of shorter-lived radionuclides for which requirements on institutional controls, waste form, and disposal methods are effective.
            (2) Classes of waste. (i) Class A waste is waste that is usually segregated from other waste classes at the disposal site. The physical form and characteristics of Class A waste must meet the minimum requirements set forth in § 61.56(a). If Class A waste also meets the stability requirements set forth in § 61.56(b), it is not necessary to segregate the waste for disposal.
            (ii) Class B waste is waste that must meet more rigorous requirements on waste form to ensure stability after disposal. The physical form and characteristics of Class B waste must meet both the minimum and stability requirements set forth in § 61.56.
            (iii) Class C waste is waste that not only must meet more rigorous requirements on waste form to ensure stability but also requires additional measures at the disposal facility to protect against inadvertent intrusion. The physical form and characteristics of Class C waste must meet both the minimum and stability requirements set forth in § 61.56.
            (iv) Waste that is not generally acceptable for near-surface disposal is waste for which form and disposal methods must be different, and in general more stringent, than those specified for Class C waste. In the absence of specific requirements in this part, such waste must be disposed of in a geologic repository as defined in part 60 or 63 of this chapter unless proposals for disposal of such waste in a disposal site licensed pursuant to this part are approved by the Commission.
            (3) Classification determined by long-lived radionuclides. If radioactive waste contains only radionuclides listed in Table 1, classification shall be determined as follows:
            (i) If the concentration does not exceed 0.1 times the value in Table 1, the waste is Class A.
            (ii) If the concentration exceeds 0.1 times the value in Table 1 but does not exceed the value in Table 1, the waste is Class C.
            (iii) If the concentration exceeds the value in Table 1, the waste is not generally acceptable for near-surface disposal.
            (iv) For wastes containing mixtures of radionuclides listed in Table 1, the total concentration shall be determined by the sum of fractions rule described in paragraph (a)(7) of this section.

              Table 1

                Radionuclide
                Concentration curies per cubic meter

                C-14
                8

                C-14 in activated metal
                80

                Ni-59 in activated metal
                220

                Nb-94 in activated metal
                0.2

                Tc-99
                3

                I-129
                0.08

                Alpha emitting transuranic nuclides with half-life greater than 5 years

                  1 100

                Pu-241

                  1 3,500

                Cm-242

                  1 20,000

                1 Units are nanocuries per gram.

            (4) Classification determined by short-lived radionuclides. If radioactive waste does not contain any of the radionuclides listed in Table 1, classification shall be determined based on the concentrations shown in Table 2. However, as specified in paragraph (a)(6) of this section, if radioactive waste does not contain any nuclides listed in either Table 1 or 2, it is Class A.
            (i) If the concentration does not exceed the value in Column 1, the waste is Class A.
            (ii) If the concentration exceeds the value in Column 1, but does not exceed the value in Column 2, the waste is Class B.
            (iii) If the concentration exceeds the value in Column 2, but does not exceed the value in Column 3, the waste is Class C.
            (iv) If the concentration exceeds the value in Column 3, the waste is not generally acceptable for near-surface disposal.
            (v) For wastes containing mixtures of the nuclides listed in Table 2, the total concentration shall be determined by the sum of fractions rule described in paragraph (a)(7) of this section.

              Table 2

                Radionuclide
                Concentration, curies per cubic meter
                Col. 1
                Col. 2
                Col. 3

                Total of all nuclides with less than 5 year half-life
                700
                (1)
                (1)

                H-3
                40
                (1)
                (1)

                Co-60
                700
                (1)
                (1)

                Ni-63
                3.5
                70
                700

                Ni-63 in activated metal
                35
                700
                7000

                Sr-90
                0.04
                150
                7000

                Cs-137
                1
                44
                4600

                1 There are no limits established for these radionuclides in Class B or C wastes. Practical considerations such as the effects of external radiation and internal heat generation on transportation, handling, and disposal will limit the concentrations for these wastes. These wastes shall be Class B unless the concentrations of other nuclides in Table 2 determine the waste to be Class C independent of these nuclides.

            (5) Classification determined by both long- and short-lived radionuclides. If radioactive waste contains a mixture of radionuclides, some of which are listed in Table 1, and some of which are listed in Table 2, classification shall be determined as follows:
            (i) If the concentration of a nuclide listed in Table 1 does not exceed 0.1 times the value listed in Table 1, the class shall be that determined by the concentration of nuclides listed in Table 2.
            (ii) If the concentration of a nuclide listed in Table 1 exceeds 0.1 times the value listed in Table 1 but does not exceed the value in Table 1, the waste shall be Class C, provided the concentration of nuclides listed in Table 2 does not exceed the value shown in Column 3 of Table 2.
            (6) Classification of wastes with radionuclides other than those listed in Tables 1 and 2. If radioactive waste does not contain any nuclides listed in either Table 1 or 2, it is Class A.
            (7) The sum of the fractions rule for mixtures of radionuclides. For determining classification for waste that contains a mixture of radionuclides, it is necessary to determine the sum of fractions by dividing each nuclide's concentration by the appropriate limit and adding the resulting values. The appropriate limits must all be taken from the same column of the same table. The sum of the fractions for the column must be less than 1.0 if the waste class is to be determined by that column. Example: A waste contains Sr-90 in a concentration of 50 Ci/m3 and Cs-137 in a concentration of 22 Ci/m3. Since the concentrations both exceed the values in Column 1, Table 2, they must be compared to Column 2 values. For Sr-90 fraction 50/150 = 0.33; for Cs-137 fraction, 22/44 = 0.5; the sum of the fractions = 0.83. Since the sum is less than 1.0, the waste is Class B.
            (8) Determination of concentrations in wastes. The concentration of a radionuclide may be determined by indirect methods such as use of scaling factors which relate the inferred concentration of one radionuclide to another that is measured, or radionuclide material accountability, if there is reasonable assurance that the indirect methods can be correlated with actual measurements. The concentration of a radionuclide may be averaged over the volume of the waste, or weight of the waste if the units are expressed as nanocuries per gram.
            [47 FR 57463, Dec. 27, 1982, as amended at 54 FR 22583, May 25, 1989; 66 FR 55792, Nov. 2, 2001]

            § 61.56
            Waste characteristics.
            (a) The following requirements are minimum requirements for all classes of waste and are intended to facilitate handling at the disposal site and provide protection of health and safety of personnel at the disposal site.
            (1) Waste must not be packaged for disposal in cardboard or fiberboard boxes.
            (2) Liquid waste must be solidified or packaged in sufficient absorbent material to absorb twice the volume of the liquid.
            (3) Solid waste containing liquid shall contain as little free standing and noncorrosive liquid as is reasonably achievable, but in no case shall the liquid exceed 1% of the volume.
            (4) Waste must not be readily capable of detonation or of explosive decomposition or reaction at normal pressures and temperatures, or of explosive reaction with water.
            (5) Waste must not contain, or be capable of generating, quantities of toxic gases, vapors, or fumes harmful to persons transporting, handling, or disposing of the waste. This does not apply to radioactive gaseous waste packaged in accordance with paragraph (a)(7) of this section.
            (6) Waste must not be pyrophoric. Pyrophoric materials contained in waste shall be treated, prepared, and packaged to be nonflammable.

            (7) Waste in a gaseous form must be packaged at a pressure that does not exceed 1.5 atmospheres at 20 °C. Total activity must not exceed 100 curies per container.
            (8) Waste containing hazardous, biological, pathogenic, or infectious material must be treated to reduce to the maximum extent practicable the potential hazard from the non-radiological materials.
            (b) The requirements in this section are intended to provide stability of the waste. Stability is intended to ensure that the waste does not structurally degrade and affect overall stability of the site through slumping, collapse, or other failure of the disposal unit and thereby lead to water infiltration. Stability is also a factor in limiting exposure to an inadvertent intruder, since it provides a recognizable and nondispersible waste.
            (1) Waste must have structural stability. A structurally stable waste form will generally maintain its physical dimensions and its form, under the expected disposal conditions such as weight of overburden and compaction equipment, the presence of moisture, and microbial activity, and internal factors such as radiation effects and chemical changes. Structural stability can be provided by the waste form itself, processing the waste to a stable form, or placing the waste in a disposal container or structure that provides stability after disposal.
            (2) Notwithstanding the provisions in § 61.56(a) (2) and (3), liquid wastes, or wastes containing liquid, must be converted into a form that contains as little free standing and noncorrosive liquid as is reasonably achievable, but in no case shall the liquid exceed 1% of the volume of the waste when the waste is in a disposal container designed to ensure stability, or 0.5% of the volume of the waste for waste processed to a stable form.
            (3) Void spaces within the waste and between the waste and its package must be reduced to the extent practicable.

            § 61.57
            Labeling.
            Each package of waste must be clearly labeled to identify whether it is Class A waste, Class B waste, or Class C waste, in accordance with § 61.55.

            § 61.58
            Alternative requirements for waste classification and characteristics.
            The Commission may, upon request or on its own initiative, authorize other provisions for the classification and characteristics of waste on a specific basis, if, after evaluation, of the specific characteristics of the waste, disposal site, and method of disposal, it finds reasonable assurance of compliance with the performance objectives in subpart C of this part.

            § 61.59
            Institutional requirements.
            (a) Land ownership. Disposal of radioactive waste received from other persons may be permitted only on land owned in fee by the Federal or a State government.
            (b) Institutional control. The land owner or custodial agency shall carry out an institutional control program to physically control access to the disposal site following transfer of control of the disposal site from the disposal site operator. The institutional control program must also include, but not be limited to, carrying out an environmental monitoring program at the disposal site, periodic surveillance, minor custodial care, and other requirements as determined by the Commission; and administration of funds to cover the costs for these activities. The period of institutional controls will be determined by the Commission, but institutional controls may not be relied upon for more than 100 years following transfer of control of the disposal site to the owner.

          Subpart E—Financial Assurances

            § 61.61
            Applicant qualifications and assurances.
            Each applicant shall show that it either possesses the necessary funds or has reasonable assurance of obtaining the necessary funds, or by a combination of the two, to cover the estimated costs of conducting all licensed activities over the planned operating life of the project, including costs of construction and disposal.

            § 61.62
            Funding for disposal site closure and stabilization.
            (a) The applicant shall provide assurance that sufficient funds will be available to carry out disposal site closure and stabilization, including: (1) Decontamination or dismantlement of land disposal facility structures; and (2) closure and stabilization of the disposal site so that following transfer of the disposal site to the site owner, the need for ongoing active maintenance is eliminated to the extent practicable and only minor custodial care, surveillance, and monitoring are required. These assurances shall be based on Commission-approved cost estimates reflecting the Commission-approved plan for disposal site closure and stabilization. The applicant's cost estimates must take into account total capital costs that would be incurred if an independent contractor were hired to perform the closure and stabilization work.
            (b) In order to avoid unnecessary duplication and expense, the Commission will accept financial sureties that have been consolidated with earmarked financial or surety arrangements established to meet requirements of other Federal or State agencies and/or local governing bodies for such decontamination, closure and stabilization. The Commission will accept this arrangement only if they are considered adequate to satisfy these requirements and that the portion of the surety which covers the closure of the disposal site is clearly identified and committed for use in accomplishing these activities.
            (c) The licensee's surety mechanism will be annually reviewed by the Commission to assure that sufficient funds are available for completion of the closure plan, assuming that the work has to be performed by an independent contractor.
            (d) The amount of surety liability should change in accordance with the predicted cost of future closure and stabilization. Factors affecting closure and stabilization cost estimates include: inflation; increases in the amount of disturbed land; changes in engineering plans; closure and stabilization that has already been accomplished and any other conditions affecting costs. This will yield a surety that is at least sufficient at all times to cover the costs of closure of the disposal units that are expected to be used before the next license renewal.
            (e) The term of the surety mechanism must be open ended unless it can be demonstrated that another arrangement would provide an equivalent level of assurance. This assurance could be provided with a surety mechanism which is written for a specified period of time (e.g., five years) yet which must be automatically renewed unless the party who issues the surety notifies the Commission and the beneficiary (the site owner) and the principal (the licensee) not less than 90 days prior to the renewal date of its intention not to renew. In such a situation the licensee must submit a replacement surety within 30 days after notification of cancellation. If the licensee fails to provide a replacement surety acceptable to the Commission, the site owner may collect on the original surety.
            (f) Proof of forfeiture must not be necessary to collect the surety so that in the event that the licensee could not provide an acceptable replacement surety within the required time, the surety shall be automatically collected prior to its expiration. The conditions described above would have to be clearly stated on any surety instrument which is not open-ended, and must be agreed to by all parties. Liability under the surety mechanism must remain in effect until the closure and stabilization program has been completed and approved by the Commission and the license has been transferred to the site owner.

            (g) Financial surety arrangements generally acceptable to the Commission include: surety bonds, cash deposits, certificates of deposits, deposits of government securities, escrow accounts, irrevocable letters or lines of credit, trust funds, and combinations of the above or such other types of arrangements as may be approved by the Commission. However, self-insurance, or any arrangement which essentially constitutes pledging the assets of the licensee, will not satisfy the surety requirement for private sector applicants since this provides no additional assurance other than that which already exists through license requirements.

            § 61.63
            Financial assurances for institutional controls.
            (a) Prior to the issuance of the license, the applicant shall provide for Commission review and approval a copy of a binding arrangement, such as a lease, between the applicant and the disposal site owner that ensures that sufficient funds will be available to cover the costs of monitoring and any required maintenance during the institutional control period. The binding arrangement will be reviewed periodically by the Commission to ensure that changes in inflation, technology and disposal facility operations are reflected in the arrangements.
            (b) Subsequent changes to the binding arrangement specified in paragraph (a) of this section relevant to institutional control shall be submitted to the Commission for approval.

          Subpart F—Participation by State Governments and Indian Tribes

            § 61.70
            Scope.
            This subpart describes mechanisms through which the Commission will implement a formal request from a State or tribal government to participate in the review of a license application for a land disposal facility. Nothing in this subpart may be construed to bar the State or tribal governing body from participating in subsequent Commission proceedings concerning the license application as provided under Federal law and regulations.

            § 61.71
            State and Tribal government consultation.
            Upon request of a State or tribal governing body, the Director shall make available Commission staff to discuss with representatives of the State or tribal governing body information submitted by the applicant, applicable Commission regulations, licensing procedures, potential schedules, and the type and scope of State activities in the license review permitted by law. In addition, staff shall be made available to consult and cooperate with the State or tribal governing body in developing proposals for participation in the license review.

            § 61.72
            Filing of proposals for State and Tribal participation.
            (a) A State or tribal governing body whose interest is affected by a near-surface disposal facility at the proposed site may submit to the Director a proposal for participation in the review of a license application. Proposals must be submitted within the following time periods:

            (1) For the State in which the disposal facility will be located, or any State that is member of an interstate compact that includes the State in which the disposal facility is located, no later than 45 days following publication in the Federal Register of the notice of tendering of an application submitted under § 61.20.

            (2) For any other State, or for a tribal governing body, no later than 120 days following publication in the Federal Register of the notice of tendering of an application submitted under § 61.20.
            (b) Proposals for participation in the licensing process must be made in writing and must be signed by the Governor of the State or the official otherwise provided for by State or tribal law.
            (c) At a minimum, proposals must contain each of the following items of information:
            (1) A general description of how the State or Tribe wishes to participate in the licensing process specifically identifying those issues it wishes to review.
            (2) A description of material and information which the State or Tribe plans to submit to the Commission for consideration in the licensing process. A tentative schedule referencing steps in the review and calendar dates for planned submittals should be included.
            (3) A description of any work that the State or Tribe proposes to perform for the Commission in support of the licensing process.
            (4) A description of State or Tribal plans to facilitate local government and citizen participation.

            (5) A preliminary estimate of the types and extent of impacts which the State expects, should a disposal facility be located as proposed.

            (6) If desired, any requests for educational or information services (seminars, public meetings) or other actions from the Commission such as establishment of additional Public Document Rooms or exchange of State personnel under the Intergovernmental Personnel Act.

            § 61.73
            Commission approval of proposals.
            (a) Upon receipt of a proposal submitted in accordance with § 61.72, the Director shall arrange for a meeting between the representatives of the State or Tribal governing body and the Commission staff to discuss the proposal and to ensure full and effective participation by the State or Tribe in the Commission's license review.
            (b) If requested by a State or Tribal governing body, the Director may approve all or any part of a proposal if the Director determines that:
            (1) The proposed activities are within the scope of Commission statutory responsibility and the type and magnitude of impacts which the State or Tribe may bear are sufficient to justify their participation; and
            (2) The proposed activities will contribute productively to the licensing review.
            (c) The decision of the Director will be transmitted in writing to the governor or the designated official of the Tribal governing body.
            (d) Participation by a State or Indian Tribe shall not affect their rights to participate in an adjudicatory hearing as provided by part 2 of this chapter.

          Subpart G—Records, Reports, Tests, and Inspections

            § 61.80
            Maintenance of records, reports, and transfers.
            (a) Each licensee shall maintain any records and make any reports in connection with the licensed activities as may be required by the conditions of the license or by the rules, regulations, and orders of the Commission.
            (b) Records which are required by the regulations in this part or by license conditions must be maintained for a period specified by the appropriate regulations in this chapter or by license condition. If a retention period is not otherwise specified, these records must be maintained and transferred to the officials specified in paragraph (e) of this section as a condition of license termination unless the Commission otherwise authorizes their disposition.
            (c) Records which must be maintained pursuant to this part may be the original or a reproduced copy or a microform if this reproduced copy or microform is capable of producing copy that is clear and legible at the end of the required retention period. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, specifications, must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.
            (d) If there is a conflict between the Commission's regulations in this part, license condition, or other written Commission approval or authorization pertaining to the retention period for the same type of record, the longest retention period specified takes precedence.
            (e) Notwithstanding paragraphs (a) through (d) of this section, the licensee shall record the location and the quantity of radioactive wastes contained in the disposal site and transfer these records upon license termination to the chief executive of the nearest municipality, the chief executive of the county in which the facility is located, the county zoning board or land development and planning agency, the State governor and other State, local, and Federal governmental agencies as designated by the Commission at the time of license termination.

            (f) Following receipt and acceptance of a shipment of radioactive waste, the licensee shall record the date that the shipment is received at the disposal facility, the date of disposal of the waste, a traceable shipment manifest number, a description of any engineered barrier or structural overpack provided for disposal of the waste, the location of disposal at the disposal site, the containment integrity of the waste disposal containers as received, any discrepancies between materials listed on the manifest and those received, the volume of any pallets, bracing, or other shipping or onsite generated materials that are contaminated, and are disposed of as contaminated or suspect materials, and any evidence of leaking or damaged disposal containers or radiation or contamination levels in excess of limits specified in Department of Transportation and Commission regulations. The licensee shall briefly describe any repackaging operations of any of the disposal containers included in the shipment, plus any other information required by the Commission as a license condition. The licensee shall retain these records until the Commission transfers or terminates the license that authorizes the activities described in this section.
            (g) Each licensee shall comply with the safeguards reporting requirements of §§ 30.55, 40.64, 74.13, and 74.15 of this chapter if the quantities or activities of materials received or transferred exceed the limits of these sections. Inventory reports required by these sections are not required for materials after disposal.
            (h) Each licensee authorized to dispose of radioactive waste received from other persons shall file a copy of its financial report or a certified financial statement annually with the Commission in order to update the information base for determining financial qualifications.
            (i)(1) Each licensee authorized to dispose of waste materials received from other persons under this part shall submit annual reports to the Director, Office of Nuclear Material Safety and Safeguards, by an appropriate method listed in § 60.4 of this chapter, with a copy to the appropriate NRC Regional Office shown in appendix D to part 20 of this chapter. Reports must be submitted by the end of the first calendar quarter of each year for the preceding year.
            (2) The reports shall include (i) specification of the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in airborne effluents during the preceding year, (ii) the results of the environmental monitoring program, (iii) a summary of licensee disposal unit survey and maintenance activities, (iv) a summary, by waste class, of activities and quantities of radionuclides disposed of, (v) any instances in which observed site characteristics were significantly different from those described in the application for a license; and (vi) any other information the Commission may require. If the quantities of radioactive materials released during the reporting period, monitoring results, or maintenance performed are significantly different from those expected in the materials previously reviewed as part of the licensing action, the report must cover this specifically.
            (j) Each licensee shall report in accordance with the requirements of § 70.52 of this chapter.
            (k) Any transfer of byproduct, source, and special nuclear materials by the licensee is subject to the requirements in §§ 30.41, 40.51, and 70.42 of this chapter. Byproduct, source and special nuclear material means materials as defined in these parts, respectively.
            (l) In addition to the other requirements of this section, the licensee shall store, or have stored, manifest and other information pertaining to receipt and disposal of radioactive waste in an electronic recordkeeping system.
            (1) The manifest information that must be electronically stored is—
            (i) That required in 10 CFR part 20, appendix G, with the exception of shipper and carrier telephone numbers and shipper and consignee certifications; and
            (ii) That information required in paragraph (f) of this section.
            (2) As specified in facility license conditions, the licensee shall report the stored information, or subsets of this information, on a computer-readable medium.
            [47 FR 57463, Dec. 27, 1982, as amended at 52 FR 31612, Aug. 21, 1987; 53 FR 19251, May 27, 1988; 58 FR 33891, June 22, 1993; 60 FR 15666, Mar. 27, 1995; 67 FR 78141, Dec. 23, 2002; 68 FR 58814, Oct. 10, 2003; 73 FR 5725, Jan. 31, 2008; 79 FR 75740, Dec. 19, 2014]

            § 61.81
            Tests at land disposal facilities.

            (a) Each licensee shall perform, or permit the Commission to perform, any tests as the Commission deems appropriate or necessary for the administration of the regulations in this part, including tests of:
            (1) Radioactive wastes and facilities used for the receipt, storage, treatment, handling and disposal of radioactive wastes.
            (2) Radiation detection and monitoring instruments; and
            (3) Other equipment and devices used in connection with the receipt, possession, handling, treatment, storage, or disposal of radioactive waste.

            § 61.82
            Commission inspections of land disposal facilities.
            (a) Each licensee shall afford to the Commission at all reasonable times opportunity to inspect radioactive waste not yet disposed of, and the premises, equipment, operations, and facilities in which radioactive wastes are received, possessed, handled, treated, stored, or disposed of.
            (b) Each licensee shall make available to the Commission for inspection, upon reasonable notice, records kept by it pursuant to the regulations in this chapter. Authorized representatives of the Commission may copy and take away copies of, for the Commission's use, any record required to be kept pursuant to this part.

            § 61.83
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of—
            (1) The Atomic Energy Act of 1954, as amended;
            (2) Title II of the Energy Reorganization Act of 1974, as amended; or
            (3) A regulation or order issued pursuant to those Acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:
            (1) For violations of—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section;
            (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.
            [57 FR 55077, Nov. 24, 1992]

            § 61.84
            Criminal penalties.
            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 61 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in part 61 that are not issued under sections 161b, 161i, or 161o for the purposes of Section 223 are as follows: §§ 61.1, 61.2, 61.4, 61.5, 61.6, 61.7, 61.8, 61.10, 61.11, 61.12, 61.13, 61.14, 61.15, 61.16, 61.20, 61.21, 61.22, 61.23, 61.26, 61.30, 61.31, 61.50, 61.51, 61.54, 61.55, 61.58, 61.59, 61.61, 61.63, 61.70, 61.71, 61.72, 61.73, 61.83, and 61.84.
            [57 FR 55077, Nov. 24, 1992]

        Pt. 62
        PART 62—CRITERIA AND PROCEDURES FOR EMERGENCY ACCESS TO NON-FEDERAL AND REGIONAL LOW-LEVEL WASTE DISPOSAL FACILITIES

            Subpart A—General Provisions
            Sec.
            62.1
            Purpose and scope.
            62.2
            Definitions.
            62.3
            Communications.
            62.4
            Interpretations.
            62.5
            Specific exemptions.
            62.8
            Information collection requirements: OMB approval.

            Subpart B—Request for a Commission Determination
            62.11
            Filing and distribution of a determination request.
            62.12
            Contents of a request for emergency access: General information.
            62.13

            Contents of a request for emergency access: Alternatives.

            62.14
            Contents of a request for an extension of emergency access.
            62.15
            Additional information.
            62.16
            Withdrawal of a determination request.
            62.17
            Elimination of repetition.
            62.18
            Denial of request.

            Subpart C—Issuance of a Commission Determination
            62.21
            Determination for granting emergency access.
            62.22
            Notice of issuance of a determination.
            62.23
            Determination for granting temporary emergency access.
            62.24
            Extension of emergency access.
            62.25
            Criteria for a Commission determination.
            62.26
            Criteria for designating a disposal facility.

            Subpart D—Termination of Emergency Access
            62.31
            Termination of emergency access.

          Authority:
          Atomic Energy Act of 1954, sec. 161 (42 U.S.C. 2201); Energy Reorganization Act of 1974, secs. 201 (42 U.S.C. 5841); Low-Level Radioactive Waste Policy Amendments Act of 1985, secs. 2, 6 (42 U.S.C. 2021b, 2021f); 44 U.S.C. 3504 note.

          Source:
          54 FR 5420, Feb. 3, 1989, unless otherwise noted.

          Subpart A—General Provisions

            § 62.1
            Purpose and scope.
            (a) The regulations in this part establish for specific low-level radioactive waste:
            (1) Criteria and procedures for granting emergency access under section 6 of the Low-Level Radioactive Waste Policy Amendments Act of 1985 (42 U.S.C. 2021) to any non-Federal or regional low-level radioactive waste (LLW) disposal facility or to any non-Federal disposal facility within a State that is not a member of a Compact, and
            (2) The terms and conditions upon which the Commission will grant this emergency access.
            (b) The regulations in this part apply to all persons as defined by this regulation, who have been denied access to existing regional or non-Federal low-level radioactive waste disposal facilities and who submit a request to the Commission for a determination pursuant to this part.
            (c) The regulations in this part apply only to the LLW that the States have the responsibility to dispose of pursuant to section 3(1)(a) of the Act.

            § 62.2
            Definitions.
            As used in this part:

              Act means the Low-Level Radioactive Waste Policy Amendments Act of 1985 (Pub. L. 99-240).

              Agreement State means a State that—
            (1) Has entered into an agreement with the Nuclear Regulatory Commission under section 274 of the Atomic Energy Act of 1954 (42 U.S.C. 2021); and
            (2) Has authority to regulate the disposal of low-level radioactive waste under such agreement.

              Commission means the Nuclear Regulatory Commission or its duly authorized representatives.

              Compact means a Compact entered into by two or more States pursuant to the Low-Level Radioactive Waste Policy Amendments Act of 1985.

              Compact Commission means the regional commission, committee, or board established in a Compact to administer such Compact.

              Disposal means the permanent isolation of low-level radioactive waste pursuant to the requirements established by the Nuclear Regulatory Commission under applicable laws, or by an Agreement State if such isolation occurs in this Agreement State.

              Emergency access means access to an operating non-Federal or regional low-level radioactive waste disposal facility or facilities for a period not to exceed 180 days, which is granted by NRC to a generator of low-level radioactive waste who has been denied the use of those facilities.

              Extension of emergency access means an extension of the access that had been previously granted by NRC to an operating non-Federal or regional low-level radioactive waste disposal facility or facilities for a period not to exceed 180 days.

              Low-level radioactive waste (LLW) means radioactive material that—

            (1) Is not high-level radioactive waste, spent nuclear fuel, or byproduct material (as defined in paragraphs (2), (3), and (4) of the definition of Byproduct Material set forth in § 20.1003 of this chapter); and

            (2) The NRC, consistent with existing law and in accordance with paragraph (1) of this definition, classifies as low-level radioactive waste.

              Non-Federal disposal facility means a low-level radioactive waste disposal facility that is commercially operated or is operated by a State.

              Person means any individual, corporation, partnership, firm, association, trust, State, public or private institution, group or agency who is an NRC or NRC Agreement State licensed generator of low-level radioactive waste within the scope of § 62.1(c) of this part; any Governor (or for any State without a Governor, the chief executive officer of the State) on behalf of any NRC or NRC Agreement State licensed generator or generators of low-level radioactive waste within the scope of § 62.1(c) of this part located in his or her State; or their duly authorized representative, legal successor, or agent.

              Regional disposal facility means a non-Federal low-level radioactive waste disposal facility in operation on January 1, 1985, or subsequently established and operated under a compact.

              State means any State of the United States, the District of Columbia, and the Commonwealth of Puerto Rico.

              Temporary emergency access means access that is granted at NRC's discretion under § 62.23 of this part upon determining that access is necessary to eliminate an immediate and serious threat to the public health and safety or the common defense and security. Such access expires 45 days after the granting and cannot be extended.
            [54 FR 5420, Feb. 3, 1989, as amended at 72 FR 55933, Oct. 1, 2007]

            § 62.3
            Communications.

            Except where otherwise specified, all communications and reports concerning the regulations in this part and applications filed under them should be sent by mail addressed: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.
            [73 FR 5725, Jan. 31, 2008, as amended at 74 FR 62683, Dec. 1, 2009; 79 FR 75740, Dec. 19, 2014; 80 FR 74980, Dec. 1, 2015]

            § 62.4
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be considered binding on the Commission.

            § 62.5
            Specific exemptions.
            The Commission may, upon application of any interested person or upon its own initiative, grant an exemption from the requirements of the regulations in this part that it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

            § 62.8
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0143.
            (b) The approved information collection requirements contained in this part appear in §§ 62.11, 62.12, 62.13, 62.14, and 62.15.
            [54 FR 5420, Feb. 3, 1989, as amended at 62 FR 52188, Oct. 6, 1997]

          Subpart B—Request for a Commission Determination

            § 62.11
            Filing and distribution of a determination request.
            (a) The person submitting a request for a Commission determination shall file a signed original of the request with the Commission at the address specified in § 62.3 of this part, with a copy also provided to the appropriate Regional Administrator at the address specified in appendix D to part 20 of this chapter. The request must be signed by the person requesting the determination or the person's authorized representative under oath or affirmation.

            (b) Upon receipt of a request for a determination, the Secretary of the Commission shall publish a notice acknowledging receipt of the request in the Federal Register. The notice must require that public comment on the request be submitted within 10 days of the publication date of the notice. A copy of the request will be made available for inspection or copying at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public Document Room. The Secretary of the Commission shall also transmit a copy of the request to the U.S. Department of Energy, to the Governors of the States of the Compact region where the waste is generated, to the Governors of the States with operating non-Federal low-level radioactive waste disposal facilities, to the Compact Commissions with operating regional low-level radioactive waste disposal facilities, and to the Governors of the States in the Compact Commissions with operating disposal facilities.
            (c) Upon receipt of a request for a determination based on a serious and immediate threat to the common defense and security, the Commission will notify DOD and/or DOE and provide a copy of the request as needed for their consideration.
            (d) Fees applicable to a request for a Commission determination under this part will be determined in accordance with the procedures set forth for special projects under category 12 of § 170.31 of this chapter.
            (e) In the event that the allocations or limitations established in section 5(b) or 6(h) of the Act are met at all operating non-Federal or regional LLW disposal facilities, the Commission may suspend the processing or acceptance of requests for emergency access determinations until additional LLW disposal capacity is authorized by Congress.
            [54 FR 5420, Feb. 3, 1989, as amended at 64 FR 48954, Sept. 9, 1999; 68 FR 58814, Oct. 10, 2003]

            § 62.12
            Contents of a request for emergency access: General information.
            A request for a Commission determination under this part must include the following information for each generator to which the request applies:
            (a) Name and address of the person making the request;
            (b) Name and address of the person(s) or company(ies) generating the low-level radioactive waste for which the determination is sought;
            (c) A statement indicating whether the generator is basing the request on the grounds of a serious and immediate threat to the public health and safety or the common defense and security;
            (d) Certification that the radioactive waste for which emergency access is requested is low-level radioactive waste within § 62.1(c) of this part;
            (e) The low-level waste generation facility(ies) producing the waste for which the request is being made;
            (f) A description of the activity that generated the waste;
            (g) Name of the disposal facility or facilities which had been receiving the waste stream of concern before the generator was denied access;
            (h) A description of the low-level radioactive waste for which emergency access is requested, including—
            (1) The characteristics and composition of the waste, including, but not limited to—

            (i) Type of waste (e.g. solidified oil, scintillation fluid, failed equipment);

            (ii) Principal chemical composition;
            (iii) Physical state (solid, liquid, gas);
            (iv) Type of solidification media; and
            (v) Concentrations and percentages of any hazardous or toxic chemicals, chelating agents, or infectious or biological agents associated with the waste;
            (2) The radiological characteristics of the waste such as—
            (i) The classification of the waste in accordance with 61.55;
            (ii) A list of the radionuclides present or potentially present in the waste, their concentration or contamination levels, and total quantity;
            (iii) Distribution of the radionuclides within the waste (surface or volume distribution);
            (iv) Amount of transuranics (nanocuries/gram);
            (3) The minimum volume of the waste requiring emergency access to eliminate the threat to the public health and safety or the common defense and security;
            (4) The time duration for which emergency access is requested (not to exceed 180 days);
            (5) Type of disposal container or packaging (55 gallon drum, box, liner, etc.); and
            (6) Description of the volume reduction and waste minimization techniques applied to the waste which assure that it is reduced to the maximum extent practicable, and the actual reduction in volume that occurred;
            (i) Basis for requesting the determination set out in this part, including—
            (1) The circumstances that led to the denial of access to existing low-level radioactive waste disposal facilities;
            (2) A description of the situation that is responsible for creating the serious and immediate threat to the public health and safety or the common defense and security, including the date when the need for emergency access was identified;
            (3) A chronology and description of the actions taken by the person requesting emergency access to prevent the need for making such a request, including consideration of all alternatives set forth in § 62.13 of this part, and any supporting documentation as appropriate;
            (4) An explanation of the impacts of the waste on the public health and safety or the common defense and security if emergency access is not granted, and the basis for concluding that these impacts constitute a serious and immediate threat to the public health and safety or the common defense and security. The impacts to the public health and safety or the common defense and security should also be addressed if the generator's services, including research activities, were to be curtailed, either for a limited period of time or indefinitely;
            (5) Other consequences if emergency access is not granted;
            (j) Steps taken by the person requesting emergency access to correct the situation requiring emergency access and the person's plans to eliminate the need for additional or future emergency access requests;
            (k) Documentation certifying that access has been denied;
            (l) Documentation that the waste for which emergency access is requested could not otherwise qualify for disposal pursuant to the Unusual Volumes provision (Section 5(c)(5) of the Act) or is not simultaneously under consideration by the Department of Energy (DOE) for access through the Unusual Volumes allocation;
            (m) Date by which access is required;
            (n) Any other information which the Commission should consider in making its determination.

            § 62.13
            Contents of a request for emergency access: Alternatives.
            (a) A request for emergency access under this part must include information on alternatives to emergency access. The request shall include a discussion of the consideration given to any alternatives, including, but not limited to, the following:
            (1) Storage of low-level radioactive waste at the site of generation;
            (2) Storage of low-level radioactive waste in a licensed storage facility;
            (3) Obtaining access to a disposal facility by voluntary agreement;

            (4) Purchasing disposal capacity available for assignment pursuant to the Act;

            (5) Requesting disposal at a Federal low-level radioactive waste disposal facility in the case of a Federal or defense related generator of LLW;
            (6) Reducing the volume of the waste;
            (7) Ceasing activities that generate low-level radioactive waste; and
            (8) Other alternatives identified under paragraph (b) of this section.
            (b) The request must identify all of the alternatives to emergency access considered, including any that would require State or Compact action, or any others that are not specified in paragraph (a) of this section. The request should also include a description of the process used to identify the alternatives, a description of the factors that were considered in identifying and evaluating them, a chronology of actions taken to identify and implement alternatives during the process, and a discussion of any actions that were considered, but not implemented.
            (c) The evaluation of each alternative must consider:
            (1) Its potential for mitigating the serious and immediate threat to public health and safety or the common defense and security posed by lack of access to disposal;
            (2) The adverse effects on public health and safety and the common defense and security, if any, of implementing each alternative, including the curtailment or cessation of any essential services affecting the public health and safety or the common defense and security;
            (3) The technical and economic feasibility of each alternative including the person's financial capability to implement the alternatives;
            (4) Any other pertinent societal costs and benefits;
            (5) Impacts to the environment;
            (6) Any legal impediments to implementation of each alternative, including whether the alternatives will comply with applicable NRC and NRC Agreement States regulatory requirements; and
            (7) The time required to develop and implement each alternative.
            (d) The request must include the basis for:
            (1) Rejecting each alternative; and
            (2) Concluding that no alternative is available.

            § 62.14
            Contents of a request for an extension of emergency access.
            A request for an extension of emergency access must include:
            (a) Updates of the information required in §§ 62.12 and 62.13; and
            (b) Documentation that the generator of the low-level radioactive waste granted emergency access and the State in which the low-level radioactive waste was generated have diligently, though unsuccessfully, acted during the period of the initial grant to eliminate the need for emergency access. Documentation must include:
            (1) An identification of additional alternatives that have been evaluated during the period of the initial grant, and
            (2) A discussion of any reevaluation of previously considered alternatives, including verification of continued attempts to gain access to a disposal facility by voluntary agreement.

            § 62.15
            Additional information.
            (a) The Commission may require additional information from a person making a request for a Commission determination under this part concerning any portion of the request.
            (b) The Commission shall deny a request for a Commission determination under this part if the person making the request fails to respond to a request for additional information under paragraph (a) of this section within ten (10) days from the date of the request for additional information, or any other time that the Commission may specify. This denial will not prejudice the right of the person making the request to file another request for a Commission determination under this part.

            § 62.16
            Withdrawal of a determination request.
            (a) A person may withdraw a request for a Commission determination under this part without prejudice at any time prior to the issuance of an initial determination under § 62.21 of this part.

            (b) The Secretary of the Commission will cause to be published in the Federal Register a notice of the withdrawal of a request for a Commission determination under this part.

            § 62.17
            Elimination of repetition.
            In any request under this part, the person making the request may incorporate by reference information contained in a previous application, Statement, or report filed with the Commission provided that these references are updated, clear, and specific.

            § 62.18
            Denial of request.
            If a request for a determination is based on circumstances that are too remote and speculative to allow an informed determination, the Commission may deny the request.

          Subpart C—Issuance of a Commission Determination

            § 62.21
            Determination for granting emergency access.
            (a) Not later than (45) days after the receipt of a request for a Commission determination under this part from any generator of low-level radioactive waste, or any Governor on behalf of any generator or generators located in his or her State, the Commission shall determine whether—
            (1) Emergency access to a regional disposal facility or a non-Federal disposal facility within a State that is not a member of a Compact for specific low-level radioactive waste is necessary because of an immediate and serious threat—
            (i) To the public health and safety or
            (ii) The common defense and security; and
            (2) The threat cannot be mitigated by any alternative consistent with the public health and safety, including those identified in § 62.13.
            (b) In making a determination under this section, the Commission shall be guided by the criteria set forth in § 62.25 of this part.
            (c) A determination under this section must be in writing and contain a full explanation of the facts upon which the determination is based and the reasons for granting or denying the request. An affirmative determination must designate an appropriate non-Federal or regional LLW disposal facility or facilities for the disposal of wastes, specifically describe the low-level radioactive waste as to source, physical and radiological characteristics, and the minimum volume and duration (not to exceed 180 days) necessary to eliminate the immediate threat to public health and safety or the common defense and security. It may also contain conditions upon which the determination is dependent.

            § 62.22
            Notice of issuance of a determination.
            (a) Upon the issuance of a Commission determination the Secretary of the Commission will notify in writing the following persons of the final determination: The person making the request, the Governor of the State in which the low-level radioactive waste requiring emergency access was generated, the Governor of the State in which the designated disposal facility is located, and if pertinent, the appropriate Compact Commission for such approval as is specified as necessary in section 6(g) of the Act. For the Governor of the State in which the designated disposal facility is located and for the appropriate Compact Commission, the notification must set forth the reasons that emergency access was granted and specifically describe the low-level radioactive waste as to source, physical and radiological characteristics, and the minimum volume and duration (not to exceed 180 days) necessary to alleviate the immediate and serious threat to public health and safety or the common defense and security. For the Governor of the State in which the low-level waste was generated, the notification must indicate that no extension of emergency access will be granted under § 62.24 of this part absent diligent State and generator action during the period of the initial grant.

            (b) The Secretary of the Commission will cause to be published in the Federal Register a notice of the issuance of the determination.

            (c) The Secretary of the Commission shall make a copy of the final determination available for inspection at the NRC Web site, http://www.nrc.gov.

            [54 FR 5420, Feb. 3, 1989, as amended at 64 FR 48954, Sept. 9, 1999]

            § 62.23
            Determination for granting temporary emergency access.
            (a) The Commission may grant temporary emergency access to an appropriate non-Federal or regional disposal facility or facilities provided that the determination required under § 62.21(a)(1) of this part is made;
            (b) The notification procedures under § 62.22 of this part are complied with; and
            (c) The temporary emergency access duration will not exceed forty-five (45) days.

            § 62.24
            Extension of emergency access.
            (a) After the receipt of a request from any generator of low-level waste, or any Governor on behalf of any generator or generators in his or her State, for an extension of emergency access that was initially granted under § 62.21, the Commission shall make an initial determination of whether—
            (1) Emergency access continues to be necessary because of an immediate and serious threat to the public health and safety or the common defense and security;
            (2) The threat cannot be mitigated by any alternative that is consistent with public health and safety; and
            (3) The generator of low-level waste and the State have diligently though unsuccessfully acted during the period of the initial grant to eliminate the need for emergency access.
            (b) After making a determination pursuant to paragraph (a) of this section, the requirements specified in §§ 62.21(c) and 62.22 of this part, must be followed.

            § 62.25
            Criteria for a Commission determination.
            (a) In making the determination required by § 62.21(a) of this part, the Commission will determine whether the circumstances described in the request for emergency access create a serious and immediate threat to the public health and safety or the common defense and security.
            (b) In making the determination that a serious and immediate threat exists to the public health and safety, the Commission will consider, notwithstanding the availability of any alternative identified in § 62.13 of this part:
            (1) The nature and extent of the radiation hazard that would result from the denial of emergency access, including consideration of—
            (i) The standards for radiation protection contained in part 20 of this chapter;
            (ii) Any standards governing the release of radioactive materials to the general environment that are applicable to the facility that generated the low level waste; and
            (iii) Any other Commission requirements specifically applicable to the facility or activity that is the subject of the emergency access request; and
            (2) The extent to which essential services affecting the public health and safety (such as medical, therapeutic, diagnostic, or research activities) will be disrupted by the denial of emergency access.
            (c) For purposes of granting temporary emergency access under § 62.23 of this part, the Commission will consider the criteria contained in the Commission's Policy Statement (45 FR 10950, February 24, 1977) for determining whether an event at a facility or activity licensed or otherwise regulated by the Commission is an abnormal occurrence within the purview of section 208 of the Energy Reorganization Act of 1974.
            (d) In making the determination that a serious and immediate threat to the common defense and security exists, the Commission will consider, notwithstanding the availability of any alternative identified in § 62.13 of this part:
            (1) Whether the activity generating the wastes is necessary to the protection of the common defense and security, and
            (2) Whether the lack of access to a disposal site would result in a significant disruption in that activity that would seriously threaten the common defense and security.

            The Commission will consider the views of the Department of Defense (DOD) and or the Department of Energy (DOE) regarding the importance of the activities responsible for generating the LLW to the common defense and security, when evaluating requests based all, or in part, on a serious and immediate threat to the common defense and security.

            (e) In making the determination required by § 62.21(a)(2) of this part, the Commission will consider whether the person submitting the request—
            (1) Has identified and evaluated any alternative that could mitigate the need for emergency access; and
            (2) Has considered all pertinent factors in its evaluation of alternatives including state-of-the-art technology and impacts on public health and safety.
            (f) In making the determination required by § 62.21(a)(2) of this part, the Commission will consider implementation of an alternative to be unreasonable if:
            (1) It adversely affects public health and safety, the environment, or the common defense and security; or
            (2) It results in a significant curtailment or cessation of essential services, affecting public health and safety or the common defense and security; or
            (3) It is beyond the technical and economic capabilities of the person requesting emergency access; or
            (4) Implementation of the alternative would conflict with applicable State or local or Federal laws and regulations; or
            (5) It cannot be implemented in a timely manner.
            (g) The Commission shall make an affirmative determination under § 62.21(a) of this part only if all of the alternatives that were considered are found to be unreasonable.
            (h) As part of its mandated evaluation of the alternatives that were considered by the generator, the Commission shall consider the characteristics of the wastes (including: physical properties, chemical properties, radioactivity, pathogenicity, infectiousness, and toxicity, pyrophoricity, and explosive potential); condition of current container; potential for contaminating the disposal site; the technologies or combination of technologies available for treatment of the waste (including incinerators; evaporators-crystallizers; fluidized bed dryers; thin film evaporators; extruders, evaporators; and Compactors); the suitability of volume reduction equipment to the circumstances (specific activity considerations, actual volume reduction factors, generation of secondary wastes, equipment contamination, effluent releases, worker exposure, and equipment availability); and the administrative controls which could be applied, in making a determination whether waste to be delivered for disposal under this part has been reduced in volume to the maximum extent practicable using available technology.

            § 62.26
            Criteria for designating a disposal facility.
            (a) The Commission shall designate an appropriate non-Federal or regional disposal facility if an affirmative determination is made pursuant to §§ 62.21, 62.23, or 62.24 of this part.
            (b) The Commission will exclude a disposal facility from consideration if:
            (1) The low-level radioactive wastes of the generator do not meet the criteria established by the license agreement or the license agreement of the facility; or
            (2) The disposal facility is in excess of its approved capacity; or
            (3) Granting emergency access would delay the closing of the disposal facility pursuant to plans established before the receipt of the request for emergency access; or
            (4) The volume of waste requiring emergency access exceeds 20 percent of the total volume of low-level radioactive waste accepted for disposal at the facility during the previous calendar year.
            (c) If, after applying the exclusionary criteria in paragraph (b) of this section, more than one disposal facility is identified as appropriate for designation, the Commission will then consider additional factors in designating a facility or facilities including—
            (1) Type of waste and its characteristics,
            (2) Previous disposal practices,
            (3) Transportation
            (4) Radiological effects,
            (5) Site capability for handling waste,
            (6) The volume of emergency access waste previously accepted by each site both for the particular year and overall, and
            (7) Any other considerations deemed appropriate by the Commission.

            (d) The Commission, in making its designation, will also consider any information submitted by the operating non-Federal or regional LLW disposal sites, or any information submitted by the public in response to a Federal Register notice requesting comment, as provided in paragraph (b) of § 62.11 of this part.

          Subpart D—Termination of Emergency Access

            § 62.31
            Termination of emergency access.
            (a) The Commission may terminate a grant of emergency access when emergency access is no longer necessary to eliminate an immediate threat to public health and safety or the common defense and security.
            (b) The Commission may terminate a grant of emergency access if an applicant has provided inaccurate information in its application for emergency access or if the applicant has failed to comply with this part or any conditions set by the Commission pursuant to this part.

        Pt. 63
        PART 63—DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN A GEOLOGIC REPOSITORY AT YUCCA MOUNTAIN, NEVADA

            Subpart A—General Provisions
            Sec.
            63.1
            Purpose and scope.
            63.2
            Definitions.
            63.3
            License required.
            63.4
            Communications and records.
            63.5
            Interpretations.
            63.6
            Exemptions.
            63.7
            License not required for certain preliminary activities.
            63.8
            Information collection requirements: OMB approval.
            63.9
            Employee protection.
            63.10
            Completeness and accuracy of information.
            63.11
            Deliberate misconduct.

            Subpart B—Licenses

              Preapplication Review
              63.15
              Site characterization.
              63.16
              Review of site characterization activities.

              License Application
              63.21
              Content of application.
              63.22
              Filing and distribution of application.
              63.23
              Elimination of repetition.
              63.24
              Updating of application and environmental impact statement.

              Construction Authorization
              63.31
              Construction authorization.
              63.32
              Conditions of construction authorization.
              63.33
              Amendment of construction authorization.

              License Issuance and Amendment
              63.41
              Standards for issuance of a license.
              63.42
              Conditions of license.
              63.43
              License specification.
              63.44
              Changes, tests, and experiments.
              63.45
              Amendment of license.
              63.46
              Particular activities requiring license amendment.

              US/IAEA Safeguards Agreement
              63.47
              Facility information and verification.

              Permanent Closure
              63.51
              License amendment for permanent closure.
              63.52
              Termination of license.

            Subpart C—Participation by State Government, Affected Units of Local Government, and Affected Indian Tribes
            63.61
            Provision of information.
            63.62
            Site review.
            63.63
            Participation in license reviews.
            63.64
            Notice to State.
            63.65
            Representation.

            Subpart D—Records, Reports, Tests, and Inspections
            63.71
            Records and reports.
            63.72
            Construction records.
            63.73
            Reports of deficiencies.
            63.74
            Tests.
            63.75
            Inspections.
            63.78
            Material control and accounting records and reports.

            Subpart E—Technical Criteria
            63.101
            Purpose and nature of findings.
            63.102
            Concepts.

              Preclosure Performance Objectives
              63.111
              Performance objectives for the geologic repository operations area through permanent closure.

              Preclosure Safety Analysis
              63.112
              Requirements for preclosure safety analysis of the geologic repository operations area.

              Postclosure Performance Objectives
              63.113
              Performance objectives for the geologic repository after permanent closure.

              Postclosure Performance Assessment
              63.114
              Requirements for performance assessment.
              63.115
              Requirements for multiple barriers.

              Land Ownership and Control
              63.121
              Requirements for ownership and control of interests in land.

            Subpart F—Performance Confirmation Program
            63.131
            General requirements.
            63.132
            Confirmation of geotechnical and design parameters.
            63.133
            Design testing.
            63.134
            Monitoring and testing waste packages.

            Subpart G—Quality Assurance
            63.141
            Scope.
            63.142
            Quality assurance criteria.
            63.143
            Implementation.
            63.144
            Quality assurance program change.

            Subpart H—Training and Certification of Personnel
            63.151
            General requirements.
            63.152
            Training and certification program.
            63.153
            Physical requirements.

            Subpart I—Emergency Planning Criteria
            63.161
            Emergency plan for the geologic repository operations area through permanent closure.

            Subpart J—Violations
            63.171
            Violations.
            63.172
            Criminal penalties.

            Subpart K—Preclosure Public Health and Environmental Standards
            63.201
            Purpose and scope.
            63.202
            Definitions for Subpart K.
            63.203
            Implementation of Subpart K.
            63.204
            Preclosure standard.

            Subpart L—Postclosure Public Health and Environmental Standards
            63.301
            Purpose and scope.
            63.302
            Definitions for Subpart L.
            63.303
            Implementation of Subpart L.
            63.304
            Reasonable expectation.
            63.305
            Required characteristics of the reference biosphere.

              Postclosure Individual Protection Standard
              63.311
              Individual protection standard after permanent closure.
              63.312
              Required characteristics of the reasonably maximally exposed individual.

              Human Intrusion Standard
              63.321
              Individual protection standard for human intrusion.
              63.322
              Human intrusion scenario.

              Groundwater Protection Standards
              63.331
              Separate standards for protection of groundwater.
              63.332
              Representative volume.

              Additional Provisions
              63.342
              Limits on performance assessments.
              63.343
              Severability of individual protection and groundwater protection standards.

          Authority:
          Atomic Energy Act of 1954, secs. 51, 53, 62, 63, 65, 81, 161, 182, 183, 223, 234 (42 U.S.C. 2071, 2073, 2092, 2093, 2095, 2111, 2201, 2232, 2233, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 42 U.S.C. 2021a; National Environmental Policy Act of 1969 (42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 114, 117, 121 (42 U.S.C. 10134, 10137, 10141); 44 U.S.C. 3504 note.

          Source:
          66 FR 55792, Nov. 2, 2001, unless otherwise noted.

          Editorial Note:
          Nomenclature changes to part 63 appear at 81 FR 86909, Dec. 2, 2016.

          Subpart A—General Provisions

            § 63.1
            Purpose and scope.
            This part prescribes rules governing the licensing (including issuance of a construction authorization) of the U.S. Department of Energy to receive and possess source, special nuclear, and byproduct material at a geologic repository operations area sited, constructed, or operated at Yucca Mountain, Nevada, in accordance with the Nuclear Waste Policy Act of 1982, as amended, and the Energy Policy Act of 1992. As provided in 10 CFR 60.1, the regulations in part 60 of this chapter do not apply to any activity licensed under another part of this chapter. This part also gives notice to all persons who knowingly provide to any licensee, applicant, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licensee's or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 63.11.
            [69 FR 2280, Jan. 14, 2004]

            § 63.2
            Definitions.
            As used in this part:

              Affected Indian Tribe means any Indian Tribe within whose reservation boundaries a repository for high-level radioactive waste or spent fuel is proposed to be located; or whose Federally-defined possessory or usage rights to other lands outside of the reservation's boundaries arising out of Congressionally-ratified treaties or other Federal law may be substantially and adversely affected by the location of the facility if the Secretary of the Interior finds, on the petition of the appropriate governmental officials of the Tribe, that the effects are both substantial and adverse to the Tribe.

              Barrier means any material, structure, or feature that, for a period to be determined by NRC, prevents or substantially reduces the rate of movement of water or radionuclides from the Yucca Mountain repository to the accessible environment, or prevents the release or substantially reduces the release rate of radionuclides from the waste. For example, a barrier may be a geologic feature, an engineered structure, a canister, a waste form with physical and chemical characteristics that significantly decrease the mobility of radionuclides, or a material placed over and around the waste, provided that the material substantially delays movement of water or radionuclides.

              Commencement of construction means clearing of land, surface or subsurface excavation, or other substantial action that would adversely affect the environment of a site. It does not include changes desirable for the temporary use of the land for public recreational uses, site characterization activities, other preconstruction monitoring and investigation necessary to establish background information related to the suitability of the Yucca Mountain site or to the protection of environmental values, or procurement or manufacture of components of the geologic repository operations area.

              Commission means the Nuclear Regulatory Commission or its duly authorized representatives.

              Containment means the confinement of radioactive waste within a designated boundary.

              Design bases means that information that identifies the specific functions to be performed by a structure, system, or component of a facility and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be constraints derived from generally accepted “state-of-the-art” practices for achieving functional goals or requirements derived from analysis (based on calculation or experiments) of the effects of a postulated event under which a structure, system, or component must meet its functional goals. The values for controlling parameters for external events include:
            (1) Estimates of severe natural events to be used for deriving design bases that will be based on consideration of historical data on the associated parameters, physical data, or analysis of upper limits of the physical processes involved; and
            (2) Estimates of severe external human-induced events to be used for deriving design bases, that will be based on analysis of human activity in the region, taking into account the site characteristics and the risks associated with the event.

              Director means the Director of the Nuclear Regulatory Commission's Office of Nuclear Material Safety and Safeguards.

              Disposal means the emplacement of radioactive waste in a geologic repository with the intent of leaving it there permanently.

              DOE means the U.S. Department of Energy or its duly authorized representatives.

              Engineered barrier system means the waste packages, including engineered components and systems other than the waste package (e.g., drip shields), and the underground facility.

              Event sequence means a series of actions and/or occurrences within the natural and engineered components of a geologic repository operations area that could potentially lead to exposure of individuals to radiation. An event sequence includes one or more initiating events and associated combinations of repository system component failures, including those produced by the action or inaction of operating personnel. Those event sequences that are expected to occur one or more times before permanent closure of the geologic repository operations area are referred to as Category 1 event sequences. Other event sequences that have at least one chance in 10,000 of occurring before permanent closure are referred to as Category 2 event sequences.

              Geologic repository means a system that is intended to be used for, or may be used for, the disposal of radioactive wastes in excavated geologic media. A geologic repository includes the engineered barrier system and the portion of the geologic setting that provides isolation of the radioactive waste.

              Geologic repository operations area means a high-level radioactive waste facility that is part of a geologic repository, including both surface and subsurface areas, where waste handling activities are conducted.

              Geologic setting means the geologic, hydrologic, and geochemical systems of the region in which a geologic repository is or may be located.

              High-level radioactive waste or HLW means:
            (1) The highly radioactive material resulting from the reprocessing of spent nuclear fuel, including liquid waste produced directly in reprocessing and any solid material derived from such liquid waste that contains fission products in sufficient concentrations;
            (2) Irradiated reactor fuel; and
            (3) Other highly radioactive material that the Commission, consistent with existing law, determines by rule requires permanent isolation.

              HLW facility means a facility subject to the licensing and related regulatory authority of the Commission pursuant to sections 202(3) and 202(4) of the Energy Reorganization Act of 1974 (88 Stat. 1244). 1

                1 These are DOE “facilities used primarily for the receipt and storage of high-level radioactive wastes resulting from activities licensed under such Act (the Atomic Energy Act)” and “Retrievable Surface Storage Facilities and other facilities authorized for the express purpose of subsequent long-term storage of high-level radioactive wastes generated by (DOE), which are not used for, or are part of, research and development activities.”

              Host rock means the geologic medium in which the waste is emplaced.

              Important to safety, with reference to structures, systems, and components, means those engineered features of the geologic repository operations area whose function is:
            (1) To provide reasonable assurance that high-level waste can be received, handled, packaged, stored, emplaced, and retrieved without exceeding the requirements of § 63.111(b)(1) for Category 1 event sequences; or
            (2) To prevent or mitigate Category 2 event sequences that could result in radiological exposures exceeding the values specified at § 63.111(b)(2) to any individual located on or beyond any point on the boundary of the site.

              Important to waste isolation, with reference to design of the engineered barrier system and characterization of natural barriers, means those engineered and natural barriers whose function is to provide a reasonable expectation that high-level waste can be disposed of without exceeding the requirements of § 63.113(b) and (c).

              Initiating event means a natural or human induced event that causes an event sequence.

              Isolation means inhibiting the transport of radioactive material to:
            (1) The location of the reasonably maximally exposed individual so that radiological exposures will not exceed the requirements of § 63.113(b); and
            (2) The accessible environment so that releases of radionuclides into the accessible environment will not exceed the requirements of § 63.113(c).

              Performance assessment means an analysis that:
            (1) Identifies the features, events, processes (except human intrusion), and sequences of events and processes (except human intrusion) that might affect the Yucca Mountain disposal system and their probabilities of occurring;
            (2) Examines the effects of those features, events, processes, and sequences of events and processes upon the performance of the Yucca Mountain disposal system; and

            (3) Estimates the dose incurred by the reasonably maximally exposed individual, including the associated uncertainties, as a result of releases caused by all significant features, events, processes, and sequences of events and processes, weighted by their probability of occurrence.

              Performance confirmation means the program of tests, experiments, and analyses that is conducted to evaluate the adequacy of the information used to demonstrate compliance with the performance objectives in subpart E of this part.

              Permanent closure means final backfilling of the underground facility, if appropriate, and the sealing of shafts, ramps, and boreholes.

              Preclosure safety analysis means a systematic examination of the site; the design; and the potential hazards, initiating events and event sequences and their consequences (e.g., radiological exposures to workers and the public). The analysis identifies structures, systems, and components important to safety.

              Public Document Room means the place at One White Flint North, 11555 Rockville Pike, Room O-1F13, Rockville, MD, at which records of the Commission will ordinarily be made available for public inspection and any other place, the location of which has been published in the Federal Register, at which public records of the Commission pertaining to a geologic repository at the Yucca Mountain site are made available for public inspection.

              Publicly Available Records System (PARS) Library means the electronic library generated by the NRC's Agencywide Documents Access and Management System (ADAMS) to provide access to public documents. PARS has full text documents which can be searched using specific fields and parameters. The public can search, download, print, create reports, and order documents online. The PARS Library contains publicly available documents created or received by NRC since November 1, 1999, as well as some older documents that the NRC has retrofit into the collection. PARS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm.html.

              Radioactive waste or waste means HLW and radioactive materials other than HLW that are received for emplacement in a geologic repository.

              Reasonably maximally exposed individual means the hypothetical person meeting the criteria specified at § 63.312.

              Reference biosphere means the description of the environment inhabited by the reasonably maximally exposed individual. The reference biosphere comprises the set of specific biotic and abiotic characteristics of the environment, including, but not necessarily limited to, climate, topography, soils, flora, fauna, and human activities.

              Restricted area means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set aside as a restricted area.

              Retrieval means the act of permanently removing radioactive waste from the underground location at which the waste had been previously emplaced for disposal.

              Saturated zone means that part of the earth's crust beneath the regional water table in which statistically all voids, large and small, are filled with water under pressure greater than atmospheric.

              Site means that area surrounding the geologic repository operations area for which DOE exercises authority over its use in accordance with the provisions of this part.

              Site characterization means the program of exploration and research, both in the laboratory and in the field, undertaken to establish the geologic conditions and the ranges of those parameters of the Yucca Mountain site, and the surrounding region to the extent necessary, relevant to the procedures under this part. Site characterization includes borings, surface excavations, excavation of exploratory shafts and/or ramps, limited subsurface lateral excavations and borings, and in situ testing at depth needed to determine the suitability of the site for a geologic repository.

              Total effective dose equivalent (TEDE) means the sum of the effective dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

              Underground facility means the underground structure, backfill materials, if any, and openings that penetrate the underground structure (e.g., ramps, shafts, and boreholes, including their seals).

              Unrestricted area means an area, access to which is neither limited nor controlled by the licensee.

              Unsaturated zone means the zone between the land surface and the regional water table. Generally, fluid pressure in this zone is less than atmospheric pressure, and some of the voids may contain air or other gases at atmospheric pressure. Beneath flooded areas or in perched water bodies, the fluid pressure locally may be greater than atmospheric.

              Waste form means the radioactive waste materials and any encapsulating or stabilizing matrix.

              Waste package means the waste form and any containers, shielding, packing, and other absorbent materials immediately surrounding an individual waste container.

              Water table means that surface in a groundwater body, separating the unsaturated zone from the saturated zone, at which the water pressure is atmospheric.
            [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003; 74 FR 10828, Mar. 13, 2009]

            § 63.3
            License required.
            (a) DOE may not receive nor possess source, special nuclear, or byproduct material at a geologic repository operations area at the Yucca Mountain site except as authorized by a license issued by the Commission under this part.
            (b) DOE may not begin construction of a geologic repository operations area at the Yucca Mountain site unless it has filed an application with the Commission and has obtained construction authorization as provided in this part. Failure to comply with this requirement is grounds for denial of a license.

            § 63.4
            Communications and records.
            (a) Except as otherwise specified, in this part or in subpart J of part 2 of this chapter, all communications and reports concerning the regulations in this part and applications filed under them should be sent to the NRC as follows:
            (1) By mail addressed: ATTN: Document Control Desk; Director, Office of Nuclear Material Safety and Safeguards; U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001;
            (2) By hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; ATTN: Document Control Desk: Director, Office of Nuclear Material Safety and Safeguards; or,

            (3) Where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.
            (b) Each record required by this part must be legible throughout the retention period specified by each Commission regulation. The record may be the original or a reproduced copy or a microform if the copy or microform is authenticated by authorized personnel and the microform is capable of producing a clear copy throughout the required retention period. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.
            [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003; 74 FR 62683, Dec. 1, 2009; 80 FR 74981, Dec. 1, 2015]

            § 63.5
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel is binding on the Commission.

            § 63.6
            Exemptions.
            The Commission may, upon application by DOE, any interested person, or upon its own initiative, grant an exemption from the requirements of this part if it determines that the exemption is authorized by law, does not endanger life nor property nor the common defense and security, and is otherwise in the public interest.

            § 63.7
            License not required for certain preliminary activities.
            The requirement for a license set forth in § 63.3(a) is not applicable to the extent that DOE receives and possesses source, special nuclear, and byproduct material at a geologic repository at the Yucca Mountain site:
            (a) For purposes of site characterization; or
            (b) For use, during site characterization or construction, as components of radiographic, radiation monitoring, or similar equipment or instrumentation.

            § 63.8
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501, et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0199.
            (b) The approved information collection requirements contained in this Part appear in §§ 63.47, 63.62, 63.63, and 63.65.
            (c) In § 63.47, Form N-71 and associated forms are approved under control number 3150-0056, and DOC/NRC Forms AP-1, AP-A, and associated forms are approved under control numbers 0694-0135.
            [66 FR 55792, Nov. 2, 2001, as amended at 73 FR 78606, Dec. 23, 2008]

            § 63.9
            Employee protection.
            (a) Discrimination by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant, against an employee, for engaging in certain protected activities, is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act or the Energy Reorganization Act.
            (1) The protected activities include but are not limited to:
            (i) Providing the Commission, or his or her employer, information about alleged violations of either of the statutes named in paragraph (a) of this section or possible violations of requirements imposed under either of those aforementioned statutes;
            (ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) of this section, or under these requirements, if the employee has identified the alleged illegality to the employer;
            (iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;
            (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) of this section;
            (v) Assisting or participating in, or is about to assist or participate in, these activities.
            (2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.

            (3) This section does not apply to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.
            (b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.
            (c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant may be grounds for—
            (1) Denial, revocation, or suspension of the license;
            (2) Imposition of a civil penalty on the licensee, applicant, or a contractor or subcontractor of the licensee or applicant; or
            (3) Other enforcement action.
            (d) Actions taken by an employer, or others, that adversely affect an employee, may be predicated on nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.
            (e)(1) Each licensee and each applicant for a license shall prominently post the revision of NRC Form 3, “Notice to Employees,” referenced in § 19.11(c) of this chapter. This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, and for 30 days following license termination.

            (2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, via email to Forms.Resource@nrc.gov, or by visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.

            (f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision that would prohibit, restrict, or otherwise discourage an employee from participating in a protected activity as defined in paragraph (a)(1) of this section, including, but not limited to, providing information to NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.
            [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003; 72 FR 63974, Nov. 14, 2007; 73 FR 30459, May 28, 2008; 79 FR 66605, Nov. 10, 2014]

            § 63.10
            Completeness and accuracy of information.
            (a) Information provided to the Commission by an applicant for a license or by a licensee, or information required by statute, or required by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee must be complete and accurate in all material respects.

            (b) The applicant or licensee shall notify the Commission of information identified by the applicant or licensee as having, for the regulated activity, a significant implication for public health and safety or common defense and security. An applicant or licensee violates this paragraph only if the applicant or licensee fails to notify the Commission of information that the applicant or licensee has identified as having a significant implication for public health and safety or common defense and security. Notification must be provided to the Director of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, within 2 working days of identifying the information. This requirement is not applicable to information that is already required to be provided to the Commission by other reporting or updating requirements.

            § 63.11
            Deliberate misconduct.
            (a) Any licensee, applicant for a license, employee of a licensee or applicant; or any contractor (including a supplier or consultant), subcontractor, employee of a contractor or subcontractor of any licensee or applicant for a license, who knowingly provides to any licensee, applicant, contractor, or subcontractor, any components, equipment, materials, or other goods or services that relate to a licensee's or applicant's activities in this part, may not:
            (1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission; or
            (2) Deliberately submit to NRC, a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to NRC.
            (b) A person who violates paragraph (a)(1) or (a)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.
            (c) For purposes of paragraph (a)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:
            (1) Would cause a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license issued by the Commission; or
            (2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor.

          Subpart B—Licenses

            Preapplication Review

              § 63.15
              Site characterization.
              (a) DOE shall conduct a program of site characterization with respect to the Yucca Mountain site before it submits an application for a license to be issued under this part.
              (b) DOE shall conduct the investigations to obtain the required information in a manner that limits adverse effects on the long-term performance of the geologic repository at Yucca Mountain to the extent practical.

              § 63.16
              Review of site characterization activities. 2

                  2 In addition to the review of site characterization activities specified in this section, the Commission contemplates an ongoing review of other information on site investigation and site characterization, to allow early identification of potential licensing issues for timely resolution at the staff level.

              (a) If DOE's planned site characterization activities include onsite testing with radioactive material, including radioactive tracers, the Commission shall determine whether the proposed use of such radioactive material is necessary to provide data for the preparation of the environmental reports required by law and for an application to be submitted under § 63.22.

              (b) During the conduct of site characterization activities at the Yucca Mountain site, DOE shall report the nature and extent of the activities, the information that has been developed, and the progress of waste form and waste package research and development to the Commission not less than once every 6 months. The semiannual reports must include the results of site characterization studies, the identification of new issues, plans for additional studies to resolve new issues, elimination of planned studies no longer necessary, identification of decision points reached, and modifications to schedules, where appropriate. DOE shall also report its progress in developing the design of a geologic repository operations area appropriate for the area being characterized, noting when key design parameters or features that depend on the results of site characterization will be established. Other topics related to site characterization must also be covered if requested by the Director.
              (c) During the conduct of site characterization activities at the Yucca Mountain site, NRC staff shall be permitted to visit and inspect the locations at which such activities are carried out and to observe excavations, borings, and in situ tests, as they are done.
              (d) The Director may comment at any time in writing to DOE, expressing current views on any aspect of site characterization or performance assessment at the Yucca Mountain site. In particular, the Director shall comment whenever he or she determines that there are substantial grounds for making recommendations or stating objections to DOE's site characterization program. The Director shall invite public comment on any comments that the Director makes to DOE on review of the DOE semiannual reports or on any other comments that the Director makes to DOE on site characterization and performance assessment by placing the comments in a public forum to allow the public to comment on them after the Director's comments are sent to DOE.
              (e) The Director shall transmit copies of all comments to DOE made by the Director under this section to the Governor and legislature of the State of Nevada and to the governing body of any affected Indian Tribe.
              (f) The NRC shall place all correspondence between DOE and NRC resulting from the requirements of this section, including the reports described in paragraph (b) of this section, in the Publicly Available Records System (PARS) Library.
              (g) The activities described in paragraphs (a) through (f) of this section constitute informal conference between a prospective applicant and the NRC staff, as described in § 2.101(a)(1) of this chapter, and are not part of a proceeding under the Atomic Energy Act of 1954, as amended. Accordingly, the issuance of the Director's comments made under this section does not constitute a commitment to issue any authorization or license, or in any way affect the authority of the Commission, Atomic Safety and Licensing Board, other presiding officers, or the Director, in any such proceeding.
              [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003]

            License Application

              § 63.21
              Content of application.
              (a) An application consists of general information and a Safety Analysis Report. An environmental impact statement must be prepared in accordance with the Nuclear Waste Policy Act of 1982, as amended, and must accompany the application. Any Restricted Data or National Security Information must be separated from unclassified information. The application must be as complete as possible in the light of information that is reasonably available at the time of docketing.
              (b) The general information must include:
              (1) A general description of the proposed geologic repository at the Yucca Mountain site, identifying the location of the geologic repository operations area, the general character of the proposed activities, and the basis for the exercise of the Commission's licensing authority.
              (2) Proposed schedules for construction, receipt of waste, and emplacement of wastes at the proposed geologic repository operations area.
              (3) A description of the detailed security measures for physical protection of high-level radioactive waste in accordance with § 73.51 of this chapter. This plan must include the design for physical protection, the licensee's safeguards contingency plan, and security organization personnel training and qualification plan. The plan must list tests, inspections, audits, and other means to be used to demonstrate compliance with such requirements.

              (4) A description of the material control and accounting program to meet the requirements of § 63.78.

              (5) A description of work conducted to characterize the Yucca Mountain site.
              (c) The Safety Analysis Report must include:
              (1) A description of the Yucca Mountain site, with appropriate attention to those features, events, and processes of the site that might affect design of the geologic repository operations area and performance of the geologic repository. The description of the site must include information regarding features, events, and processes outside of the site to the extent the information is relevant and material to safety or performance of the geologic repository. The information referred to in this paragraph must include:
              (i) The location of the geologic repository operations area with respect to the boundary of the site;
              (ii) Information regarding the geology, hydrology, and geochemistry of the site, including geomechanical properties and conditions of the host rock;
              (iii) Information regarding surface water hydrology, climatology, and meteorology of the site; and
              (iv) Information regarding the location of the reasonably maximally exposed individual, and regarding local human behaviors and characteristics, as needed to support selection of conceptual models and parameters used for the reference biosphere and reasonably maximally exposed individual.
              (2) Information relative to materials of construction of the geologic repository operations area (including geologic media, general arrangement, and approximate dimensions), and codes and standards that DOE proposes to apply to the design and construction of the geologic repository operations area.
              (3) A description and discussion of the design of the various components of the geologic repository operations area and the engineered barrier system including:
              (i) Dimensions, material properties, specifications, analytical and design methods used along with any applicable codes and standards;
              (ii) The design criteria used and their relationships to the preclosure and postclosure performance objectives specified at § 63.111(b), § 63.113(b), and § 63.113(c); and
              (iii) The design bases and their relation to the design criteria.
              (4) A description of the kind, amount, and specifications of the radioactive material proposed to be received and possessed at the geologic repository operations area at the Yucca Mountain site.
              (5) A preclosure safety analysis of the geologic repository operations area, for the period before permanent closure, to ensure compliance with § 63.111(a), as required by § 63.111(c). For the purposes of this analysis, it is assumed that operations at the geologic repository operations area will be carried out at the maximum capacity and rate of receipt of radioactive waste stated in the application.
              (6) A description of the program for control and monitoring of radioactive effluents and occupational radiological exposures to maintain such effluents and exposures in accordance with the requirements of § 63.111.
              (7) A description of plans for retrieval and alternate storage of the radioactive wastes, should retrieval be necessary.
              (8) A description of design considerations that are intended to facilitate permanent closure and decontamination or decontamination and dismantlement of surface facilities.
              (9) An assessment to determine the degree to which those features, events, and processes of the site that are expected to materially affect compliance with § 63.113—whether beneficial or potentially adverse to performance of the geologic repository—have been characterized, and the extent to which they affect waste isolation. Investigations must extend from the surface to a depth sufficient to determine principal pathways for radionuclide migration from the underground facility. Specific features, events, and processes of the geologic setting must be investigated outside of the site if they affect performance of the geologic repository.

              (10) An assessment of the anticipated response of the geomechanical, hydrogeologic, and geochemical systems to the range of design thermal loadings under consideration, given the pattern of fractures and other discontinuities and the heat transfer properties of the rock mass and water.
              (11) An assessment of the ability of the proposed geologic repository to limit radiological exposures to the reasonably maximally exposed individual for the period after permanent closure, as required by § 63.113(b).
              (12) An assessment of the ability of the proposed geologic repository to limit releases of radionuclides into the accessible environment as required by § 63.113(c).
              (13) An assessment of the ability of the proposed geologic repository to limit radiological exposures to the reasonably maximally exposed individual for the period after permanent closure in the event of human intrusion into the engineered barrier system as required by § 63.113(d).
              (14) An evaluation of the natural features of the geologic setting and design features of the engineered barrier system that are considered barriers important to waste isolation as required by § 63.115.
              (15) An explanation of measures used to support the models used to provide the information required in paragraphs (c)(9) through (c)(14) of this section. Analyses and models that will be used to assess performance of the geologic repository must be supported by using an appropriate combination of such methods as field tests, in situ tests, laboratory tests that are representative of field conditions, monitoring data, and natural analog studies.
              (16) An identification of those structures, systems, and components of the geologic repository, both surface and subsurface, that require research and development to confirm the adequacy of design. For structures, systems, and components important to safety and for the engineered and natural barriers important to waste isolation, DOE shall provide a detailed description of the programs designed to resolve safety questions, including a schedule indicating when these questions would be resolved.
              (17) A description of the performance confirmation program that meets the requirements of subpart F of this part.
              (18) An identification and justification for the selection of those variables, conditions, or other items that are determined to be probable subjects of license specifications. Special attention must be given to those items that may significantly influence the final design.
              (19) An explanation of how expert elicitation was used.
              (20) A description of the quality assurance program to be applied to the structures, systems, and components important to safety and to the engineered and natural barriers important to waste isolation. The description of the quality assurance program must include a discussion of how the applicable requirements of § 63.142 will be satisfied.
              (21) A description of the plan for responding to, and recovering from, radiological emergencies that may occur at any time before permanent closure and decontamination or decontamination and dismantlement of surface facilities, as required by § 63.161.
              (22) The following information concerning activities at the geologic repository operations area:
              (i) The organizational structure of DOE as it pertains to construction and operation of the geologic repository operations area, including a description of any delegations of authority and assignments of responsibilities, whether in the form of regulations, administrative directives, contract provisions, or otherwise.
              (ii) Identification of key positions that are assigned responsibility for safety at and operation of the geologic repository operations area.
              (iii) Personnel qualifications and training requirements.
              (iv) Plans for startup activities and startup testing.
              (v) Plans for conduct of normal activities, including maintenance, surveillance, and periodic testing of structures, systems, and components of the geologic repository operations area.
              (vi) Plans for permanent closure and plans for the decontamination or decontamination and dismantlement of surface facilities.

              (vii) Plans for any uses of the geologic repository operations area at the Yucca Mountain site for purposes other than disposal of radioactive wastes, with an analysis of the effects, if any, that such uses may have on the operation of the structures, systems, and components important to safety and the engineered and natural barriers important to waste isolation.
              (23) A description of the program to be used to maintain the records described in §§ 63.71 and 63.72.
              (24) A description of the controls that DOE will apply to restrict access and to regulate land use at the Yucca Mountain site and adjacent areas, including a conceptual design of monuments that would be used to identify the site after permanent closure.
              (d) The applicant for a license to receive and possess source, special nuclear, and byproduct material at a geologic repository at Yucca Mountain, Nevada, shall protect Safeguards Information in accordance with the requirements in § 73.21, and the requirements in § 73.22, or § 73.23 of this chapter, as applicable, and shall protect classified information in accordance with the requirements of parts 25 and 95 of this chapter, as applicable.
              [66 FR 55792, Nov. 2, 2001, as amended at 73 FR 63572, Oct. 24, 2008]

              § 63.22
              Filing and distribution of application.
              (a) An application for a construction authorization for a high-level radioactive waste repository at a geologic repository operations area at Yucca Mountain, and an application for a license to receive and possess source, special nuclear, or byproduct material at a geologic repository operations area at the Yucca Mountain site that has been characterized, any amendments to the application, and an accompanying environmental impact statement and any supplements, must be signed by the Secretary of Energy or the Secretary's authorized representative and must be filed with the Director in triplicate on paper and optical storage media.
              (b) DOE shall submit 30 additional copies, on paper and optical storage media, of each portion of the application and any amendments, and each environmental impact statement and any supplements. DOE shall maintain the capability to generate additional copies for distribution in accordance with written instructions from the Director or the Director's designee.
              (c) On notification of the appointment of an Atomic Safety and Licensing Board, DOE shall update the application, eliminating all superseded information, and supplement the environmental impact statement if necessary, and serve the updated application and environmental impact statement (as it may have been supplemented) as directed by the Board. Any subsequent amendments to the application or supplements to the environmental impact statement must be served in the same manner.
              (d) When an application, and any amendment to it is filed, copies on paper and optical storage media must be made available in appropriate locations near the proposed geologic repository operations areas at the Yucca Mountain site for inspection by the public. These copies must be updated as amendments to the application are made. The environmental impact statement and any supplements to it must be made available in the same manner. An updated copy of the application, and the environmental impact statement and supplements, must be produced at any public hearing held by the Commission on the application for use by any party to the proceeding.
              (e) DOE shall certify that the updated copies of the application, and the environmental impact statement as it may have been supplemented, as referred to in paragraphs (c) and (d) of this section, contain the current contents of these documents submitted as required by this part.
              [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003; 69 FR 2280, Jan. 14, 2004]

              § 63.23
              Elimination of repetition.
              In its application or environmental impact statement, DOE may incorporate, by reference, information contained in previous applications, statements, or reports filed with the Commission, if the references are clear and specific and copies of the information incorporated are made available to the public locations near the site of the proposed geologic repository, as specified in § 63.22(d).

              § 63.24
              Updating of application and environmental impact statement.
              (a) The application must be as complete as possible in light of the information that is reasonably available at the time of docketing.
              (b) DOE shall update its application in a timely manner so as to permit the Commission to review, before issuance of a license—
              (1) Additional geologic, geophysical, geochemical, hydrologic, meteorologic, materials, design, and other data obtained during construction;
              (2) Conformance of construction of structures, systems, and components with the design;
              (3) Results of research programs carried out to confirm the adequacy of designs, conceptual models, parameter values, and estimates of performance of the geologic repository.
              (4) Other information bearing on the Commission's issuance of a license that was not available at the time a construction authorization was issued.
              (c) DOE shall supplement its environmental impact statement in a timely manner so as to take into account the environmental impacts of any substantial changes in its proposed actions or any significant new circumstances or information relevant to environmental concerns bearing on the proposed action or its impacts.

            Construction Authorization

              § 63.31
              Construction authorization.
              On review and consideration of an application and environmental impact statement submitted under this part, the Commission may authorize construction of a geologic repository operations area at the Yucca Mountain site if it determines:
              (a) Safety. (1) That there is reasonable assurance that the types and amounts of radioactive materials described in the application can be received and possessed in a geologic repository operations area of the design proposed without unreasonable risk to the health and safety of the public; and
              (2) That there is reasonable expectation that the materials can be disposed of without unreasonable risk to the health and safety of the public.
              (3) In arriving at these determinations, the Commission shall consider whether—
              (i) DOE has described the proposed geologic repository as specified at § 63.21;
              (ii) The site and design comply with the performance objectives and requirements contained in subpart E of this part;
              (iii) DOE's quality assurance program complies with the requirements of subpart G of this part;
              (iv) DOE's personnel training program complies with the criteria contained in subpart H of this part;
              (v) DOE's emergency plan complies with the criteria contained in subpart I of this part; and
              (vi) DOE's proposed operating procedures to protect health and to minimize danger to life or property are adequate.
              (b) Common defense and security. That there is reasonable assurance that the activities proposed in the application will not be inimical to the common defense and security.
              (c) Environmental. That, after weighing the environmental, economic, technical, and other benefits against environmental costs, and considering available alternatives, the action called for is the issuance of the construction authorization, with any appropriate conditions to protect environmental values.

              § 63.32
              Conditions of construction authorization.
              (a) In a construction authorization for a geologic repository operations area at the Yucca Mountain site, the Commission shall include any conditions it considers necessary to protect the health and safety of the public, the common defense and security, or environmental values.
              (b) The Commission shall incorporate provisions in the construction authorization requiring DOE to furnish periodic or special reports regarding:
              (1) Progress of construction;
              (2) Any data about the site, obtained during construction, that are not within the predicted limits on which the facility design was based;

              (3) Any deficiencies, in design and construction, that, if uncorrected, could adversely affect safety at any future time; and
              (4) Results of research and development programs being conducted to resolve safety questions.
              (c) The construction authorization for a geologic repository operations area at the Yucca Mountain site will include restrictions on subsequent changes to the features of the geologic repository and the procedures authorized. The restrictions that may be imposed under this paragraph can include measures to prevent adverse effects on the geologic setting as well as measures related to the design and construction of the geologic repository operations area. These restrictions will fall into three categories of descending importance to public health and safety, as follows:
              (1) Those features and procedures that may not be changed without—
              (i) 60 days prior notice to the Commission;
              (ii) 30 days notice of opportunity for a prior hearing; and
              (iii) Prior Commission approval;
              (2) Those features and procedures that may not be changed without—
              (i) 60 days prior notice to the Commission; and
              (ii) Prior Commission approval; and
              (3) Those features and procedures that may not be changed without 60 days notice to the Commission. Features and procedures falling in this paragraph section may not be changed without prior Commission approval if the Commission, after having received the required notice, so orders.
              (d) A construction authorization must be subject to the limitation that a license to receive and possess source, special nuclear, or byproduct material at the Yucca Mountain site geologic repository operations area may not be issued by the Commission until;
              (1) DOE has updated its application, as specified at § 63.24; and
              (2) The Commission has made the findings stated in § 63.41.

              § 63.33
              Amendment of construction authorization.
              (a) An application for amendment of a construction authorization must be filed with the Commission that fully describes any desired changes and follows, as far as applicable, the content requirements prescribed in § 63.21.
              (b) In determining whether an amendment of a construction authorization will be approved, the Commission will be guided by the considerations that govern the issuance of the initial construction authorization, to the extent applicable.

            License Issuance and Amendment

              § 63.41
              Standards for issuance of a license.
              A license to receive and possess source, special nuclear, or byproduct material at a geologic repository operations area at the Yucca Mountain site may be issued by the Commission on finding that—
              (a) Construction of the geologic repository operations area has been substantially completed in conformity with the application as amended, the provisions of the Atomic Energy Act, and the rules and regulations of the Commission. Construction may be considered substantially complete for the purposes of this paragraph if the construction of—
              (1) Surface and interconnecting structures, systems, and components; and
              (2) Any underground storage space required for initial operation, are substantially complete.
              (b) The activities to be conducted at the geologic repository operations area will be in conformity with the application as amended, the provisions of the Atomic Energy Act and the Energy Reorganization Act, and the rules and regulations of the Commission.
              (c) The issuance of the license will not be inimical to the common defense and security and will not constitute an unreasonable risk to the health and safety of the public.
              (d) Adequate protective measures can and will be taken in the event of a radiological emergency at any time before permanent closure and decontamination or decontamination and dismantlement of surface facilities.
              (e) All applicable requirements of part 51 of this chapter have been satisfied.

              § 63.42
              Conditions of license.
              (a) The Commission shall include any conditions, including license specifications, it considers necessary to protect the health and safety of the public, the common defense and security, and environmental values in a license issued under this part.
              (b) Whether stated in the license or not, the following are considered to be conditions in every license issued:
              (1) The license is subject to revocation, suspension, modification, or amendment for cause, as provided by the Atomic Energy Act and the Commission's regulations.
              (2) DOE shall, at any time while the license is in effect, on written request of the Commission, submit written statements to enable the Commission to determine whether or not the license should be modified, suspended, or revoked.
              (3) The license is subject to the provisions of the Atomic Energy Act now or hereafter in effect and to all rules, regulations, and orders of the Commission. The terms and conditions of the license are subject to amendment, revision, or modification, by reason of amendments to or by reason of rules, regulations, and orders issued in accordance with the terms of the Atomic Energy Act.
              (c) Each license includes the provisions set forth in section 183 b-d, inclusive, of the Atomic Energy Act, whether or not these provisions are expressly set forth in the license.
              (d) A license issued under this part includes the provisions set forth in section 114(d) of the Nuclear Waste Policy Act, as amended, defining the quantity of solidified high-level radioactive waste and spent nuclear fuel, until such time as a second repository is in operation, whether or not these provisions are expressly set forth in the license.
              (e) The licensee (Department of Energy) shall ensure that Safeguards Information is protected against unauthorized disclosure in accordance with the requirements in § 73.21, and the requirements in § 73.22, or § 73.23 of this chapter, as applicable, and shall protect classified information in accordance with the requirements of parts 25 and 95 of this chapter, as applicable.
              [66 FR 55792, Nov. 2, 2001, as amended at 73 FR 63572, Oct. 24, 2008]

              § 63.43
              License specification.
              (a) A license issued under this part includes license conditions derived from the analyses and evaluations included in the application, including amendments made before a license is issued, together with any additional conditions the Commission finds appropriate.
              (b) License conditions include items in the following categories:
              (1) Restrictions as to the physical and chemical form and radioisotopic content of radioactive waste.
              (2) Restrictions as to size, shape, and materials and methods of construction of radioactive waste packaging.
              (3) Restrictions as to the amount of waste permitted per unit volume of storage space, considering the physical characteristics of both the waste and the host rock.
              (4) Requirements relating to test, calibration, or inspection, to assure that the foregoing restrictions are observed.
              (5) Controls to be applied to restrict access and to avoid disturbance to the site and to areas outside the site where conditions may affect compliance with §§ 63.111 and 63.113.
              (6) Administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure that activities at the facility are conducted in a safe manner and in conformity with the other license specifications.

              § 63.44
              Changes, tests, and experiments.
              (a) Definitions for the purposes of this section:
              (1) Change means a modification or addition to, or removal from, the geologic repository operations area design or procedures that affects a design function, event sequence, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished.

              (2) Departure from a method of evaluation described in the Safety Analysis Report (SAR) (as updated) used in establishing the preclosure safety analyses or performance assessment means:
              (i) Changing any of the elements of the method described in the SAR (as updated) unless the results of the analysis are conservative or essentially the same; or
              (ii) Changing from a method described in the SAR to another method unless that method has been approved by NRC for the intended application, addition or removal.
              (3) Safety Analysis Report (SAR) (as updated) means the Safety Analysis Report for the geologic repository, submitted in accordance with § 63.21, as updated in accordance with § 63.24.
              (4) Geologic repository operations area as described in the SAR (as updated) means:
              (i) The structures, systems, and components important to safety or barriers important to waste isolation that are described in the SAR (as updated); and
              (ii) The design and performance requirements for such structures, systems, and components described in the SAR (as updated).
              (5) Procedures as described in the SAR (as updated) means those procedures that contain information described in the SAR (as updated) such as how structures, systems, and components important to safety, or important to waste isolation, are operated or controlled.
              (6) Tests or experiments not described in the SAR (as updated) means any condition where the geologic repository operations area or any of its structures, systems, and components important to safety, or important to waste isolation, are utilized, controlled, or altered in a manner which is either:
              (i) Outside the reference bounds of the design bases as described in the SAR (as updated); or
              (ii) Inconsistent with the analyses or descriptions in the SAR (as updated).
              (b)(1) DOE may make changes in the geologic repository operations area as described in the SAR (as updated), make changes in the procedures as described in the SAR (as updated), and conduct tests or experiments not described in the SAR (as updated), without obtaining either an amendment of construction authorization under § 63.33 or a license amendment under § 63.45, if:
              (i) A change in the conditions incorporated in the construction authorization or license is not required; and
              (ii) The change, test, or experiment does not meet any of the criteria in paragraph (b)(2) of this section.
              (2) DOE shall obtain an amendment of construction authorization under § 63.33 or a license amendment under § 63.45, before implementing a change, test, or experiment if it would:
              (i) Result in more than a minimal increase in the frequency of occurrence of an event sequence previously evaluated in the SAR (as updated);
              (ii) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of structures, systems, components important to safety, or important to waste isolation, which were previously evaluated in the SAR (as updated);
              (iii) Result in more than a minimal increase in the consequences of an event sequence previously evaluated in the SAR (as updated);
              (iv) Result in more than a minimal increase in the consequences of malfunction of structures, systems, components important to safety, or important to waste isolation, which were previously evaluated in the SAR (as updated);
              (v) Create the possibility for an event sequence, or of a pathway for release of radionuclides, of a different type than any evaluated previously in the SAR (as updated);
              (vi) Create the possibility for a malfunction of structures, systems, and components important to safety, or important to waste isolation, with a different result than any evaluated previously in the SAR (as updated);
              (vii) Result in a departure from a method of evaluation described in the SAR (as updated) used in establishing the preclosure safety analysis or the performance assessment.

              (3) In implementing this paragraph, the SAR (as updated) is considered to include SAR changes resulting from evaluations performed pursuant to this section and from safety analyses performed under § 63.33 or § 63.45, as applicable, after the last Safety Analysis Report was updated under § 63.24.
              (4) The provisions in this section do not apply to changes to the geologic repository operations area or procedures when the applicable regulations establish more specific criteria for accomplishing such changes.
              (c)(1) DOE shall maintain records of changes in the geologic repository operations area at the Yucca Mountain site, of changes in procedures, and of tests and experiments made under paragraph (b) of this section. These records must include a written evaluation that provides the bases for the determination that the change, test, or experiment does not require an amendment of construction authorization or license amendment under paragraph (b) of this section.
              (2) No less frequently than every 24 months, DOE shall prepare a report containing a brief description of such changes, tests, and experiments, including a summary of the evaluation of each. These written reports must be sent to the NRC using an appropriate method listed in § 63.4; addressed: ATTN: Document Control Desk; Director, Office of Nuclear Material Safety and Safeguards; U.S. Nuclear Regulatory Commission, Washington, DC 20555-001; and DOE shall furnish the report to the appropriate NRC Regional Office shown in appendix D to part 20 of this chapter. Any report submitted under this paragraph must be made a part of the public record of the licensing proceedings.
              (d) Changes to the quality assurance program description required by § 63.21(c)(20) must be processed in accordance with § 63.144.
              [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003]

              § 63.45
              Amendment of license.
              (a) An application for amendment of a license may be filed with the Commission fully describing the changes desired and following as far as applicable the format prescribed for license applications.
              (b) In determining whether an amendment of a license will be approved, the Commission will be guided by the considerations that govern the issuance of the initial license, to the extent applicable.

              § 63.46
              Particular activities requiring license amendment.
              (a) Unless expressly authorized in the license, a license amendment is required for any of the following activities:
              (1) Any action that would make emplaced high-level radioactive waste irretrievable or that would substantially increase the difficulty of retrieving the emplaced waste;
              (2) Dismantling of structures;
              (3) Removal or reduction of controls applied to restrict access to or avoid disturbance of the site and to areas outside the site where conditions may affect compliance with §§ 63.111 and 63.113;
              (4) Destruction or disposal of records required to be maintained under the provisions of this part;
              (5) Any substantial change to the design or operating procedures from that specified in the license, except as authorized in § 63.44; and
              (6) Permanent closure.
              (b) An application for an amendment must be filed, and will be reviewed, as specified in § 63.45.

            US/IAEA Safeguards Agreement

              § 63.47
              Facility information and verification.
              (a) In response to a written request by the Commission, each applicant for a construction authorization or license and each recipient of a construction authorization or a license shall submit facility information, as described in § 75.10 of this chapter, on Form N-71 and associated forms and site information on DOC/NRC Form AP-A and associated forms;
              (b) As required by the Additional Protocol, applicants and licensees specified in paragraph (a) of this section shall submit location information described in § 75.11 of this chapter on DOC/NRC Form AP-1 and associated forms;

              (c) Shall permit verification thereof by the International Atomic Energy Agency (IAEA) and take other action as necessary to implement the US/IAEA Safeguards Agreement, as described in Part 75 of this chapter.
              [73 FR 78606, Dec. 23, 2008]

            Permanent Closure

              § 63.51
              License amendment for permanent closure.
              (a) DOE shall submit an application to amend the license before permanent closure of a geologic repository at the Yucca Mountain site. The submission must consist of an update of the license application submitted under §§ 63.21 and 63.22, including:
              (1) An update of the assessment of the performance of the geologic repository for the period after permanent closure. The updated assessment must include any performance confirmation data collected under the program required by subpart F, and pertinent to compliance with § 63.113.
              (2) A description of the program for post-permanent closure monitoring of the geologic repository.
              (3) A detailed description of the measures to be employed—such as land use controls, construction of monuments, and preservation of records—to regulate or prevent activities that could impair the long-term isolation of emplaced waste within the geologic repository and to assure that relevant information will be preserved for the use of future generations. As a minimum, these measures must include:
              (i) Identification of the site and geologic repository operations area by monuments that have been designed, fabricated, and emplaced to be as permanent as is practicable;
              (ii) Placement of records in the archives and land record systems of local, State, and Federal government agencies, and archives elsewhere in the world, that would be likely to be consulted by potential human intruders—such records to identify the location of the geologic repository operations area, including the underground facility, boreholes, shafts and ramps, and the boundaries of the site, and the nature and hazard of the waste; and
              (iii) A program for continued oversight, to prevent any activity at the site that poses an unreasonable risk of breaching the geologic repository's engineered barriers; or increasing the exposure of individual members of the public to radiation beyond allowable limits.
              (4) Geologic, geophysical, geochemical, hydrologic, and other site data that are obtained during the operational period, pertinent to compliance with § 63.113.
              (5) The results of tests, experiments, and any other analyses relating to backfill of excavated areas, shaft, borehole, or ramp sealing, drip shields, waste packages, interactions between natural and engineered systems, and any other tests, experiments, or analyses pertinent to compliance with § 63.113.
              (6) Any substantial revision of plans for permanent closure.
              (7) Other information bearing on permanent closure that was not available at the time a license was issued.
              (b) If necessary, to take into account the environmental impact of any substantial changes in the permanent closure activities proposed to be carried out or any significant new information regarding the environmental impacts of permanent closure, DOE shall also supplement its environmental impact statement and submit this statement, as supplemented, with the application for license amendment.

              § 63.52
              Termination of license.
              (a) Following permanent closure and the decontamination or decontamination and dismantlement of surface facilities at the Yucca Mountain site, DOE may apply for an amendment to terminate the license.
              (b) The application must be filed and will be reviewed in accordance with the provisions of § 63.45 and this section.
              (c) A license may be terminated only when the Commission finds with respect to the geologic repository:
              (1) That the final disposition of radioactive wastes has been made in conformance with DOE's plan, as amended and approved as part of the license.

              (2) That the final state of the geologic repository operations area conforms to DOE's plans for permanent closure and DOE's plans for the decontamination or decontamination and dismantlement of surface facilities, as amended and approved as part of the license.

              (3) That the termination of the license is authorized by law, including sections 57, 62, and 81 of the Atomic Energy Act, as amended.

          Subpart C—Participation by State Government, Affected Units of Local Government, and Affected Indian Tribes

            § 63.61
            Provision of information.
            (a) The Director shall provide the Governor and the Nevada State legislature, affected units of local government, and the governing body of any affected Indian Tribe, with timely and complete information regarding determinations or plans made by the Commission with respect to the Yucca Mountain site. Information must be provided concerning the site characterization, siting, development, design, licensing, construction, operation, regulation, permanent closure, or decontamination and dismantlement of surface facilities of the geologic repository operations area at the site.
            (b) Notwithstanding paragraph (a) of this section, the Director is not required to distribute any document to any entity if, with respect to the document, that entity or its counsel is included on a service list prepared under part 2 of this chapter.
            (c) The NRC shall place communications by the Director under this section in the Publicly Available Records System (PARS) Library and furnish copies to DOE.
            [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003]

            § 63.62
            Site review.
            (a) The Director shall make the NRC staff available to consult with representatives of the State of Nevada, affected units of local government, and affected Indian Tribes regarding the status of site characterization at the Yucca Mountain site.
            (b) Requests for consultation must be made in writing to the Director.
            (c) Consultation under this section may include:
            (1) Keeping the parties informed of the Director's views on the progress of site characterization.
            (2) Review of applicable NRC regulations, licensing procedures, schedules, and opportunities for State, affected units of local government, and Tribe participation in the Commission's regulatory activities.
            (3) Cooperation in development of proposals for State, affected units of local government, and Tribal participation in license reviews.

            § 63.63
            Participation in license reviews.
            (a) The State, affected units of local government, and affected Indian Tribes may participate in license reviews as provided in subpart J of part 2 of this chapter.
            (b) In addition, a State, or an affected unit of local government, or an affected Indian Tribe may submit a proposal to the Director to facilitate its participation in the review of the license application. The proposal may be submitted at any time and must contain a description and schedule of how the State, or affected unit of local government, or affected Indian Tribe wishes to participate in the review, or what services or activities the State, or affected unit of local government, or affected Indian Tribe wishes the NRC to carry out, and how the services or activities proposed to be carried out by the NRC would contribute to this participation. The proposal may include educational or information services (seminars, public meetings) or other actions on the part of NRC, such as establishing additional public document rooms or employment or exchange of State personnel under the Intergovernmental Personnel Act.
            (c) The Director shall arrange for a meeting between the representatives of the State, or affected unit of local government, or affected Indian Tribe and the NRC staff, to discuss any proposal submitted under paragraph (b) of this section, with a view to identifying any modifications that may contribute to the effective participation by such State, or affected unit of local government, or Tribe.

            (d) Subject to the availability of funds, the Director shall approve all or any part of a proposal, as it may be modified through the meeting described in paragraph (c) of this section, if it is determined that:
            (1) The proposed activities are suitable in light of the type and magnitude of impacts that the State, or affected unit of local government, or affected Indian Tribe may bear;
            (2) The proposed activities—
            (i) Will enhance communications between NRC and the State, or affected unit of local government, or affected Indian Tribe;
            (ii) Will make a productive and timely contribution to the review; and
            (iii) Are authorized by law.
            (e) The Director shall advise the State, or affected unit of local government, or affected Indian Tribe whether its proposal has been accepted or denied. If all or any part of a proposal is denied, the Director shall state the reason for the denial.
            (f) The NRC shall place all proposals submitted under this section, and responses to them, in the Publicly Available Records System (PARS) Library.
            [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58816, Oct. 10, 2003]

            § 63.64
            Notice to State.
            If the Governor and legislature of the State of Nevada have jointly designated, on their behalf, a single person or entity to receive notice and information from the Commission under this part, the Commission will provide the notice and information to the jointly designated person or entity instead of the Governor and legislature separately.

            § 63.65
            Representation.
            Any person who acts under this subpart as a representative for the State of Nevada (or for the Governor or legislature of Nevada), for an affected unit of local government, or for an affected Indian Tribe shall include in the request or other submission, or at the request of the Commission, a statement of the basis of his or her authority to act in this capacity.

          Subpart D—Records, Reports, Tests, and Inspections

            § 63.71
            Records and reports.
            (a) DOE shall maintain records and make reports in connection with the licensed activity that are required by the conditions of the license or by rules, regulations, and orders of the Commission, as authorized by the Atomic Energy Act and the Energy Reorganization Act.
            (b) Records of the receipt, handling, and disposition of radioactive waste at a geologic repository operations area at the Yucca Mountain site must contain sufficient information to provide a complete history of the movement of the waste from the shipper through all phases of storage and disposal. DOE shall retain these records in a manner that ensures their usability for future generations in accordance with § 63.51(a)(3).

            § 63.72
            Construction records.
            (a) DOE shall maintain records of construction of the geologic repository operations area at the Yucca Mountain site in a manner that ensures their usability for future generations in accordance with § 63.51(a)(3).
            (b) The records required under paragraph (a) of this section must include at least the following—
            (1) Surveys of the underground facility excavations, shafts, ramps, and boreholes referenced to readily identifiable surface features or monuments;
            (2) A description of the materials encountered;
            (3) Geologic maps and geologic cross-sections;
            (4) Locations and amount of seepage;
            (5) Details of equipment, methods, progress, and sequence of work;
            (6) Construction problems;
            (7) Anomalous conditions encountered;
            (8) Instrument locations, readings, and analysis;
            (9) Location and description of structural support systems;
            (10) Location and description of dewatering systems;

            (11) Details, methods of emplacement, and location of seals used; and

            (12) Facility design records (e.g, design specifications and “as built” drawings).

            § 63.73
            Reports of deficiencies.
            (a) DOE shall promptly notify the Commission of each deficiency found in the characteristics of the Yucca Mountain site, and design, and construction of the geologic repository operations area that, were it to remain uncorrected, could—
            (1) Adversely affect safety at any future time;
            (2) Represent a significant deviation from the design criteria and design basis stated in the design application; or
            (3) Represent a deviation from the conditions stated in the terms of a construction authorization or the license, including license specifications.
            (b) DOE shall implement a program for evaluating and reporting deviations and failures to comply, to identify defects and failures to comply associated with substantial safety hazards, based on the applicable requirements in 10 CFR 50.55(e) as it applies to the construction authorization and design of the geologic repository operations area at the Yucca Mountain site.
            (c) DOE shall implement a program of reporting specific events and conditions that is the same as that specified in 10 CFR 72.75.
            (d) The requisite notification must be as specified in the applicable regulation. By an appropriate method listed in § 63.4 of this chapter, written reports must be submitted to NRC addressed: ATTN: Document Control Desk; Director, Office of Nuclear Material Safety and Safeguards; U.S. Nuclear Regulatory Commission, Washington, DC 20555-001; and to the NRC onsite representative. DOE shall also furnish the report to the appropriate NRC Regional Office shown in appendix D to part 20 of this chapter.
            [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58815, Oct. 10, 2003]

            § 63.74
            Tests.
            (a) DOE shall perform, or permit the Commission to perform, those tests the Commission considers appropriate or necessary for the administration of the regulations in this part. This may include tests of—
            (1) Radioactive waste,
            (2) The geologic repository, including portions of the geologic setting and the structures, systems, and components constructed or placed therein,
            (3) Radiation detection and monitoring instruments, and
            (4) Other equipment and devices used in connection with the receipt, handling, or storage of radioactive waste.
            (b) The tests required under this section must include a performance confirmation program carried out in accordance with subpart F of this part.

            § 63.75
            Inspections.
            (a) DOE shall allow the Commission to inspect the premises of the geologic repository operations area at the Yucca Mountain site and adjacent areas to which DOE has rights of access.
            (b) DOE shall make available to the Commission for inspection, on reasonable notice, records kept by DOE pertaining to activities under this part.
            (c)(1) DOE shall, on requests by the Director, Office of Nuclear Material Safety and Safeguards, provide rent-free office space for the exclusive use of the Commission inspection personnel. Heat, air-conditioning, light, electrical outlets, and janitorial services must be furnished by DOE. The office must be convenient to and have full access to the facility and must provide the inspector both visual and acoustic privacy.

            (2) The space provided must be adequate to accommodate two full-time inspectors, and other transient NRC personnel and will be generally commensurate with other office facilities at the Yucca Mountain site geologic repository operations area. A space of 250 square feet either within the geologic repository operations area's office complex or in an office trailer or other onsite space at the geologic repository operations area is suggested as a guide. For locations at which activities are carried out under licenses issued under other parts of this chapter, additional space may be requested to accommodate additional full-time inspectors. The office space provided is subject to the approval of the Director, Office of Nuclear Material Safety and Safeguards. All furniture, supplies, and communication equipment will be furnished by the Commission.
            (3) DOE shall afford any NRC resident inspector assigned to the Yucca Mountain site or other NRC inspectors identified by the Regional Administrator as likely to inspect the Yucca Mountain facility, immediate unfettered access, equivalent to access provided regular employees, after proper identification and compliance with applicable access control measures for security, radiological protection, and personal safety.

            § 63.78
            Material control and accounting records and reports.
            DOE shall implement a program of material control and accounting (and accidental criticality reporting) that is the same as that specified in §§ 72.72, 72.74, 72.76, and 72.78 of this chapter.

          Subpart E—Technical Criteria

            § 63.101
            Purpose and nature of findings.
            (a)(1) Subpart B prescribes the standards for issuance of a license to receive and possess source, special nuclear, or byproduct material at a geologic repository operations area at the Yucca Mountain site. In particular, § 63.41(c) requires a finding that the issuance of a license will not constitute an unreasonable risk to the health and safety of the public. The purpose of this subpart is to set out the performance objectives for postclosure performance of the geologic repository and other criteria that, if satisfied, support a finding of no unreasonable risk. Postclosure performance objectives for the geologic repository include a requirement to limit radiological exposures to the reasonably maximally exposed individual, a requirement to limit releases of radionuclides to the accessible environment to protect groundwater, and a requirement to limit radiological exposures to the reasonably maximally exposed individual in the event of human intrusion (see § 63.113(b), (c), and (d), respectively).
            (2) Although the postclosure performance objectives specified at § 63.113 are generally stated in unqualified terms, it is not expected that complete assurance that the requirements will be met can be presented. A reasonable expectation, on the basis of the record before the Commission, that the postclosure performance objectives will be met, is the general standard required. Proof that the geologic repository will conform with the objectives for postclosure performance is not to be had in the ordinary sense of the word because of the uncertainties inherent in the understanding of the evolution of the geologic setting, biosphere, and engineered barrier system. For such long-term performance, what is required is reasonable expectation, making allowance for the time period, hazards, and uncertainties involved, that the outcome will conform with the objectives for postclosure performance for the geologic repository. Demonstrating compliance will involve the use of complex predictive models that are supported by limited data from field and laboratory tests, site-specific monitoring, and natural analog studies that may be supplemented with prevalent expert judgment. Compliance demonstrations should not exclude important parameters from assessments and analyses simply because they are difficult to precisely quantify to a high degree of confidence. The performance assessments and analyses should focus upon the full range of defensible and reasonable parameter distributions rather than only upon extreme physical situations and parameter values. Further, in reaching a determination of reasonable expectation, the Commission may supplement numerical analyses with qualitative judgments including, for example, consideration of the degree of diversity among the multiple barriers as a measure of the resiliency of the geologic repository.

            (b) Subpart B lists findings that must be made in support of an authorization to construct a geologic repository operations area at the Yucca Mountain site. Prior to closure, § 63.31(a)(1) requires a finding that there is reasonable assurance that the types and amounts of radioactive materials described in the application can be received, possessed, and stored in a geologic repository operations area of the design proposed without unreasonable risk to the health and safety of the public. After permanent closure, § 63.31(a)(2) requires the Commission to consider whether there is a reasonable expectation the site and design comply with the postclosure performance objectives. Once again, although the criteria may be written in unqualified terms, the demonstration of compliance must take uncertainties and gaps in knowledge into account so that the Commission can make the specified finding with respect to paragraph (a)(2) of § 63.31.

            § 63.102
            Concepts.
            This section provides a functional overview of this Subpart E. In the event of any inconsistency, the definitions in § 63.2 prevail.
            (a) The HLW facility at the Yucca Mountain site. NRC exercises licensing and related regulatory authority over those facilities described in section 202 (3) and (4) of the Energy Reorganization Act of 1974, including the site at Yucca Mountain, as designated by the Energy Policy Act of 1992.
            (b) The geologic repository operations area. (1) These regulations deal with the exercise of authority with respect to a particular class of HLW facility—namely, a geologic repository operations area at Yucca Mountain.
            (2) A geologic repository operations area consists of those surface and subsurface areas of the site that are part of a geologic repository where radioactive waste handling activities are conducted. The underground structure, backfill materials, if any, and openings that penetrate the underground structure (e.g., ramps, shafts and boreholes, including their seals), are designated the underground facility.

            (3) The exercise of Commission authority requires that the geologic repository operations area be used for storage (which includes disposal) of high-level radioactive wastes (HLW).
            (4) HLW includes irradiated reactor fuel as well as reprocessing wastes. However, if DOE proposes to use the geologic repository operations area for storage of radioactive waste other than HLW, the storage of this radioactive waste is subject to the requirements of this part.
            (c) Stages in the licensing process. There are several stages in the licensing process. The site characterization stage, when the performance confirmation program is started, begins before submission of a license application, and may result in consequences requiring evaluation in the license review. The construction stage would follow after the issuance of a construction authorization. A period of operations follows the Commission's issuance of a license. The period of operations includes the time during which emplacement of wastes occurs; any subsequent period before permanent closure during which the emplaced wastes are retrievable; and permanent closure, which includes sealing openings to the repository. Permanent closure represents the end of the performance confirmation program; final backfilling of the underground facility, if appropriate; and the sealing of shafts, ramps, and boreholes.
            (d) Areas related to isolation. Although the activities subject to regulation under this part are those to be carried out at the geologic repository operations area, the licensing process also considers characteristics of adjacent areas that are defined in other ways. There must be an area surrounding the geologic repository operations area, that could include either a portion or all of the site, within which DOE shall exercise specified controls to prevent adverse human actions after permanent closure. There is an area, designated the geologic setting, which includes the geologic, hydrologic, and geochemical systems of the region in which the site and geologic repository operations area are located. The geologic repository operations area, plus the portion of the geologic setting that provides isolation of the radioactive waste, make up the geologic repository.
            (e) Performance objectives through permanent closure. Before permanent closure, the geologic repository operations area is required to limit radiation levels and radiological exposures, in both restricted and unrestricted areas, and releases of radioactive materials to unrestricted areas, as specified at § 63.111(a).
            (f) Preclosure safety analysis. Section 63.111 includes performance objectives for the geologic repository operations area for the period before permanent closure and decontamination or permanent closure, decontamination, and dismantlement of surface facilities. The preclosure safety analysis is a systematic examination of the site; the design; and the potential hazards, initiating events and their resulting event sequences and potential radiological exposures to workers and the public. Initiating events are to be considered for inclusion in the preclosure safety analysis for determining event sequences only if they are reasonable (i.e., based on the characteristics of the geologic setting and the human environment, and consistent with precedents adopted for nuclear facilities with comparable or higher risks to workers and the public). The analysis identifies structures, systems, and components important to safety.
            (g) Performance objectives after permanent closure. After permanent closure, the geologic repository is required to:
            (1) Limit radiological exposures to the reasonably maximally exposed individual, as specified at § 63.113(b);
            (2) Limit releases of radionuclides to the accessible environment to protect groundwater, as specified at § 63.113(c); and
            (3) Limit radiological exposures to the reasonably maximally exposed individual in the event of human intrusion, as specified at § 63.113(d).
            (h) Multiple barriers. Section 63.113(a) requires that the geologic repository include multiple barriers, both natural and engineered. Geologic disposal of HLW is predicated on the expectation that one or more aspects of the geologic setting will be capable of contributing to the isolation of radioactive waste and thus be a barrier important to waste isolation. Although there is an extensive geologic record ranging from thousands to millions of years, this record is subject to interpretation and includes many uncertainties. In addition, there are uncertainties in the isolation capability and performance of engineered barriers. Although the composition and configuration of engineered structures (barriers) can be defined with a degree of precision not possible for natural barriers, it is recognized that except for a few archaeologic and natural analogs, there is a limited experience base for the performance of complex, engineered structures over periods longer than a few hundred years, considering the uncertainty in characterizing and modeling individual barriers. These uncertainties are addressed by requiring the use of a multiple barrier approach; specifically, an engineered barrier system is required in addition to the natural barriers provided by the geologic setting. The performance assessment provides an evaluation of the repository performance based on credible models and parameters including the consideration of uncertainty in the behavior of the repository system. Thus the performance assessment results reflect the capability of each of the barriers to cope with a variety of challenges (e.g., combinations of parameters leading to less favorable performance for individual barriers and combinations of barriers). A description of each barrier's capability (e.g., retardation of radionuclides in the saturated zone, waste package lifetime, matrix diffusion in the unsaturated zone), as reflected in the performance assessment, provides an understanding of how the natural barriers and the engineered barrier system work in combination to enhance the resiliency of the geologic repository. The Commission believes that this understanding can increase confidence that the postclosure performance objectives specified at § 63.113(b) and (c) will be achieved and that DOE's design includes a system of multiple barriers.
            (i) Reference biosphere and reasonably maximally exposed individual. The performance assessment will estimate the amount of radioactive material released to water or air at various locations and times in the future. To estimate the potential for future human exposures resulting from release of radioactive material from a geologic repository at Yucca Mountain, it is necessary to make certain assumptions about the location and characteristics of the reasonably maximally exposed individual. The environment inhabited by the reasonably maximally exposed individual, along with associated human exposure pathways and parameters, make up the reference biosphere, as described in § 63.305. The reasonably maximally exposed individual, as a hypothetical person living in a community with characteristics of the Town of Amargosa Valley, is a representative person using water with average concentrations of radionuclides as described at § 63.312. The reasonably maximally exposed individual is selected to represent those persons in the vicinity of Yucca Mountain who are reasonably expected to receive the greatest exposure to radioactive material released from a geologic repository at Yucca Mountain. Characteristics of the reference biosphere and the reasonably maximally exposed individual are to be based on current human behavior and biospheric conditions in the region, as described in § 63.305 and § 63.312.
            (j) Performance assessment. Demonstrating compliance with the postclosure performance objective specified at § 63.113(b) requires a performance assessment to quantitatively estimate radiological exposures to the reasonably maximally exposed individual at any time during the compliance period. The performance assessment is a systematic analysis that identifies the features, events, and processes (i.e., specific conditions or attributes of the geologic setting, degradation, deterioration, or alteration processes of engineered barriers, and interactions between the natural and engineered barriers) that might affect performance of the geologic repository; examines their effects on performance; and estimates the radiological exposures to the reasonably maximally exposed individual. The features, events, and processes considered in the performance assessment should represent a wide range of both beneficial and potentially adverse effects on performance (e.g., beneficial effects of radionuclide sorption; potentially adverse effects of fracture flow or a criticality event). Those features, events, and processes expected to materially affect compliance with § 63.113(b) or be potentially adverse to performance are included, while events (event classes or scenario classes) that are very unlikely (less than one chance in 10,000 over 10,000 years) can be excluded from the analysis. An event class consists of all possible specific initiating events that are caused by a common natural process (e.g., the event class for seismicity includes the range of credible earthquakes for the Yucca Mountain site). Radiological exposures to the reasonably maximally exposed individual are estimated using the selected features, events, and processes, and incorporating the probability that the estimated exposures will occur. Additionally, performance assessment methods are appropriate for use in demonstrating compliance with the postclosure performance objectives for groundwater protection and human intrusion, and are subject to the requirements for performance assessments specified at § 63.114 and applicable criteria in Subpart L (e.g., criteria for evaluating compliance with groundwater protection and individual protection standards).
            (k) Institutional controls. Active and passive institutional controls will be maintained over the Yucca Mountain site, and are expected to reduce significantly, but not eliminate, the potential for human activity that could inadvertently cause or accelerate the release of radioactive material. However, because it is not possible to make scientifically sound forecasts of the long-term reliability of institutional controls, it is not appropriate to include consideration of human intrusion into a fully risk-based performance assessment for purposes of evaluating the ability of the geologic repository to achieve the performance objective at § 63.113(b). Hence, human intrusion is addressed in a stylized manner as described in paragraph (l) of this section.
            (l) Human intrusion. In contrast to events unrelated to human activity, the probability and characteristics of human intrusion occurring many hundreds or thousands of years into the future cannot be estimated by examining either the historic or geologic record. Rather than speculating on the nature and probability of future intrusion, it is more useful to assess how resilient the geologic repository would be against a human intrusion event. Although the consequences of an assumed intrusion event would be a separate analysis, the analysis is similar to the performance assessment required by § 63.113(b) but subject to specific requirements for evaluation of human intrusion specified at §§ 63.321, 63.322 and 63.342 of subpart L of this part.
            (m) Performance confirmation. A performance confirmation program will be conducted to evaluate the adequacy of assumptions, data, and analyses that led to the findings that permitted construction of the repository and subsequent emplacement of the wastes. Key geotechnical and design parameters, including any interactions between natural and engineered systems and components, will be monitored throughout site characterization, construction, emplacement, and operation to identify any significant changes in the conditions assumed in the license application that may affect compliance with the performance objectives specified at § 63.113(b) and (c).
            (n) Groundwater protection. Separate groundwater protection standards are designed to protect the groundwater resources in the vicinity of Yucca Mountain. These standards, specified at § 63.331, require the estimation of groundwater concentrations in the representative volume of water. Depending on the radionuclide, the estimated concentrations must either be below a specified concentration or result in an annual, drinking water dose to the whole body or any organ of no greater than 0.04 mSv (4 mrem). Although the estimation of radionuclide concentrations in the representative volume would be a separate analysis, the analysis is similar to the performance assessment required by § 63.113(b) but subject to specific requirements for evaluation of groundwater protection specified at §§ 63.331, 63.332 and 63.342 of subpart L of this part.
            (o) Implementation of TEDE. When external exposure is determined by measurement with an external personal monitoring device, the deep-dose equivalent must be used in place of the effective dose equivalent, unless the effective dose equivalent is determined by a dosimetry method approved by the NRC. The assigned deep-dose equivalent must be for the part of the body receiving the highest exposure. The assigned shallow-dose equivalent must be the dose averaged over the contiguous 10 square centimeters of skin receiving the highest exposure. The radiation and organ or tissue weighting factors in Appendix A of 40 CFR part 197 are to be used to calculate TEDE. After the effective date of this regulation, the Commission may allow DOE to use updated factors, which have been issued by consensus scientific organizations and incorporated by EPA into Federal radiation guidance. Additionally, as scientific models and methodologies for estimating doses are updated, DOE may use the most current and appropriate (e.g., those accepted by the International Commission on Radiological Protection) scientific models and methodologies to calculate the TEDE. The weighting factors used in the calculation of TEDE must be consistent with the methodology used to perform the calculation.
            [66 FR 55792, Nov. 2, 2001, as amended at 74 FR 10828, Mar. 13, 2009]

            Preclosure Performance Objectives

              § 63.111
              Performance objectives for the geologic repository operations area through permanent closure.
              (a) Protection against radiation exposures and releases of radioactive material. (1) The geologic repository operations area must meet the requirements of part 20 of this chapter.
              (2) During normal operations, and for Category 1 event sequences, the annual TEDE (hereafter referred to as “dose”) to any real member of the public located beyond the boundary of the site may not exceed the preclosure standard specified at § 63.204.
              (b) Numerical guides for design objectives. (1) The geologic repository operations area must be designed so that, taking into consideration Category 1 event sequences and until permanent closure has been completed, the aggregate radiation exposures and the aggregate radiation levels in both restricted and unrestricted areas, and the aggregate releases of radioactive materials to unrestricted areas, will be maintained within the limits specified in paragraph (a) of this section.

              (2) The geologic repository operations area must be designed so that, taking into consideration any single Category 2 event sequence and until permanent closure has been completed, no individual located on, or beyond, any point on the boundary of the site will receive, as a result of the single Category 2 event sequence, the more limiting of a TEDE of 0.05 Sv (5 rem), or the sum of the deep dose equivalent and the committed dose equivalent to any individual organ or tissue (other than the lens of the eye) of 0.5 Sv (50 rem). The lens dose equivalent may not exceed 0.15 Sv (15 rem), and the shallow dose equivalent to skin may not exceed 0.5 Sv (50 rem).
              (c) Preclosure safety analysis. A preclosure safety analysis of the geologic repository operations area that meets the requirements specified at § 63.112 must be performed. This analysis must demonstrate that:
              (1) The requirements of § 63.111(a) will be met; and
              (2) The design meets the requirements of § 63.111(b).
              (d) Performance confirmation. The geologic repository operations area must be designed so as to permit implementation of a performance confirmation program that meets the requirements of subpart F of this part.
              (e) Retrievability of waste. (1) The geologic repository operations area must be designed to preserve the option of waste retrieval throughout the period during which wastes are being emplaced and thereafter, until the completion of a performance confirmation program and Commission review of the information obtained from such a program. To satisfy this objective, the geologic repository operations area must be designed so that any or all of the emplaced waste could be retrieved on a reasonable schedule starting at any time up to 50 years after waste emplacement operations are initiated, unless a different time period is approved or specified by the Commission. This different time period may be established on a case-by-case basis consistent with the emplacement schedule and the planned performance confirmation program.
              (2) This requirement may not preclude decisions by the Commission to allow backfilling part, or all of, or permanent closure of the geologic repository operations area, before the end of the period of design for retrievability.
              (3) For purposes of paragraph (e) of this section, a reasonable schedule for retrieval is one that would permit retrieval in about the same time as that required to construct the geologic repository operations area and emplace waste.

            Preclosure Safety Analysis

              § 63.112
              Requirements for preclosure safety analysis of the geologic repository operations area.
              The preclosure safety analysis of the geologic repository operations area must include:
              (a) A general description of the structures, systems, components, equipment, and process activities at the geologic repository operations area;
              (b) An identification and systematic analysis of naturally occurring and human-induced hazards at the geologic repository operations area, including a comprehensive identification of potential event sequences;
              (c) Data pertaining to the Yucca Mountain site, and the surrounding region to the extent necessary, used to identify naturally occurring and human-induced hazards at the geologic repository operations area;
              (d) The technical basis for either inclusion or exclusion of specific, naturally occurring and human-induced hazards in the safety analysis;
              (e) An analysis of the performance of the structures, systems, and components to identify those that are important to safety. This analysis identifies and describes the controls that are relied on to limit or prevent potential event sequences or mitigate their consequences. This analysis also identifies measures taken to ensure the availability of safety systems. The analysis required in this paragraph must include, but not necessarily be limited to, consideration of—
              (1) Means to limit concentration of radioactive material in air;
              (2) Means to limit the time required to perform work in the vicinity of radioactive materials;
              (3) Suitable shielding;
              (4) Means to monitor and control the dispersal of radioactive contamination;

              (5) Means to control access to high radiation areas or airborne radioactivity areas;

              (6) Means to prevent and control criticality;
              (7) Radiation alarm system to warn of significant increases of radiation levels, concentrations of radioactive material in air, and increased radioactivity in effluents;
              (8) Ability of structures, systems, and components to perform their intended safety functions, assuming the occurrence of event sequences;
              (9) Explosion and fire detection systems and appropriate suppression systems;
              (10) Means to control radioactive waste and radioactive effluents, and permit prompt termination of operations and evacuation of personnel during an emergency;
              (11) Means to provide reliable and timely emergency power to instruments, utility service systems, and operating systems important to safety if there is a loss of primary electric power;
              (12) Means to provide redundant systems necessary to maintain, with adequate capacity, the ability of utility services important to safety; and
              (13) Means to inspect, test, and maintain structures, systems, and components important to safety, as necessary, to ensure their continued functioning and readiness.
              (f) A description and discussion of the design, both surface and subsurface, of the geologic repository operations area, including—
              (1) The relationship between design criteria and the requirements specified at § 63.111(a) and (b); and
              (2) The design bases and their relation to the design criteria.

            Postclosure Performance Objectives

              § 63.113
              Performance objectives for the geologic repository after permanent closure.
              (a) The geologic repository must include multiple barriers, consisting of both natural barriers and an engineered barrier system.
              (b) The engineered barrier system must be designed so that, working in combination with natural barriers, radiological exposures to the reasonably maximally exposed individual are within the limits specified at § 63.311 of subpart L of this part. Compliance with this paragraph must be demonstrated through a performance assessment that meets the requirements specified at § 63.114 of this subpart, and §§ 63.303, 63.305, 63.312 and 63.342 of Subpart L of this part.
              (c) The engineered barrier system must be designed so that, working in combination with natural barriers, releases of radionuclides into the accessible environment are within the limits specified at § 63.331 of subpart L of this part. Compliance with this paragraph must be demonstrated through a performance assessment that meets the requirements specified at § 63.114 of this subpart and §§ 63.303, 63.332 and 63.342 of subpart L of this part.
              (d) The ability of the geologic repository to limit radiological exposures to the reasonably maximally exposed individual, in the event of human intrusion into the engineered barrier system, must be demonstrated through an analysis that meets the requirements at §§ 63.321 and 63.322 of subpart L of this part. Estimating radiological exposures to the reasonably maximally exposed individual requires a performance assessment that meets the requirements specified at § 63.114 of this subpart, and §§ 63.303, 63.305, 63.312 and 63.342 of subpart L of this part.

            Postclosure Performance Assessment

              § 63.114
              Requirements for performance assessment.
              (a) Any performance assessment used to demonstrate compliance with § 63.113 for 10,000 years after disposal must:
              (1) Include data related to the geology, hydrology, and geochemistry (including disruptive processes and events) of the Yucca Mountain site, and the surrounding region to the extent necessary, and information on the design of the engineered barrier system used to define, for 10,000 years after disposal, parameters and conceptual models used in the assessment.

              (2) Account for uncertainties and variabilities in parameter values, for 10,000 years after disposal, and provide for the technical basis for parameter ranges, probability distributions, or bounding values used in the performance assessment.

              (3) Consider alternative conceptual models of features and processes, for 10,000 years after disposal, that are consistent with available data and current scientific understanding and evaluate the effects that alternative conceptual models have on the performance of the geologic repository.
              (4) Consider only features, events, and processes consistent with the limits on performance assessment specified at § 63.342.
              (5) Provide the technical basis for either inclusion or exclusion of specific features, events, and processes in the performance assessment. Specific features, events, and processes must be evaluated in detail if the magnitude and time of the resulting radiological exposures to the reasonably maximally exposed individual, or radionuclide releases to the accessible environment, for 10,000 years after disposal, would be significantly changed by their omission.
              (6) Provide the technical basis for either inclusion or exclusion of degradation, deterioration, or alteration processes of engineered barriers in the performance assessment, including those processes that would adversely affect the performance of natural barriers. Degradation, deterioration, or alteration processes of engineered barriers must be evaluated in detail if the magnitude and time of the resulting radiological exposures to the reasonably maximally exposed individual, or radionuclide releases to the accessible environment, for 10,000 years after disposal, would be significantly changed by their omission.
              (7) Provide the technical basis for models used to represent the 10,000 years after disposal in the performance assessment, such as comparisons made with outputs of detailed process-level models and/or empirical observations (e.g., laboratory testing, field investigations, and natural analogs).
              (b) The performance assessment methods used to satisfy the requirements of paragraph (a) of this section are considered sufficient for the performance assessment for the period of time after 10,000 years and through the period of geologic stability.
              [74 FR 10828, Mar. 13, 2009]

              § 63.115
              Requirements for multiple barriers.
              Demonstration of compliance with § 63.113(a) must:
              (a) Identify those design features of the engineered barrier system, and natural features of the geologic setting, that are considered barriers important to waste isolation.
              (b) Describe the capability of barriers, identified as important to waste isolation, to isolate waste, taking into account uncertainties in characterizing and modeling the behavior of the barriers.
              (c) Provide the technical basis for the description of the capability of barriers, identified as important to waste isolation, to isolate waste. The technical basis for each barrier's capability shall be based on and consistent with the technical basis for the performance assessments used to demonstrate compliance with § 63.113(b) and (c).

            Land Ownership and Control

              § 63.121
              Requirements for ownership and control of interests in land.
              (a) Ownership of land. (1) The geologic repository operations area must be located in and on lands that are either acquired lands under the jurisdiction and control of DOE, or lands permanently withdrawn and reserved for its use.
              (2) These lands must be held free and clear of all encumbrances, if significant, such as:
              (i) Rights arising under the general mining laws;
              (ii) Easements for right-of-way; and
              (iii) All other rights arising under lease, rights of entry, deed, patent, mortgage, appropriation, prescription, or otherwise.
              (b) Additional controls for permanent closure. Appropriate controls must be established outside of the geologic repository operations area. DOE shall exercise any jurisdiction and control over surface and subsurface estates necessary to prevent adverse human actions that could significantly reduce the geologic repository's ability to achieve isolation. The rights of DOE may take the form of appropriate possessory interests, servitudes, or withdrawals from location or patent under the general mining laws.
              (c) Additional controls through permanent closure. Appropriate controls must be established outside the geologic repository operations area. DOE shall exercise any jurisdiction or control of activities necessary to ensure the requirements at § 63.111(a) and (b) are met. Control includes the authority to exclude members of the public, if necessary.
              (d) Water rights. (1) DOE shall also have obtained such water rights as may be needed to accomplish the purpose of the geologic repository operations area.
              (2) Water rights are included in the additional controls to be established under paragraph (b) of this section.

          Subpart F—Performance Confirmation Program

            § 63.131
            General requirements.
            (a) The performance confirmation program must provide data that indicate, where practicable, whether:
            (1) Actual subsurface conditions encountered and changes in those conditions during construction and waste emplacement operations are within the limits assumed in the licensing review; and
            (2) Natural and engineered systems and components required for repository operation, and that are designed or assumed to operate as barriers after permanent closure, are functioning as intended and anticipated.
            (b) The program must have been started during site characterization, and it will continue until permanent closure.
            (c) The program must include in situ monitoring, laboratory and field testing, and in situ experiments, as may be appropriate to provide the data required by paragraph (a) of this section.
            (d) The program must be implemented so that:
            (1) It does not adversely affect the ability of the geologic and engineered elements of the geologic repository to meet the performance objectives.
            (2) It provides baseline information and analysis of that information on those parameters and natural processes pertaining to the geologic setting that may be changed by site characterization, construction, and operational activities.
            (3) It monitors and analyzes changes from the baseline condition of parameters that could affect the performance of a geologic repository.

            § 63.132
            Confirmation of geotechnical and design parameters.
            (a) During repository construction and operation, a continuing program of surveillance, measurement, testing, and geologic mapping must be conducted to ensure that geotechnical and design parameters are confirmed and to ensure that appropriate action is taken to inform the Commission of design changes needed to accommodate actual field conditions encountered.
            (b) Subsurface conditions must be monitored and evaluated against design assumptions.
            (c) Specific geotechnical and design parameters to be measured or observed, including any interactions between natural and engineered systems and components, must be identified in the performance confirmation plan.
            (d) These measurements and observations must be compared with the original design bases and assumptions. If significant differences exist between the measurements and observations and the original design bases and assumptions, the need for modifications to the design or in construction methods must be determined and these differences, their significance to repository performance, and the recommended changes reported to the Commission.
            (e) In situ monitoring of the thermomechanical response of the underground facility must be conducted until permanent closure, to ensure that the performance of the geologic and engineering features is within design limits.

            § 63.133
            Design testing.

            (a) During the early or developmental stages of construction, a program for testing of engineered systems and components used in the design, such as, for example, borehole and shaft seals, backfill, and drip shields, as well as the thermal interaction effects of the waste packages, backfill, drip shields, rock, and unsaturated zone and saturated zone water, must be conducted.
            (b) The testing must be initiated as early as practicable.
            (c) If backfill is included in the repository design, a test must be conducted to evaluate the effectiveness of backfill placement and compaction procedures against design requirements before permanent backfill placement is begun.
            (d) Tests must be conducted to evaluate the effectiveness of borehole, shaft, and ramp seals before full-scale operation proceeds to seal boreholes, shafts, and ramps.

            § 63.134
            Monitoring and testing waste packages.
            (a) A program must be established at the geologic repository operations area for monitoring the condition of the waste packages. Waste packages chosen for the program must be representative of those to be emplaced in the underground facility.
            (b) Consistent with safe operation at the geologic repository operations area, the environment of the waste packages selected for the waste package monitoring program must be representative of the environment in which the wastes are to be emplaced.
            (c) The waste package monitoring program must include laboratory experiments that focus on the internal condition of the waste packages. To the extent practical, the environment experienced by the emplaced waste packages within the underground facility during the waste package monitoring program must be duplicated in the laboratory experiments.
            (d) The waste package monitoring program must continue as long as practical up to the time of permanent closure.

          Subpart G—Quality Assurance

            § 63.141
            Scope.
            As used in this part, quality assurance comprises all those planned and systematic actions necessary to provide adequate confidence that the geologic repository and its structures, systems, or components will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to the physical characteristics of a material, structure, component, or system that provide a means to control the quality of the material, structure, component, or system to predetermined requirements.

            § 63.142
            Quality assurance criteria.
            (a) Introduction and Applicability. DOE is required by § 63.21(c)(20) to include in its safety analysis report a description of the quality assurance program to be applied to all structures, systems, and components important to safety, to design and characterization of barriers important to waste isolation, and to related activities. These activities include: site characterization; acquisition, control, and analyses of samples and data; tests and experiments; scientific studies; facility and equipment design and construction; facility operation; performance confirmation; permanent closure; and decontamination and dismantling of surface facilities. The description must indicate how the applicable quality assurance requirements will be satisfied. DOE shall include information pertaining to the managerial and administrative controls to be used to ensure safe operation in its safety analysis report. High-level waste repositories include structures, systems, and components that prevent or mitigate the consequences of postulated event sequences or that are important to waste isolation capabilities that could cause undue risk to the health and safety of the public. The pertinent requirements of this subpart apply to all activities that are important to waste isolation and important to safety functions of those structures, systems, and components. These activities include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, modifying, site characterization, performance confirmation, permanent closure, decontamination, and dismantling of surface facilities.
            (b) Organization. DOE shall establish and execute a quality assurance program. DOE may delegate to others, such as contractors, agents, or consultants, the work of establishing and executing the quality assurance program, or any part of it, but DOE retains responsibility for it.
            (1) The authority and duties of persons and organizations performing activities affecting the functions of structures, systems, and components that are important to waste isolation and important to safety must be clearly established and delineated in writing. These activities include both the performing functions of attaining quality objectives and the quality assurance functions. The quality assurance functions are those of:
            (i) Assuring that an appropriate quality assurance program is established and effectively executed; and
            (ii) Verifying that activities important to waste isolation and important to safety functions have been correctly performed by checking, auditing, and inspection of structures, systems, and components.
            (2) The persons and organizations performing quality assurance functions shall have sufficient authority and organizational freedom to identify quality problems; to initiate, recommend, or provide solutions; and to verify implementation of solutions. The persons and organizations performing quality assurance functions shall report to a management level so that the required authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations, are provided.
            (3) Because of the many variables involved, such as the number of personnel, the type of activity being performed, and the location or locations where activities are performed, the organizational structure for executing the quality assurance program may take various forms provided that the persons and organizations assigned the quality assurance functions have this required authority and organizational freedom. Irrespective of the organizational structure, the individual(s) assigned the responsibility for assuring effective execution of any portion of the quality assurance program at any location where activities subject to 10 CFR part 63 are being performed must have direct access to the levels of management as may be necessary to perform this function.
            (c) Quality assurance program. DOE shall establish a quality assurance program that complies with the requirements of this subpart at the earliest practicable time, consistent with the schedule for accomplishing the activities. This program must be documented by written policies, procedures, or instructions and must be carried out throughout facility life in accordance with those policies, procedures, or instructions.
            (1) DOE shall identify the structures, systems, and components to be covered by the quality assurance program and the major organizations participating in the program, together with the designated functions of these organizations. The quality assurance program must control activities affecting the quality of the identified structures, systems, and components, to an extent consistent with their importance to safety.
            (2) Activities affecting quality must be accomplished under suitably controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanness; and assurance that all prerequisites for the given activity have been satisfied.
            (3) The program must take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test. The program must provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained.
            (4) DOE shall regularly review the status and adequacy of the quality assurance program. Management of other organizations participating in the quality assurance program shall regularly review the status and adequacy of that part of the quality assurance program which they are executing.
            (d) Design control. (1) DOE shall establish measures to assure that applicable regulatory requirements and the design basis, as defined in § 63.2 and as specified in the license application, for those structures, systems, and components to which this subpart applies, are correctly translated into specifications, drawings, procedures, and instructions. These measures must assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Measures must also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are important to waste isolation and important to safety functions of the structures, systems and components.
            (2) DOE shall establish measures to identify and control design interfaces and for coordination among participating design organizations. These measures must include the establishment of procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces.
            (i) The design control measures must provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. The verifying or checking process must be performed by individuals or groups other than those who performed the original design. These individuals may be from the same organization. If a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, it must include suitable qualifications testing of a prototype unit under the most adverse design conditions. Design control measures must be applied to items such as: criticality physics, stress, thermal, hydraulic, and preclosure and postclosure analyses; compatibility of materials; accessibility for inservice inspection, maintenance and repair; and delineation of acceptance criteria for inspections and tests.
            (ii) Design changes, including field changes, must be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization.
            (e) Procurement document control. DOE shall establish measures to assure that applicable regulatory requirements, design bases, and other requirements necessary to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, and services, whether purchased by the licensee or applicant or by its contractors or subcontractors. To the extent necessary, procurement documents must require contractors or subcontractors to provide a quality assurance program consistent with the pertinent provisions of this section.
            (f) Instructions, procedures, and drawings. Activities affecting quality must be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and must be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings must include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
            (g) Document control. DOE shall establish measures to control the issuance of documents, such as instructions, procedures, and drawings, including changes to them that prescribe all activities affecting quality. These measures must assure that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel and are distributed to and used at the location where the prescribed activity is performed. Changes to documents must be reviewed and approved by the same organizations that performed the original review and approval unless the applicant designates another responsible organization.
            (h) Control of purchased material, equipment, and services. DOE shall establish measures to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents.

            (1) These measures must include appropriate provisions for source evaluation and selection, objective evidence of quality furnished by the contractor or subcontractor, inspection at the contractor or subcontractor source, and examination of products upon delivery.
            (2) Documentary evidence that material and equipment conform to the procurement requirements must be available at the high-level waste repository site before the material and equipment are installed or used. This documentary evidence must be retained at the high-level waste repository site and be sufficient to identify the specific requirements, such as codes, standards, or specifications, met by the purchased material and equipment.
            (3) The effectiveness of the control of quality by contractors and subcontractors must be assessed by the licensee or applicant or designee at intervals consistent with the importance, complexity, and quantity of the product or services.
            (i) Identification and control of materials, parts, and components. Measures must be established for the identification and control of materials, parts, and components, including partially fabricated assemblies. These measures must assure that identification of the item is maintained by heat number, part number, serial number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item. These identification and control measures must be designed to prevent the use of incorrect or defective material, parts, and components.
            (j) Control of special processes. DOE shall establish measures to assure that special processes, including welding, heat treating, and nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
            (k) Inspection. DOE shall establish and execute a program for inspection of activities affecting quality to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. The inspection must be performed by individuals other than those who performed the activity being inspected.
            (1) Examinations, measurements, or tests of material or products processed must be performed for each work operation where necessary to assure quality. If inspection of processed material or products is impossible or disadvantageous, indirect control by monitoring processing methods, equipment, and personnel must be provided. Both inspection and process monitoring must be provided when control is inadequate without both.
            (2) If mandatory inspection hold points that require witnessing or inspecting by the applicant's designated representative and beyond which work may not proceed without the consent of its designated representative are required, the specific hold points must be indicated in appropriate documents.
            (l) Test control. DOE shall establish a test program to assure that all testing required to demonstrate that structures, systems, and components important to safety will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
            (1) The test program must include, as appropriate, proof tests prior to installation, preoperational tests, and operational tests during repository operation, of structures, systems, and components.
            (2) Test procedures must include provisions for assuring that all prerequisites for the given test have been met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions.
            (3) Test results must be documented and evaluated to assure that test requirements have been satisfied.
            (m) Control of measuring and test equipment. DOE shall establish measures to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits.

            (n) Handling, storage, and shipping. DOE shall establish measures to control the handling, storage, shipping, cleaning and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration. When necessary for particular products, special protective environments, such as inert gas atmosphere, specific moisture content levels, and temperature levels, must be specified and provided.
            (o) Inspection, test, and operating status. DOE shall establish measures to indicate the status of inspections and tests performed on individual items of the high-level waste repository by markings such as stamps, tags, labels, routing cards, or other suitable means. These measures must provide for the identification of items that have satisfactorily passed required inspections and tests, where necessary to preclude inadvertent bypassing of such inspections and tests. Measures must also be established for indicating the operating status of structures, systems, and components of the high-level waste repository, such as by tagging valves and switches, to prevent inadvertent operation.
            (p) Nonconforming materials, parts, or components. DOE shall establish measures to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. These measures must include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Nonconforming items must be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures.
            (q) Corrective action. DOE shall establish measures to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. If significant conditions are adverse to quality, the measures must assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken must be documented and reported to appropriate levels of management.
            (r) Quality assurance records. DOE shall maintain sufficient records to furnish evidence of activities affecting quality.
            (1) The records must include at least the following: Operating logs and the results of reviews, inspections, tests, audits, monitoring of work performance, and materials analyses.
            (2) The records must also include closely-related data such as qualifications of personnel, procedures, and equipment.
            (3) Inspection and test records must, at a minimum, identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted.
            (4) Records must be identifiable and retrievable. Consistent with applicable regulatory requirements, the applicant shall establish requirements concerning record retention, such as duration, location, and assigned responsibility.
            (s) Audits. DOE shall carry out a comprehensive system of planned and periodic audits to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program. The audits must be performed in accordance with the written procedures or check lists by appropriately trained personnel not having direct responsibilities in the areas being audited. Audit results must be documented and reviewed by management having responsibility in the area audited. Followup action, including reaudit of deficient areas, must be taken where indicated.

            § 63.143
            Implementation.
            DOE shall implement a quality assurance program based on the criteria required by § 63.142.

            § 63.144
            Quality assurance program change.

            Changes to DOE's NRC-approved Safety Analysis Report quality assurance program description are processed as follows:

            (a) DOE may change a previously accepted quality assurance program description included or referenced in the Safety Analysis Report without prior NRC approval, if the change does not reduce the commitments in the program description previously accepted by the NRC. Changes to the quality assurance program description that do not reduce the commitments must be submitted every 24 months, in accordance with paragraph (b)(1) of this section. In addition to quality assurance program changes involving administrative improvements and clarifications, spelling corrections, punctuation, or editorial items, the following changes are not considered reductions in commitment:
            (1) The use of a quality assurance standard approved by the NRC which is more recent than the quality assurance standard in DOE's current quality assurance program at the time of the change;
            (2) The use of generic organizational position titles that clearly denote the position function, supplemented as necessary by descriptive text, rather than specific titles;
            (3) The use of generic organizational charts to indicate functional relationships, authorities, and responsibilities, or alternatively, the use of descriptive text;
            (4) The elimination of quality assurance program information that duplicates language in quality assurance regulatory guides and quality assurance standards to which the licensee is committed; and
            (5) Organizational revisions that ensure that persons and organizations performing quality assurance functions continue to have the requisite authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations.
            (b) DOE shall submit changes made to the NRC-accepted Safety Analysis Report quality assurance program description that do reduce the commitments to the NRC and receive NRC approval prior to implementation, as follows:
            (1) By an appropriate method listed in § 63.4 of this chapter, the signed document must be submitted to the Nuclear Regulatory Commission, addressed: ATTN: Document Control Desk; Director, Office of Nuclear Material and Safeguards; U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and one copy to the appropriate NRC Resident Inspector, if one has been assigned to the site or facility.
            (2) The submittal of a change to the Safety Analysis Report quality assurance program description must include all pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to describe how the requirements of § 63.142 will be satisfied and continues to satisfy the criteria of § 63.142 and the Safety Analysis Report quality assurance program description previously accepted by the NRC (the letter need not provide the basis for changes that correct spelling, punctuation, or editorial items).
            (3) DOE shall maintain records of quality assurance program changes that do reduce commitments.
            [66 FR 55792, Nov. 2, 2001, as amended at 68 FR 58816, Oct. 10, 2003]

          Subpart H—Training and Certification of Personnel

            § 63.151
            General requirements.
            Operations of systems and components that have been identified as important to safety in the Safety Analysis Report and in the license must be performed only by trained and certified personnel or by personnel under the direct visual supervision of an individual with training and certification in such operation. Supervisory personnel who direct operations that are important to safety must also be certified in such operations.

            § 63.152
            Training and certification program.
            DOE shall establish a program for training, proficiency testing, certification, and requalification of operating and supervisory personnel.

            § 63.153
            Physical requirements.
            The physical condition and the general health of personnel certified for operations that are important to safety may not be such as might cause operational errors that could endanger the public health and safety. Any condition that might cause impaired judgment or motor coordination must be considered in the selection of personnel for activities that are important to safety. These conditions need not categorically disqualify a person, so long as appropriate provisions are made to accommodate the conditions.

          Subpart I—Emergency Planning Criteria

            § 63.161
            Emergency plan for the geologic repository operations area through permanent closure.
            DOE shall develop and be prepared to implement a plan to cope with radiological accidents that may occur at the geologic repository operations area, at any time before permanent closure and decontamination or decontamination and dismantlement of surface facilities. The emergency plan must be based on the criteria of § 72.32(b) of this chapter.

          Subpart J—Violations

            § 63.171
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of—
            (1) The Atomic Energy Act of 1954, as amended;
            (2) Title II of the Energy Reorganization Act of 1974, as amended; or
            (3) A regulation or order issued under those Acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:
            (1) For violations of—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any rule, regulation, or order issued under the sections specified in paragraph (b)(1)(i) of this section;
            (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.

            § 63.172
            Criminal penalties.
            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in this part 63 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in this part 63 that are not issued under sections 161b, 161i, or 161o for the purposes of Section 223 are as follows: §§ 63.1, 63.2, 63.5, 63.6, 63.7, 63.8, 63.15, 63.16, 63.21, 63.22, 63.23, 63.24, 63.31, 63.32, 63.33, 63.41, 63.42, 63.43, 63.45, 63.46, 63.51, 63.52, 63.61, 63.62, 63.63, 63.64, 63.65, 63.101, 63.102, 63.111, 63.112, 63.113, 63.114, 63.115, 63.121, 63.131, 63.132, 63.133, 63.134, 63.141, 63.142, 63.143, 63.153, 63.161, 63.171, 63.172, 63.201, 63.202, 63.203, 63.204, 63.301, 63.302, 63.303, 63.304, 63.305, 63.311, 63.312, 63.321, 63.322, 63.331, 63.332, 63.341, and 63.342.

          Subpart K—Preclosure Public Health and Environmental Standards

            § 63.201
            Purpose and scope.
            This subpart covers the storage of radioactive material by DOE in the Yucca Mountain repository and on the Yucca Mountain site. For the purposes of demonstrating compliance with this subpart, to the extent there may be any conflict with the requirements specified in this subpart and the requirements contained in Subparts A-J of this regulation, including definitions, the requirements in this subpart shall take precedence.

            § 63.202
            Definitions for Subpart K.

              General environment means everywhere outside the Yucca Mountain site, the Nellis Air Force Range, and the Nevada Test Site.

              Member of the public means anyone who is not a radiation worker for purposes of worker protection.

              Radioactive material means matter composed of or containing radionuclides subject to the Atomic Energy Act of 1954, as amended (42 U.S.C. sec. 2014 et seq.). Radioactive material includes, but is not limited to, high-level radioactive waste and spent nuclear fuel.

              Spent nuclear fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, the constituent elements of which have not been separated by reprocessing.

              Storage means retention (and any associated activity, operation, or process necessary to carry out successful retention) of radioactive material with the intent or capability to readily access or retrieve such material.

              Yucca Mountain repository means the excavated portion of the facility constructed underground within the Yucca Mountain site.

              Yucca Mountain site means:
            (1) The site recommended by the Secretary of DOE to the President under section 112(b)(1)(B) of the Nuclear Waste Policy Act of 1982 (42 U.S.C. 10132(b)(1)(B)) on May 27, 1986; or
            (2) The area under the control of DOE for the use of Yucca Mountain activities at the time of licensing, if the site designated under the Nuclear Waste Policy Act is amended by Congress prior to the time of licensing.

            § 63.203
            Implementation of Subpart K.
            DOE must demonstrate that normal operations at the Yucca Mountain site will and do occur in compliance with this subpart before the Commission grants or continues a license for DOE to receive and possess radioactive material within the Yucca Mountain site.

            § 63.204
            Preclosure standard.
            DOE must ensure that no member of the public in the general environment receives more than an annual dose of 0.15 mSv (15 mrem) from the combination of:
            (a) Management and storage (as defined in 40 CFR 191.2) of radioactive material that:
            (1) Is subject to 40 CFR 191.3(a); and
            (2) Occurs outside of the Yucca Mountain repository but within the Yucca Mountain site; and
            (b) Storage (as defined in § 63.202) of radioactive material inside the Yucca Mountain repository.

          Subpart L—Postclosure Public Health and Environmental Standards

            § 63.301
            Purpose and scope.
            This subpart covers the disposal of radioactive material in the Yucca Mountain repository by DOE. For the purposes of demonstrating compliance with this subpart, to the extent that there may be any conflict with the requirements specified in this subpart and the requirements contained in Subparts A-J of this part, including definitions, the requirements in this subpart shall take precedence.

            § 63.302
            Definitions for Subpart L.
            All definitions in subpart K of this part, and the following:

              Accessible environment means any point outside of the controlled area, including:
            (1) The atmosphere (including the atmosphere above the surface area of the controlled area);
            (2) Land surfaces;
            (3) Surface waters;
            (4) Oceans; and
            (5) The lithosphere.

              Aquifer means a water-bearing underground geological formation, group of formations, or part of a formation (excluding perched water bodies) that can yield a significant amount of groundwater to a well or spring.

              Controlled area means:
            (1) The surface area, identified by passive institutional controls, that encompasses no more than 300 square kilometers. It must not extend farther:
            (i) South than 36°40′13.6661″ North latitude, in the predominant direction of groundwater flow; and
            (ii) Than five kilometers from the repository footprint in any other direction; and
            (2) The subsurface underlying the surface area.

              Disposal means the emplacement of radioactive material into the Yucca Mountain disposal system with the intent of isolating it for as long as reasonably possible and with no intent of recovery, whether or not the design of the disposal system permits the ready recovery of the material. Disposal of radioactive material in the Yucca Mountain disposal system begins when all of the ramps and other openings into the Yucca Mountain repository are sealed.

              Groundwater means water that is below the land surface and in a saturated zone.

              Human intrusion means breaching of any portion of the Yucca Mountain disposal system, within the repository footprint, by any human activity.

              Passive institutional controls means:
            (1) Markers, as permanent as practicable, placed on the Earth's surface;
            (2) Public records and archives;
            (3) Government ownership and regulations regarding land or resource use; and
            (4) Other reasonable methods of preserving knowledge about the location, design, and contents of the Yucca Mountain disposal system.

              Peak dose means the highest annual dose projected to be received by the reasonably maximally exposed individual.

              Period of geologic stability means the time during which the variability of geologic characteristics and their future behavior in and around the Yucca Mountain site can be bounded, that is, they can be projected within a reasonable range of possibilities. This period is defined to end at 1 million years after disposal.

              Plume of contamination means that volume of groundwater in the predominant direction of groundwater flow that contains radioactive contamination from releases from the Yucca Mountain repository. It does not include releases from any other potential sources on or near the Nevada Test Site.

              Repository footprint means the outline of the outermost locations of where the waste is emplaced in the Yucca Mountain repository.

              Slice of the plume means a cross-section of the plume of contamination with sufficient thickness parallel to the prevalent direction of flow of the plume that it contains the representative volume.

              Total dissolved solids means the total dissolved (filterable) solids in water as determined by use of the method specified in 40 CFR part 136.

              Undisturbed performance means that human intrusion or the occurrence of unlikely natural features, events, and processes do not disturb the disposal system.

              Undisturbed Yucca Mountain disposal system means that the Yucca Mountain disposal system is not affected by human intrusion.

              Waste means any radioactive material emplaced for disposal into the Yucca Mountain repository.

              Well-capture zone means the volume from which a well pumping at a defined rate is withdrawing water from an aquifer. The dimensions of the well-capture zone are determined by the pumping rate in combination with aquifer characteristics assumed for calculations, such as hydraulic conductivity, gradient, and the screened interval.

              Yucca Mountain disposal system means the combination of underground engineered and natural barriers within the controlled area that prevents or substantially reduces releases from the waste.
            [66 FR 55792, Nov. 2, 2001, as amended at 74 FR 10829, Mar. 13, 2009]

            § 63.303
            Implementation of Subpart L.
            (a) Compliance is based upon the arithmetic mean of the projected doses from DOE's performance assessments for the period within 1 million years after disposal, with:
            (1) Sections 63.311(a)(1) and 63.311(a)(2); and
            (2) Sections 63.321(b)(1), 63.321(b)(2), and 63.331, if performance assessment is used to demonstrate compliance with either or both of these sections.
            (b) [Reserved]
            [74 FR 10829, Mar. 13, 2009]

            § 63.304
            Reasonable expectation.

              Reasonable expectation means that the Commission is satisfied that compliance will be achieved based upon the full record before it. Characteristics of reasonable expectation include that it:

            (1) Requires less than absolute proof because absolute proof is impossible to attain for disposal due to the uncertainty of projecting long-term performance;
            (2) Accounts for the inherently greater uncertainties in making long-term projections of the performance of the Yucca Mountain disposal system;
            (3) Does not exclude important parameters from assessments and analyses simply because they are difficult to precisely quantify to a high degree of confidence; and
            (4) Focuses performance assessments and analyses on the full range of defensible and reasonable parameter distributions rather than only upon extreme physical situations and parameter values.

            § 63.305
            Required characteristics of the reference biosphere.
            (a) Features, events, and processes that describe the reference biosphere must be consistent with present knowledge of the conditions in the region surrounding the Yucca Mountain site.
            (b) DOE should not project changes in society, the biosphere (other than climate), human biology, or increases or decreases of human knowledge or technology. In all analyses done to demonstrate compliance with this part, DOE must assume that all of those factors remain constant as they are at the time of submission of the license application.
            (c) DOE must vary factors related to the geology, hydrology, and climate based upon cautious, but reasonable assumptions of the changes in these factors that could affect the Yucca Mountain disposal system during the period of geologic stability, consistent with the requirements for performance assessments specified at § 63.342.
            (d) Biosphere pathways must be consistent with arid or semi-arid conditions.
            [66 FR 55792, Nov. 2, 2001, as amended at 74 FR 10829, Mar. 13, 2009]

            Postclosure Individual Protection Standard

              § 63.311
              Individual protection standard after permanent closure.
              (a) DOE must demonstrate, using performance assessment, that there is a reasonable expectation that the reasonably maximally exposed individual receives no more than the following annual dose from releases from the undisturbed Yucca Mountain disposal system:
              (1) 0.15 mSv (15 mrem) for 10,000 years following disposal; and
              (2) 1.0 mSv (100 mrem) after 10,000 years, but within the period of geologic stability.
              (b) DOE's performance assessment must include all potential pathways of radionuclide transport and exposure.
              [74 FR 10829, Mar. 13, 2009]

              § 63.312
              Required characteristics of the reasonably maximally exposed individual.
              The reasonably maximally exposed individual is a hypothetical person who meets the following criteria:
              (a) Lives in the accessible environment above the highest concentration of radionuclides in the plume of contamination;
              (b) Has a diet and living style representative of the people who now reside in the Town of Amargosa Valley, Nevada. DOE must use projections based upon surveys of the people residing in the Town of Amargosa Valley, Nevada, to determine their current diets and living styles and use the mean values of these factors in the assessments conducted for §§ 63.311 and 63.321;
              (c) Uses well water with average concentrations of radionuclides based on an annual water demand of 3000 acre-feet;
              (d) Drinks 2 liters of water per day from wells drilled into the groundwater at the location specified in paragraph (a) of this section; and
              (e) Is an adult with metabolic and physiological considerations consistent with present knowledge of adults.

            Human Intrusion Standard

              § 63.321
              Individual protection standard for human intrusion.
              (a) DOE must determine the earliest time after disposal that the waste package would degrade sufficiently that a human intrusion (see § 63.322) could occur without recognition by the drillers.
              (b) DOE must demonstrate that there is a reasonable expectation that the reasonably maximally exposed individual receives, as a result of the human intrusion, no more than the following annual dose:
              (1) 0.15 mSv (15 mrem) for 10,000 years following disposal; and
              (2) 1.0 mSv (100 mrem) after 10,000 years, but within the period of geologic stability.
              (c) DOE's analysis must include all potential environmental pathways of radionuclide transport and exposure, subject to the requirements of § 63.322.
              [74 FR 10829, Mar. 13, 2009]

              § 63.322
              Human intrusion scenario.
              For the purposes of the analysis of human intrusion, DOE must make the following assumptions:
              (a) There is a single human intrusion as a result of exploratory drilling for groundwater;
              (b) The intruders drill a borehole directly through a degraded waste package into the uppermost aquifer underlying the Yucca Mountain repository;
              (c) The drillers use the common techniques and practices that are currently employed in exploratory drilling for groundwater in the region surrounding Yucca Mountain;
              (d) Careful sealing of the borehole does not occur, instead natural degradation processes gradually modify the borehole;
              (e) No particulate waste material falls into the borehole;
              (f) The exposure scenario includes only those radionuclides transported to the saturated zone by water (e.g., water enters the waste package, releases radionuclides, and transports radionuclides by way of the borehole to the saturated zone); and
              (g) No releases are included which are caused by unlikely natural processes and events.

            Groundwater Protection Standards

              § 63.331
              Separate standards for protection of groundwater.
              DOE must demonstrate that there is a reasonable expectation that, for 10,000 years of undisturbed performance after disposal, releases of radionuclides from waste in the Yucca Mountain disposal system into the accessible environment will not cause the level of radioactivity in the representative volume of groundwater to exceed the limits in the following Table 1:

                Table 1—Limits on Radionuclides in the Representative Volume

                  Radionuclide or type of radiation emitted
                  Limit
                  Is natural backgroundincluded?

                  Combined radium-226 and radium-228
                  5 picocuries per liter
                  Yes.

                  Gross alpha activity (including radium-226 but excluding radon and uranium)
                  15 picocuries per liter
                  Yes.

                  Combined beta and photon emitting radionuclides
                  0.04 mSv (4 mrem) per year to the whole body or any organ, based on drinking 2 liters of water per day from the representative volume
                  No.

              § 63.332
              Representative volume.

              (a) The representative volume is the volume of groundwater that would be withdrawn annually from an aquifer containing less than 10,000 milligrams of total dissolved solids per liter of water to supply a given water demand. DOE must project the concentration of radionuclides released from the Yucca Mountain disposal system that will be in the representative volume. DOE must use the projected concentrations to demonstrate a reasonable expectation that the Yucca Mountain disposal system complies with § 63.331. The DOE must make the following assumptions concerning the representative volume:

              (1) It includes the highest concentration level in the plume of contamination in the accessible environment;
              (2) Its position and dimensions in the aquifer are determined using average hydrologic characteristics which have cautious, but reasonable, values representative of the aquifers along the radionuclide migration path from the Yucca Mountain repository to the accessible environment as determined by site characterization; and
              (3) It contains 3,000 acre-feet of water (about 3,714,450,000 liters or 977,486,000 gallons).
              (b) DOE must use one of two alternative methods for determining the dimensions of the representative volume. The DOE must propose its chosen method, and any underlying assumptions, to NRC for approval.
              (1) DOE may calculate the dimensions as a well-capture zone. If DOE uses this approach, it must assume that the:
              (i) Water supply well(s) has (have) characteristics consistent with public water supply wells in the Town of Amargosa Valley, Nevada, for example, well-bore size and length of the screened intervals;
              (ii) Screened interval(s) include(s) the highest concentration in the plume of contamination in the accessible environment; and
              (iii) Pumping rates and the placement of the well(s) must be set to produce an annual withdrawal equal to the representative volume and to tap the highest concentration within the plume of contamination.
              (2) DOE may calculate the dimensions as a slice of the plume. If DOE uses this approach, it must:
              (i) Propose, for approval, where the location of the edge of the plume of contamination occurs. For example, the place where the concentration of radionuclides reaches 0.1% of the level of the highest concentration in the accessible environment;
              (ii) Assume that the slice of the plume is perpendicular to the prevalent direction of flow of the aquifer; and
              (iii) Assume that the volume of groundwater contained within the slice of the plume equals the representative volume.

            Additional Provisions

              § 63.342
              Limits on performance assessments.

              (a) DOE's performance assessments conducted to show compliance with §§ 63.311(a)(1), 63.321(b)(1), and 63.331 shall not include consideration of very unlikely features, events, or processes, i.e., those that are estimated to have less than one chance in 100,000,000 per year of occurring. In addition, DOE's performance assessments need not evaluate the impacts resulting from any features, events, and processes or sequences of events and processes with a higher chance of occurring if the results of the performance assessments would not be changed significantly in the initial 10,000-year period after disposal.

              (b) For performance assessments conducted to show compliance with §§ 63.321(b)(1) and 63.331, DOE's performance assessments shall exclude the unlikely features, events, and processes, or sequences of events and processes, i.e., those that are estimated to have less than one chance in 100,000 per year of occurring and at least one chance in 100,000,000 per year of occurring.
              (c) For performance assessments conducted to show compliance with §§ 63.311(a)(2) and 63.321(b)(2), DOE's performance assessments shall project the continued effects of the features, events, and processes included in paragraph (a) of this section beyond the 10,000-year post-disposal period through the period of geologic stability. DOE must evaluate all of the features, events, or processes included in paragraph (a) of this section, and also:
              (1) DOE must assess the effects of seismic and igneous activity scenarios, subject to the probability limits in paragraph (a) of this section for very unlikely features, events, and processes, or sequences of events and processes. Performance assessments conducted to show compliance with § 63.321(b)(2) are also subject to the probability limits in paragraph (b) of this section for unlikely features, events, and processes, or sequences of events and processes.

              (i) The seismic analysis may be limited to the effects caused by damage to the drifts in the repository, failure of the waste packages, and changes in the elevation of the water table under Yucca Mountain (i.e., the magnitude of the water table rise under Yucca Mountain).
              (ii) The igneous activity analysis may be limited to the effects of a volcanic event directly intersecting the repository. The igneous event may be limited to that causing damage to the waste packages directly, causing releases of radionuclides to the biosphere, atmosphere, or groundwater.
              (2) DOE must assess the effects of climate change. The climate change analysis may be limited to the effects of increased water flow through the repository as a result of climate change, and the resulting transport and release of radionuclides to the accessible environment. The nature and degree of climate change may be represented by constant-in-time climate conditions. The analysis may commence at 10,000 years after disposal and shall extend through the period of geologic stability. The constant-in-time values to be used to represent climate change are to be the spatial average of the deep percolation rate within the area bounded by the repository footprint. The constant-in-time deep percolation rates to be used to represent climate change shall be based on a lognormal distribution with an arithmetic mean of 41 mm/year (1.6 in./year) and a standard deviation of 33 mm/year (1.3 in./year). The lognormal distribution is to be truncated so that the deep percolation rates vary between 10 and 100 mm/year (0.39 and 3.9 in./year).
              (3) DOE must assess the effects of general corrosion on engineered barriers. DOE may use a constant representative corrosion rate throughout the period of geologic stability or a distribution of corrosion rates correlated to other repository parameters.
              [74 FR 10829, Mar. 13, 2009]

              § 63.343
              Severability of individual protection and groundwater protection standards.
              The individual protection and groundwater protection standards are severable.

        Pt. 70
        PART 70—DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL

            Subpart A—General Provisions
            Sec.
            70.1
            Purpose.
            70.2
            Scope.
            70.3
            License requirements.
            70.4
            Definitions.
            70.5
            Communications.
            70.6
            Interpretations.
            70.7
            Employee protection.
            70.8
            Information collection requirements: OMB approval.
            70.9
            Completeness and accuracy of information.
            70.10
            Deliberate misconduct.

            Subpart B—Exemptions
            70.11
            Persons using special nuclear material under certain Department of Energy and Nuclear Regulatory Commission contracts.
            70.12
            Carriers.
            70.13
            Department of Defense.
            70.14
            Foreign military aircraft.
            70.17
            Specific exemptions.

            Subpart C—General Licenses
            70.18
            Types of licenses.
            70.19
            General license for calibration or reference sources.
            70.20
            General license to own special nuclear material.
            70.20a
            General license to possess special nuclear material for transport.
            70.20b
            General license for carriers of transient shipments of formula quantities of strategic special nuclear material, special nuclear material of moderate strategic significance, special nuclear material of low strategic significance, and irradiated reactor fuel.

            Subpart D—License Applications
            70.21
            Filing.
            70.22
            Contents of applications.
            70.23
            Requirements for the approval of applications.
            70.23a
            Hearing required for uranium enrichment facility.
            70.24
            Criticality accident requirements.
            70.25
            Financial assurance and recordkeeping for decommissioning.

            Subpart E—Licenses
            70.31
            Issuance of licenses.
            70.32
            Conditions of licenses.
            70.33
            Applications for renewal of licenses.
            70.34
            Amendment of licenses.
            70.35

            Commission action on applications to renew or amend.

            70.36
            Inalienability of licenses.
            70.37
            Disclaimer of warranties.
            70.38
            Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas.
            70.39
            Specific licenses for the manufacture or initial transfer of calibration or reference sources.
            70.40
            Ineligibility of certain applicants.

            Subpart F—Acquisition, Use and Transfer of Special Nuclear Material, Creditors' Rights
            70.41
            Authorized use of special nuclear material.
            70.42
            Transfer of special nuclear material.
            70.44
            Creditor regulations.

            Subpart G—Special Nuclear Material Control, Records, Reports and Inspections
            70.50
            Reporting requirements.
            70.51
            Records requirements.
            70.52
            Reports of accidental criticality.
            70.55
            Inspections.
            70.56
            Tests.
            70.59
            Effluent monitoring reporting requirements.

            Subpart H—Additional Requirements for Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material
            70.60
            Applicability.
            70.61
            Performance requirements.
            70.62
            Safety program and integrated safety analysis.
            70.64
            Requirements for new facilities or new processes at existing facilities.
            70.65
            Additional content of applications.
            70.66
            Additional requirements for approval of license application.
            70.72
            Facility changes and change process.
            70.73
            Renewal of licenses.
            70.74
            Additional reporting requirements.
            70.76
            Backfitting.

            Subpart I—Modification and Revocation of Licenses
            70.81
            Modification and revocation of licenses.
            70.82
            Suspension and operation in war or national emergency.

            Subpart J—Enforcement
            70.91
            Violations.
            70.92
            Criminal penalties.
            Appendix A to Part 70—Reportable Safety Events

          Authority:
          Atomic Energy Act of 1954, secs. 51, 53, 57(d), 108, 122, 161, 182, 183, 184, 186, 187, 193, 223, 234, 274, 1701 (42 U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201, 2232, 2233, 2234, 2236, 2237, 2243, 2273, 2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
          Sections 70.1(c) and 70.20a(b) also issued under secs. 135, 141, Pub. L. 97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
          Section 70.21(g) also issued under Atomic Energy Act sec. 122 (42 U.S.C. 2152).
          Section 70.31 also issued under Atomic Energy Act sec. 57(d) (42 U.S.C. 2077(d)).
          Sections 70.36 and 70.44 also issued under Atomic Energy Act sec. 184 (42 U.S.C. 2234).
          Section 70.81 also issued under Atomic Energy Act secs. 186, 187 (42 U.S.C. 2236, 2237).
          Section 70.82 also issued under Atomic Energy Act sec. 108 (42 U.S.C. 2138).

          Source:
          21 FR 764, Feb. 3, 1956, unless otherwise noted.

          Editorial Note:
          Nomenclature changes to part 70 appear at 81 FR 86910, Dec. 2, 2016.

          Subpart A—General Provisions

            § 70.1
            Purpose.
            (a) Except as provided in paragraphs (c) and (d) of this section, the regulations of this part establish procedures and criteria for the issuance of licenses to receive title to, own, acquire, deliver, receive, possess, use, and transfer special nuclear material; and establish and provide for the terms and conditions upon which the Commission will issue such licenses.
            (b) The regulations contained in this part are issued pursuant to the Atomic Energy Act of 1954, as amended (68 Stat. 919) and Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242).
            (c) The regulations in part 72 of this chapter establish requirements, procedures, and criteria for the issuance of licenses to possess:
            (1) Spent fuel, power reactor-related Greater than Class C (GTCC) waste, and other radioactive materials associated with spent fuel storage in an independent spent fuel storage installation (ISFSI), or

            (2) Spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive materials associated with the storage in a monitored retrievable storage installation (MRS), and the terms and conditions under which the Commission will issue such licenses.

            (d) As provided in part 76 of this chapter, the regulations of this part establish procedures and criteria for physical security and material control and accounting for the issuance of a certificate of compliance or the approval of a compliance plan.
            (e) As provided in the Atomic Energy Act of 1954, as amended, the regulations in this part establish requirements, procedures, and criteria for the issuance of licenses to uranium enrichment facilities.
            [21 FR 764, Feb. 3, 1956, as amended at 32 FR 4056, Mar. 15, 1967; 40 FR 8791, Mar. 3, 1975; 43 FR 6924, Feb. 17, 1978; 45 FR 74712, Nov. 12, 1980; 53 FR 31682, Aug. 19, 1988; 59 FR 48960, Sept. 23, 1994; 62 FR 6669, Feb. 12, 1997; 66 FR 51838, Oct. 11, 2001]

            § 70.2
            Scope.
            Except as provided in §§ 70.11 to 70.13, inclusive, the regulations in this part apply to all persons in the United States. This part also gives notice to all persons who knowingly provide to any licensee, applicant, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licensee's or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 70.10.
            [63 FR 1898, Jan. 13, 1998]

            § 70.3
            License requirements.
            No person subject to the regulations in this part shall receive title to, own, acquire, deliver, receive, possess, use, or transfer special nuclear material except as authorized in a license issued by the Commission pursuant to these regulations.
            [32 FR 2562, Feb. 7, 1967, as amended at 43 FR 6924, Feb. 17, 1978]

            § 70.4
            Definitions.

              Act means the Atomic Energy Act of 1954 (68 Stat 919), including any amendments thereto;

              Acute, as used in this part, means a single radiation dose or chemical exposure event or multiple radiation dose or chemical exposure events occurring within a short time (24 hours or less).

              Agreement State as designated in part 150 of this chapter means any State with which the Commission has entered into an effective agreement under subsection 274b. of the Act. Non-agreement State means any other State.

              Alert means events may occur, are in progress, or have occurred that could lead to a release of radioactive material[s] but that the release is not expected to require a response by an offsite response organization to protect persons offsite.

              Atomic energy means all forms of energy released in the course of nuclear fission or nuclear transformation;

              Atomic weapon means any device utilizing atomic energy, exclusive of the means for transporting or propelling the device (where such means is a separable and divisible part of the device), the principal purpose of which is for use as, or for development of, a weapon, a weapon prototype, or a weapon test device;

              Available and reliable to perform their function when needed, as used in subpart H of this part, means that, based on the analyzed, credible conditions in the integrated safety analysis, items relied on for safety will perform their intended safety function when needed, and management measures will be implemented that ensure compliance with the performance requirements of § 70.61 of this part, considering factors such as necessary maintenance, operating limits, common-cause failures, and the likelihood and consequences of failure or degradation of the items and measures.

              Commencement of construction means taking any action defined as “construction” or any other activity at the site of a facility subject to the regulations in this part that has a reasonable nexus to:
            (1) Radiological health and safety; or
            (2) Common defense and security.

              Commission means the Nuclear Regulatory Commission or its duly authorized representatives;

              Common defense and security means the common defense and security of the United States;

              Configuration management (CM) means a management measure that provides oversight and control of design information, safety information, and records of modifications (both temporary and permanent) that might impact the ability of items relied on for safety to perform their functions when needed.

              Construction means the installation of foundations, or in-place assembly, erection, fabrication, or testing for any structure, system, or component of a facility or activity subject to the regulations in this part that are related to radiological safety or security. The term “construction” does not include:
            (1) Changes for temporary use of the land for public recreational purposes;
            (2) Site exploration, including necessary borings to determine foundation conditions or other preconstruction monitoring to establish background information related to the suitability of the site, the environmental impacts of construction or operation, or the protection of environmental values;
            (3) Preparation of the site for construction of the facility, including clearing of the site, grading, installation of drainage, erosion and other environmental mitigation measures, and construction of temporary roads and borrow areas;
            (4) Erection of fences and other access control measures that are not related to the safe use of, or security of, radiological materials subject to this part;
            (5) Excavation;
            (6) Erection of support buildings (e.g., construction equipment storage sheds, warehouse and shop facilities, utilities, concrete mixing plants, docking and unloading facilities, and office buildings) for use in connection with the construction of the facility;
            (7) Building of service facilities (e.g., paved roads, parking lots, railroad spurs, exterior utility and lighting systems, potable water systems, sanitary sewerage treatment facilities, and transmission lines);
            (8) Procurement or fabrication of components or portions of the proposed facility occurring at other than the final, in-place location at the facility; or
            (9) Taking any other action that has no reasonable nexus to:
            (i) Radiological health and safety, or
            (ii) Common defense and security.

              Contiguous sites means licensee controlled locations, deemed by the Commission to be in close enough proximity to each other, that the special nuclear material must be considered in the aggregate for the purpose of physical protection.

              Corporation means the United States Enrichment Corporation (USEC), or its successor, a Corporation that is authorized by statute to lease the gaseous diffusion enrichment plants in Paducah, Kentucky, and Piketon, Ohio, from the Department of Energy, or any person authorized to operate one or both of the gaseous diffusion plants, or other facilities, pursuant to a plan for the privatization of USEC that is approved by the President.

              Critical mass of special nuclear material (SNM), as used in Subpart H, means special nuclear material in a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of uranium-235 is present; 450 grams of any combination thereof; or one-half such quantities if massive moderators or reflectors made of graphite, heavy water, or beryllium may be present.

              Decommission means to remove a facility or site safely from service and reduce residual radioactivity to a level that permits—
            (1) Release of the property for unrestricted use and termination of the license; or
            (2) Release of the property under restricted conditions and termination of the license.

              Department and Department of Energy means the Department of Energy Organization Act (Pub. L. 95-91, 91 Stat. 565, 42 U.S.C. 7101 et seq.), to the extent that the Department, or its duly authorized representatives, exercises functions formerly vested in the U.S. Atomic Energy Commission, its Chairman, members, officers and components and transferred to the U.S. Energy Research and Development Administration and to the Administrator thereof pursuant to sections 104(b), (c) and (d) of the Energy Reorganization Act of 1974 (Pub. L. 93-438, 88 Stat. 1233 at 1237, 42 U.S.C. 5814) and retransferred to the Secretary of Energy pursuant to section 301(a) of the Department of Energy Organization Act (Pub. L. 95-91, 91 Stat. 565 at 577-578, 42 U.S.C. 7151).

              Double contingency principle means that process designs should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.

              Effective dose equivalent means the sum of the products of the dose equivalent to the body organ or tissue and the weighting factors applicable to each of the body organs or tissues that are irradiated. Weighting factors are: 0.25 for gonads, 0.15 for breast, 0.12 for red bone marrow, 0.12 for lungs, 0.03 for thyroid, 0.03 for bone surface, and 0.06 for each of the other five organs receiving the highest dose equivalent.

              Effective kilograms of special nuclear material means: (1) For plutonium and uranium-233 their weight in kilograms; (2) For uranium with an enrichment in the isotope U-235 of 0.01 (1%) and above, its element weight in kilograms multiplied by the square of its enrichment expressed as a decimal weight fraction; and (3) For uranium with an enrichment in the isotope U-235 below 0.01 (1%), by its element weight in kilograms multiplied by 0.0001.

              Formula quantity means strategic special nuclear material in any combination in a quantity of 5000 grams or more computed by the formula, grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium). This class of material is sometimes referred to as a Category I quantity of material.

              Government agency means any executive department, commission, independent establishment, corporation, wholly or partly owned by the United States of America which is an instrumentality of the United States, or any board, bureau, division, service, office, officer, authority, administration, or other establishment in the executive branch of the Government;

              Hazardous chemicals produced from licensed materials means substances having licensed material as precursor compound(s) or substances that physically or chemically interact with licensed materials; and that are toxic, explosive, flammable, corrosive, or reactive to the extent that they can endanger life or health if not adequately controlled. These include substances commingled with licensed material, and include substances such as hydrogen fluoride that is produced by the reaction of uranium hexafluoride and water, but do not include substances prior to process addition to licensed material or after process separation from licensed material.

              Integrated safety analysis (ISA) means a systematic analysis to identify facility and external hazards and their potential for initiating accident sequences, the potential accident sequences, their likelihood and consequences, and the items relied on for safety. As used here, integrated means joint consideration of, and protection from, all relevant hazards, including radiological, nuclear criticality, fire, and chemical. However, with respect to compliance with the regulations of this part, the NRC requirement is limited to consideration of the effects of all relevant hazards on radiological safety, prevention of nuclear criticality accidents, or chemical hazards directly associated with NRC licensed radioactive material. An ISA can be performed process by process, but all processes must be integrated, and process interactions considered.

              Integrated safety analysis summary means a document or documents submitted with the license application, license amendment application, license renewal application, or pursuant to § 70.62(c)(3)(ii) that provides a synopsis of the results of the integrated safety analysis and contains the information specified in § 70.65(b). The ISA Summary can be submitted as one document for the entire facility, or as multiple documents that cover all portions and processes of the facility.

              Items relied on for safety mean structures, systems, equipment, components, and activities of personnel that are relied on to prevent potential accidents at a facility that could exceed the performance requirements in § 70.61 or to mitigate their potential consequences. This does not limit the licensee from identifying additional structures, systems, equipment, components, or activities of personnel (i.e.,
              beyond those in the minimum set necessary for compliance with the performance requirements) as items relied on for safety.

              License, except where otherwise specified, means a license issued pursuant to the regulations in this part;

              Management measures mean the functions performed by the licensee, generally on a continuing basis, that are applied to items relied on for safety, to ensure the items are available and reliable to perform their functions when needed. Management measures include configuration management, maintenance, training and qualifications, procedures, audits and assessments, incident investigations, records management, and other quality assurance elements.

              Person means (1) any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, Government agency other than the Commission or the Department, except that the Department shall be considered a person within the meaning of the regulations in this part to the extent that its facilities and activities are subject to the licensing and related regulatory authority of the Commission pursuant to section 202 of the Energy Reorganization Act of 1974 (88 Stat. 1244), any State or any political subdivision of or any political entity within a State, any foreign government or nation or any political subdivision of any such government or nation, or other entity; and (2) any legal successor, representative, agent, or agency of the foregoing;

              Plutonium processing and fuel fabrication plant means a plant in which the following operations or activities are conducted: (1) Operations for manufacture of reactor fuel containing plutonium including any of the following: (i) Preparation of fuel material; (ii) formation of fuel material into desired shapes; (iii) application of protective cladding; (iv) recovery of scrap material; and (v) storage associated with such operations; or (2) Research and development activities involving any of the operations described in paragraph (1) of this definition except for research and development activities utilizing unsubstantial amounts of plutonium.

              Principal activities, as used in this part, means activities authorized by the license which are essential to achieving the purpose(s) for which the license was issued or amended. Storage during which no licensed material is accessed for use or disposal and activities incidental to decontamination or decommissioning are not principal activities.

              Produce, when used in relation to special nuclear material, means (1) to manufacture, make, produce, or refine special nuclear material; (2) to separate special nuclear material from other substances in which such material may be contained; or (3) to make or to produce new special nuclear material;

              Research and development means (1) theoretical analysis, exploration, or experimentation; or (2) the extension of investigative findings and theories of a scientific or technical nature into practical application for experimental and demonstration purposes, including the experimental production and testing of models, devices, equipment, materials, and processes;

              Restricted Data means all data concerning (1) design, manufacture or utilization of atomic weapons; (2) the production of special nuclear material; or (3) the use of special nuclear material in the production of energy, but shall not include data declassified or removed from the Restricted Data category pursuant to section 142 of the Act;

              Sealed source means any special nuclear material that is encased in a capsule designed to prevent leakage or escape of the special nuclear material.

              Site Area emergency means events may occur, are in progress, or have occurred that could lead to a significant release of radioactive material and that could require a response by offsite response organizations to protect persons offsite.

              Source material means source material as defined in section 11z. of the Act and in the regulations contained in part 40 of this chapter;

              Special nuclear material means (1) plutonium, uranium 233, uranium enriched in the isotope 233 or in the isotope 235, and any other material which the Commission, pursuant to the provisions of section 51 of the act, determines to be special nuclear material, but does not include source material; or (2) any material artificially enriched by any of the foregoing but does not include source material;

              Special nuclear material of low strategic significance means:
            (1) Less than an amount of special nuclear material of moderate strategic significance as defined in paragraph (1) of the definition of strategic nuclear material of moderate strategic significance in this section, but more than 15 grams of uranium-235 (contained in uranium enriched to 20 percent or more in U-235 isotope) or 15 grams of uranium-233 or 15 grams of plutonium or the combination of 15 grams when computed by the equation, grams = (grams contained U-235) + (grams plutonium) + (grams U-233); or
            (2) Less than 10,000 grams but more than 1,000 grams of uranium-235 (contained in uranium enriched to 10 percent or more but less than 20 percent in the U-235 isotope); or
            (3) 10,000 grams or more of uranium-235 (contained in uranium enriched above natural but less than 10 percent in the U-235 isotope).
            This class of material is sometimes referred to as a Category III quantity of material.

              Special nuclear material of moderate strategic significance means:
            (1) Less than a formula quantity of strategic special nuclear material but more than 1,000 grams of uranium-235 (contained in uranium enriched to 20 percent or more in the U-235 isotope) or more than 500 grams of uranium-233 or plutonium, or in a combined quantity of more than 1,000 grams when computed by the equation, grams = (grams contained U-235) + 2 (grams U-233 + grams plutonium); or
            (2) 10,000 grams or more of uranium-235 (contained in uranium enriched to 10 percent or more but less than 20 percent in the U-235 isotope).
            This class of material is sometimes referred to as a Category II quantity of material.

              Special nuclear material scrap means the various forms of special nuclear material generated during chemical and mechanical processing, other than recycle material and normal process intermediates, which are unsuitable for use in their present form, but all or part of which will be used after further processing.

              Strategic special nuclear material means uranium-235 (contained in uranium enriched to 20 percent or more in the U235 isotope), uranium-233, or plutonium.

              Transient shipment means a shipment of nuclear material, originating and terminating in foreign countries, on a vessel or aircraft which stops at a United States port.

              Unacceptable performance deficiencies mean deficiencies in the items relied on for safety or the management measures that need to be corrected to ensure an adequate level of protection as defined in 10 CFR 70.61(b), (c), or (d).

              United States, when used in a geographical sense, includes Puerto Rico and all territories and possessions of the United States.

              Uranium enrichment facility means:
            (1) Any facility used for separating the isotopes of uranium or enriching uranium in the isotope 235, except laboratory scale facilities designed or used for experimental or analytical purposes only; or
            (2) Any equipment or device, or important component part especially designed for such equipment or device, capable of separating the isotopes of uranium or enriching uranium in the isotope 235.

              Worker, when used in Subpart H of this Part, means an individual who receives an occupational dose as defined in 10 CFR 20.1003.
            [21 FR 764, Feb. 3, 1956]

              Editorial Note:
              For Federal Register citations affecting § 70.4, see the List of CFR Sections Affected, which appears in the Finding Aids section of the printed volume and at www.govinfo.gov.

            § 70.5
            Communications.
            (a) Unless otherwise specified or covered under the regional licensing program as provided in paragraph (b) of this section, any communication or report concerning the regulations in this part and any application filed under these regulations may be submitted to the Commission as follows:

            (1) By mail addressed to: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards or Director, Office of Nuclear Security and Incident Response, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
            (2) By hand delivery to the Director, Office of Nuclear Material Safety and Safeguards or Director, Office of Nuclear Security and Incident Response at the NRC's offices at 11555 Rockville Pike, Rockville, Maryland.

            (3) Where practicable, by electronic submission, for example, via Electronic Information Exchange, and CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.
            (4) Classified communications shall be transmitted to the NRC Headquarters' classified mailing address as specified in appendix A to part 73 of this chapter or delivered by hand in accordance with paragraph (a)(2) of this section.
            (b) The Commission has delegated to the four Regional Administrators licensing authority for selected parts of its decentralized licensing program for nuclear materials as described in paragraph (b)(1) of this section. Any communication, report, or application covered under this licensing program must be submitted to the appropriate Regional Administrator. The Administrators' jurisdictions and mailing addresses are listed in paragraph (b)(2) of this section.
            (1) The delegated licensing program includes authority to issue, renew, amend, cancel, modify, suspend, or revoke licenses for nuclear materials issued pursuant to 10 CFR parts 30 through 36, 39, 40, and 70 to all persons for academic, medical, and industrial uses, with the following exceptions:
            (i) Activities in the fuel cycle and special nuclear material in quantities sufficient to constitute a critical mass in any room or area. This exception does not apply to license modifications relating to termination of special nuclear material licenses that authorize possession of larger quantities when the case is referred for action from NRC's Headquarters to the Regional Administrators.
            (ii) Health and safety design review of sealed sources and devices and approval, for licensing purposes, of sealed sources and devices.
            (iii) Processing of source material for extracting of metallic compounds (including Zirconium, Hafnium, Tantalum, Titanium, Niobium, etc.).
            (iv) Distribution of products containing radioactive material under §§ 32.11 through 32.30 and 40.52 of this chapter to persons exempt from licensing requirements.
            (v) New uses or techniques for use of byproduct, source, or special nuclear material.
            (vi) Reviews pursuant to § 70.32(c).
            (vii) Uranium enrichment facilities.
            (2) Submissions—(i) Region I. The regional licensing program involves all Federal facilities in the region and non-Federal licensees in the following Region I non-Agreement States and the District of Columbia: Connecticut, Delaware, and Vermont. All mailed or hand-delivered inquiries, communications, and applications for a new license or an amendment or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region I, Nuclear Material Section B, 2100 Renaissance Boulevard, Suite 100, King of Prussia, PA 19406-2713; where email is appropriate it should be addressed to RidsRgn1MailCenter.Resource@nrc.gov.

            (ii) Region II. The regional licensing program involves all Federal facilities in the region and non-Federal licensees in the following Region II non-Agreement States and territories: West Virginia, Puerto Rico, and the Virgin Islands. All mailed or hand-delivered inquiries, communications, and applications for a new license or an amendment, renewal, or termination request of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region I, Nuclear Material Section B, 475 Allendale Road, King of Prussia, PA 19406-1415; where e-mail is appropriate it should be addressed to RidsRgn1MailCenter.Resource@nrc.gov.

            (iii) Region III. (A) The regional licensing program for mining and milling involves all Federal facilities in the region, and non-Federal licensees in the Region III non-Agreement States of Indiana, Michigan, Missouri and Region III Agreement States of Minnesota, Wisconsin, and Iowa. All mailed or hand-delivered inquiries, communications, and applications for a new license or an amendment, renewal, or termination request of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region III, Material Licensing Section, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; where e-mail is appropriate it should be addressed to RidsRgn3MailCenter.Resource@nrc.gov.

            (B) Otherwise, the regional licensing program involves all Federal facilities in the region and non-Federal licensees in the following Region III non-Agreement States: Indiana, Michigan, and Missouri. All mailed or hand-delivered inquiries, communications, and applications for a new license or an amendment, or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region III, Material Licensing Section, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; where e-mail is appropriate it should be addressed to RidsRgn3MailCenter.Resource@nrc.gov. Outside of this jurisdiction, concerning the licensing program involving mining and milling, the Agreement States of Illinois and Ohio should be contacted.
            (iv) Region IV. (A) The regional licensing program for mining and milling involves all Federal facilities in the region, and non-Federal licensees in the Region IV non-Agreement States and territory of Alaska, Hawaii, Idaho, Montana, South Dakota, Wyoming and Guam and Region IV Agreement States of Oregon, California, Nevada, New Mexico, Louisiana, Mississippi, Arkansas, Oklahoma, Kansas, Nebraska, and North Dakota. All mailed or hand-delivered inquiries, communications, and applications for a new license or an amendment or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region IV, Division of Nuclear Materials Safety, 1600 E. Lamar Blvd., Arlington, TX 76011-4511; where email is appropriate, it should be addressed to RidsRgn4MailCenter@nrc.gov.

            (B) Otherwise, the regional licensing program involves all Federal facilities in the region and non-Federal licensees in the following Region IV non-Agreement States and territory: Alaska, Hawaii, Idaho, Montana, South Dakota, Wyoming, and Guam. All mailed or hand-delivered inquiries, communications, and applications for a new license or an amendment or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region IV, Division of Nuclear Materials Safety, 1600 E. Lamar Blvd., Arlington, TX 76011-4511; where email is appropriate, it should be addressed to RidsRgn4MailCenter@nrc.gov. Outside of this jurisdiction, concerning the licensing program involving mining and milling, the Agreement States of Colorado, Utah, Texas and Washington should be contacted.
            [48 FR 16032, Apr. 14, 1983]

              Editorial Note:
              For Federal Register citations affecting § 70.5, see the List of CFR Sections Affected, which appears in the Finding Aids section of the printed volume and at www.govinfo.gov.

            § 70.6
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

            § 70.7
            Employee protection.
            (a) Discrimination by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act or the Energy Reorganization Act.
            (1) The protected activities include but are not limited to:
            (i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) introductory text of this section or possible violations of requirements imposed under either of those statutes;
            (ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) introductory text or under these requirements if the employee has identified the alleged illegality to the employer;
            (iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;
            (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) introductory text.
            (v) Assisting or participating in, or is about to assist or participate in, these activities.
            (2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.
            (3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.
            (b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.
            (c) A violation of paragraphs (a), (e), or (f) of this section by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant may be grounds for—
            (1) Denial, revocation, or suspension of the license.
            (2) Imposition of a civil penalty on the licensee, applicant, or a contractor or subcontractor of the licensee or applicant.
            (3) Other enforcement action.
            (d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.

            (e)(1) Each specific licensee, each applicant for a specific license, and each general licensee subject to part 19 shall prominently post the revision of NRC Form 3, “Notice to Employees,” referenced in 10 CFR 19.11(c).
            (2) The posting of NRC Form 3 must be at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, and for 30 days following license termination.

            (3) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, via email to Forms.Resource@nrc.gov, or by visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.

            (f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.
            [58 FR 52413, Oct. 8, 1993, as amended at 60 FR 24552, May 9, 1995; 61 FR 6765, Feb. 22, 1996; 68 FR 58816, Oct. 10, 2003; 72 FR 63974, Nov. 14, 2007; 73 FR 30459, May 28, 2008; 79 FR 66605, Nov. 10, 2014]

            § 70.8
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the office of Management and Budget (OMB) for approval as required by the Paperwork reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0009.
            (b) The approved information collection requirements contained in this part appear in §§ 70.9, 70.17, 70.19, 70.20a, 70.20b, 70.21, 70.22, 70.24, 70.25, 70.32, 70.33, 70.34, 70.38, 70.39, 70.42, 70.50, 70.51, 70.52, 70.59, 70.61, 70.62, 70.64, 70.65, 70.72, 70.73, 70.74, and Appendix A.
            (c) This part contains information collection requirements in addition to those approved under the control number specified in paragraph (a) of this section. These information collection requirements and the control numbers under which they are approved are as follows:
            (1) In § 70.21(g), Form N-71 and associated forms are approved under control number 3150-0056.
            (2) In § 70.38, NRC form 314 is approved under control number 3150-0028.
            (3) In § 70.21(g), DOC/NRC Forms AP-1, AP-A, and associated forms are approved under control number 0694-0135.
            [49 FR 19628, May 9, 1984, as amended at 52 FR 19305, May 22, 1987; 56 FR 40769, Aug. 16, 1991; 57 FR 18392, Apr. 30, 1992; 58 FR 39634, July 26, 1993; 62 FR 52189, Oct. 6, 1997; 65 FR 56225, Sept. 18, 2000; 67 FR 78142, Dec. 23, 2003; 73 FR 78606, Dec. 23, 2008]

            § 70.9
            Completeness and accuracy of information.
            (a) Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.

            (b) Each applicant or licensee shall notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant or licensee violates this paragraph only if the applicant or licensee fails to notify the Commission of information that the applicant or licensee has identified as having a significant implication for public health and safety or common defense and security. Notification shall be provided to the Administrator of the appropriate Regional Office within two working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.
            [52 FR 49373, Dec. 31, 1987]

            § 70.10
            Deliberate misconduct.
            (a) Any licensee, applicant for a license, employee of a licensee or applicant; or any contractor (including a supplier or consultant), subcontractor, employee of a contractor or subcontractor of any licensee or applicant for a license, who knowingly provides to any licensee, applicant, contractor, or subcontractor, any components, equipment, materials, or other goods or services that relate to a licensee's or applicant's activities in this part, may not:
            (1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission; or
            (2) Deliberately submit to the NRC, a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.
            (b) A person who violates paragraph (a)(1) or (a)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.
            (c) For the purposes of paragraph (a)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:
            (1) Would cause a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license issued by the Commission; or
            (2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor.
            [63 FR 1899, Jan. 13, 1998]

          Subpart B—Exemptions

            § 70.11
            Persons using special nuclear material under certain Department of Energy and Nuclear Regulatory Commission contracts.
            Except to the extent that Department facilities or activities of the types subject to licensing pursuant to section 202 of the Energy Reorganization Act of 1974 are involved, any prime contractor of the Department is exempt from the requirements for a license set forth in section 53 of the Act and from the regulations in this part to the extent that such contractor, under his prime contract with the Department receives title to, owns, acquires, delivers, receives, possesses, uses, or transfers special nuclear material for:

            (a) The performance of work for the Department at a United States Government-owned or controlled site, including the transportation of special nuclear material to or from such site and the performance of contract services during temporary interruptions of such transportation; (b) research in, or development, manufacture, storage, testing or transportation of, atomic weapons or components thereof; or (c) the use or operation of nuclear reactors or other nuclear devices in a United States Government-owned vehicle or vessel. In addition to the foregoing exemptions, and subject to the requirement for licensing of Department facilities and activities pursuant to section 202 of the Energy Reorganization Act of 1974, any prime contractor or subcontractor of the Department or the Commission is exempt from the requirements for a license set forth in section 53 of the Act and from the regulations in this part to the extent that such prime contractor or subcontractor receives title to, owns, acquires, delivers, receives, possesses, uses, or transfers special nuclear material under his prime contract or subcontract when the Commission determines that the exemption of the prime contractor or subcontractor is authorized by law; and that, under the terms of the contract or subcontract there is adequate assurance that the work thereunder can be accomplished without undue risk to the public health and safety.
            [40 FR 14085, Mar. 28, 1975; 40 FR 16047, Apr. 9, 1975, as amended at 43 FR 6924, Feb. 17, 1978; 65 FR 54950, Sept. 12, 2000]

            § 70.12
            Carriers.
            Common and contract carriers, freight forwarders, warehousemen, and the U.S. Postal Service are exempt from the regulations in this part to the extent that they transport special nuclear material in the regular course of carriage for another or storage incident thereto. This exemption does not apply to the storage in transit or transport of material by persons covered by the general license issued under § 70.20a and § 70.20b.
            [46 FR 12696, Feb. 18, 1981]

            § 70.13
            Department of Defense.
            The regulations in this part do not apply to the Department of Defense to the extent that the Department receives, possesses and uses special nuclear material in accordance with the direction of the President pursuant to section 91 of the Act.

            § 70.14
            Foreign military aircraft.
            The regulations in this part do not apply to persons who carry special nuclear material (other than plutonium) in aircraft of the armed forces of foreign nations subject to 49 U.S.C. 40103(d).
            [71 FR 15012, Mar. 27, 2006]

            § 70.17
            Specific exemptions.
            (a) The Commission may, upon application of any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.
            (b) [Reserved]
            (c) The DOE is exempt from the requirements of the regulations in this part to the extent that its activities are subject to the requirements of part 60 or part 63 of this chapter.
            (d) Except as specifically provided in part 61 of this chapter, any licensee is exempt from the requirements of the regulations in this part to the extent that its activities are subject to the requirements of part 61 of this chapter.
            [37 FR 5749, Mar. 21, 1972, as amended at 45 FR 65536, Oct. 3, 1980; 46 FR 13987, Feb. 25, 1981; 47 FR 57481, Dec. 27, 1982; Redesignated at 65 FR 56225, Sept. 18, 2000, as amended at 66 FR 55815, Nov. 2, 2001]

          Subpart C—General Licenses

            § 70.18
            Types of licenses.
            Licenses for special nuclear material are of two types: general and specific. Any general license provided in this part is effective without the filing of applications with the Commission or the issuance of licensing documents to particular persons. Specific licenses are issued to named persons upon applications filed pursuant to the regulations in this part.
            [29 FR 5884, May 5, 1964]

            § 70.19
            General license for calibration or reference sources.
            (a) A general license is hereby issued to those persons listed below to receive title to, own, acquire, deliver, receive, possess, use and transfer in accordance with the provisions of paragraphs (b) and (c) of this section, plutonium in the form of calibration or reference sources:
            (1) Any person in a non-agreement State who holds a specific license issued by the Commission or the Atomic Energy Commission which authorizes him to receive, possess, use and transfer byproduct material, source material, or special nuclear material;
            (2) Any Government agency as defined in § 70.4 that holds a specific license issued by the Commission that authorizes it to receive, possess, use, or transfer byproduct material, source material, or special nuclear material; and
            (3) Any person in an agreement State who holds a specific license issued by the Commission or the Atomic Energy Commission which authorizes him to receive, possess, use and transfer special nuclear material.

            (b) The general license in paragraph (a) of this section applies only to calibration or reference sources which have been manufactured or initially transferred in accordance with the specifications contained in a specific license issued pursuant to § 70.39 or in accordance with the specifications contained in a specific license issued by an agreement State which authorizes manufacture of the sources for distribution to persons generally licensed by the agreement State.
            (c) The general license in paragraph (a) of this section is subject to the provisions of §§ 70.32, 70.50, 70.55, 70.56, 70.91, 70.81, and 70.82; the provisions of §§ 74.11 and 74.19 of this chapter; and to the provisions of parts 19, 20, and 21 of this chapter. In addition, persons who receive title to own, acquire, deliver, receive, possess, use or transfer one or more calibration or reference sources under this general license:
            (1) Shall not possess at any one time, at any one location of storage or use, more than 5 microcuries of plutonium in such sources;
            (2) Shall not receive, possess, use or transfer such source unless the source, or the storage container, bears a label which includes the following statement or a substantially similar statement which contains the information called for in the following statement: 1

                1 Sources generally licensed under this section prior to January 19, 1975 may bear labels authorized by the regulations in effect on January 1, 1975.

              The receipt, possession, use and transfer of this source, Model ___, Serial No. ___, are subject to a general license and the regulations of the United States Nuclear Regulatory Commission or of a State with which the Commission has entered into an agreement for the exercise of regulatory authority. Do not remove this label.
              caution—radioactive material—this source contains plutonium. do not touch radioactive portion of this source.
              (Name of Manufacturer or Initial Transferor)

            (3) Shall not transfer, abandon, or dispose of such source except by transfer to a person authorized by a license from the Commission or the Atomic Energy Commission or an Agreement State to receive the source.
            (4) Shall store such source, except when the source is being used, in a closed container adequately designed and constructed to contain plutonium which might otherwise escape during storage.
            (5) Shall not use such source for any purpose other than the calibration of radiation detectors or the standardization of other sources.
            (d) The general license in paragraph (a) of this section does not authorize the manufacture, import, or export of calibration or reference sources containing plutonium.
            [29 FR 5884, May 5, 1964, as amended at 32 FR 8124, June 7, 1967; 38 FR 22221, Aug. 17, 1973; 40 FR 8792, Mar. 3, 1975; 42 FR 28896, June 6, 1977; 43 FR 6924, Feb. 17, 1978; 48 FR 32329, July 15, 1983; 56 FR 40769, Aug. 16, 1991; 57 FR 33428, July 29, 1992; 67 FR 78142, Dec. 23, 2002; 72 FR 35145, June 27, 2007]

            § 70.20
            General license to own special nuclear material.
            A general license is hereby issued to receive title to and own special nuclear material without regard to quantity. Notwithstanding any other provision of this chapter, a general licensee under this section is not authorized to acquire, deliver, receive, possess, use, transfer, import, or export special nuclear material, except as authorized in a specific license.
            [33 FR 9810, July 9, 1968]

            § 70.20a
            General license to possess special nuclear material for transport.
            (a) A general license is issued to any person to possess formula quantities of strategic special nuclear material of the types and quantities subject to the requirements of §§ 73.20, 73.25, 73.26 and 73.27 of this chapter, and irradiated reactor fuel containing material of the types and quantities subject to the requirements of § 73.37 of this chapter, in the regular course of carriage for another or storage incident. Carriers generally licensed under § 70.20b are exempt from the requirements of this section. Carriers of irradiated reactor fuel for the United States Department of Energy are also exempt from the requirements of this section. The general license is subject to the applicable provisions of §§ 70.7 (a) through (e), 70.32 (a) and (b), and §§ 70.42, 70.52, 70.55, 70.91, 70.81, 70.82 and 10 CFR 74.11.

            (b) Notwithstanding any other provision of this chapter, the general license issued under this section does not authorize any person to conduct any activity that would be authorized by a license issued pursuant to parts 30 through 36, 39, 40, 50, 72, 110, or other sections of this part.
            (c) Notwithstanding any other provision of this chapter, the duties of a general licensee under this section while in possession of formula quantities of strategic special nuclear material or irradiated reactor fuel in the regular course of carriage for another or storage incident thereto shall be limited to providing for the physical protection of such material against theft or sabotage. Unless otherwise provided by this section, a general license under this section is not subject to the requirements of parts 19, 20, 70 and 73.
            (d) Any person who possesses formula quantities of strategic special nuclear material under this general license:
            (1) Shall have submitted and received approval of a transportation security plan. The security plan shall outline the procedures that will be used to meet the requirements of §§ 73.20, 73.25, 73.26, 73.27 and 73.70(g) of this chapter including a plan for the selection, qualification, and training of armed escorts, or the specification and design of a specially designed truck or trailer as appropriate.
            (2) Shall assure that the transportation is in accordance with the applicable physical protection requirements of §§ 73.20, 73.25, 73.26, 73.27 and 73.70(g) of this chapter and the applicable approved transportation security plan.
            (3) Shall be subject to part 26 and § 73.80 of this chapter.
            (e) Any person who possesses irradiated reactor fuel under this general license shall:
            (1) Assure or receive certification from the shipper that the transportation is in accordance with the applicable physical protection requirements of § 73.37 of this chapter; and
            (2) Comply with the reporting requirements of § 73.71 of this chapter.
            [44 FR 26851, May 8, 1979, as amended at 44 FR 68186, Nov. 28, 1979; 46 FR 12696, Feb. 18, 1981; 47 FR 30458, July 14, 1982; 53 FR 31682, Aug. 19, 1988; 58 FR 7737, Feb. 9, 1993; 58 FR 31471, June 3, 1993; 67 FR 78142, Dec. 23, 2002; 72 FR 35145, June 27, 2007]

            § 70.20b
            General license for carriers of transient shipments of formula quantities of strategic special nuclear material, special nuclear material of moderate strategic significance, special nuclear material of low strategic significance, and irradiated reactor fuel.
            (a) A general license is hereby issued to any person to possess transient shipments of the following kinds and quantities of special nuclear material:
            (1) A formula quantity of special nuclear material of the types and quantities subject to the requirements of §§ 73.20, 73.25, 73.26, and 73.27 of this chapter.
            (2) Special nuclear material of moderate and low strategic significance of the types and quantities subject to the requirements of § 73.67 of this chapter.
            (3) Irradiated reactor fuel of the type and quantity subject to the requirements of § 73.37 of this chapter.
            (b) Persons generally licensed under this section are exempt from the requirements of parts 19 and 20 of this chapter and the requirements of this part, except §§ 70.32 (a) and (b), 70.52, 70.55, 70.91, 70.81, and 70.82.
            (c) Persons generally licensed under this section to possess a transient shipment of special nuclear material of the kind and quantity specified in paragraph (a)(1) of this section shall provide physical protection for that shipment in accordance with or equivalent to §§ 73.20(a), 73.20(b), 73.25, and 73.71(b) of this chapter from the time a shipment enters a United States port until it exits that or another United States port.
            (d) Persons generally licensed under this section to possess a transient shipment of special nuclear material of moderate or low strategic significance of the kind and quantity specified in paragraph (a)(2) of this section shall provide physical protection for that shipment in accordance with or equivalent to § 73.67 of this chapter and shall comply with the requirements of § 73.71(b) of this chapter.

            (e) Persons generally licensed under this section to possess a transient shipment of irradiated reactor fuel of the kind and quantity specified in paragraph (a)(3) of this section shall provide physical protection for that shipment in accordance with or equivalent to § 73.37 of this chapter and shall comply with the requirements of § 73.71(b) of this chapter.
            (f)(1) Persons generally licensed under this section, who plan to carry transient shipments with scheduled stops at United States ports, shall notify in writing the Director, Office of Nuclear Security and Incident Response, using an appropriate method listed in § 70.5(a). Classified notifications shall be sent to the NRC headquarters classified mailing address listed in appendix A to part 73 of this chapter.
            (2) A person generally licensed under this section shall assure that:
            (i) The notification will be received at least 10 days before transport of the shipment commences at the shipping facility;
            (ii) The NRC Headquarters Operations Center shall be notified by telephone at least 2 days before commencement of the shipment at the numbers listed in appendix A to part 73 of this chapter. Classified notifications shall be made by secure telephone.
            (iii) The NRC Headquarters Operations Center shall be notified by telephone of schedule changes greater than ±6 hours at the numbers listed in appendix A to part 73 of this chapter. Classified notifications shall be made by secure telephone.
            (3) Persons who are generally licensed under paragraph (a)(1) of this section must include the information listed in paragraphs (f)(3)(i) through (ix) of this section. Persons who are generally licensed under § 70.20b(a)(2) and § 70.20b(a)(3) must include the information listed in paragraphs (f)(3) (i) through (viii) of this section.
            (i) Location of all scheduled stops in United States territory;
            (ii) Arrival and departure times for all scheduled stops in United States territory;
            (iii) The type of transport vehicle;
            (iv) A physical description of the shipment (elements, isotopes, and enrichments);
            (v) The number and types of containers;
            (vi) The name and telephone number of the carrier's representative at each stopover location in United States territory;
            (vii) The estimated time and date that shipment will commence and that each country (other than the United States) along the route is scheduled to be entered;
            (viii) For shipments between countries that are not party to the Convention on the Physical Protection of Nuclear Material, provide assurances, as far as is practicable, that this nuclear material will be protected during international transport at levels described in Annex I to that Convention (see appendices E and F of part 73 of this chapter); and
            (ix) A physical protection plan for implementing the requirement of § 70.20b(c), which will include the use of armed personnel to protect the shipment during the time the shipment is in a United States port.
            (g) Persons generally licensed under this section making unscheduled stops at United States ports, immediately after the decision to make an unscheduled stop, shall:
            (1) Provide to the Director, Office of Security Policy and Incident Response, the information required under paragraph (f) of this section.
            (2) In the case of persons generally licensed under paragraph (a)(1) of this section, arrange for local law enforcement authorities or trained and qualified private guards to protect the shipment during the stop.
            (3) In the case of persons generally licensed under paragraph (a)(2) of this section, arrange for the shipment to be protected as required in § 73.67(e) of this chapter.
            (4) In the case of persons generally licensed under paragraph (a)(3) of this section, arrange for the shipment to be protected as required in § 73.37(e) of this chapter.
            (5) Implement these arrangements within a reasonable time after the arrival of the shipment at a United States port to remain in effect until the shipment exits that or another United States port.
            [52 FR 9652, Mar. 26, 1987, as amended at 60 FR 24552, May 9, 1995; 67 FR 3585, Jan. 25, 2002; 68 FR 14529, Mar. 26, 2003; 68 FR 23575, May 5, 2003; 68 FR 58817, Oct. 10, 2003; 72 FR 35145, June 27, 2007; 74 FR 62683, Dec. 1, 2009; 83 FR 58723, Nov. 21, 2018]

          Subpart D—License Applications

            § 70.21
            Filing.
            (a)(1) A person may apply for a license to possess and use special nuclear material in a plutonium processing or fuel fabrication plant, or for a uranium enrichment facility license, by filing the application with the Director of the NRC's Office of Nuclear Material Safety and Safeguards in accordance with the instructions in § 70.5(a). If the application is on paper or CD-ROM, only one copy need be provided. If the application is to be submitted electronically, see guidance for electronic submissions to the Commission.
            (2) A person may apply for any other license issued under this part, by filing the application in accordance with the instructions in § 70.5(a). If the application is on paper, only one copy need be provided. If the application is to be submitted electronically, see guidance for electronic submissions to the Commission.
            (3) Information contained in previous applications, statements, or reports filed with the Commission may be incorporated by reference if the references are clear and specific.
            (b) An application for license filed pursuant to the regulations in this part will be considered also as an application for licenses authorizing other activities for which licenses are required by the Act, provided the application specifies the additional activities for which licenses are requested and complies with regulations of the Commission as to applications for such licenses.
            (c) Any application which contains Restricted Data shall be prepared in such manner that all Restricted Data are separated from the unclassified information.
            (d) Applications and documents submitted to the Commission in connection with applications may be made available for public inspection in accordance with the provisions of the regulations contained in part 2 of this chapter.
            (e) Each application for a special nuclear material license, other than a license exempted from part 170 of this chapter, shall be accompanied by the fee prescribed in § 170.31 of this chapter. No fee will be required to accompany an application for renewal or amendment of a license, except as provided in § 170.31 of this chapter.
            (f) An application for a license to possess and use special nuclear material for processing and fuel fabrication, scrap recovery or conversion of uranium hexafluoride, or for the conduct of any other activity which the Commission has determined pursuant to subpart A of part 51 of this chapter will significantly affect the quality of the environment shall be filed at least 9 months prior to commencement of construction of the plant or facility in which the activity will be conducted, and shall be accompanied by an Environmental Report required under subpart A of part 51 of this chapter.
            (g)(1) In response to a written request by the Commission, each applicant for a construction authorization or license and each recipient of a construction authorization or a license to possess and use special nuclear material shall submit facility information, as described in § 75.10 of this chapter, on Form N-71 and associated forms and site information on DOC/NRC Form AP-A and associated forms;
            (2) As required by the Additional Protocol, applicants and licensees specified in paragraph (a) of this section shall submit location information described in § 75.11 of this chapter on DOC/NRC Form AP-1 and associated forms; and
            (3) Shall permit verification thereof by the International Atomic Energy Agency (IAEA) and take other action as necessary to implement the US/IAEA Safeguards Agreement, as described in Part 75 of this chapter.
            (h) A license application for a uranium enrichment facility must be accompanied by an Environmental Report required under subpart A of part 51 of this chapter.
            [21 FR 764, Feb. 3, 1956, as amended at 23 FR 1122, Feb. 21, 1958; 31 FR 4670, Mar. 19, 1966; 34 FR 19546, Dec. 11, 1969; 36 FR 146, Jan. 6, 1971; 37 FR 5749, Mar. 21, 1972; 49 FR 9406, Mar. 12, 1984; 49 FR 19628, 19632, May 9, 1984; 49 FR 21699, May 23, 1984; 57 FR 18392, Apr. 30, 1992; 68 FR 58817, Oct. 10, 2003; 73 FR 78606, Dec. 23, 2008]

            § 70.22
            Contents of applications.
            (a) Each application for a license shall contain the following information:
            (1) The full name, address, age (if an individual), and citizenship of the applicant and the names and addresses of three personal references. If the applicant is a corporation or other entity, it shall indicate the State where it was incorporated or organized, the location of the principal office, the names, addresses, and citizenship of its principal officers, and shall include information known to the applicant concerning the control or ownership, if any, exercised over the applicant by any alien, foreign corporation, or foreign government;
            (2) The activity for which the special nuclear material is requested, or in which special nuclear material will be produced, the place at which the activity is to be performed and the general plan for carrying out the activity;
            (3) The period of time for which the license is requested;
            (4) The name, amount, and specifications (including the chemical and physical form and, where applicable, isotopic content) of the special nuclear material the applicant proposes to use or produce;
            (5) [Reserved]
            (6) The technical qualifications, including training and experience of the applicant and members of his staff to engage in the proposed activities in accordance with the regulations in this chapter;
            (7) A description of equipment and facilities which will be used by the applicant to protect health and minimize danger to life or property (such as handling devices, working areas, shields, measuring and monitoring instruments, devices for the disposal of radioactive effluents and wastes, storage facilities, criticality accident alarm systems, etc.);

            (8) Proposed procedures to protect health and minimize danger to life or property (such as procedures to avoid accidental criticality, procedures for personnel monitoring and waste disposal, post-criticality accident emergency procedures, etc.).

              Note:
              Where the nature of the proposed activities is such as to require consideration of the applicant's financial qualifications to engage in the proposed activities in accordance with the regulations in this chapter, the Commission may request the applicant to submit information with respect to his financial qualifications.

            (9) As provided by § 70.25, certain applications for specific licenses filed under this part must contain a proposed decommissioning funding plan or a certification of financial assurance for decommissioning. In the case of renewal applications submitted on or before July 27, 1990, this submittal may follow the renewal application but must be submitted on or before July 27, 1990.
            (b) Each application for a license to possess special nuclear material, to possess equipment capable of enriching uranium, to operate an uranium enrichment facility, to possess and use at any one time and location special nuclear material in a quantity exceeding one effective kilogram, except for applications for use as sealed sources and for those uses involved in the operation of a nuclear reactor licensed pursuant to part 50 of this chapter and those involved in a waste disposal operation, must contain a full description of the applicant's program for control and accounting of such special nuclear material or enrichment equipment that will be in the applicant's possession under license to show how compliance with the requirements of §§ 74.31, 74.33, 74.41, or 74.51 of this chapter, as applicable, will be accomplished.
            (c) [Reserved]
            (d) The Commission may at any time after the filing of the original application, and before the expiration of the license, require further statements in order to enable the Commission to determine whether the application should be granted or denied or whether a license should be modified or revoked. All applications and statements shall be signed by the applicant or licensee or a corporate officer thereof.
            (e) Each application and statement shall contain complete and accurate disclosure as to all matters and things required to be disclosed.

            (f) Each application for a license to possess and use special nuclear material in a plutonium processing and fuel fabrication plant shall contain, in addition to the other information required by this section, a description of the plantsite, a description and safety assessment of the design bases of the principal structure, systems, and components of the plant, including provisions for protection against natural phenomena, and a description of the quality assurance program to be applied to the design, fabrication, construction, testing and operation of the structures, systems, and components of the plant. 2

                2 The description of the quality assurance program should include a discussion of how the criteria in appendix B of part 50 of this chapter will be met.

            (g)(1) Each application for a license that would authorize the transport or delivery to a carrier for transport of special nuclear material in an amount specified in § 73.1(b)(2) of this chapter must include (i) a description of the plan for physical protection of special nuclear material in transit in accordance with §§ 73.20, 73.25, 73.26, 73.27, and 73.67 (a), (e), and (g) for 10 kg or more of special nuclear material of low strategic significance, and § 73.70(g) of this chapter including, as appropriate, a plan for the selection, qualification, and training of armed escorts, or the specification and design of a specially designed truck or trailer, and (ii) a licensee safeguards contingency plan or response procedures, as appropriate, for dealing with threats, thefts, and radiological sabotage relating to the special nuclear material in transit.
            (2) Each application for such a license involving formula quantities of strategic special nuclear material must include the first four categories of information contained in the applicant's safeguards contingency plan. (The first four categories of information, as set forth in appendix C to part 73 of this chapter, are Background, Generic Planning Base, Licensee Planning Base, and Responsibility Matrix. The fifth category of information, Procedures, does not have to be submitted for approval.)
            (3) The licensee shall retain this discription of the plan for physical protection of special nuclear material in transit and the safeguards contingency plan or safeguards response procedures and each change to the plan or procedures as a record for a period of three years following the date on which the licensee last possessed the appropriate type and quantity of special nuclear material requiring this record under each license.
            (h)(1) Each application for a license to possess or use, at any site or contiguous sites subject to licensee control, a formula quantity of strategic special nuclear material, as defined in § 70.4, other than a license for possession or use of this material in the operation of a nuclear reactor licensed pursuant to part 50 of this chapter, must include a physical security plan. The plan must describe how the applicant will meet the applicable requirements of part 73 of this chapter in the conduct of the activity to be licensed, including the identification and description of jobs as required by 10 CFR 11.11(a). The plan must list tests, inspections, audits, and other means to be used to demonstrate compliance with the requirements of 10 CFR parts 11 and 73, if applicable.
            (2) The licensee shall retain a copy of this physical security plan and each change to the plan as a record for a period of three years following the date on which the licensee last possessed the appropriate type and quantity of special nuclear material requiring this record under each license.
            (i)(1) Each application to possess enriched uranium or plutonium for which a criticality accident alarm system is required, uranium hexafluoride in excess of 50 kilograms in a single container or 1000 kilograms total, or in excess of 2 curies of plutonium in unsealed form or on foils or plated sources, must contain either:
            (i) An evaluation showing that the maximum dose to a member of the public offsite due to a release of radioactive materials would not exceed 1 rem effective dose equivalent or an intake of 2 milligrams of soluble uranium, or
            (ii) An emergency plan for responding to the radiological hazards of an accidental release of special nuclear material and to any associated chemical hazards directly incident thereto.

            (2) One or more of the following factors may be used to support an evaluation submitted under paragraph (i)(1)(i) of this section:

            (i) The radioactive material is physically separated so that only a portion could be involved in an accident;
            (ii) All or part of the radioactive material is not subject to release during an accident or to criticality because of the way it is stored or packaged;
            (iii) In the case of fires or explosions, the release fraction would be lower than 0.001 due to the chemical or physical form of the material;
            (iv) The solubility of the material released would reduce the dose received;
            (v) The facility design or engineered safety features in the facility would cause the release fraction to be lower than 0.001;
            (vi) Operating restrictions or procedures would prevent a release large enough to cause a member of the public offsite to receive a dose exceeding 1 rem effective dose equivalent; or
            (vii) Other factors appropriate for the specific facility.
            (3) Emergency plans submitted under paragraph (i)(1)(ii) of this section must include the following information:
            (i) Facility description. A brief description of the licensee's facility and area near the site.
            (ii) Types of accidents. An identification of each type of radioactive materials accident for which protective actions may be needed.
            (iii) Classification of accidents. A classification system for classifying accidents as alerts or site area emergencies.
            (iv) Detection of accidents. Identification of the means of detecting each type of accident in a timely manner.
            (v) Mitigation of consequences. A brief description of the means and equipment for mitigating the consequences of each type of accident, including those provided to protect workers onsite, and a description of the program for maintaining the equipment.
            (vi) Assessment of releases. A brief description of the methods and equipment to assess releases of radioactive materials.
            (vii) Responsibilities. A brief description of the responsibilities of licensee personnel should an accident occur, including identification of personnel responsible for promptly notifying offsite response organizations and the NRC; also responsibilities for developing, maintaining, and updating the plan.
            (viii) Notification and coordination. A commitment to and a brief description of the means to promptly notify offsite response organizations and request offsite assistance, including medical assistance for the treatment of contaminated injured onsite workers when appropriate. A control point must be established. The notification and coordination must be planned so that unavailability of some personnel, parts of the facility, and some equipment will not prevent the notification and coordination. The licensee shall also commit to notify the NRC operations center immediately after notification of the appropriate offsite response organizations and not later than one hour after the licensee declares an emergency. 1

                1 These reporting requirements do not superceed or release licensees of complying with the requirements under the Emergency Planning and Community Right-to-Know Act of 1986, Title III, Pub. L. 99-499 or other state or federal reporting requirements.

            (ix) Information to be communicated. A brief description of the types of information on facility status, radioactive releases, and recommended protective actions, if necessary, to be given to offsite response organizations and to the NRC.
            (x) Training. A brief description of the frequency, performance objectives and plans for the training that the licensee will provide workers on how to respond to an emergency including any special instructions and orientation tours the licensee would offer to fire, police, medical and other emergency personnel. The training shall familiarize personnel with site-specific emergency procedures. Also, the training shall thoroughly prepare site personnel for their responsibilities in the event of accident scenarios postulated as most probable for the specific site, including the use of team training for such scenarios.
            (xi) Safe shutdown. A brief description of the means of restoring the facility to a safe condition after an accident.

            (xii) Exercises. Provisions for conducting quarterly communications checks with offsite response organizations and biennial onsite exercises to test response to simulated emergencies. Quarterly communications checks with offsite response organizations must include the check and update of all necessary telephone numbers. The licensee shall invite offsite response organizations to participate in the biennial exercises. Participation of offsite response organizations in biennial exercises although recommended is not required. Exercises must use accident scenarios postulated as most probable for the specific site and the scenarios shall not be known to most exercise participants. The licensee shall critique each exercise using individuals not having direct implementation responsibility for the plan. Critiques of exercises must evaluate the appropriateness of the plan, emergency procedures, facilities, equipment, training of personnel, and overall effectiveness of the response. Deficiencies found by the critiques must be corrected.
            (xiii) Hazardous chemicals. A certification that the applicant has met its responsibilities under the Emergency Planning and Community Right-to-Know Act of 1986, Title III, Pub. L. 99-499, if applicable to the applicant's activities at the proposed place of use of the special nuclear material.
            (4) The licensee shall allow the offsite response organizations expected to respond in case of an accident 60 days to comment on the licensee's emergency plan before submitting it to NRC. The licensee shall provide any comments received within the 60 days to the NRC with the emergency plan.
            (j)(1) Each application for a license to possess or use at any site or contiguous sites subject to control by the licensee uranium-235 (contained in uranium enriched to 20 percent or more in the uranium-235 isotope), uranium-233, or plutonium alone or in any combination in a quantity of 5,000 grams or more computed by the formula, grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium) other than a license for possession or use of this material in the operation of a nuclear reactor licensed pursuant to part 50 of this chapter, must include a licensee safeguards contingency plan for dealing with threats, thefts, and radiological sabotage, as defined in part 73 of this chapter, relating to nuclear facilities licensed under part 50 of this chapter or to the possession of special nuclear material licensed under this part.
            (2) Each application for such a license must include the first four categories of information contained in the applicant's safeguards contingency plan. (The first four categories of information, as set forth in appendix C to part 73 of this chapter, are Background, Generic Planning Base, Licensee Planning Base, and Responsibility Matrix.) The fifth category of information, Procedures, does not have to be submitted for approval.
            (3) The licensee shall retain a copy of this safeguards contingency plan as a record until the Commission terminates each license obtained by this application or any application for renewal of a license and retain each change to the plan as a record for three years after the date of the change.
            (k) Each application for a license to possess or use at any site or contiguous sites subject to licensee control, special nuclear material of moderate strategic significance or 10 kg or more of special nuclear material of low strategic significance as defined under § 70.4, other than a license for possession or use of this material in the operation of a nuclear power reactor licensed pursuant to part 50 of this chapter, must include a physical security plan that demonstrates how the applicant plans to meet the requirements of paragraphs (d), (e), (f), and (g) of § 73.67 of this chapter, as appropriate. The licensee shall retain a copy of this physical security plan as a record for the period during which the licensee possesses the appropriate type and quantity of special nuclear material under each license, and if any portion of the plan is superseded, retain that superseded portion of the plan for 3 years after the effective date of the change.

            (l) Each applicant for a license shall protect Safeguards Information against unauthorized disclosure in accordance with the requirements in § 73.21 and the requirements of § 73.22, or 73.23 of this chapter, as applicable, and shall protect classified information in accordance with the requirements of parts 25 and 95 of this chapter, as applicable.
            (m) Each application for a license to possess equipment capable of enriching uranium or operate an enrichment facility, and produce, possess, or use more than one effective kilogram of special nuclear material at any site or contiguous sites subject to control by the applicant, must contain a full description of the applicant's security program to protect against theft, and to protect against unauthorized viewing of classified enrichment equipment, and unauthorized disclosure of classified matter in accordance with the requirements of 10 CFR parts 25 and 95.
            (n) A license application that involves the use of special nuclear material in a uranium enrichment facility must include the applicant's provisions for liability insurance.
            [21 FR 764, Feb. 3, 1956]

              Editorial Note:
              For Federal Register citations affecting § 70.22, see the List of CFR Sections Affected, which appears in the Finding Aids section of the printed volume and at www.govinfo.gov.

            § 70.23
            Requirements for the approval of applications.
            (a) An application for a license will be approved if the Commission determines that:
            (1) The special nuclear material is to be used for the conduct of research or development activities of a type specified in section 31 of the Act, 1
               in activities licensed by the Commission under section 103 or 104 of the Act, or for such other uses as the Commission determines to be appropriate to carry out the purposes of the Act;

                1 The types of research and development activities specified in section 31 are those relating to:
              (1) Nuclear processes;
              (2) The theory and production of atomic energy, including processes, materials, and devices related to such production;
              (3) Utilization of special nuclear material and radioactive material for medical, biological, agricultural, health or military purposes;
              (4) Utilization of special nuclear material, atomic energy, and radioactive material and processes entailed in the utilization or production of atomic energy or such material for all other purposes, including industrial use, the generation of usable energy, and the demonstration of the practical value of utilization or production facilities for industrial or commercial purposes; and
              (5) The protection of health and the promotion of safety during research and production activities.

            (2) The applicant is qualified by reason of training and experience to use the material for the purpose requested in accordance with the regulations in this chapter;
            (3) The applicant's proposed equipment and facilities are adequate to protect health and minimize danger to life or property;
            (4) The applicant's proposed procedures to protect health and to minimize danger to life or property are adequate;
            (5) Where the nature of the proposed activities is such as to require consideration by the Commission, that the applicant appears to be financially qualified to engage in the proposed activities in accordance with the regulations in this part;
            (6) Where the applicant is required to submit a summary description of the fundamental material controls provided in his procedures for the control of and accounting for special nuclear material pursuant to § 70.22 (b), the applicant's proposed controls are adequate;

            (7) Where the proposed activity is processing and fuel fabrication, scrap recovery, conversion of uranium hexafluoride, uranium enrichment facility construction and operation, or any other activity which the NRC determines will significantly affect the quality of the environment, the Director of Nuclear Material Safety and Safeguards or his/her designee, before commencement of construction of the plant or facility in which the activity will be conducted, on the basis of information filed and evaluations made pursuant to subpart A of part 51 of this chapter, has concluded, after weighing the environmental, economic, technical, and other benefits against environmental costs and considering available alternatives, that the action called for is the issuance of the proposed license, with any appropriate conditions to protect environmental values. Commencement of construction prior to this conclusion is grounds for denial to possess and use special nuclear material in the plant or facility. Commencement of construction as defined in section 70.4 may include non-construction activities if the activity has a reasonable nexus to radiological safety and security.
            (8) Where the proposed activity is the operation of a plutonium processing and fuel fabrication plant, construction of the principal structures, systems, and components approved pursuant to paragraph (b) of this section has been completed in accordance with the application;
            (9) Where the applicant is required to submit a plan for physical protection of special nuclear material in transit pursuant to § 70.22(g), of this chapter, the applicant's plan is adequate;
            (10) Where the applicant is required to submit a physical security plan pursuant to § 70.22(h), the applicant's proposed plan is adequate;
            (11) Where the proposed activity is processing and fuel fabrication, scrap recovery, conversion of uranium hexafluoride, or involves the use of special nuclear material in a uranium enrichment facility, the applicant's proposed emergency plan is adequate.
            (12) Where the proposed activity is use of special nuclear material in a uranium enrichment facility, the applicable provisions of part 140 of this chapter have been satisfied.
            (b) The Commission will approve construction of the principal structures, systems, and components of a plutonium processing and fuel fabrication plant on the basis of information filed pursuant to § 70.22(f) when the Commission has determined that the design bases of the principal structures, systems, and components, and the quality assurance program provide reasonable assurance of protection against natural phenomena and the consequences of potential accidents. 3
               Failure to obtain Commission approval prior to beginning of such construction may be grounds for denial of a license to possess and use special nuclear material in a plutonium processing and fuel fabrication plant.

                3 The criteria in appendix B of part 50 of this chapter will be used by the Commission in determining the adequacy of the quality assurance program.

            [36 FR 17574, Sept. 2, 1971, as amended at 37 FR 5749, Mar. 21, 1972; 38 FR 30534, 30538, Nov. 6, 1973; 39 FR 26286, July 18, 1974; 42 FR 17126, Mar. 31, 1977; 43 FR 6924, Feb. 17, 1978; 49 FR 9406, Mar. 12, 1984; 54 FR 14064, Apr. 7, 1989; 57 FR 18392, Apr. 30, 1992; 67 FR 78142, Dec. 23, 2002; 76 FR 56966, Sept. 15, 2011]

            § 70.23a
            Hearing required for uranium enrichment facility.

            The Commission will hold a hearing under 10 CFR part 2, subparts A, C, G, and I, on each application for issuance of a license for construction and operation of a uranium enrichment facility. The Commission will publish public notice of the hearing in the Federal Register at least thirty (30) days before the hearing.
            [69 FR 2280, Jan. 14, 2004]

            § 70.24
            Criticality accident requirements.
            (a) Each licensee authorized to possess special nuclear material in a quantity exceeding 700 grams of contained uranium-235, 520 grams of uranium-233, 450 grams of plutonium, 1,500 grams of contained uranium-235 if no uranium enriched to more than 4 percent by weight of uranium-235 is present, 450 grams of any combination thereof, or one-half such quantities if massive moderators or reflectors made of graphite, heavy water or beryllium may be present, shall maintain in each area in which such licensed special nuclear material is handled, used, or stored, a monitoring system meeting the requirements of either paragraph (a)(1) or (a)(2), as appropriate, and using gamma- or neutron-sensitive radiation detectors which will energize clearly audible alarm signals if accidental criticality occurs. This section is not intended to require underwater monitoring when special nuclear material is handled or stored beneath water shielding or to require monitoring systems when special nuclear material is being transported when packaged in accordance with the requirements of part 71 of this chapter.

            (1) The monitoring system shall be capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of 2 meters from the reacting material within one minute. Coverage of all areas shall be provided by two detectors.
            (2) Persons licensed prior to December 6, 1974, to possess special nuclear material subject to this section may maintain a monitoring system capable of detecting a criticality which generates radiation levels of 300 rems per hour one foot from the source of the radiation. The monitoring devices in the system shall have a preset alarm point of not less than 5 millirems per hour (in order to avoid false alarms) nor more than 20 millirems per hour. In no event may any such device be farther than 120 feet from the special nuclear material being handled, used, or stored; lesser distances may be necessary to meet the requirements of this paragraph (a)(2) on account of intervening shielding or other pertinent factors.
            (3) The licensee shall maintain emergency procedures for each area in which this licensed special nuclear material is handled, used, or stored to ensure that all personnel withdraw to an area of safety upon the sounding of the alarm. These procedures must include the conduct of drills to familiarize personnel with the evacuation plan, and designation of responsible individuals for determining the cause of the alarm, and placement of radiation survey instruments in accessible locations for use in such an emergency. The licensee shall retain a copy of current procedures for each area as a record for as long as licensed special nuclear material is handled, used, or stored in the area. The licensee shall retain any superseded portion of the procedures for three years after the portion is superseded.
            (b) Each licensee authorized to possess special nuclear material in quantities in excess of those specified in paragraph (a) shall:
            (1) Provide the means for identifying quickly which individuals have received doses of 10 rads or more.
            (2) Maintain facilities and supplies at the site for decontamination of personnel, arrangements for the services of a physician and other medical personnel qualified to handle radiation emergencies, arrangements for transportation of injured or contaminated individuals to treatment facilities, and arrangements for treatment of individuals at treatment facilities outside the site boundary.
            (c) Holders of licenses for construction or operation of a nuclear reactor issued pursuant to part 50 of this chapter, except critical assembly reactors, are exempt for the requirements of paragraph (b) of this section with respect to special nuclear material used or to be used in the reactor.
            (d)(1) The requirements in paragraphs (a) through (c) of this section do not apply to a holder of a construction permit or operating license for a nuclear power reactor issued under part 50 of this chapter or a combined license issued under part 52 of this chapter, if the holder complies with the requirements of paragraph (b) of 10 CFR 50.68.
            (2) An exemption from § 70.24 held by a licensee who thereafter elects to comply with requirements of paragraph (b) of 10 CFR 50.68 does not exempt that licensee from complying with any of the requirements in § 50.68, but shall be ineffective so long as the licensee elects to comply with § 50.68.
            [39 FR 39021, Nov. 5, 1974, as amended at 41 FR 31522, July 29, 1976; 53 FR 19252, May 27, 1988; 62 FR 63828, Dec. 3, 1997; 63 FR 9403, Feb. 25, 1998; 63 FR 63130, Nov. 12, 1998]

            § 70.25
            Financial assurance and recordkeeping for decommissioning.
            (a) Each applicant for a specific license of the types described in paragraphs (a) (1) and (2) of this section shall submit a decommissioning funding plan as described in paragraph (e) of this section.
            (1) A specific license for a uranium enrichment facility;

            (2) A specific license authorizing the possession and use of unsealed special nuclear material in quantities exceeding 105 times the applicable quantities set forth in appendix B to part 30. A decommissioning funding plan must also be submitted when a combination of isotopes is involved if R divided by 105 is greater than 1 (unity rule), where R is the sum of the ratios of the quantity of each isotope to the applicable value in appendix B to part 30.

            (b) Each applicant for a specific license authorizing possession and use of unsealed special nuclear material in quantities specified in paragraph (d) of this section shall either—
            (1) Submit a decommissioning funding plan as described in paragraph (e) of this section; or
            (2) Submit a certification that financial assurance for decommissioning has been provided in the amount prescribed by paragraph (d) of this section using one of the methods described in paragraph (f) of this section. For an applicant, this certification may state that the appropriate assurance will be obtained after the application has been approved and the license issued but before the receipt of licensed material. If the applicant defers execution of the financial instrument until after the license has been issued, a signed original of the financial instrument obtained to satisfy the requirements of paragraph (f) of this section must be submitted to NRC before receipt of licensed material. If the applicant does not defer execution of the financial instrument, the applicant shall submit to NRC, as part of the certification, a signed original of the financial instrument obtained to satisfy the requirements of paragraph (f) of this section.
            (c)(1) Each holder of a specific license issued on or after July 27, 1990, which is of a type described in paragraph (a) or (b) of this section, shall provide financial assurance for decommissioning in accordance with the criteria set forth in this section.
            (2) Each holder of a specific license issued before July 27, 1990, and of a type described in paragraph (a) of this section shall submit a decommissioning funding plan as described in paragraph (e) of this section or a certification of financial assurance for decommissioning in an amount at least equal to $1,125,000 in accordance with the criteria set forth in this section. If the licensee submits the certification of financial assurance rather than a decommissioning funding plan, the licensee shall include a decommissioning funding plan in any application for license renewal.
            (3) Each holder of a specific license issued before July 27, 1990, and of a type described in paragraph (b) of this section shall submit, on or before July 27, 1990, a decommissioning funding plan, described in paragraph (e) of this section, or a certification of financial assurance for decommissioning in accordance with the criteria set forth in this section.
            (4) Any licensee who has submitted an application before July 27, 1990, for renewal of license in accordance with § 70.33 shall provide financial assurance for decommissioning in accordance with paragraphs (a) and (b) of this section. This assurance must be submitted when this rule becomes effective November 24, 1995.
            (5) If, in surveys made under 10 CFR 20.1501(a), residual radioactivity in the facility and environment, including the subsurface, is detected at levels that would, if left uncorrected, prevent the site from meeting the 10 CFR 20.1402 criteria for unrestricted use, the licensee must submit a decommissioning funding plan within one year of when the survey is completed.
            (d) Table of required amounts of financial assurance for decommissioning by quantity of material. Licensees required to submit the $1,125,000 amount must do so by December 2, 2004. Licensees required to submit the $225,000 amount must do so by June 2, 2005. Licensees having possession limits exceeding the upper bounds of this table must base financial assurance on a decommissioning funding plan.

                Greater than 104 but less than or equal to 105 times the applicable quantities of appendix B to part 30. (For a combination of isotopes, if R, as defined in § 70.25(a), divided by 104 is greater than 1 but R divided by 105 is less than or equal to 1.)
                $1,125,000

                Greater than 103 but less than or equal to 104 times the applicable quantities of appendix B to part 30. (For a combination of isotopes, if R, as defined in § 70.25(a), divided by 103 is greater than 1 but R divided by 104 is less than or equal to 1.)
                $225,000

            (e)(1) Each decommissioning funding plan must be submitted for review and approval and must contain—
            (i) A detailed cost estimate for decommissioning, in an amount reflecting:
            (A) The cost of an independent contractor to perform all decommissioning activities;
            (B) The cost of meeting the 10 CFR 20.1402 criteria for unrestricted use, provided that, if the applicant or licensee can demonstrate its ability to meet the provisions of 10 CFR 20.1403, the cost estimate may be based on meeting the 10 CFR 20.1403 criteria;
            (C) The volume of onsite subsurface material containing residual radioactivity that will require remediation; and
            (D) An adequate contingency factor.
            (ii) Identification of and justification for using the key assumptions contained in the DCE;
            (iii) A description of the method of assuring funds for decommissioning from paragraph (f) of this section, including means for adjusting cost estimates and associated funding levels periodically over the life of the facility;
            (iv) A certification by the licensee that financial assurance for decommissioning has been provided in the amount of the cost estimate for decommissioning; and
            (v) A signed original, or, if permitted, a copy, of the financial instrument obtained to satisfy the requirements of paragraph (f) of this section (unless a previously submitted and accepted financial instrument continues to cover the cost estimate for decommissioning).
            (2) At the time of license renewal and at intervals not to exceed 3 years, the decommissioning funding plan must be resubmitted with adjustments as necessary to account for changes in costs and the extent of contamination. If the amount of financial assurance will be adjusted downward, this can not be done until the updated decommissioning funding plan is approved. The decommissioning funding plan must update the information submitted with the original or prior approved plan, and must specifically consider the effect of the following events on decommissioning costs:
            (i) Spills of radioactive material producing additional residual radioactivity in onsite subsurface material;
            (ii) Waste inventory increasing above the amount previously estimated;
            (iii) Waste disposal costs increasing above the amount previously estimated;
            (iv) Facility modifications;
            (v) Changes in authorized possession limits;
            (vi) Actual remediation costs that exceed the previous cost estimate;
            (vii) Onsite disposal; and
            (viii) Use of a settling pond.
            (f) The financial instrument must include the licensee's name, license number, and docket number; and the name, address, and other contact information of the issuer, and, if a trust is used, the trustee. When any of the foregoing information changes, the licensee must, within 30 days, submit financial instruments reflecting such changes. Financial assurance for decommissioning must be provided by one or more of the following methods:
            (1) Prepayment. Prepayment is the deposit before the start of operation into an account segregated from licensee assets and outside the licensee's administrative control of cash or liquid assets such that the amount of funds would be sufficient to pay decommissioning costs. Prepayment must be made into a trust account, and the trustee and the trust must be acceptable to the Commission.
            (2) A surety method, insurance, or other guarantee method. These methods guarantee that decommissioning costs will be paid. A surety method may be in the form of a surety bond, or letter of credit. A parent company guarantee of funds for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix A to part 30 of this chapter. For commercial corporations that issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix C to part 30 of this chapter. For commercial companies that do not issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs may be used if the guarantee and test are as contained in appendix D to part 30 of this chapter. For nonprofit entities, such as colleges, universities, and nonprofit hospitals, a guarantee of funds by the applicant or licensee may be used if the guarantee and test are as contained in appendix E to part 30 of this chapter. Except for an external sinking fund, a parent company guarantee or a guarantee by the applicant or licensee may not be used in combination with any other financial methods used to satisfy the requirements of this section. A guarantee by the applicant or licensee may not be used in any situation where the applicant or licensee has a parent company holding majority control of the voting stock of the company. Any surety method or insurance used to provide financial assurance for decommissioning must contain the following conditions:
            (i) The surety method or insurance must be open-ended or, if written for a specified term, such as five years, must be renewed automatically unless 90 days or more prior to the renewal date, the issurer notifies the Commission, the beneficiary, and the licensee of its intention not to renew. The surety method or insurance must also provide that the full face amount be paid to the beneficiary automatically prior to the expiration without proof of forfeiture if the licensee fails to provide a replacement acceptable to the Commission within 30 days after receipt of notification of cancellation.
            (ii) The surety method or insurance must be payable to a trust established for decommissioning costs. The trustee and trust must be acceptable to the Commission. An acceptable trustee includes an appropriate State or Federal government agency or an entity which has the authority to act as a trustee and whose trust operations are regulated and examined by a Federal or State agency.
            (iii) The surety method or insurance must remain in effect until the Commission has terminated the license.
            (3) An external sinking fund in which deposits are made at least annually, coupled with a surety method, insurance, or other guarantee method, the value of which may decrease by the amount being accumulated in the sinking fund. An external sinking fund is a fund established and maintained by setting aside funds periodically in an account segregated from licensee assets and outside the licensee's administrative control in which the total amount of funds would be sufficient to pay decommissioning costs at the time termination of operation is expected. An external sinking fund must be in the form of a trust. If the other guarantee method is used, no surety or insurance may be combined with the external sinking fund. The surety, insurance, or other guarantee provisions must be as stated in paragraph (f)(2) of this section.
            (4) In the case of Federal, State, or local government licensees, a statement of intent containing a cost estimate for decommissioning or an amount based on the Table in paragraph (d) of this section, and indicating that funds for decommissioning will be obtained when necessary.
            (5) When a governmental entity is assuming custody and ownership of a site, an arrangement that is deemed acceptable by such governmental entity.
            (g) Each person licensed under this part shall keep records of information important to the decommissioning of a facility in an identified location until the site is released for unrestricted use. If records important to the decommissioning of a facility are kept for other purposes, reference to these records and their locations may be used. Information the Commission considers important to decommissioning consists of—
            (1) Records of spills or other unusual occurrences involving the spread of contamination in and around the facility, equipment, or site. These records may be limited to instances when contamination remains after any cleanup procedures or when there is reasonable likelihood that contaminants may have spread to inaccessible areas as in the case of possible seepage into porous materials such as concrete. These records must include any known information on identification of involved nuclides, quantities, forms, and concentrations.

            (2) As-built drawings and modifications of structures and equipment in restricted areas where radioactive materials are used and/or stored and of locations of possible inaccessible contamination such as buried pipes which may be subject to contamination. If required drawings are referenced, each relevant document need not be indexed individually. If drawings are not available, the licensee shall substitute appropriate records of available information concerning these areas and locations.
            (3) Except for areas containing only sealed sources (provided the sources have not leaked or no contamination remains after cleanup of any leak), a list contained in a single document and updated every 2 years, of the following:
            (i) All areas designated and formerly designated as restricted areas as defined under 10 CFR 20.1003 (For requirements prior to January 1, 1994, see 10 CFR 20.3 as contained in the CFR edition revised as of January 1, 1993.);
            (ii) All areas outside of restricted areas that require documentation under § 70.25(g)(1);
            (iii) All areas outside of restricted areas where current and previous wastes have been buried as documented under 10 CFR 20.2108; and
            (iv) All areas outside of restricted areas that contain material such that, if the license expired, the licensee would be required to either decontaminate the area to meet the criteria for decommissioning in 10 CFR part 20, subpart E, or apply for approval for disposal under 10 CFR 20.2002.
            (4) Records of the cost estimate performed for the decommissioning funding plan or of the amount certified for decommissioning, and records of the funding method used for assuring funds if either a funding plan or certification is used.
            (h) In providing financial assurance under this section, each licensee must use the financial assurance funds only for decommissioning activities and each licensee must monitor the balance of funds held to account for market variations. The licensee must replenish the funds, and report such actions to the NRC, as follows:
            (1) If, at the end of a calendar quarter, the fund balance is below the amount necessary to cover the cost of decommissioning, but is not below 75 percent of the cost, the licensee must increase the balance to cover the cost, and must do so within 30 days after the end of the calendar quarter.
            (2) If, at any time, the fund balance falls below 75 percent of the amount necessary to cover the cost of decommissioning, the licensee must increase the balance to cover the cost, and must do so within 30 days of the occurrence.
            (3) Within 30 days of taking the actions required by paragraph (h)(1) or (h)(2) of this section, the licensee must provide a written report of such actions to the Director, Office of Nuclear Material Safety and Safeguards, and state the new balance of the fund.
            [53 FR 24053, June 27, 1988, as amended at 56 FR 23474, May 21, 1991; 57 FR 18393, Apr. 30, 1992; 58 FR 39634, July 26, 1993; 58 FR 67662, Dec. 22, 1993; 58 FR 68731, Dec. 29, 1993; 59 FR 1618, Jan. 12, 1994; 60 FR 38239, July 26, 1995; 61 FR 24675, May 16, 1996; 62 FR 39091, July 21, 1997; 63 FR 29544, June 1, 1998; 68 FR 57337, Oct. 3, 2003; 76 FR 35572, June 17, 2011; 78 FR 34250, June 7, 2013; 78 FR 75450, Dec. 12, 2013; 79 FR 75740, Dec. 19, 2014]

          Subpart E—Licenses

            § 70.31
            Issuance of licenses.
            (a) Upon a determination that an application meets the requirements of the act and of the regulations of the Commission, the Commission will issue a license in such form and containing such conditions and limitations as it deems appropriate or necessary to effectuate the purposes of the act.
            (b) [Reserved]
            (c) Each license issued to a person for use of special nuclear material in activities in which special nuclear material will be produced shall (subject to the provisions of § 70.41(b)) be deemed to authorize such person to receive title to, own, acquire, receive, possess, use, and transfer the special nuclear material produced in the course of such authorized activities.

            (d) No license will be issued by the Commission to any person within the United States if the Commission finds that the issuance of such license would be inimical to the common defense and security or would constitute an unreasonable risk to the health and safety of the public.

            (e) No license to construct and operate a uranium enrichment facility may be issued until a hearing pursuant to 10 CFR part 2, subparts G and I, is completed and decision issued on the application.
            [21 FR 764, Feb. 3, 1956, as amended at 32 FR 2563, Feb. 7, 1967; 32 FR 4056, Mar. 15, 1967; 43 FR 6925, Feb. 17, 1978; 57 FR 18393, Apr. 30, 1992]

            § 70.32
            Conditions of licenses.
            (a) Each license shall contain and be subject to the following conditions:
            (1) [Reserved]
            (2) No right to the special nuclear material shall be conferred by the license except as defined by the license;
            (3) Neither the license nor any right under the license shall be assigned or otherwise transferred in violation of the provisions of the Act;
            (4) All special nuclear material shall be subject to the right of recapture or control reserved by section 108 and to all other provisions of the Act;
            (5) No special nuclear material may be used in any utilization or production facility except in accordance with the provisions of the Act;
            (6) The licensee shall not use the special nuclear material to construct an atomic weapon or any component of an atomic weapon;
            (7) Except to the extent that the indemnification and limitation of liability provisions of part 140 of this chapter apply, the licensee will hold the United States and the Department harmless from any damages resulting from the use or possession of special nuclear material leased from the Department by the licensee;
            (8) The license shall be subject to and the licensee shall observe, all applicable rules, regulations and orders of the Commission.
            (9)(i) Each licensee shall notify the appropriate NRC Regional Administrator, in writing, immediately following the filing of a voluntary or involuntary petition for bankruptcy under any Chapter of Title 11 (Bankruptcy) of the United States Code by or against:
            (A) The licensee;
            (B) An entity (as that term is defined in 11 U.S.C. 101(14)) controlling the licensee or listing the license or licensee as property of the estate; or
            (C) An affiliate (as that term is defined in 11 U.S.C. 101(a)) of the licensee.
            (ii) This notification must indicate:
            (A) The bankruptcy court in which the petition for bankruptcy was filed; and
            (B) The date of the filing of the petition.
            (b) The Commission may incorporate in any license such additional conditions and requirements with respect to the licensee's ownership, receipt, possession, use, and transfer of special nuclear material as it deems appropriate or necessary in order to:
            (1) Promote the common defense and security;
            (2) Protect health or to minimize danger to life or property;
            (3) Protect restricted data;
            (4) Guard against the loss or diversion of special nuclear material;
            (5) Require such reports and the keeping of such records, and to provide for such inspections, of activities under the license as may be necessary or appropriate to effectuate the purposes of the act and regulations thereunder.
            (c)(1) Each license authorizing the possession and use at any one time and location of uranium source material at an uranium enrichment facility or special nuclear material in a quantity exceeding one effective kilogram, except for use as sealed sources and those uses involved in the operation of a nuclear reactor licensed pursuant to part 50 of this chapter and those involved in a waste disposal operation, shall contain and be subject to a condition requiring the licensee to maintain and follow:
            (i) The program for control and accounting of uranium source material at an uranium enrichment facility and special nuclear material at all applicable facilities as implemented pursuant to § 70.22(b), or §§ 74.31(b), 74.33(b), 74.41(b), or 74.51(c) of this chapter, as appropriate;

            (ii) The measurement control program for uranium source material at an uranium enrichment facility and for special nuclear material at all applicable facilities as implemented pursuant to §§ 74.31(b), 74.33(b), 74.45(c), or 74.59(e) of this chapter, as appropriate; and

            (iii) Other material control procedures as the Commission determines to be essential for the safeguarding of uranium source material at an uranium enrichment facility or of special nuclear material and providing that the licensee shall make no change that would decrease the effectiveness of the material control and accounting program implemented pursuant to § 70.22(b), or §§ 74.31(b), 74.33(b), 74.41(b), or 74.51(c) of this chapter, and the measurement control program implemented pursuant to §§ 74.31(b), 74.33(b), 74.41(b), or 74.59(e) of this chapter without the prior approval of the Commission. A licensee desiring to make changes that would decrease the effectiveness of its material control and accounting program or its measurement control program shall submit an application for amendment to its license pursuant to § 70.34.
            (2) The licensee shall maintain records of changes to the material control and accounting program made without prior Commission approval for a period of 5 years from the date of the change. Licensees located in all four Regions as indicated in appendix A of part 73 of this chapter shall furnish to the Director, Office of Nuclear Material Safety and Safeguards, using an appropriate method listed in § 70.5(a), a report containing a description of each change within:
            (i) Two months of the change if it pertains to uranium-233, uranium-235 contained in uranium enriched 20 percent or more in the uranium-235 isotope, or plutonium, except plutonium containing 80 percent or more by weight of the isotope Pu-238, and
            (ii) Six months of the change if it pertains to uranium enriched less than 20 percent in the uranium-235 isotope, or plutonium containing 80 percent or more by weight of the isotope Pu-238.
            (d) The licensee shall make no change which would decrease the effectiveness of the plan for physical protection of special nuclear material in transit prepared pursuant to § 70.22(g) or § 73.20(c) of this chapter without the prior approval of the Commission. A licensee desiring to make such changes shall submit an application for a change in the technical specifications incorporated in his or her license, if any, or for an amendment to the license pursuant to § 50.90 or § 70.34 of this chapter, as appropriate. The licensee may make changes to the plan for physical protection of special nuclear material without prior Commission approval if these changes do not decrease the effectiveness of the plan. The licensee shall retain a copy of the plan as a record for the period during which the licensee possesses a formula quantity of special nuclear material requiring this record under each license and each change to the plan for three years from the effective date of the change. Within two months after each change, a report containing a description of the change must be furnished to the Director of the NRC's Office of Nuclear Material Safety and Safeguards, using an appropriate method listed in § 70.5(a); and a copy must be sent to the appropriate NRC Regional Office shown in appendix A to part 73 of this chapter.
            (e) The licensee shall make no change which would decrease the effectiveness of a security plan prepared pursuant to §§ 70.22(h), 70.22(k), or 73.20(c) without the prior approval of the Commission. A licensee desiring to make such a change shall submit an application for an amendment to its license pursuant to § 70.34. The licensee shall maintain records of changes to the plan made without prior Commission approval, for three years from the effective date of the change, and shall, within two months after the change is made, furnish a report containing a description of each change to the Director, Office of Nuclear Material Safety and Safeguards; the report may be sent using an appropriate method listed in § 70.5(a), and a copy of the report must be sent to the appropriate NRC Regional Office shown in appendix A to part 73 of this chapter.
            (f) [Reserved]

            (g) The licensee shall prepare and maintain safeguards contingency plan procedures in accordance with appendix C to part 73 of this chapter for bringing about the actions and decisions contained in the Responsibility Matrix of its safeguards contingency plan. The licensee shall retain the current safeguards contingency plan procedures as a record for the entire period during which the licensee possesses the appropriate type and quantity of special nuclear material under each license for which the procedures were developed and, if any portion of the plan is superseded, retain that superseded portion for 3 years after the effective date of the change. The licensee shall not make a change that would decrease the safeguards effectiveness of the first four categories of information (i.e., Background, Generic Planning Base, Licensee Planning Base, and Responsibility Matrix) contained in any licensee safeguards contingency plan prepared pursuant to §§ 70.22(g), 70.22(j), 72.184, 73.20(c), 73.26(e)(1), 73.46(h)(1), or 73.50(g)(1) of this chapter without the prior approval of the NRC. A licensee desiring to make such a change shall submit an application for an amendment to its license pursuant to § 70.34. The licensee may make changes to the licensee safeguards contingency plan without prior NRC approval if the changes do not decrease the safeguards effectiveness of the plan. The licensee shall maintain each change to the plan made without prior approval as a record during the period for which possession of a formula quantity of special nuclear material is authorized under a license and retain the superseded portion for 3 years after the effective date of the change, and shall, within 60 days after the change is made, furnish a report containing a description of each change to the Director of Nuclear Material Safety and Safeguards; the report may be sent using an appropriate method listed in § 70.5(a), and a copy of the report must be sent to the Regional Administrator of the appropriate NRC Regional Office as specified in appendix A to part 73 of this chapter.
            (h) [Reserved]
            (i) Licensees required to submit emergency plans in accordance with § 70.22(i) shall follow the emergency plan approved by the Commission. The licensee may change the approved plan without Commission approval if the changes do not decrease the effectiveness of the plan. Within six months after each change is made, the licensee shall, using an appropriate method listed in § 70.5(a), furnish the Director, Office of Nuclear Material Safety and Safeguards, a copy of each change, with copies to the appropriate NRC Regional Office specified in appendix D to part 20 of this chapter and to affected offsite response organizations. Proposed changes that decrease the effectiveness of the approved emergency plan may not be implemented without prior application to and prior approval by the Commission.
            (j) Each licensee who possesses special nuclear material, or who transports, or delivers to a carrier for transport, a formula quantity of strategic special nuclear material, special nuclear material of moderate strategic significance, or special nuclear material of low strategic significance, or more than 100 grams of irradiated reactor fuel shall ensure that Safeguards Information is protected against unauthorized disclosure in accordance with the requirements in § 73.21 and the requirements of § 73.22 or § 73.23 of this chapter, as applicable, and shall protect classified information in accordance with the requirements of parts 25 and 95 of this chapter, as applicable.

            (k) No person may commence operation of a uranium enrichment facility until the Commission verifies through inspection that the facility has been constructed in accordance with the requirements of the license. The Commission shall publish notice of the inspection results in the Federal Register.

            [21 FR 764, Feb. 3, 1956]

              Editorial Note:
              For Federal Register citations affecting § 70.32, see the List of CFR Sections Affected, which appears in the Finding Aids section of the printed volume and at www.govinfo.gov.

            § 70.33
            Applications for renewal of licenses.
            Applications for renewal of a license should be filed in accordance with §§ 70.21 and 70.22. Information contained in previous applications, statements or reports filed with the Commission under the license may be incorporated by reference, provided that such references are clear and specific.
            [75 FR 73944, Nov. 30, 2010]

            § 70.34
            Amendment of licenses.
            Applications for amendment of a license shall be filed in accordance with § 70.21(a) and shall specify the respects in which the licensee desires his license to be amended and the grounds for such amendment.

            § 70.35
            Commission action on applications to renew or amend.
            In considering an application by a licensee to renew or amend his license, the Commission will apply the criteria set forth in § 70.23.

            § 70.36
            Inalienability of licenses.
            (a) No license granted under the regulations in this part and no right to possess or utilize special nuclear material granted by any license issued pursuant to the regulations in this part shall be transferred, assigned or in any manner disposed of, either voluntarily or involuntarily, directly or indirectly, through transfer of control of any license to any person unless the Commission shall after securing full information, find that the transfer is in accordance with the provisions of the Act, and shall give its consent in writing.
            (b) An application for transfer of license must include:
            (1) The identity, technical and financial qualifications of the proposed transferee; and
            (2) Financial assurance for decommissioning information required by § 70.25.
            [21 FR 764, Feb. 3, 1956, as amended at 35 FR 11461, July 17, 1970; 76 FR 35573, June 17, 2011]

            § 70.37
            Disclaimer of warranties.
            Neither the Government nor the Commission makes any warranty or other representation that special nuclear material (a) will not result in injury or damage when used for purposes approved by the Commission, (b) will accomplish the results for which it is requested and approved by the Commission, or (c) is safe for any other use.

            § 70.38
            Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas.
            (a) Each specific license expires at the end of the day on the expiration date stated in the license unless the licensee has filed an application for renewal under § 70.33 not less than 30 days before the expiration date stated in the existing license. If an application for renewal has been filed at least 30 days before the expiration date stated in the existing license, the existing license expires at the end of the day on which the Commission makes a final determination to deny the renewal application or, if the determination states an expiration date, the expiration date stated in the determination.
            (b) Each specific license revoked by the Commission expires at the end of the day on the date of the Commission's final determination to revoke the license, or on the expiration date stated in the determination, or as otherwise provided by Commission Order.
            (c) Each specific license continues in effect, beyond the expiration date if necessary, with respect to possession of special nuclear material until the Commission notifies the licensee in writing that the license is terminated. During this time, the licensee shall—
            (1) Limit actions involving special nuclear material to those related to decommissioning; and
            (2) Continue to control entry to restricted areas until they are suitable for release in accordance with NRC requirements.

            (d) Within 60 days of the occurrence of any of the following, consistent with the administrative directions in § 70.5, each licensee shall provide notification to the NRC in writing and either begin decommissioning its site, or any separate building or outdoor area that contains residual radioactivity, so that the building or outdoor area is suitable for release in accordance with NRC requirements, or submit within 12 months of notification a decommissioning plan, if required by paragraph (g)(1) of this section, and begin decommissioning upon approval of that plan if—

            (1) The license has expired pursuant to paragraph (a) or (b) of this section; or
            (2) The licensee has decided to permanently cease principal activities, as defined in this part, at the entire site or in any separate building or outdoor area; or
            (3) No principal activities under the license have been conducted for a period of 24 months; or
            (4) No principal activities have been conducted for a period of 24 months in any separate building or outdoor area that contains residual radioactivity such that the building or outdoor area is unsuitable for release in accordance with NRC requirements.
            (e) Coincident with the notification required by paragraph (d) of this section, the licensee shall maintain in effect all decommissioning financial assurances established by the licensee pursuant to § 70.25 in conjunction with a license issuance or renewal or as required by this section. The amount of the financial assurance must be increased, or may be decreased, as appropriate, to cover the detailed cost estimate for decommissioning established pursuant to paragraph (g)(4)(v) of this section.
            (1) Any licensee who has not provided financial assurance to cover the detailed cost estimate submitted with the decommissioning plan shall do so when this rule becomes effective November 24, 1995.
            (2) Following approval of the decommissioning plan, a licensee may reduce the amount of the financial assurance as decommissioning proceeds and radiological contamination is reduced at the site with the approval of the Commission.
            (f) The Commission may grant a request to delay or postpone initiation of the decommissioning process if the Commission determines that this relief is not detrimental to the public health and safety and is otherwise in the public interest. The request must be submitted no later than 30 days before notification pursuant to paragraph (d) of this section. The schedule for decommissioning set forth in paragraph (d) of this section may not commence until the Commission has made a determination on the request.
            (g)(1) A decommissioning plan must be submitted if required by license condition or if the procedures and activities necessary to carry out decommissioning of the site or separate building or outdoor area have not been previously approved by the Commission and these procedures could increase potential health and safety impacts to workers or to the public, such as in any of the following cases:
            (i) Procedures would involve techniques not applied routinely during cleanup or maintenance operations;
            (ii) Workers would be entering areas not normally occupied where surface contamination and radiation levels are significantly higher than routinely encountered during operation;
            (iii) Procedures could result in significantly greater airborne concentrations of radioactive materials than are present during operation; or
            (iv) Procedures could result in significantly greater releases of radioactive material to the environment than those associated with operation.
            (2) The Commission may approve an alternate schedule for submittal of a decommissioning plan required pursuant to paragraph (d) of this section if the Commission determines that the alternative schedule is necessary to the effective conduct of decommissioning operations and presents no undue risk from radiation to the public health and safety and is otherwise in the public interest.
            (3) The procedures listed in paragraph (g)(1) of this section may not be carried out prior to approval of the decommissioning plan.
            (4) The proposed decommissioning plan for the site or separate building or outdoor area must include:
            (i) A description of the conditions of the site or separate building or outdoor area sufficient to evaluate the acceptability of the plan;
            (ii) A description of planned decommissioning activities;
            (iii) A description of methods used to ensure protection of workers and the environment against radiation hazards during decommissioning;
            (iv) A description of the planned final radiation survey; and

            (v) An updated detailed cost estimate for decommissioning, comparison of that estimate with present funds set aside for decommissioning, and a plan for assuring the availability of adequate funds for completion of decommissioning.
            (vi) A description of the physical security plan and material control and accounting plan provisions in place during decommissioning.
            (vii) For decommissioning plans calling for completion of decommissioning later than 24 months after plan approval, a justification for the delay based on the criteria in paragraph (i) of this section.
            (5) The proposed decommissioning plan will be approved by the Commission if the information therein demonstrates that the decommissioning will be completed as soon as practical and that the health and safety of workers and the public will be adequately protected.
            (h)(1) Except as provided in paragraph (i) of this section, licensees shall complete decommissioning of the site or separate building or outdoor area as soon as practicable but no later than 24 months following the initiation of decommissioning.
            (2) Except as provided in paragraph (i) of this section, when decommissioning involves the entire site, the licensee shall request license termination as soon as practicable but no later than 24 months following the initiation of decommissioning.
            (i) The Commission may approve a request for an alternate schedule for completion of decommissioning of the site or separate building or outdoor area, and license termination if appropriate, if the Commission determines that the alternative is warranted by consideration of the following:
            (1) Whether it is technically feasible to complete decommissioning within the allotted 24-month period;
            (2) Whether sufficient waste disposal capacity is available to allow completion of decommissioning within the allotted 24-month period;
            (3) Whether a significant volume reduction in wastes requiring disposal will be achieved by allowing short-lived radionuclides to decay;
            (4) Whether a significant reduction in radiation exposure to workers can be achieved by allowing short-lived radionuclides to decay; and
            (5) Other site-specific factors which the Commission may consider appropriate on a case-by-case basis, such as regulatory requirements of other government agencies, lawsuits, groundwater treatment activities, monitored natural groundwater restoration, actions that could result in more environmental harm than deferred cleanup, and other factors beyond the control of the licensee.
            (j) As the final step in decommissioning, the licensee shall—
            (1) Certify the disposition of all licensed material, including accumulated wastes, by submitting a completed NRC Form 314 or equivalent information; and
            (2) Conduct a radiation survey of the premises where the licensed activities were carried out and submit a report of the results of this survey, unless the licensee demonstrates in some other manner that the premises are suitable for release in accordance with the criteria for decommissioning in 10 CFR part 20, subpart E. The licensee shall, as appropriate—
            (i) Report levels of gamma radiation in units of millisieverts (microroentgen) per hour at one meter from surfaces, and report levels of radioactivity, including alpha and beta, in units of megabecquerels (disintegrations per minute or microcuries) per 100 square centimeters removable and fixed for surfaces, megabecquerels (microcuries) per milliliter for water, and becquerels (picocuries) per gram for solids such as soils or concrete; and
            (ii) Specify the survey instrument(s) used and certify that each instrument is properly calibrated and tested.
            (k) Specific licenses, including expired licenses, will be terminated by written notice to the licensee when the Commission determines that:
            (1) Special nuclear material has been properly disposed;
            (2) Reasonable effort has been made to eliminate residual radioactive contamination, if present; and

            (3)(i) A radiation survey has been performed which demonstrates that the premises are suitable for release in accordance with the criteria for decommissioning in 10 CFR part 20, subpart E; or
            (ii) Other information submitted by the licensee is sufficient to demonstrate that the premises are suitable for release in accordance with the criteria for decommissioning in 10 CFR part 20, subpart E.
            (4) Records required by § 70.51(a) have been received.
            [59 FR 36037, July 15, 1994, as amended at 60 FR 38240, July 26, 1995; 61 FR 1115, Jan. 16, 1996; 61 FR 24675, May 16, 1996; 61 FR 29637, 29638, June 12, 1996; 62 FR 39091, July 21, 1997; 66 FR 24049, May 11, 2001; 73 FR 42675, July 23, 2008; 83 FR 58465, Nov. 20, 2018]

            § 70.39
            Specific licenses for the manufacture or initial transfer of calibration or reference sources.
            (a) An application for a specific license to manufacture or initially transfer calibration or reference sources containing plutonium, for distribution to persons generally licensed under § 70.19, will be approved if:
            (1) The applicant satisfies the general requirements of § 70.23.
            (2) The applicant submits sufficient information regarding each type of calibration or reference source pertinent to evaluation of the potential radiation exposure, including:
            (i) Chemical and physical form and maximum quantity of plutonium in the source;
            (ii) Details of construction and design;
            (iii) Details of the method of incorporation and binding of the plutonium in the source;
            (iv) Procedures for and results of prototype testing of sources, which are designed to contain more than 0.005 microcurie of plutonium, to demonstrate that the plutonium contained in each source will not be released or be removed from the source under normal conditions of use;
            (v) Details of quality control procedures to be followed in manufacture of the source;
            (vi) Description of labeling to be affixed to the source or the storage container for the source;
            (vii) Any additional information, including experimental studies and tests, required by the Commission to facilitate a determination of the safety of the source.
            (3) Each source will contain no more than 5 microcuries of plutonium.
            (4) The Commission determines, with respect to any type of source containing more than 0.005 microcurie of plutonium, that:
            (i) The method of incorporation and binding of the plutonium in the source is such that the plutonium will not be released or be removed from the source under normal conditions of use and handling of the source; and
            (ii) The source has been subjected to and has satisfactorily passed the prototype tests prescribed by paragraph (a)(5) of this section.
            (5) For any type of source which is designed to contain more than 0.005 microcurie of plutonium, the applicant has conducted prototype tests, in the order listed, on each of five prototypes of such source, which contains more than 0.005 microcurie of plutonium, as follows:
            (i) Initial measurement. The quantity of radioactive material deposited on the source shall be measured by direct counting of the source.
            (ii) Dry wipe test. The entire radioactive surface of the source shall be wiped with filter paper with the application of moderate finger pressure. Removal of radioactive material from the source shall be determined by measuring the radioactivity on the filter paper or by direct measurement of the radioactivity on the source following the dry wipe.
            (iii) Wet wipe test. The entire radioactive surface of the source shall be wiped with filter paper, moistened with water, with the application of moderate finger pressure. Removal of radioactive material from the source shall be determined by measuring the radioactivity on the filter paper after it has dried or by direct measurement of the radioactivity on the source following the wet wipe.
            (iv) Water soak test. The source shall be immersed in water at room temperature for a period of 24 consecutive hours. The source shall then be removed from the water. Removal of radioactive material from the source shall be determined by direct measurement of the radioactivity on the source after it has dried or by measuring the radioactivity in the residue obtained by evaporation of the water in which the source was immersed.
            (v) Dry wipe test. On completion of the preceding tests in paragraphs (a)(5)(i) through (iv) of this section, the dry wipe test described in paragraph (a)(5)(ii) of this section shall be repeated.
            (vi) Observations. Removal of more than 0.005 microcurie of radioactivity in any test prescribed by this paragraph shall be cause for rejection of the source design. Results of prototype tests submitted to the Commission shall be given in terms of radioactivity in microcuries and percent of removal from the total amount of radioactive material deposited on the source.
            (b) Each person licensed under this section shall affix to each source, or storage container for the source, a label which shall contain sufficient information relative to safe use and storage of the source and shall include the following statement or a substantially similar statement which contains the information called for in the following statement. 1

                1 Sources generally licensed under this section prior to January 19, 1975 may bear labels authorized by the regulations in effect on January 1, 1975.

              The receipt, possession, use and transfer of this source, Model ___, Serial No. ___, are subject to a general license and the regulations of the United States Nuclear Regulatory Commission or of a State with which the Commission has entered into an agreement for the exercise of regulatory authority. Do not remove this label.

                  caution—radioactive material—this source contains plutonium. do not touch radioactive portion of this source.

              (Name of Manufacturer or Initial Transferor)

            (c) Each person licensed under this section shall perform a dry wipe test upon each source containing more than 0.1 microcurie of plutonium prior to transferring the source to a general licensee under § 70.19. This test shall be performed by wiping the entire radioactive surface of the source with a filter paper with the application of moderate finger pressure. The radioactivity on the paper shall be measured by using radiation detection instrumentation capable of detecting 0.005 microcurie of plutonium. If any such test discloses more than 0.005 microcurie of radioactive material, the source shall be deemed to be leaking or losing plutonium and shall not be transferred to a general licensee under § 70.19.
            [29 FR 5884, May 5, 1964, as amended at 32 FR 2563, Feb. 7, 1967; 38 FR 1272, Jan. 11, 1973; 40 FR 8792, Mar. 3, 1975; 42 FR 43966, Sept. 1, 1977; 43 FR 6925, Feb. 17, 1978]

            § 70.40
            Ineligibility of certain applicants.
            A license may not be issued to the Corporation if the Commission determines that:
            (a) The Corporation is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government; or
            (b) The issuance of such a license would be inimical to—
            (1) The common defense and security of the United States; or
            (2) The maintenance of a reliable and economical domestic source of enrichment services.
            [62 FR 6669, Feb. 12, 1997]

          Subpart F—Acquisition, Use and Transfer of Special Nuclear Material, Creditors' Rights

            § 70.41
            Authorized use of special nuclear material.
            (a) Each licensee shall confine his possession and use of special nuclear material to the locations and purposes authorized in his license. Except as otherwise provided in the license, each license issued pursuant to the regulations in this part shall carry with it the right to receive title to, own, acquire, receive, possess and use special nuclear material. Preparation for shipment and transport of special nuclear material shall be in accordance with the provisions of part 71 of this chapter.

            (b) The possession, use and transfer of any special nuclear material produced by a licensee, in connection with or as a result of use of special nuclear material received under his license, shall be subject to the provisions of the license and the regulations in this part.
            [21 FR 764, Feb. 3, 1956, as amended at 38 FR 33970, Dec. 10, 1973; 43 FR 6925, Feb. 17, 1978]

            § 70.42
            Transfer of special nuclear material.
            (a) No licensee shall transfer special nuclear material except as authorized pursuant to this section.
            (b) Except as otherwise provided in his license and subject to the provisions of paragraphs (c) and (d) of this section, any licensee may transfer special nuclear material:
            (1) To the Department;
            (2) To the agency in any Agreement State which regulates radioactive materials pursuant to an agreement with the Commission or the Atomic Energy Commission under section 274 of the Act, if the quantity transferred is not sufficient to form a critical mass;
            (3) To any person exempt from the licensing requirements of the Act and regulations in this part, to the extent permitted under such exemption;
            (4) To any person in an Agreement State, subject to the jurisdiction of that State, who has been exempted from the licensing requirements and regulations of that State, to the extent permitted under such exemption;
            (5) To any person authorized to receive such special nuclear material under terms of a specific license or a general license or their equivalents issued by the Commission or an Agreement State;
            (6) To any person abroad pursuant to an export license issued under part 110 of this chapter; or
            (7) As otherwise authorized by the Commission in writing.
            (c) Before transferring special nuclear material to a specific licensee of the Commission or an Agreement State or to a general licensee who is required to register with the Commission or with an Agreement State prior to receipt of the special nuclear material, the licensee transferring the material shall verify that the transferee's license authorizes receipt of the type, form, and quantity of special nuclear material to be transferred.
            (d) The following methods for the verification required by paragraph (c) of this section are acceptable:
            (1) The transferor may have in his or her possession, and read, a current copy of the transferee's specific license or registration certificate. The transferor shall retain a copy of each license or certificate for three years from the date that it was obtained.
            (2) The transferor may have in its possession a written certification by the transferee that the transferee is authorized by license or registration certificate to receive the type, form, and quantity of special nuclear material to be transferred, specifying the license or registration certificate number, issuing agency, and expiration date. The transferor shall retain the written certification as a record for three years from the date of receipt of the certification;
            (3) For emergency shipments the transferor may accept oral certification by the transferee that he or she is authorized by license or registration certification to receive the type, form, and quantity of special nuclear material to be transferred, specifying the license or registration certificate number, issuing agency, and expiration date, provided that the oral certification is confirmed in writing within ten days. The transferor shall retain the written confirmation of the oral certification for three years from the date of receipt of the confirmation;
            (4) The transferor may obtain other sources of information compiled by a reporting service from official records of the Commission or the licensing agency of an Agreement State as to the identity of licensees and the scope and expiration dates of licenses and registrations. The transferor shall retain the compilation of information as a record for three years from the date that it was obtained; or

            (5) When none of the methods of verification described in paragraphs (d) (1) to (4) of this section are readily available or when a transferor desires to verify that information received by one of these methods is correct or up-to-date, the transferor may obtain and record confirmation from the Commission or the licensing agency of an Agreement State that the transferee is licensed to receive the special nuclear material. The transferor shall retain the record of confirmation for three years from the date the record is made.
            [38 FR 33970, Dec. 10, 1973, as amended at 40 FR 8792, Mar. 3, 1975; 43 FR 6925, Feb. 21, 1978; 53 FR 19253, May 27, 1988]

            § 70.44
            Creditor regulations.

            (a) Pursuant to section 184 of the Act, the Commission consents, without individual application, to the creation of any mortgage, pledge, or other lien upon any special nuclear material, not owned by the United States, which is subject to licensing: Provided:

            (1) That the rights of any creditor so secured may be exercised only in compliance with and subject to the same requirements and restrictions as would apply to the licensee pursuant to the provisions of the license, the Atomic Energy Act of 1954, as amended, and regulations issued by the Commission pursuant to said Act; and
            (2) That no creditor so secured may take possession of the special nuclear material pursuant to the provisions of this section prior to either the issuance of a license by the Commission authorizing such possession or the transfer of a license pursuant to § 70.36.
            (b) Nothing contained in this section shall be deemed to affect the means of acquiring, or the priority of, any tax lien or other lien provided by law.
            (c) As used in this section, creditor includes, without implied limitation, the trustee under any mortgage, pledge, or lien on special nuclear material made to secure any creditor, any trustee or receiver of the special nuclear material appointed by a court of competent jurisdiction in any action brought for the benefit of any creditor secured by such mortgage, pledge, or lien, any purchaser of such special nuclear material at the sale thereof upon foreclosure of such mortgage, pledge, or lien or upon exercise of any power of sale contained therein, or any assignee of any such purchaser.
            [32 FR 2563, Feb. 7, 1967, as amended at 35 FR 11461, July 17, 1970]

          Subpart G—Special Nuclear Material Control, Records, Reports and Inspections

            § 70.50
            Reporting requirements.
            (a) Immediate report. Each licensee shall notify the NRC as soon as possible but not later than 4 hours after the discovery of an event that prevents immediate protective actions necessary to avoid exposures to radiation or radioactive materials that could exceed regulatory limits or releases of licensed material that could exceed regulatory limits (events may include fires, explosions, toxic gas releases, etc.).
            (b) Twenty-four hour report. Each licensee shall notify the NRC within 24 hours after the discovery of any of the following events involving licensed material:
            (1) An unplanned contamination event that:
            (i) Requires access to the contaminated area, by workers or the public, to be restricted for more than 24 hours by imposing additional radiological controls or by prohibiting entry into the area;
            (ii) Involves a quantity of material greater than five times the lowest annual limit on intake specified in Appendix B of §§ 20.1001-20.2401 of 10 CFR part 20 for the material; and
            (iii) Has access to the area restricted for a reason other than to allow isotopes with a half-life of less than 24 hours to decay prior to decontamination.
            (2) An event in which equipment is disabled or fails to function as designed when:
            (i) The equipment is required by regulation or licensee condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident;
            (ii) The equipment is required to be available and operable when it is disabled or fails to function; and

            (iii) No redundant equipment is available and operable to perform the required safety function.

            (3) An event that requires unplanned medical treatment at a medical facility of an individual with spreadable radioactive contamination on the individual's clothing or body.
            (4) An unplanned fire or explosion damaging any licensed material or any device, container, or equipment containing licensed material when:
            (i) The quantity of material involved is greater than five times the lowest annual limit on intake specified in appendix B of §§ 20.1001-20.2401 of 10 CFR part 20 for the material; and
            (ii) The damage affects the integrity of the licensed material or its container.
            (c) Preparation and submission of reports. Reports made by licensees in response to the requirements of this section must be made as follows:
            (1) Licensees shall make reports required by paragraphs (a) and (b) of this section, and by § 70.74 and Appendix A of this part, if applicable, by telephone to the NRC Operations Center. 1
               To the extent that the information is available at the time of notification, the information provided in these reports must include:

                1 The commercial telephone number for the NRC Operations Center is (301) 816-5100.

            (i) Caller's name, position title, and call-back telephone number;
            (ii) Date, time, and exact location of the event;
            (iii) Description of the event, including:
            (A) Radiological or chemical hazards involved, including isotopes, quantities, and chemical and physical form of any material released;
            (B) Actual or potential health and safety consequences to the workers, the public, and the environment, including relevant chemical and radiation data for actual personnel exposures to radiation or radioactive materials or hazardous chemicals produced from licensed materials (e.g., level of radiation exposure, concentration of chemicals, and duration of exposure);
            (C) The sequence of occurrences leading to the event, including degradation or failure of structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences; and
            (D) Whether the remaining structures, systems, equipment, components, and activities of personnel relied on to prevent potential accidents or mitigate their consequences are available and reliable to perform their function;
            (iv) External conditions affecting the event;
            (v) Additional actions taken by the licensee in response to the event;
            (vi) Status of the event (e.g., whether the event is on-going or was terminated);
            (vii) Current and planned site status, including any declared emergency class;
            (viii) Notifications, related to the event, that were made or are planned to any local, State, or other Federal agencies;
            (ix) Status of any press releases, related to the event, that were made or are planned.
            (2) Written report. Each licensee that makes a report required by paragraph (a) or (b) of this section shall submit a written follow-up report within 30 days of the initial report. Written reports prepared pursuant to other regulations may be submitted to fulfill this requirement if the report contains all the necessary information, and the appropriate distribution is made. These written reports must be sent to the NRC's Document Control Desk, using an appropriate method listed in § 70.5(a), with a copy to the appropriate NRC regional office listed in appendix D to part 20 of this chapter. The reports must include the following:
            (i) Complete applicable information required by § 70.50(c)(1);
            (ii) The probable cause of the event, including all factors that contributed to the event and the manufacturer and model number (if applicable) of any equipment that failed or malfunctioned;
            (iii) Corrective actions taken or planned to prevent occurrence of similar or identical events in the future and the results of any evaluations or assessments; and

            (iv) For licensees subject to Subpart H of this part, whether the event was identified and evaluated in the Integrated Safety Analysis.
            (d) The provisions of § 70.50 do not apply to licensees subject to § 50.72. They do apply to those Part 50 licensees possessing material licensed under Part 70 that are not subject to the notification requirements in § 50.72.
            [56 FR 40769, Aug. 16, 1991; 56 FR 64980, Dec. 13, 1991, as amended at 59 FR 14087, Mar. 25, 1994; 65 FR 56226, Sept. 18, 2000; 68 FR 58817, Oct. 10, 2003; 79 FR 57725, Sept. 26, 2014]

            § 70.51
            Records requirements.
            (a) Before license termination, licensees shall forward the following records to the appropriate NRC Regional Office:
            (1) Records of disposal of licensed material made under 10 CFR 20.2002 (including burials authorized before January 28, 1981 1
              ), 20.2003, 20.2004, 20.2005;

                1 A previous § 20.304 permitted burial of small quantities of licensed materials in soil before January 28, 1981, without specific Commission authorization. See § 20.304 contained in the 10 CFR, parts 0 to 199, edition revised as of January 1, 1981.

            (2) Records required by 10 CFR 20.2103(b)(4); and
            (3) Records required by § 70.25(g).
            (b) If licensed activities are transferred or assigned in accordance with § 70.32(a)(3), the licensee shall transfer the following records to the new licensee and the new licensee will be responsible for maintaining these records until the license is terminated:
            (1) Records of disposal of licensed material made under 10 CFR 20.2002 (including burials authorized before January 28, 1981 1), 20.2003, 20.2004, 20.2005;
            (2) Records required by 10 CFR 20.2103(b)(4); and
            (3) Records required by § 70.25(g).
            (c)(1) Records which must be maintained pursuant to this part may be the original or a reproduced copy, or microform if the reproduced copy or microform is duly authenticated by authorized personnel, and the microform is capable of producing a clear and legible copy after storage for the period specified by Commission regulations. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications, must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.
            (2) If there is a conflict between the Commission's regulations in this part, license condition, or other written Commission approval or authorization pertaining to the retention period for the same type of record, the retention period specified in the regulations in this part for these records shall apply unless the Commission, under § 70.17 has granted a specific exemption from the record retention requirements specified in the regulations in this part.
            [67 FR 78142, Dec. 23, 2002, as amended at 72 FR 35145, June 27, 2007]

            § 70.52
            Reports of accidental criticality.
            (a) Each licensee shall notify the NRC Operations Center 1
               within one hour after discovery of any case of accidental criticality.

                1 Commercial telephone number of the NRC Operations Center is (301) 816-5100.

            (b) This notification must be made to the NRC Operations Center via the Emergency Notification System if the licensee is party to that system. If the Emergency Notification System is inoperative or unavailable, the licensee shall make the required notification via commercial telephonic service or other dedicated telephonic system or any other method that will ensure that a report is received by the NRC Operations Center within one hour.
            [67 FR 78143, Dec. 23, 2002]

            § 70.55
            Inspections.
            (a) Each licensee shall afford to the Commission at all reasonable times opportunity to inspect special nuclear material and the premises and facilities wherein special nuclear material is used, produced, or stored.

            (b) Each licensee shall make available to the Commission for inspection, upon reasonable notice, records kept by the licensee pertaining to his receipt, possession, use, acquisition, import, export, or transfer of special nuclear material.
            (c)(1) In the case of fuel cycle facilities where nuclear reactor fuel is fabricated or processed each licensee shall upon request by the Director, Office of Nuclear Material Safety and Safeguards or the appropriate NRC Regional Administrator, provide rent-free office space for the exclusive use of Commission inspection personnel. Heat, air conditioning, light, electrical outlets and janitorial services shall be furnished by each licensee. The office shall be convenient to and have full access to the facility and, shall provide the inspector both visual and acoustic privacy.
            (2) For a site with a single fuel facility licensed pursuant to part 70, the space provided shall be adequate to accommodate a full-time inspector, a part-time secretary and transient NRC personnel and will be generally commensurate with other office facilities at the site. A space of 250 square feet either within the site's office complex or in an office trailer or other on site space is suggested as a guide. For sites containing multiple fuel facilities, additional space may be requested to accommodate additional full-time inspector(s). The office space that is provided shall be subject to the approval of the Director, Office of Nuclear Material Safety and Safeguards or the appropriate NRC Regional Administrator. All furniture, supplies and communication equipment will be furnished by the Commission.
            (3) The licensee shall afford any NRC resident inspector assigned to that site or other NRC inspectors identified by the Director, Office of Nuclear Material Safety and Safeguards, as likely to inspect the facility, immediate unfettered access, equivalent to access provided regular plant employees, following proper identification and compliance with applicable access control measures for security, radiological protection, and personal safety.
            [21 FR 764, Feb. 3, 1956. Redesignated at 25 FR 1607, Feb. 25, 1960, and 25 FR 12730, Dec. 13, 1960, and amended at 32 FR 2563, Feb. 7, 1967; 44 FR 47919, Aug. 16, 1979; 52 FR 31612, Aug. 21, 1987; 54 FR 6877, Feb. 15, 1989; 55 FR 5979, Feb. 21, 1990]

            § 70.56
            Tests.
            Each licensee shall perform, or permit the Commission to perform, such tests as the Commission deems appropriate or necessary for the administration of the regulations in this part, including tests of (a) special nuclear material, (b) facilities wherein special nuclear material is utilized, produced or stored, (c) radiation detection and monitoring instruments, and (d) other equipment and devices used in connection with the production, utilization or storage of special nuclear material.
            [21 FR 764, Feb. 3, 1956. Redesignated at 25 FR 1607, Feb. 25, 1960, and 25 FR 12730, Dec. 13, 1960]

            § 70.59
            Effluent monitoring reporting requirements.

            Within 60 days after January 1 and July 1 of each year, and using an appropriate method listed in § 70.5(a), each licensee authorized to possess and use special nuclear material for processing and fuel fabrication, scrap recovery, conversion of uranium hexafluoride, or in a uranium enrichment facility shall submit a report addressed: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, with a copy to the appropriate NRC Regional Office shown in appendix D to part 20 of this chapter. The report must specify the quantity of each of the principal radionuclides released to unrestricted areas in liquid and gaseous effluents during the previous six months of operation, and such other information as the Commission may require to estimate maximum potential annual radiation doses to the public resulting from effluent releases. If quantities of radioactive materials released during the reporting periods are significantly above the licensee's design objectives previously reviewed as part of the licensing action, the report must cover this specifically. On the basis of these reports and any additional information the Commission may obtain from the licensee or others, the Commission may from time to time require the licensee to take such action as the Commission deems appropriate.
            [68 FR 58817, Oct. 10, 2003]

          Subpart H—Additional Requirements for Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material

            Source:
            65 FR 56226, Sept. 18, 2000, unless otherwise noted.

            § 70.60
            Applicability.
            The regulations in § 70.61 through § 70.76 apply, in addition to other applicable Commission regulations, to each applicant or licensee that is or plans to be authorized to possess greater than a critical mass of special nuclear material, and engaged in enriched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium enrichment, enriched uranium hexafluoride conversion, plutonium processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap recovery of special nuclear material, or any other activity that the Commission determines could significantly affect public health and safety. The regulations in § 70.61 through § 70.76 do not apply to decommissioning activities performed pursuant to other applicable Commission regulations including § 70.25 and § 70.38 of this part. Also, the regulations in § 70.61 through § 70.76 do not apply to activities that are certified by the Commission pursuant to part 76 of this chapter or licensed by the Commission pursuant to other parts of this chapter. Unless specifically addressed in § 70.61 through § 70.76, implementation by current licensees of the Subpart H requirements shall be completed no later than the time of the ISA Summary submittal required in § 70.62(c)(3)(ii).

            § 70.61
            Performance requirements.
            (a) Each applicant or licensee shall evaluate, in the integrated safety analysis performed in accordance with § 70.62, its compliance with the performance requirements in paragraphs (b), (c), and (d) of this section.
            (b) The risk of each credible high-consequence event must be limited. Engineered controls, administrative controls, or both, shall be applied to the extent needed to reduce the likelihood of occurrence of the event so that, upon implementation of such controls, the event is highly unlikely or its consequences are less severe than those in paragrahs (b)(1)-(4) of this section. High consequence events are those internally or externally initiated events that result in:
            (1) An acute worker dose of 1 Sv (100 rem) or greater total effective dose equivalent;
            (2) An acute dose of 0.25 Sv (25 rem) or greater total effective dose equivalent to any individual located outside the controlled area identified pursuant to paragraph (f) of this section;
            (3) An intake of 30 mg or greater of uranium in soluble form by any individual located outside the controlled area identified pursuant to paragraph (f) of this section; or
            (4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that:
            (i) Could endanger the life of a worker, or
            (ii) Could lead to irreversible or other serious, long-lasting health effects to any individual located outside the controlled area identified pursuant to paragraph (f) of this section. If an applicant possesses or plans to possess quantities of material capable of such chemical exposures, then the applicant shall propose appropriate quantitative standards for these health effects, as part of the information submitted pursuant to § 70.65 of this subpart.

            (c) The risk of each credible intermediate-consequence event must be limited. Engineered controls, administrative controls, or both shall be applied to the extent needed so that, upon implementation of such controls, the event is unlikely or its consequences are less than those in paragraphs (c)(1)-(4) of this section. Intermediate consequence events are those internally or externally initiated events that are not high consequence events, that result in:
            (1) An acute worker dose of 0.25 Sv (25 rem) or greater total effective dose equivalent;
            (2) An acute dose of 0.05 Sv (5 rem) or greater total effective dose equivalent to any individual located outside the controlled area identified pursuant to paragraph (f) of this section;
            (3) A 24-hour averaged release of radioactive material outside the restricted area in concentrations exceeding 5000 times the values in Table 2 of Appendix B to Part 20; or
            (4) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that:
            (i) Could lead to irreversible or other serious, long-lasting health effects to a worker, or
            (ii) Could cause mild transient health effects to any individual located outside the controlled area as specified in paragraph (f) of this section. If an applicant possesses or plans to possess quantities of material capable of such chemical exposures, then the applicant shall propose appropriate quantitative standards for these health effects, as part of the information submitted pursuant to § 70.65 of this subpart.
            (d) In addition to complying with paragraphs (b) and (c) of this section, the risk of nuclear criticality accidents must be limited by assuring that under normal and credible abnormal conditions, all nuclear processes are subcritical, including use of an approved margin of subcriticality for safety. Preventive controls and measures must be the primary means of protection against nuclear criticality accidents.
            (e) Each engineered or administrative control or control system necessary to comply with paragraphs (b), (c), or (d) of this section shall be designated as an item relied on for safety. The safety program, established and maintained pursuant to § 70.62 of this subpart, shall ensure that each item relied on for safety will be available and reliable to perform its intended function when needed and in the context of the performance requirements of this section.
            (f) Each licensee must establish a controlled area, as defined in § 20.1003. In addition, the licensee must retain the authority to exclude or remove personnel and property from the area. For the purpose of complying with the performance requirements of this section, individuals who are not workers, as defined in § 70.4, may be permitted to perform ongoing activities (e.g., at a facility not related to the licensed activities) in the controlled area, if the licensee:
            (1) Demonstrates and documents, in the integrated safety analysis, that the risk for those individuals at the location of their activities does not exceed the performance requirements of paragraphs (b)(2), (b)(3), (b)(4)(ii), (c)(2), and (c)(4)(ii) of this section; or
            (2) Provides training that satisfies 10 CFR 19.12(a)(1)-(5) to these individuals and ensures that they are aware of the risks associated with accidents involving the licensed activities as determined by the integrated safety analysis, and conspicuously posts and maintains notices stating where the information in 10 CFR 19.11(a) may be examined by these individuals. Under these conditions, the performance requirements for workers specified in paragraphs (b) and (c) of this section may be applied to these individuals.

            § 70.62
            Safety program and integrated safety analysis.
            (a) Safety program. (1) Each licensee or applicant shall establish and maintain a safety program that demonstrates compliance with the performance requirements of § 70.61. The safety program may be graded such that management measures applied are graded commensurate with the reduction of the risk attributable to that item. Three elements of this safety program; namely, process safety information, integrated safety analysis, and management measures, are described in paragraphs (b) through (d) of this section.

            (2) Each licensee or applicant shall establish and maintain records that demonstrate compliance with the requirements of paragraphs (b) through (d) of this section.

            (3) Each licensee or applicant shall maintain records of failures readily retrievable and available for NRC inspection, documenting each discovery that an item relied on for safety or management measure has failed to perform its function upon demand or has degraded such that the performance requirements of § 70.61 are not satisfied. These records must identify the item relied on for safety or management measure that has failed and the safety function affected, the date of discovery, date (or estimated date) of the failure, duration (or estimated duration) of the time that the item was unable to perform its function, any other affected items relied on for safety or management measures and their safety function, affected processes, cause of the failure, whether the failure was in the context of the performance requirements or upon demand or both, and any corrective or compensatory action that was taken. A failure must be recorded at the time of discovery and the record of that failure updated promptly upon the conclusion of each failure investigation of an item relied on for safety or management measure.
            (b) Process safety information. Each licensee or applicant shall maintain process safety information to enable the performance and maintenance of an integrated safety analysis. This process safety information must include information pertaining to the hazards of the materials used or produced in the process, information pertaining to the technology of the process, and information pertaining to the equipment in the process.
            (c) Integrated safety analysis. (1) Each licensee or applicant shall conduct and maintain an integrated safety analysis, that is of appropriate detail for the complexity of the process, that identifies:
            (i) Radiological hazards related to possessing or processing licensed material at its facility;
            (ii) Chemical hazards of licensed material and hazardous chemicals produced from licensed material;
            (iii) Facility hazards that could affect the safety of licensed materials and thus present an increased radiological risk;
            (iv) Potential accident sequences caused by process deviations or other events internal to the facility and credible external events, including natural phenomena;
            (v) The consequence and the likelihood of occurrence of each potential accident sequence identified pursuant to paragraph (c)(1)(iv) of this section, and the methods used to determine the consequences and likelihoods; and
            (vi) Each item relied on for safety identified pursuant to § 70.61(e) of this subpart, the characteristics of its preventive, mitigative, or other safety function, and the assumptions and conditions under which the item is relied upon to support compliance with the performance requirements of § 70.61.
            (2) Integrated safety analysis team qualifications. To assure the adequacy of the integrated safety analysis, the analysis must be performed by a team with expertise in engineering and process operations. The team shall include at least one person who has experience and knowledge specific to each process being evaluated, and persons who have experience in nuclear criticality safety, radiation safety, fire safety, and chemical process safety. One member of the team must be knowledgeable in the specific integrated safety analysis methodology being used.
            (3) Requirements for existing licensees. Individuals holding an NRC license on September 18, 2000 shall, with regard to existing licensed activities:
            (i) By April 18, 2001, submit for NRC approval, a plan that describes the integrated safety analysis approach that will be used, the processes that will be analyzed, and the schedule for completing the analysis of each process.

            (ii) By October 18, 2004, or in accordance with the approved plan submitted under § 70.62(c)(3)(i), complete an integrated safety analysis, correct all unacceptable performance deficiencies, and submit, for NRC approval, an integrated safety analysis summary, including a description of the management measures, in accordance with § 70.65. The Commission may approve a request for an alternative schedule for completing the correction of unacceptable performance deficiencies if the Commission determines that the alternative is warranted by consideration of the following:
            (A) Adequate compensatory measures have been established;
            (B) Whether it is technically feasible to complete the correction of the unacceptable performance deficiency within the allotted 4-year period;
            (C) Other site-specific factors which the Commission may consider appropriate on a case-by-case basis and that are beyond the control of the licensee.
            (iii) Pending the correction of unacceptable performance deficiencies identified during the conduct of the integrated safety analysis, the licensee shall implement appropriate compensatory measures to ensure adequate protection.
            (d) Management measures. Each applicant or licensee shall establish management measures to ensure compliance with the performance requirements of § 70.61. The measures applied to a particular engineered or administrative control or control system may be graded commensurate with the reduction of the risk attributable to that control or control system. The management measures shall ensure that engineered and administrative controls and control systems that are identified as items relied on for safety pursuant to § 70.61(e) of this subpart are designed, implemented, and maintained, as necessary, to ensure they are available and reliable to perform their function when needed, to comply with the performance requirements of § 70.61 of this subpart.

            § 70.64
            Requirements for new facilities or new processes at existing facilities.
            (a) Baseline design criteria. Each prospective applicant or licensee shall address the following baseline design criteria in the design of new facilities. Each existing licensee shall address the following baseline design criteria in the design of new processes at existing facilities that require a license amendment under § 70.72. The baseline design criteria must be applied to the design of new facilities and new processes, but do not require retrofits to existing facilities or existing processes (e.g., those housing or adjacent to the new process); however, all facilities and processes must comply with the performance requirements in § 70.61. Licensees shall maintain the application of these criteria unless the analysis performed pursuant to § 70.62(c) demonstrates that a given item is not relied on for safety or does not require adherence to the specified criteria.
            (1) Quality standards and records. The design must be developed and implemented in accordance with management measures, to provide adequate assurance that items relied on for safety will be available and reliable to perform their function when needed. Appropriate records of these items must be maintained by or under the control of the licensee throughout the life of the facility.
            (2) Natural phenomena hazards. The design must provide for adequate protection against natural phenomena with consideration of the most severe documented historical events for the site.
            (3) Fire protection. The design must provide for adequate protection against fires and explosions.
            (4) Environmental and dynamic effects. The design must provide for adequate protection from environmental conditions and dynamic effects associated with normal operations, maintenance, testing, and postulated accidents that could lead to loss of safety functions.
            (5) Chemical protection. The design must provide for adequate protection against chemical risks produced from licensed material, facility conditions which affect the safety of licensed material, and hazardous chemicals produced from licensed material.
            (6) Emergency capability. The design must provide for emergency capability to maintain control of:
            (i) Licensed material and hazardous chemicals produced from licensed material;
            (ii) Evacuation of on-site personnel; and
            (iii) Onsite emergency facilities and services that facilitate the use of available offsite services.
            (7) Utility services. The design must provide for continued operation of essential utility services.
            (8) Inspection, testing, and maintenance. The design of items relied on for safety must provide for adequate inspection, testing, and maintenance, to ensure their availability and reliability to perform their function when needed.
            (9) Criticality control. The design must provide for criticality control including adherence to the double contingency principle.
            (10) Instrumentation and controls. The design must provide for inclusion of instrumentation and control systems to monitor and control the behavior of items relied on for safety.
            (b) Facility and system design and facility layout must be based on defense-in-depth practices. 1
               The design must incorporate, to the extent practicable:

                1 As used in § 70.64, Requirements for new facilities or new processes at existing facilities, defense-in-depth practices means a design philosophy, applied from the outset and through completion of the design, that is based on providing successive levels of protection such that health and safety will not be wholly dependent upon any single element of the design, construction, maintenance, or operation of the facility. The net effect of incorporating defense-in-depth practices is a conservatively designed facility and system that will exhibit greater tolerance to failures and external challenges. The risk insights obtained through performance of the integrated safety analysis can be then used to supplement the final design by focusing attention on the prevention and mitigation of the higher-risk potential accidents.

            (1) Preference for the selection of engineered controls over administrative controls to increase overall system reliability; and
            (2) Features that enhance safety by reducing challenges to items relied on for safety.

            § 70.65
            Additional content of applications.
            (a) In addition to the contents required by § 70.22, each application must include a description of the applicant's safety program established under § 70.62.
            (b) The integrated safety analysis summary must be submitted with the license or renewal application (and amendment application as necessary), but shall not be incorporated in the license. However, changes to the integrated safety analysis summary shall meet the conditions of § 70.72. The integrated safety analysis summary must contain:

            (1) A general description of the site with emphasis on those factors that could affect safety (i.e., meteorology, seismology);
            (2) A general description of the facility with emphasis on those areas that could affect safety, including an identification of the controlled area boundaries;
            (3) A description of each process (defined as a single reasonably simple integrated unit operation within an overall production line) analyzed in the integrated safety analysis in sufficient detail to understand the theory of operation; and, for each process, the hazards that were identified in the integrated safety analysis pursuant to § 70.62(c)(1)(i)-(iii) and a general description of the types of accident sequences;
            (4) Information that demonstrates the licensee's compliance with the performance requirements of § 70.61, including a description of the management measures; the requirements for criticality monitoring and alarms in § 70.24; and, if applicable, the requirements of § 70.64;
            (5) A description of the team, qualifications, and the methods used to perform the integrated safety analysis;
            (6) A list briefly describing each item relied on for safety which is identified pursuant to § 70.61(e) in sufficient detail to understand their functions in relation to the performance requirements of § 70.61;
            (7) A description of the proposed quantitative standards used to assess the consequences to an individual from acute chemical exposure to licensed material or chemicals produced from licensed materials which are on-site, or expected to be on-site as described in § 70.61(b)(4) and (c)(4);
            (8) A descriptive list that identifies all items relied on for safety that are the sole item preventing or mitigating an accident sequence that exceeds the performance requirements of § 70.61; and
            (9) A description of the definitions of unlikely, highly unlikely, and credible as used in the evaluations in the integrated safety analysis.

            § 70.66
            Additional requirements for approval of license application.
            (a) An application for a license from an applicant subject to subpart H will be approved if the Commission determines that the applicant has complied with the requirements of §§ 70.21, 70.22, 70.23, and 70.60 through 70.65.
            (b) Submittals by existing licensees in accordance with § 70.62(c)(3)(i) will be approved if the Commission determines that:
            (1) The integrated safety analysis approach is in accordance with the requirements of §§ 70.61, 70.62(c)(1), and 70.62(c)(2); and
            (2) The schedule is in compliance with § 70.62(c)(3)(ii).
            (c) Submittals by existing licensees in accordance with § 70.62(c)(3)(ii) will be approved if the Commission determines that:
            (1) The requirements of § 70.65(b) are satisfied; and
            (2) The performance requirements in § 70.61 (b), (c) and (d) are satisfied, based on the information in the ISA Summary, together with other information submitted to NRC or available to NRC at the licensee's site.

            § 70.72
            Facility changes and change process.
            (a) The licensee shall establish a configuration management system to evaluate, implement, and track each change to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel. This system must be documented in written procedures and must assure that the following are addressed prior to implementing any change:
            (1) The technical basis for the change;
            (2) Impact of the change on safety and health or control of licensed material;
            (3) Modifications to existing operating procedures including any necessary training or retraining before operation;
            (4) Authorization requirements for the change;
            (5) For temporary changes, the approved duration (e.g., expiration date) of the change; and
            (6) The impacts or modifications to the integrated safety analysis, integrated safety analysis summary, or other safety program information, developed in accordance with § 70.62.
            (b) Any change to site, structures, processes, systems, equipment, components, computer programs, and activities of personnel must be evaluated by the licensee as specified in paragraph (a) of this section, before the change is implemented. The evaluation of the change must determine, before the change is implemented, if an amendment to the license is required to be submitted in accordance with § 70.34.
            (c) The licensee may make changes to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel, without prior Commission approval, if the change:
            (1) Does not:
            (i) Create new types of accident sequences that, unless mitigated or prevented, would exceed the performance requirements of § 70.61 and that have not previously been described in the integrated safety analysis summary; or
            (ii) Use new processes, technologies, or control systems for which the licensee has no prior experience;
            (2) Does not remove, without at least an equivalent replacement of the safety function, an item relied on for safety that is listed in the integrated safety analysis summary and is necessary for compliance with the performance requirements of § 70.61;
            (3) Does not alter any item relied on for safety, listed in the integrated safety analysis summary, that is the sole item preventing or mitigating an accident sequence that exceeds the performance requirements of § 70.61; and
            (4) Is not otherwise prohibited by this section, license condition, or order.
            (d)(1) For changes that require pre-approval under § 70.72, the licensee shall submit an amendment request to the NRC in accordance with § 70.34 and § 70.65 of this chapter.

            (2) For changes that do not require pre-approval under § 70.72, the licensee shall submit to NRC annually, within 30 days after the end of the calendar year during which the changes occurred, a brief summary of all changes to the records required by § 70.62(a)(2) of this subpart.
            (3) For all changes that affect the integrated safety analysis summary, the licensee shall submit to NRC annually, within 30 days after the end of the calendar year during which the changes occurred, revised integrated safety analysis summary pages.
            (e) If a change covered by § 70.72 is made, the affected on-site documentation must be updated promptly.
            (f) The licensee shall maintain records of changes to its facility carried out under this section. These records must include a written evaluation that provides the bases for the determination that the changes do not require prior Commission approval under paragraph (c) or (d) of this section. These records must be maintained until termination of the license.
            [65 FR 56226, Sept. 18, 2000, as amended at 71 FR 56346, Sept. 27, 2006]

            § 70.73
            Renewal of licenses.
            Applications for renewal of a license must be filed in accordance with §§ 2.109, 70.21, 70.22, 70.33, 70.38, and 70.65 of this chapter. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference, provided that these references are clear and specific.

            § 70.74
            Additional reporting requirements.
            (a) Reports to NRC Operations Center. (1) Each licensee shall report to the NRC Operations Center the events described in Appendix A to Part 70.
            (2) Reports must be made by a knowledgeable licensee representative and by any method that will ensure compliance with the required time period for reporting.
            (3) The information provided must include a description of the event and other related information as described in § 70.50(c)(1).
            (4) Follow-up information to the reports must be provided until all information required to be reported in § 70.50(c)(1) of this subpart is complete.
            (5) Each licensee shall provide reasonable assurance that reliable communication with the NRC Operations Center is available during each event.
            (b) Written reports. Each licensee that makes a report required by paragraph (a)(1) of this section shall submit a written follow-up report within 60 days of the initial report. The written report must be sent to the NRC's Document Control Desk, using an appropriate method listed in § 70.5(a), with a copy to the appropriate NRC regional office listed in appendix D to part 20 of this chapter. The reports must include the information as described in § 70.50(c)(2)(i) through (iv).
            [65 FR 56226, Sept. 18, 2000, as amended at 79 FR 57725, Sept. 26, 2014]

            § 70.76
            Backfitting.
            (a) For each licensee, this provision shall apply to Subpart H requirements as soon as the NRC approves that licensee's ISA Summary pursuant to § 70.66. For requirements other than Subpart H, this provision applies regardless of the status of the approval of a licensee's ISA Summary.
            (1) Backfitting is defined as the modification of, or addition to, systems, structures, or components of a facility; or to the procedures or organization required to operate a facility; any of which may result from a new or amended provision in the Commission rules or the imposition of a regulatory staff position interpreting the Commission rules that is either new or different from a previous NRC staff position.
            (2) Except as provided in paragraph (a)(4) of this section, the Commission shall require a systematic and documented analysis pursuant to paragraph (b) of this section for backfits which it seeks to impose.
            (3) Except as provided in paragraph (a)(4) of this section, the Commission shall require the backfitting of a facility only when it determines, based on the analysis described in paragraph (b) of this section, that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection.

            (4) The provisions of paragraphs (a)(2) and (a)(3) of this section are inapplicable and, therefore, backfit analysis is not required and the standards in paragraph (a)(3) of this section do not apply where the Commission finds and declares, with appropriately documented evaluation for its finding, any of the following:
            (i) That a modification is necessary to bring a facility into compliance with Subpart H of this part;
            (ii) That a modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee;
            (iii) That regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and is in accord with the common defense and security; or
            (iv) That the regulatory action involves defining or redefining what level of protection to the public health and safety or common defense and security should be regarded as adequate.
            (5) The Commission shall always require the backfitting of a facility if it determines that the regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and is in accord with the common defense and security.
            (6) The documented evaluation required by paragraph (a)(4) of this section must include a statement of the objectives of and reasons for the modification and the basis for invoking the exception. If immediate effective regulatory action is required, then the documented evaluation may follow, rather than precede, the regulatory action.
            (7) If there are two or more ways to achieve compliance with a license or the rules or orders of the Commission, or with written license commitments, or there are two or more ways to reach an adequate level of protection, then ordinarily the licensee is free to choose the way that best suits its purposes. However, should it be necessary or appropriate for the Commission to prescribe a specific way to comply with its requirements or to achieve adequate protection, then cost may be a factor in selecting the way, provided that the objective of compliance or adequate protection is met.
            (b) In reaching the determination required by paragraph (a)(3) of this section, the Commission will consider how the backfit should be scheduled in light of other ongoing regulatory activities at the facility and, in addition, will consider information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed backfit:
            (1) Statement of the specific objectives that the proposed backfit is designed to achieve;
            (2) General description of the activity that would be required by the licensee in order to complete the backfit;
            (3) Potential change in the risk to the public from the accidental release of radioactive material and hazardous chemicals produced from licensed material;
            (4) Potential impact on radiological exposure or exposure to hazardous chemicals produced from licensed material of facility employees;
            (5) Installation and continuing costs associated with the backfit, including the cost of facility downtime;
            (6) The potential safety impact of changes in facility or operational complexity, including the relationship to proposed and existing regulatory requirements;
            (7) The estimated resource burden on the NRC associated with the proposed backfit and the availability of such resources;
            (8) The potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed backfit; and
            (9) Whether the proposed backfit is interim or final and, if interim, the justification for imposing the proposed backfit on an interim basis.
            (c) No license will be withheld during the pendency of backfit analyses required by the Commission's rules.
            (d) The Executive Director for Operations shall be responsible for implementation of this section, and all analyses required by this section shall be approved by the Executive Director for Operations or his or her designee.
            [65 FR 56226, Sept. 18, 2000]

          Subpart I—Modification and Revocation of Licenses

            § 70.81
            Modification and revocation of licenses.
            (a) The terms and conditions of all licenses shall be subject to amendment, revision, or modification by reason of amendments to the Atomic Energy Act of 1954, or by reason of rules, regulations or orders issued in accordance with the Act or any amendments thereto;
            (b) Any license may be revoked, suspended or modified for any material false statements in the application or any statement of fact required under section 182 of the Act or because of conditions revealed by such application or statement of fact or any report, record, or inspection or other means which would warrant the Commission to refuse to grant a license on an original application, or for failure to construct or operate a facility in accordance with the terms of the construction permit or license, the technical specifications in the application, or for violation of, or failure to observe any of the terms and conditions of the Act, or of any regulation of the Commission.
            (c) Upon revocation, suspension or modification of a license, the Commission may immediately retake possession of all special nuclear material held by the licensee. In cases found by the Commission to be of extreme importance to the national defense or security, or to the health and safety of the public, the Commission may recapture any special nuclear material held by the licensee prior to any of the procedures provided under section 551-558 of title 5 of the United States Code.
            (d) Except in cases of willfulness or those in which the public health, interest or safety requires otherwise, no license shall be modified, suspended or revoked unless, prior to the institution of proceedings therefor, facts or conduct which may warrant such action shall have been called to the attention of the licensee in writing and the licensee shall have been accorded opportunity to demonstrate or achieve compliance with all lawful requirements.
            [21 FR 764, Feb. 3, 1956, as amended at 35 FR 11461, July 17, 1970. Redesignated at 65 FR 56226, Sept. 18, 2000]

            § 70.82
            Suspension and operation in war or national emergency.
            Whenever Congress declares that a state of war or national emergency exists, the Commission, if it finds it necessary to the common defense and security may,
            (a) Suspend any license it has issued.
            (b) Order the recapture of special nuclear material.
            (c) Order the operation of any licensed facility.
            (d) Order entry into any plant or facility in order to recapture special nuclear material or to operate the facility. Just compensation shall be paid for any damages caused by recapture of special nuclear material or by operation of any facility, pursuant to this section.
            [21 FR 764, Feb. 3, 1956, as amended at 32 FR 4056, Mar. 15, 1967; 35 FR 11461, July 17, 1970. Redesignated at 65 FR 56226, Sept. 18, 2000]

          Subpart J—Enforcement

            § 70.91
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of—
            (1) The Atomic Energy Act of 1954, as amended;
            (2) Title II of the Energy Reorganization Act of 1974, as amended; or
            (3) A regulation or order issued pursuant to those Acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:
            (1) For violations of—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section;

            (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.
            [57 FR 55077, Nov. 24, 1992. Redesignated at 65 FR 56226, Sept. 18, 2000]

            § 70.92
            Criminal penalties.
            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 70 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in part 70 that are not issued under sections 161b, 161i, or 161o, for the purposes of section 223 are as follows: §§ 70.1, 70.2, 70.4, 70.5, 70.6, 70.8, 70.11, 70.12, 70.13, 70.14, 70.17, 70.18, 70.23, 70.31, 70.33, 70.34, 70.35, 70.37, 70.66, 70.73, 70.76, 70.81, 70.82, 70.63, 70.91, and 70.92.
            [57 FR 55077, Nov. 24, 1992. Redesignated and amended at 65 FR 56226, Sept. 18, 2000]

            Pt. 70, App. A
            Appendix A to Part 70—Reportable Safety Events
            Licensees must comply with reporting requirements in this appendix. As required by 10 CFR 70.74, licensees subject to the requirements in subpart H of part 70, shall report:
            (a) One hour reports. Events to be reported to the NRC Operations Center within 1 hour of discovery, supplemented with the information in 10 CFR 70.50(c)(1) as it becomes available, followed by a written report within 60 days:
            (1) An inadvertent nuclear criticality.
            (2) An acute intake by an individual of 30 mg or greater of uranium in a soluble form.
            (3) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed material that exceeds the quantitative standards established to satisfy the requirements in § 70.61(b)(4).
            (4) An event or condition such that no items relied on for safety, as documented in the Integrated Safety Analysis summary, remain available and reliable, in an accident sequence evaluated in the Integrated Safety Analysis, to perform their function:
            (i) In the context of the performance requirements in § 70.61(b) and § 70.61(c), or
            (ii) Prevent a nuclear criticality accident (i.e., loss of all controls in a particular sequence).
            (b) Twenty-four hour reports. Events to be reported to the NRC Operations Center within 24 hours of discovery, supplemented with the information in 10 CFR 70.50(c)(1) as it becomes available, followed by a written report within 60 days:
            (1) Any event or condition that results in the facility being in a state that was not analyzed, was improperly analyzed, or is different from that analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of § 70.61.
            (2) Loss or degradation of items relied on for safety that results in failure to meet the performance requirement of § 70.61.
            (3) An acute chemical exposure to an individual from licensed material or hazardous chemicals produced from licensed materials that exceeds the quantitative standards that satisfy the requirements of § 70.61(c)(4).
            (4) Any natural phenomenon or other external event, including fires internal and external to the facility, that has affected or may have affected the intended safety function or availability or reliability of one or more items relied on for safety.
            (c) Concurrent Reports. Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made, shall be reported to the NRC Operations Center concurrent to the news release or other notification.
            [65 FR 56231, Sept. 18, 2000, as amended at 75 FR 73944, Nov. 30, 2010; 79 FR 57725, Sept. 26, 2014]

        Pt. 71
        PART 71—PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL

            Subpart A—General Provisions
            Sec.
            71.0
            Purpose and scope.
            71.1
            Communications and records.
            71.2
            Interpretations.
            71.3
            Requirement for license.
            71.4
            Definitions.
            71.5
            Transportation of licensed material.
            71.6
            Information collection requirements: OMB approval.
            71.7
            Completeness and accuracy of information.
            71.8
            Deliberate misconduct.
            71.9
            Employee protection.
            71.10
            Public inspection of application.
            71.11
            Protection of Safeguards Information.

            Subpart B—Exemptions
            71.12
            Specific exemptions.
            71.13
            Exemption of physicians.
            71.14
            Exemption for low-level materials.
            71.15
            Exemption from classification as fissile material.
            71.16
            [Reserved]

            Subpart C—General Licenses
            71.17
            General license: NRC-approved package.
            71.18
            [Reserved]
            71.19
            Previously approved package.
            71.21
            General license: Use of foreign approved package.
            71.22
            General license: Fissile material.
            71.23
            General license: Plutonium-beryllium special form material.
            71.24-71.25
            [Reserved]

            Subpart D—Application for Package Approval
            71.31
            Contents of application.
            71.33
            Package description.
            71.35
            Package evaluation.
            71.37
            Quality assurance.
            71.38
            Renewal of a certificate of compliance.
            71.39
            Requirement for additional information.

            Subpart E—Package Approval Standards
            71.41
            Demonstration of compliance.
            71.43
            General standards for all packages.
            71.45
            Lifting and tie-down standards for all packages.
            71.47
            External radiation standards for all packages.
            71.51
            Additional requirements for Type B packages.
            71.53
            [Reserved]
            71.55
            General requirements for fissile material packages.
            71.57
            [Reserved]
            71.59
            Standards for arrays of fissile material packages.
            71.61

            Special requirements for Type B packages containing more than 105A2.
            71.63
            Special requirement for plutonium shipments.
            71.64
            Special requirements for plutonium air shipments.
            71.65
            Additional requirements.

            Subpart F—Package, Special Form, and LSA-III Tests
            71.70
            Incorporation by reference.
            71.71
            Normal conditions of transport.
            71.73
            Hypothetical accident conditions.
            71.74
            Accident conditions for air transport of plutonium.
            71.75
            Qualification of special form radioactive material.
            71.77
            Qualification of LSA-III Material.

            Subpart G—Operating Controls and Procedures
            71.81
            Applicability of operating controls and procedures.
            71.83
            Assumptions as to unknown properties.
            71.85
            Preliminary determinations.
            71.87
            Routine determinations.
            71.88
            Air transport of plutonium.
            71.89
            Opening instructions.
            71.91
            Records.
            71.93
            Inspection and tests.
            71.95
            Reports.
            71.97
            Advance notification of shipment of irradiated reactor fuel and nuclear waste.
            71.99
            Violations.
            71.100
            Criminal penalties.

            Subpart H—Quality Assurance
            71.101
            Quality assurance requirements.
            71.103
            Quality assurance organization.
            71.105
            Quality assurance program.
            71.106
            Changes to quality assurance program.
            71.107
            Package design control.
            71.109
            Procurement document control.
            71.111
            Instructions, procedures, and drawings.
            71.113
            Document control.
            71.115
            Control of purchased material, equipment, and services.
            71.117
            Identification and control of materials, parts, and components.
            71.119
            Control of special processes.
            71.121
            Internal inspection.
            71.123
            Test control.
            71.125
            Control of measuring and test equipment.
            71.127
            Handling, storage, and shipping control.
            71.129
            Inspection, test, and operating status.
            71.131
            Nonconforming materials, parts, or components.
            71.133
            Corrective action.
            71.135
            Quality assurance records.
            71.137
            Audits.
            Appendix A to Part 71—Determination of A1 and A2

          Authority:
          Atomic Energy Act of 1954, secs. 53, 57, 62, 63, 81, 161, 182, 183, 223, 234, 1701 (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 2201, 2232, 2233, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 180 (42 U.S.C. 10175); 44 U.S.C. 3504 note.
          Section 71.97 also issued under Sec. 301, Pub. L. 96-295, 94 Stat. 789 (42 U.S.C. 5841 note).

          Source:
          60 FR 50264, Sept. 28, 1995, unless otherwise noted.

          Subpart A—General Provisions

            Source:
            69 FR 3786, Jan. 26, 2004, unless otherwise noted.

            § 71.0
            Purpose and scope.
            (a) This part establishes—
            (1) Requirements for packaging, preparation for shipment, and transportation of licensed material; and
            (2) Procedures and standards for NRC approval of packaging and shipping procedures for fissile material and for a quantity of other licensed material in excess of a Type A quantity.
            (b) The packaging and transport of licensed material are also subject to other parts of this chapter (e.g., 10 CFR parts 20, 21, 30, 40, 70, and 73) and to the regulations of other agencies (e.g., the U.S. Department of Transportation (DOT) and the U.S. Postal Service) 1
               having jurisdiction over means of transport. The requirements of this part are in addition to, and not in substitution for, other requirements.

                1 Postal Service manual (Domestic Mail Manual), Section 124, which is incorporated by reference at 39 CFR 111.1.

            (c) The regulations in this part apply to any licensee authorized by specific or general license issued by the Commission to receive, possess, use, or transfer licensed material, if the licensee delivers that material to a carrier for transport, transports the material outside the site of usage as specified in the NRC license, or transports that material on public highways. No provision of this part authorizes possession of licensed material.
            (d)(1) Exemptions from the requirement for license in § 71.3 are specified in § 71.14. General licenses for which no NRC package approval is required are issued in §§ 71.21 through 71.23. The general license in § 71.17 requires that an NRC certificate of compliance or other package approval be issued for the package to be used under this general license.
            (2) Application for package approval must be completed in accordance with subpart D of this part, demonstrating that the design of the package to be used satisfies the package approval standards contained in subpart E of this part, as related to the tests of subpart F of this part.
            (3) A licensee transporting licensed material, or delivering licensed material to a carrier for transport, shall comply with the operating control requirements of subpart G of this part; the quality assurance requirements of subpart H of this part; and the general provisions of subpart A of this part, including DOT regulations referenced in § 71.5.
            (e) The regulations of this part apply to any person holding, or applying for, a certificate of compliance, issued pursuant to this part, for a package intended for the transportation of radioactive material, outside the confines of a licensee's facility or authorized place of use.
            (f) The regulations in this part apply to any person required to obtain a certificate of compliance, or an approved compliance plan, pursuant to part 76 of this chapter, if the person delivers radioactive material to a common or contract carrier for transport or transports the material outside the confines of the person's plant or other authorized place of use.
            (g) This part also gives notice to all persons who knowingly provide to any licensee, certificate holder, quality assurance program approval holder, applicant for a license, certificate, or quality assurance program approval, or to a contractor, or subcontractor of any of them, components, equipment, materials, or other goods or services, that relate to a licensee's, certificate holder's, quality assurance program approval holder's, or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 71.8.
            [69 FR 3786, Jan. 26, 2004, as amended at 80 FR 34011, June 12, 2015]

            § 71.1
            Communications and records.

            (a) Except where otherwise specified, all communications and reports concerning the regulations in this part and applications filed under them should be sent by mail addressed: ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the submission date falls on a Saturday, Sunday, or a Federal holiday, the next Federal working day becomes the official due date.
            (b) Each record required by this part must be legible throughout the retention period specified by each Commission regulation. The record may be the original or a reproduced copy or a microform provided that the copy or microform is authenticated by authorized personnel and that the microform is capable of producing a clear copy throughout the required retention period. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.
            [69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004, as amended at 74 FR 62683, Dec. 1, 2009; 75 FR 73945, Nov. 30, 2010; 79 FR 75741, Dec. 19, 2014; 80 FR 74981, Dec. 1, 2015; 84 FR 65645, Nov. 29, 2019]

            § 71.2
            Interpretations.
            Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission, other than a written interpretation by the General Counsel, will be recognized to be binding upon the Commission.

            § 71.3
            Requirement for license.
            Except as authorized in a general license or a specific license issued by the Commission, or as exempted in this part, no licensee may—
            (a) Deliver licensed material to a carrier for transport; or
            (b) Transport licensed material.

            § 71.4
            Definitions.
            The following terms are as defined here for the purpose of this part. To ensure compatibility with international transportation standards, all limits in this part are given in terms of dual units: The International System of Units (SI) followed or preceded by U.S. standard or customary units. The U.S. customary units are not exact equivalents but are rounded to a convenient value, providing a functionally equivalent unit. For the purpose of this part, either unit may be used.

              A
              1 means the maximum activity of special form radioactive material permitted in a Type A package. This value is either listed in Appendix A, Table A-1, of this part, or may be derived in accordance with the procedures prescribed in Appendix A of this part.

              A
              2 means the maximum activity of radioactive material, other than special form material, LSA, and SCO material, permitted in a Type A package. This value is either listed in Appendix A, Table A-1, of this part, or may be derived in accordance with the procedures prescribed in Appendix A of this part.

              Carrier means a person engaged in the transportation of passengers or property by land or water as a common, contract, or private carrier, or by civil aircraft.

              Certificate holder means a person who has been issued a certificate of compliance or other package approval by the Commission.

              Certificate of Compliance (CoC) means the certificate issued by the Commission under subpart D of this part which approves the design of a package for the transportation of radioactive material.

              Close reflection by water means immediate contact by water of sufficient thickness for maximum reflection of neutrons.

              Consignment means each shipment of a package or groups of packages or load of radioactive material offered by a shipper for transport.

              Containment system means the assembly of components of the packaging intended to retain the radioactive material during transport.

              Contamination means the presence of a radioactive substance on a surface in quantities in excess of 0.4 Bq/cm2 (1 × 10−5 µCi/cm2) for beta and gamma emitters and low toxicity alpha emitters, or 0.04 Bq/cm2 (1 × 10−6 µCi/cm2) for all other alpha emitters.
            (1) Fixed contamination means contamination that cannot be removed from a surface during normal conditions of transport.
            (2) Non-fixed contamination means contamination that can be removed from a surface during normal conditions of transport.

              Conveyance means:

            (1) For transport by public highway or rail any transport vehicle or large freight container;
            (2) For transport by water any vessel, or any hold, compartment, or defined deck area of a vessel including any transport vehicle on board the vessel; and
            (3) For transport by any aircraft.

              Criticality Safety Index (CSI) means the dimensionless number (rounded up to the next tenth) assigned to and placed on the label of a fissile material package, to designate the degree of control of accumulation of packages, overpacks or freight containers containing fissile material during transportation. Determination of the criticality safety index is described in §§ 71.22, 71.23, and 71.59. The criticality safety index for an overpack, freight container, consignment or conveyance containing fissile material packages is the arithmetic sum of the criticality safety indices of all the fissile material packages contained within the overpack, freight container, consignment or conveyance.

              Deuterium means, for the purposes of §§ 71.15 and 71.22, deuterium and any deuterium compounds, including heavy water, in which the ratio of deuterium atoms to hydrogen atoms exceeds 1:5000.

              DOT means the U.S. Department of Transportation.

              Exclusive use means the sole use by a single consignor of a conveyance for which all initial, intermediate, and final loading and unloading are carried out in accordance with the direction of the consignor or consignee. The consignor and the carrier must ensure that any loading or unloading is performed by personnel having radiological training and resources appropriate for safe handling of the consignment. The consignor must issue specific instructions, in writing, for maintenance of exclusive use shipment controls, and include them with the shipping paper information provided to the carrier by the consignor.

              Fissile material means the radionuclides uranium-233, uranium-235, plutonium-239, and plutonium-241, or any combination of these radionuclides. Fissile material means the fissile nuclides themselves, not material containing fissile nuclides. Unirradiated natural uranium and depleted uranium and natural uranium or depleted uranium, that has been irradiated in thermal reactors only, are not included in this definition. Certain exclusions from fissile material controls are provided in § 71.15.

              Graphite means, for the purposes of §§ 71.15 and 71.22, graphite with a boron equivalent content less than 5 parts per million and density greater than 1.5 grams per cubic centimeter.

              Indian tribe means an Indian or Alaska Native Tribe, band, nation, pueblo, village, or community that the Secretary of the Interior acknowledges to exist as an Indian Tribe pursuant to the Federally Recognized Indian Tribe List Act of 1994, 25 U.S.C. 5130.

              Licensed material means byproduct, source, or special nuclear material received, possessed, used, or transferred under a general or specific license issued by the Commission pursuant to the regulations in this chapter.

              Low Specific Activity (LSA) material means radioactive material with limited specific activity which is nonfissile or is excepted under § 71.15, and which satisfies the descriptions and limits set forth in the following section. Shielding materials surrounding the LSA material may not be considered in determining the estimated average specific activity of the package contents. The LSA material must be in one of three groups:
            (1) LSA-I.
            (i) Uranium and thorium ores, concentrates of uranium and thorium ores, and other ores containing naturally occurring radionuclides that are intended to be processed for the use of these radionuclides;
            (ii) Natural uranium, depleted uranium, natural thorium or their compounds or mixtures, provided they are unirradiated and in solid or liquid form;

            (iii) Radioactive material other than fissile material, for which the A2 value is unlimited; or
            (iv) Other radioactive material in which the activity is distributed throughout and the estimated average specific activity does not exceed 30 times the value for exempt material activity concentration determined in accordance with appendix A.
            (2) LSA-II.
            (i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/liter); or

            (ii) Other radioactive material in which the activity is distributed throughout and the estimated average specific activity does not exceed 10−4 A2/g for solids and gases, and 10−5 A2/g for liquids.
            (3) LSA-III. Solids (e.g., consolidated wastes, activated materials), excluding powders, that satisfy the requirements of § 71.77, in which:
            (i) The radioactive material is distributed throughout a solid or a collection of solid objects, or is essentially uniformly distributed in a solid compact binding agent (such as concrete, bitumen, ceramic, etc.);

            (ii) The radioactive material is relatively insoluble, or it is intrinsically contained in a relatively insoluble material, so that even under loss of packaging, the loss of radioactive material per package by leaching when placed in water for 7 days will not exceed 0.1 A2; and

            (iii) The estimated average specific activity of the solid, excluding any shielding material, does not exceed 2 × 10−3 A2/g.

              Low toxicity alpha emitters means natural uranium, depleted uranium, natural thorium; uranium-235, uranium-238, thorium-232, thorium-228 or thorium-230 when contained in ores or physical or chemical concentrates or tailings; or alpha emitters with a half-life of less than 10 days.

              Maximum normal operating pressure means the maximum gauge pressure that would develop in the containment system in a period of 1 year under the heat condition specified in § 71.71(c)(1), in the absence of venting, external cooling by an ancillary system, or operational controls during transport.

              Natural thorium means thorium with the naturally occurring distribution of thorium isotopes (essentially 100 weight percent thorium-232).

              Normal form radioactive material means radioactive material that has not been demonstrated to qualify as “special form radioactive material.”

              Optimum interspersed hydrogenous moderation means the presence of hydrogenous material between packages to such an extent that the maximum nuclear reactivity results.

              Package means the packaging together with its radioactive contents as presented for transport.
            (1) Fissile material package or Type AF package, Type BF package, Type B(U)F package, or Type B(M)F package means a fissile material packaging together with its fissile material contents.
            (2) Type A package means a Type A packaging together with its radioactive contents. A Type A package is defined and must comply with the DOT regulations in 49 CFR part 173.

            (3) Type B package means a Type B packaging together with its radioactive contents. On approval, a Type B package design is designated by NRC as B(U) unless the package has a maximum normal operating pressure of more than 700 kPa (100 lbs/in2) gauge or a pressure relief device that would allow the release of radioactive material to the environment under the tests specified in § 71.73 (hypothetical accident conditions), in which case it will receive a designation B(M). B(U) refers to the need for unilateral approval of international shipments; B(M) refers to the need for multilateral approval of international shipments. There is no distinction made in how packages with these designations may be used in domestic transportation. To determine their distinction for international transportation, see DOT regulations in 49 CFR Part 173. A Type B package approved before September 6, 1983, was designated only as Type B. Limitations on its use are specified in § 71.19.

              Packaging means the assembly of components necessary to ensure compliance with the packaging requirements of this part. It may consist of one or more receptacles, absorbent materials, spacing structures, thermal insulation, radiation shielding, and devices for cooling or absorbing mechanical shocks. The vehicle, tie-down system, and auxiliary equipment may be designated as part of the packaging.

              Special form radioactive material means radioactive material that satisfies the following conditions:
            (1) It is either a single solid piece or is contained in a sealed capsule that can be opened only by destroying the capsule;
            (2) The piece or capsule has at least one dimension not less than 5 mm (0.2 in); and
            (3) It satisfies the requirements of § 71.75. A special form encapsulation designed in accordance with the requirements of § 71.4 in effect on June 30, 1983 (see 10 CFR part 71, revised as of January 1, 1983), and constructed before July 1, 1985; a special form encapsulation designed in accordance with the requirements of § 71.4 in effect on March 31, 1996 (see 10 CFR part 71, revised as of January 1, 1996), and constructed before April 1, 1998; and special form material that was successfully tested before September 10, 2015 in accordance with the requirements of § 71.75(d) of this section in effect before September 10, 2015 may continue to be used. Any other special form encapsulation must meet the specifications of this definition.

              Specific activity of a radionuclide means the radioactivity of the radionuclide per unit mass of that nuclide. The specific activity of a material in which the radionuclide is essentially uniformly distributed is the radioactivity per unit mass of the material.

              Spent nuclear fuel or Spent fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, has undergone at least 1 year's decay since being used as a source of energy in a power reactor, and has not been chemically separated into its constituent elements by reprocessing. Spent fuel includes the special nuclear material, byproduct material, source material, and other radioactive materials associated with fuel assemblies.

              State means a State of the United States, the District of Columbia, the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American Samoa, and the Commonwealth of the Northern Mariana Islands.

              Surface Contaminated Object (SCO) means a solid object that is not itself classed as radioactive material, but which has radioactive material distributed on any of its surfaces. SCO must be in one of two groups with surface activity not exceeding the following limits:
            (1) SCO-I: A solid object on which:

            (i) The nonfixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 Bq/cm2 (10−4 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 0.4 Bq/cm2 (10−5 microcurie/cm2) for all other alpha emitters;
            (ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 × 104 Bq/cm2 (1.0 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 4 × 103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha emitters; and

            (iii) The nonfixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 × 104 Bq/cm2 (1 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 4 × 103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha emitters.

            (2) SCO-II: A solid object on which the limits for SCO-I are exceeded and on which:

            (i) The nonfixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 400 Bq/cm2 (10−2 microcurie/cm2) for beta and gamma and low toxicity alpha emitters or 40 Bq/cm2 (10−3 microcurie/cm2) for all other alpha emitters;
            (ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8 × 105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8 × 104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters; and
            (iii) The nonfixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8 × 105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8 × 104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters.

              Transport index (TI) means the dimensionless number (rounded up to the next tenth) placed on the label of a package, to designate the degree of control to be exercised by the carrier during transportation. The transport index is the number determined by multiplying the maximum radiation level in millisievert (mSv) per hour at 1 meter (3.3 ft) from the external surface of the package by 100 (equivalent to the maximum radiation level in millirem per hour at 1 meter (3.3 ft)).

              Tribal official means the highest ranking individual that represents Tribal leadership, such as the Chief, President, or Tribal Council leadership.

              Type A quantity means a quantity of radioactive material, the aggregate radioactivity of which does not exceed A1 for special form radioactive material, or A2, for normal form radioactive material, where A1 and A2 are given in Table A-1 of this part, or may be determined by procedures described in Appendix A of this part.

              Type B quantity means a quantity of radioactive material greater than a Type A quantity.

              Unirradiated uranium means uranium containing not more than 2 × 103 Bq of plutonium per gram of uranium-235, not more than 9 × 106 Bq of fission products per gram of uranium-235, and not more than 5 × 10−3 g of uranium-236 per gram of uranium-235.

              Uranium—natural, depleted, enriched. (1) Natural uranium means uranium (which may be chemically separated) with the naturally occurring distribution of uranium isotopes (approximately 0.711 weight percent uranium-235, and the remainder by weight essentially uranium-238).
            (2) Depleted uranium means uranium containing less uranium-235 than the naturally occurring distribution of uranium isotopes.
            (3) Enriched uranium means uranium containing more uranium-235 than the naturally occurring distribution of uranium isotopes.
            [69 FR 3786, Jan. 26, 2004, as amended at 69 FR 58038, Sept. 29, 2004; 77 FR 34204, June 11, 2012; 80 FR 34011, June 12, 2015; 80 FR 48684, Aug. 14, 2015; 80 FR 74981, Dec. 1, 2015; 82 FR 52825, Nov. 15, 2017]

            § 71.5
            Transportation of licensed material.
            (a) Each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the DOT regulations in 49 CFR parts 107, 171 through 180, and 390 through 397, appropriate to the mode of transport.
            (1) The licensee shall particularly note DOT regulations in the following areas:
            (i) Packaging—49 CFR part 173: subparts A, B, and I.
            (ii) Marking and labeling—49 CFR part 172: subpart D; and §§ 172.400 through 172.407 and §§ 172.436 through 172.441 of subpart E.
            (iii) Placarding—49 CFR part 172: subpart F, especially §§ 172.500 through 172.519 and 172.556; and appendices B and C.
            (iv) Accident reporting—49 CFR part 171: §§ 171.15 and 171.16.

            (v) Shipping papers and emergency information—49 CFR part 172: subparts C and G.

            (vi) Hazardous material employee training—49 CFR part 172: subpart H.
            (vii) Security plans—49 CFR part 172: subpart I.
            (viii) Hazardous material shipper/carrier registration—49 CFR part 107: subpart G.
            (2) The licensee shall also note DOT regulations pertaining to the following modes of transportation:
            (i) Rail—49 CFR part 174: subparts A through D and K.
            (ii) Air—49 CFR part 175.
            (iii) Vessel—49 CFR part 176: subparts A through F and M.
            (iv) Public Highway—49 CFR part 177 and parts 390 through 397.
            (b) If DOT regulations are not applicable to a shipment of licensed material, the licensee shall conform to the standards and requirements of the DOT specified in paragraph (a) of this section to the same extent as if the shipment or transportation were subject to DOT regulations. A request for modification, waiver, or exemption from those requirements, and any notification referred to in those requirements, must be filed with, or made to, the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
            [69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004]

            § 71.6
            Information collection requirements: OMB approval.

            (a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0008.
            (b) The approved information collection requirements contained in this part appear in §§ 71.5, 71.7, 71.9, 71.12, 71.17, 71.19, 71.22, 71.23, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.41, 71.47, 71.85, 71.87, 71.89, 71.91, 71.93, 71.95, 71.97, 71.101, 71.103, 71.105, 71.106, 71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, 71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137, and appendix A, paragraph II.
            [69 FR 3786, Jan. 26, 2004, as amended at 75 FR 73945, Nov. 30, 2010; 80 FR 34012, June 12, 2015]

            § 71.7
            Completeness and accuracy of information.
            (a) Information provided to the Commission by a licensee, certificate holder, or an applicant for a license or CoC; or information required by statute or by the Commission's regulations, orders, license or CoC conditions, to be maintained by the licensee or certificate holder, must be complete and accurate in all material respects.
            (b) Each licensee, certificate holder, or applicant for a license or CoC must notify the Commission of information identified by the licensee, certificate holder, or applicant for a license or CoC as having, for the regulated activity, a significant implication for public health and safety or common defense and security. A licensee, certificate holder, or an applicant for a license or CoC violates this paragraph only if the licensee, certificate holder, or applicant for a license or CoC fails to notify the Commission of information that the licensee, certificate holder, or applicant for a license or CoC has identified as having a significant implication for public health and safety or common defense and security. Notification must be provided to the Administrator of the appropriate Regional Office within 2 working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.

            § 71.8
            Deliberate misconduct.
            (a) This section applies to any—
            (1) Licensee;
            (2) Certificate holder;
            (3) Quality assurance program approval holder;
            (4) Applicant for a license, certificate, or quality assurance program approval;

            (5) Contractor (including a supplier or consultant) or subcontractor, to any person identified in paragraph (a)(4) of this section; or
            (6) Employees of any person identified in paragraphs (a)(1) through (a)(5) of this section.
            (b) A person identified in paragraph (a) of this section who knowingly provides to any entity, listed in paragraphs (a)(1) through (a)(5) of this section, any components, materials, or other goods or services that relate to a licensee's, certificate holder's, quality assurance program approval holder's, or applicant's activities subject to this part may not:
            (1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee, certificate holder, quality assurance program approval holder, or any applicant to be in violation of any rule, regulation, or order; or any term, condition or limitation of any license, certificate, or approval issued by the Commission; or
            (2) Deliberately submit to the NRC, a licensee, a certificate holder, quality assurance program approval holder, an applicant for a license, certificate or quality assurance program approval, or a licensee's, applicant's, certificate holder's, or quality assurance program approval holder's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.
            (c) A person who violates paragraph (b)(1) or (b)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.
            (d) For the purposes of paragraph (b)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:
            (1) Would cause a licensee, certificate holder, quality assurance program approval holder, or applicant for a license, certificate, or quality assurance program approval to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license or certificate issued by the Commission; or
            (2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, certificate holder, quality assurance program approval holder, applicant, or the contractor or subcontractor of any of them.

            § 71.9
            Employee protection.
            (a) Discrimination by a Commission licensee, certificate holder, an applicant for a Commission license or a CoC, or a contractor or subcontractor of any of these, against an employee for engaging in certain protected activities, is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act of 1954, as amended, or the Energy Reorganization Act of 1974, as amended.
            (1) The protected activities include, but are not limited to:
            (i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) of this section or possible violations of requirements imposed under either of those statutes;
            (ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) of this section or under these requirements if the employee has identified the alleged illegality to the employer;
            (iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;
            (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) of this section; and
            (v) Assisting or participating in, or is about to assist or participate in, these activities.
            (2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee's assistance or participation.

            (3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.
            (b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.
            (c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, certificate holder, applicant for a Commission license or a CoC, or a contractor or subcontractor of any of these may be grounds for:
            (1) Denial, revocation, or suspension of the license or the CoC;
            (2) Imposition of a civil penalty on the licensee, applicant, or a contractor or subcontractor of the licensee or applicant; or
            (3) Other enforcement action.
            (d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.
            (e)(1) Each licensee, certificate holder, and applicant for a license or CoC must prominently post the current revision of NRC Form 3, “Notice to Employees,” referenced in § 19.11(c) of this chapter. This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. The premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license or CoC, and for 30 days following license or CoC termination.

            (2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in Appendix D to Part 20 of this chapter, via email to Forms.Resource@nrc.gov, or by visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.

            (f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in a protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.
            [69 FR 3786, Jan. 26, 2004, as amended at 72 FR 63975, Nov. 14, 2007; 73 FR 30459, May 28, 2008; 79 FR 66605, Nov. 10, 2014]

            § 71.10
            Public inspection of application.
            Applications for approval of a package design under this part, which are submitted to the Commission, may be made available for public inspection, in accordance with provisions of parts 2 and 9 of this chapter. This includes an application to amend or revise an existing package design, any associated documents and drawings submitted with the application, and any responses to NRC requests for additional information.

            § 71.11
            Protection of Safeguards Information.
            Each licensee, certificate holder, or applicant for a Certificate of Compliance for a transportation package for transport of irradiated reactor fuel, strategic special nuclear material, a critical mass of special nuclear material, or byproduct material in quantities determined by the Commission through order or regulation to be significant to the public health and safety or the common defense and security, shall protect Safeguards Information against unauthorized disclosure in accordance with the requirements in § 73.21 and the requirements of § 73.22 or § 73.23 of this chapter, as applicable.
            [73 FR 63572, Oct. 24, 2008]

          Subpart B—Exemptions

            Source:
            69 FR 3786, Jan. 26, 2004, unless otherwise noted.

            § 71.12
            Specific exemptions.
            On application of any interested person or on its own initiative, the Commission may grant any exemption from the requirements of the regulations in this part that it determines is authorized by law and will not endanger life or property nor the common defense and security.

            § 71.13
            Exemption of physicians.
            Any physician licensed by a State to dispense drugs in the practice of medicine is exempt from § 71.5 with respect to transport by the physician of licensed material for use in the practice of medicine. However, any physician operating under this exemption must be licensed under 10 CFR part 35 or the equivalent Agreement State regulations.

            § 71.14
            Exemption for low-level materials.
            (a) A licensee is exempt from all the requirements of this part with respect to shipment or carriage of the following low-level materials:
            (1) Natural material and ores containing naturally occurring radionuclides that are either in their natural state, or have only been processed for purposes other than for the extraction of the radionuclides, and which are not intended to be processed for the use of these radionuclides, provided the activity concentration of the material does not exceed 10 times the applicable radionuclide activity concentration values specified in appendix A, Table A-2, or Table A-3 of this part.
            (2) Materials for which the activity concentration is not greater than the activity concentration values specified in appendix A, Table A-2, or Table A-3 of this part, or for which the consignment activity is not greater than the limit for an exempt consignment found in appendix A, Table A-2, or Table A-3 of this part.
            (3) Non-radioactive solid objects with radioactive substances present on any surfaces in quantities not in excess of the levels cited in the definition of contamination in § 71.4.
            (b) A licensee is exempt from all the requirements of this part, other than §§ 71.5 and 71.88, with respect to shipment or carriage of the following packages, provided the packages do not contain any fissile material, or the material is exempt from classification as fissile material under § 71.15:
            (1) A package that contains no more than a Type A quantity of radioactive material;
            (2) A package transported within the United States that contains no more than 0.74 TBq (20 Ci) of special form plutonium-244; or
            (3) The package contains only LSA or SCO radioactive material, provided—
            (i) That the LSA or SCO material has an external radiation dose of less than or equal to 10 mSv/h (1 rem/h), at a distance of 3 m from the unshielded material; or
            (ii) That the package contains only LSA-I or SCO-I material.
            [69 FR 3786, Jan. 26, 2004, as amended at 80 FR 34012, June 12, 2015]

            § 71.15
            Exemption from classification as fissile material.

            Fissile material meeting the requirements of at least one of the paragraphs (a) through (f) of this section are exempt from classification as fissile material and from the fissile material package standards of §§ 71.55 and 71.59, but are subject to all other requirements of this part, except as noted.

            (a) Individual package containing 2 grams or less fissile material.
            (b) Individual or bulk packaging containing 15 grams or less of fissile material provided the package has at least 200 grams of solid nonfissile material for every gram of fissile material. Lead, beryllium, graphite, and hydrogenous material enriched in deuterium may be present in the package but must not be included in determining the required mass for solid nonfissile material.
            (c)(1) Low concentrations of solid fissile material commingled with solid nonfissile material, provided that:
            (i) There is at least 2000 grams of solid nonfissile material for every gram of fissile material, and
            (ii) There is no more than 180 grams of fissile material distributed within 360 kg of contiguous nonfissile material.
            (2) Lead, beryllium, graphite, and hydrogenous material enriched in deuterium may be present in the package but must not be included in determining the required mass of solid nonfissile material.
            (d) Uranium enriched in uranium-235 to a maximum of 1 percent by weight, and with total plutonium and uranium-233 content of up to 1 percent of the mass of uranium-235, provided that the mass of any beryllium, graphite, and hydrogenous material enriched in deuterium constitutes less than 5 percent of the uranium mass, and that the fissile material is distributed homogeneously and does not form a lattice arrangement within the package.
            (e) Liquid solutions of uranyl nitrate enriched in uranium-235 to a maximum of 2 percent by mass, with a total plutonium and uranium-233 content not exceeding 0.002 percent of the mass of uranium, and with a minimum nitrogen to uranium atomic ratio (N/U) of 2. The material must be contained in at least a DOT Type A package.
            (f) Packages containing, individually, a total plutonium mass of not more than 1000 grams, of which not more than 20 percent by mass may consist of plutonium-239, plutonium-241, or any combination of these radionuclides.
            [69 FR 3786, Jan. 26, 2004, as amended at 80 FR 34012, June 12, 2015]

            § 71.16
            [Reserved]

          Subpart C—General Licenses

            Source:
            69 FR 3786, Jan. 26, 2004, unless otherwise noted.

            § 71.17
            General license: NRC-approved package.
            (a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package for which a license, certificate of compliance (CoC), or other approval has been issued by the NRC.
            (b) This general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.
            (c) Each licensee issued a general license under paragraph (a) of this section shall—
            (1) Maintain a copy of the Certificate of Compliance, or other approval of the package, and the drawings and other documents referenced in the approval relating to the use and maintenance of the packaging and to the actions to be taken before shipment;
            (2) Comply with the terms and conditions of the license, certificate, or other approval, as applicable, and the applicable requirements of subparts A, G, and H of this part; and
            (3) Submit in writing before the first use of the package to: ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards, using an appropriate method listed in § 71.1(a), the licensee's name and license number and the package identification number specified in the package approval.
            (d) This general license applies only when the package approval authorizes use of the package under this general license.
            (e) For a Type B or fissile material package, the design of which was approved by NRC before April 1, 1996, the general license is subject to the additional restrictions of § 71.19.
            [69 FR 3786, Jan. 26, 2004, as amended at 75 FR 73945, Nov. 30, 2010; 79 FR 75741, Dec. 19, 2014; 80 FR 34012, June 12, 2015; 84 FR 65645, Nov. 29, 2019]

            § 71.18
            [Reserved]

            § 71.19
            Previously approved package.
            (a) A Type B(U) package, a Type B(M) package, or a fissile material package, previously approved by the NRC but without the designation “-85” in the identification number of the NRC CoC, may be used under the general license of § 71.17 with the following additional conditions:
            (1) Fabrication of the package is satisfactorily completed by April 1, 1999, as demonstrated by application of its model number in accordance with § 71.85(c);
            (2) A package used for a shipment to a location outside the United States is subject to multilateral approval as defined in DOT regulations at 49 CFR 173.403; and
            (3) A serial number which uniquely identifies each packaging which conforms to the approved design is assigned to and legibly and durably marked on the outside of each packaging.
            (b) A Type B(U) package, a Type B(M) package, or a fissile material package previously approved by the NRC with the designation “-85” in the identification number of the NRC CoC, may be used under the general license of § 71.17 with the following additional conditions:
            (1) Fabrication of the package must be satisfactorily completed by December 31, 2006, as demonstrated by application of its model number in accordance with § 71.85(c); and
            (2) A package used for a shipment to a location outside the United States is subject to multilateral approval as defined in the DOT's regulations at 49 CFR 173.403.
            (c) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided—
            (1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified in §§ 71.71 and 71.73;
            (2) The modifications of a fissile material package are not significant, with respect to the prevention of criticality, when the package is subjected to the tests specified in §§ 71.71 and 71.73; and
            (3) The modifications to the package satisfy the requirements of this part.
            (d) NRC will revise the package identification number to designate previously approved package designs as B, BF, AF, B(U), B(M), B(U)F, B(M)F, B(U)-85, B(U)F-85, B(M)-85, B(M)F-85, or AF-85 as appropriate, and with the identification number suffix “-96” after receipt of an application demonstrating that the design meets the requirements of this part.
            [69 FR 3786, Jan. 26, 2004; 69 FR 6139, Feb. 10, 2004, as amended at 80 FR 34012, June 12, 2015]

            § 71.21
            General license: Use of foreign approved package.
            (a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package, the design of which has been approved in a foreign national competent authority certificate, that has been revalidated by the DOT as meeting the applicable requirements of 49 CFR 171.23.
            (b) Except as otherwise provided in this section, the general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the applicable provisions of subpart H of this part.
            (c) This general license applies only to shipments made to or from locations outside the United States.
            (d) Each licensee issued a general license under paragraph (a) of this section shall—
            (1) Maintain a copy of the applicable certificate, the revalidation, and the drawings and other documents referenced in the certificate, relating to the use and maintenance of the packaging and to the actions to be taken before shipment; and
            (2) Comply with the terms and conditions of the certificate and revalidation, and with the applicable requirements of subparts A, G, and H of this part.
            [69 FR 3786, Jan. 26, 2004, as amended at 80 FR 34012, June 12, 2015]

            § 71.22
            General license: Fissile material.
            (a) A general license is issued to any licensee of the Commission to transport fissile material, or to deliver fissile material to a carrier for transport, if the material is shipped in accordance with this section. The fissile material need not be contained in a package which meets the standards of subparts E and F of this part; however, the material must be contained in a Type A package. The Type A package must also meet the DOT requirements of 49 CFR 173.417(a).
            (b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.
            (c) The general license applies only when a package's contents:
            (1) Contain no more than a Type A quantity of radioactive material; and
            (2) Contain less than 500 total grams of beryllium, graphite, or hydrogenous material enriched in deuterium.
            (d) The general license applies only to packages containing fissile material that are labeled with a CSI which:
            (1) Has been determined in accordance with paragraph (e) of this section;
            (2) Has a value less than or equal to 10; and
            (3) For a shipment of multiple packages containing fissile material, the sum of the CSIs must be less than or equal to 50 (for shipment on a nonexclusive use conveyance) and less than or equal to 100 (for shipment on an exclusive use conveyance).
            (e)(1) The value for the CSI must be greater than or equal to the number calculated by the following equation:

              ER26JA04.012

            (2) The calculated CSI must be rounded up to the first decimal place;
            (3) The values of X, Y, and Z used in the CSI equation must be taken from Tables 71-1 or 71-2, as appropriate;
            (4) If Table 71-2 is used to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium must be assumed to be zero; and
            (5) Table 71-1 values for X, Y, and Z must be used to determine the CSI if:
            (i) Uranium-233 is present in the package;
            (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235;
            (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight percent enrichment; or
            (iv) Substances having a moderating effectiveness (i.e., an average hydrogen density greater than H 2O) (e.g., certain hydrocarbon oils or plastics) are present in any form, except as polyethylene used for packing or wrapping.

              Table 71-1—Mass Limits for General License Packages Containing Mixed Quantities of Fissile Material or Uranium-235 of Unknown Enrichment per § 71.22(e)

                Fissile material
                Fissile material mass mixed with moderating substances having an average hydrogen density less than or equal to H 2O (grams)
                Fissile material mass mixed with moderating substances having an average hydrogen density greater than H 2Oa (grams)

                  235 U (X)
                60
                38

                  233 U (Y)
                43
                27

                  239 Pu or 241 Pu (Z)
                37
                24

                a When mixtures of moderating substances are present, the lower mass limits shall be used if more than 15 percent of the moderating substance has an average hydrogen density greater than H2O.

              Table 71-2—Mass Limits for General License Packages Containing Uranium-235 of Known Enrichment per § 71.22(e)

                Uranium enrichment in weight percent of 235 U not exceeding
                Fissile material mass of 235 U (X) (grams)

                24
                60

                20
                63

                15
                67

                11
                72

                10
                76

                9.5
                78

                9
                81

                8.5
                82

                8
                85

                7.5
                88

                7
                90

                6.5
                93

                6
                97

                5.5
                102

                5
                108

                4.5
                114

                4
                120

                3.5
                132

                3
                150

                2.5
                180

                2
                246

                1.5
                408

                1.35
                480

                1
                1,020

                0.92
                1,800

            [69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004]

            § 71.23
            General license: Plutonium-beryllium special form material.
            (a) A general license is issued to any licensee of the Commission to transport fissile material in the form of plutonium-beryllium (Pu-Be) special form sealed sources, or to deliver Pu-Be sealed sources to a carrier for transport, if the material is shipped in accordance with this section. This material need not be contained in a package which meets the standards of subparts E and F of this part; however, the material must be contained in a Type A package. The Type A package must also meet the DOT requirements of 49 CFR 173.417(a).
            (b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.
            (c) The general license applies only when a package's contents:
            (1) Contain no more than a Type A quantity of radioactive material; and
            (2) Contain less than 1000 g of plutonium, provided that: plutonium-239, plutonium-241, or any combination of these radionuclides, constitutes less than 240 g of the total quantity of plutonium in the package.
            (d) The general license applies only to packages labeled with a CSI which:
            (1) Has been determined in accordance with paragraph (e) of this section;
            (2) Has a value less than or equal to 100; and
            (3) For a shipment of multiple packages containing Pu-Be sealed sources, the sum of the CSIs must be less than or equal to 50 (for shipment on a nonexclusive use conveyance) and less than or equal to 100 (for shipment on an exclusive use conveyance).
            (e)(1) The value for the CSI must be greater than or equal to the number calculated by the following equation:

              ER26JA04.013

            (2) The calculated CSI must be rounded up to the first decimal place.
            [69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004]

            §§ 71.24-71.25
            [Reserved]

          Subpart D—Application for Package Approval

            § 71.31
            Contents of application.

            (a) An application for an approval under this part must include, for each proposed packaging design, the following information:

            (1) A package description as required by § 71.33;
            (2) A package evaluation as required by § 71.35; and
            (3) A quality assurance program description, as required by § 71.37, or a reference to a previously approved quality assurance program.
            (b) Except as provided in § 71.19, an application for modification of a package design, whether for modification of the packaging or authorized contents, must include sufficient information to demonstrate that the proposed design satisfies the package standards in effect at the time the application is filed.
            (c) The applicant shall identify any established codes and standards proposed for use in package design, fabrication, assembly, testing, maintenance, and use. In the absence of any codes and standards, the applicant shall describe and justify the basis and rationale used to formulate the package quality assurance program.
            [60 FR 50264, Sept. 28, 1995, as amended at 80 FR 34012, June 12, 2015]

            § 71.33
            Package description.
            The application must include a description of the proposed package in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation of the package. The description must include—
            (a) With respect to the packaging—
            (1) Classification as Type B(U), Type B(M), or fissile material packaging;
            (2) Gross weight;
            (3) Model number;
            (4) Identification of the containment system;
            (5) Specific materials of construction, weights, dimensions, and fabrication methods of—
            (i) Receptacles;
            (ii) Materials specifically used as nonfissile neutron absorbers or moderators;
            (iii) Internal and external structures supporting or protecting receptacles;
            (iv) Valves, sampling ports, lifting devices, and tie-down devices; and
            (v) Structural and mechanical means for the transfer and dissipation of heat; and
            (6) Identification and volumes of any receptacles containing coolant.
            (b) With respect to the contents of the package—
            (1) Identification and maximum radioactivity of radioactive constituents;
            (2) Identification and maximum quantities of fissile constituents;
            (3) Chemical and physical form;
            (4) Extent of reflection, the amount and identity of nonfissile materials used as neutron absorbers or moderators, and the atomic ratio of moderator to fissile constituents;
            (5) Maximum normal operating pressure;
            (6) Maximum weight;
            (7) Maximum amount of decay heat; and
            (8) Identification and volumes of any coolants.

            § 71.35
            Package evaluation.
            The application must include the following:
            (a) A demonstration that the package satisfies the standards specified in subparts E and F of this part;
            (b) For a fissile material package, the allowable number of packages that may be transported in the same vehicle in accordance with § 71.59; and
            (c) For a fissile material shipment, any proposed special controls and precautions for transport, loading, unloading, and handling and any proposed special controls in case of an accident or delay.

            § 71.37
            Quality assurance.
            (a) The applicant shall describe the quality assurance program (see Subpart H of this part) for the design, fabrication, assembly, testing, maintenance, repair, modification, and use of the proposed package.
            (b) The applicant shall identify any specific provisions of the quality assurance program that are applicable to the particular package design under consideration, including a description of the leak testing procedures.

            § 71.38
            Renewal of a certificate of compliance.

            (a) Except as provided in paragraph (b) of this section, each Certificate of Compliance expires at the end of the day, in the month and year stated in the approval.

            (b) In any case in which a person, not less than 30 days before the expiration of an existing Certificate of Compliance issued pursuant to the part, has filed an application in proper form for renewal, the existing Certificate of Compliance for which the renewal application was filed shall not be deemed to have expired until final action on the application for renewal has been taken by the Commission.
            (c) In applying for renewal of an existing Certificate of Compliance, an applicant may be required to submit a consolidated application that is comprised of as few documents as possible. The consolidated application should incorporate all changes to its certificate, including changes that are incorporated by reference in the existing certificate.
            [80 FR 34012, June 12, 2015]

            § 71.39
            Requirement for additional information.
            The Commission may at any time require additional information in order to enable it to determine whether a license, certificate of compliance, or other approval should be granted, renewed, denied, modified, suspended, or revoked.

          Subpart E—Package Approval Standards

            § 71.41
            Demonstration of compliance.

            (a) The effects on a package of the tests specified in § 71.71 (“Normal conditions of transport”), and the tests specified in § 71.73 (“Hypothetical accident conditions”), and § 71.61 (“Special requirements for Type B packages containing more than 105 A2”), must be evaluated by subjecting a specimen or scale model to a specific test, or by another method of demonstration acceptable to the Commission, as appropriate for the particular feature being considered.
            (b) Taking into account the type of vehicle, the method of securing or attaching the package, and the controls to be exercised by the shipper, the Commission may permit the shipment to be evaluated together with the transporting vehicle.
            (c) Environmental and test conditions different from those specified in §§ 71.71 and 71.73 may be approved by the Commission if the controls proposed to be exercised by the shipper are demonstrated to be adequate to provide equivalent safety of the shipment.
            (d) Packages for which compliance with the other provisions of these regulations is impracticable shall not be transported except under special package authorization. Provided the applicant demonstrates that compliance with the other provisions of the regulations is impracticable and that the requisite standards of safety established by these regulations have been demonstrated through means alternative to the other provisions, a special package authorization may be approved for one-time shipments. The applicant shall demonstrate that the overall level of safety in transport for these shipments is at least equivalent to that which would be provided if all the applicable requirements had been met.
            [60 FR 50264, Sept. 28, 1995, as amended at 69 FR 3794, Jan. 26, 2004; 69 FR 58039, Sept. 29, 2004]

            § 71.43
            General standards for all packages.
            (a) The smallest overall dimension of a package may not be less than 10 cm (4 in).
            (b) The outside of a package must incorporate a feature, such as a seal, that is not readily breakable and that, while intact, would be evidence that the package has not been opened by unauthorized persons.
            (c) Each package must include a containment system securely closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package.

            (d) A package must be made of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the package contents, including possible reaction resulting from inleakage of water, to the maximum credible extent. Account must be taken of the behavior of materials under irradiation.

            (e) A package valve or other device, the failure of which would allow radioactive contents to escape, must be protected against unauthorized operation and, except for a pressure relief device, must be provided with an enclosure to retain any leakage.
            (f) A package must be designed, constructed, and prepared for shipment so that under the tests specified in § 71.71 (“Normal conditions of transport”) there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging.
            (g) A package must be designed, constructed, and prepared for transport so that in still air at 38 °C (100 °F) and in the shade, no accessible surface of a package would have a temperature exceeding 50 °C (122 °F) in a nonexclusive use shipment, or 85 °C (185 °F) in an exclusive use shipment.
            (h) A package may not incorporate a feature intended to allow continuous venting during transport.

            § 71.45
            Lifting and tie-down standards for all packages.
            (a) Any lifting attachment that is a structural part of a package must be designed with a minimum safety factor of three against yielding when used to lift the package in the intended manner, and it must be designed so that failure of any lifting device under excessive load would not impair the ability of the package to meet other requirements of this subpart. Any other structural part of the package that could be used to lift the package must be capable of being rendered inoperable for lifting the package during transport, or must be designed with strength equivalent to that required for lifting attachments.
            (b) Tie-down devices:
            (1) If there is a system of tie-down devices that is a structural part of the package, the system must be capable of withstanding, without generating stress in any material of the package in excess of its yield strength, a static force applied to the center of gravity of the package having a vertical component of 2 times the weight of the package with its contents, a horizontal component along the direction in which the vehicle travels of 10 times the weight of the package with its contents, and a horizontal component in the transverse direction of 5 times the weight of the package with its contents.
            (2) Any other structural part of the package that could be used to tie down the package must be capable of being rendered inoperable for tying down the package during transport, or must be designed with strength equivalent to that required for tie-down devices.
            (3) Each tie-down device that is a structural part of a package must be designed so that failure of the device under excessive load would not impair the ability of the package to meet other requirements of this part.

            § 71.47
            External radiation standards for all packages.
            (a) Except as provided in paragraph (b) of this section, each package of radioactive materials offered for transportation must be designed and prepared for shipment so that under conditions normally incident to transportation the radiation level does not exceed 2 mSv/h (200 mrem/h) at any point on the external surface of the package, and the transport index does not exceed 10.
            (b) A package that exceeds the radiation level limits specified in paragraph (a) of this section must be transported by exclusive use shipment only, and the radiation levels for such shipment must not exceed the following during transportation:
            (1) 2 mSv/h (200 mrem/h) on the external surface of the package, unless the following conditions are met, in which case the limit is 10 mSv/h (1000 mrem/h):
            (i) The shipment is made in a closed transport vehicle;
            (ii) The package is secured within the vehicle so that its position remains fixed during transportation; and
            (iii) There are no loading or unloading operations between the beginning and end of the transportation;

            (2) 2 mSv/h (200 mrem/h) at any point on the outer surface of the vehicle, including the top and underside of the vehicle; or in the case of a flat-bed style vehicle, at any point on the vertical planes projected from the outer edges of the vehicle, on the upper surface of the load or enclosure, if used, and on the lower external surface of the vehicle; and
            (3) 0.1 mSv/h (10 mrem/h) at any point 2 meters (80 in) from the outer lateral surfaces of the vehicle (excluding the top and underside of the vehicle); or in the case of a flat-bed style vehicle, at any point 2 meters (6.6 feet) from the vertical planes projected by the outer edges of the vehicle (excluding the top and underside of the vehicle); and
            (4) 0.02 mSv/h (2 mrem/h) in any normally occupied space, except that this provision does not apply to private carriers, if exposed personnel under their control wear radiation dosimetry devices in conformance with 10 CFR 20.1502.
            (c) For shipments made under the provisions of paragraph (b) of this section, the shipper shall provide specific written instructions to the carrier for maintenance of the exclusive use shipment controls. The instructions must be included with the shipping paper information.
            (d) The written instructions required for exclusive use shipments must be sufficient so that, when followed, they will cause the carrier to avoid actions that will unnecessarily delay delivery or unnecessarily result in increased radiation levels or radiation exposures to transport workers or members of the general public.

            § 71.51
            Additional requirements for Type B packages.
            (a) A Type B package, in addition to satisfying the requirements of §§ 71.41 through 71.47, must be designed, constructed, and prepared for shipment so that under the tests specified in:

            (1) Section 71.71 (“Normal conditions of transport”), there would be no loss or dispersal of radioactive contents—as demonstrated to a sensitivity of 10−6 A2 per hour, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging; and

            (2) Section 71.73 (“Hypothetical accident conditions”), there would be no escape of krypton-85 exceeding 10 A2 in 1 week, no escape of other radioactive material exceeding a total amount A2 in 1 week, and no external radiation dose rate exceeding 10 mSv/h (1 rem/h) at 1 m (40 in) from the external surface of the package.

            (b) Where mixtures of different radionuclides are present, the provisions of appendix A, paragraph IV of this part shall apply, except that for Krypton-85, an effective A2 value equal to 10 A2 may be used.
            (c) Compliance with the permitted activity release limits of paragraph (a) of this section may not depend on filters or on a mechanical cooling system.

            (d) For packages which contain radioactive contents with activity greater than 105 A2, the requirements of § 71.61 must be met.
            [60 FR 50264, Sept. 28, 1995, as amended at 69 FR 3794, Jan. 26, 2004; 69 FR 58039, Sept. 29, 2004]

            § 71.53
            [Reserved]

            § 71.55
            General requirements for fissile material packages.
            (a) A package used for the shipment of fissile material must be designed and constructed in accordance with §§ 71.41 through 71.47. When required by the total amount of radioactive material, a package used for the shipment of fissile material must also be designed and constructed in accordance with § 71.51.
            (b) Except as provided in paragraph (c) or (g) of this section, a package used for the shipment of fissile material must be so designed and constructed and its contents so limited that it would be subcritical if water were to leak into the containment system, or liquid contents were to leak out of the containment system so that, under the following conditions, maximum reactivity of the fissile material would be attained:
            (1) The most reactive credible configuration consistent with the chemical and physical form of the material;
            (2) Moderation by water to the most reactive credible extent; and

            (3) Close full reflection of the containment system by water on all sides, or such greater reflection of the containment system as may additionally be provided by the surrounding material of the packaging.

            (c) The Commission may approve exceptions to the requirements of paragraph (b) of this section if the package incorporates special design features that ensure that no single packaging error would permit leakage, and if appropriate measures are taken before each shipment to ensure that the containment system does not leak.
            (d) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in § 71.71 (“Normal conditions of transport”)—
            (1) The contents would be subcritical;
            (2) The geometric form of the package contents would not be substantially altered;
            (3) There would be no leakage of water into the containment system unless, in the evaluation of undamaged packages under § 71.59(a)(1), it has been assumed that moderation is present to such an extent as to cause maximum reactivity consistent with the chemical and physical form of the material; and
            (4) There will be no substantial reduction in the effectiveness of the packaging, including:
            (i) No more than 5 percent reduction in the total effective volume of the packaging on which nuclear safety is assessed;
            (ii) No more than 5 percent reduction in the effective spacing between the fissile contents and the outer surface of the packaging; and
            (iii) No occurrence of an aperture in the outer surface of the packaging large enough to permit the entry of a 10 cm (4 in) cube.
            (e) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in § 71.73 (“Hypothetical accident conditions”), the package would be subcritical. For this determination, it must be assumed that:
            (1) The fissile material is in the most reactive credible configuration consistent with the damaged condition of the package and the chemical and physical form of the contents;
            (2) Water moderation occurs to the most reactive credible extent consistent with the damaged condition of the package and the chemical and physical form of the contents; and
            (3) There is full reflection by water on all sides, as close as is consistent with the damaged condition of the package.
            (f) For fissile material package designs to be transported by air:
            (1) The package must be designed and constructed, and its contents limited so that it would be subcritical, assuming reflection by 20 cm (7.9 in) of water but no water inleakage, when subjected to sequential application of:
            (i) The free drop test in § 71.73(c)(1);
            (ii) The crush test in § 71.73(c)(2);
            (iii) A puncture test, for packages of 250 kg or more, consisting of a free drop of the specimen through a distance of 3 m (120 in) in a position for which maximum damage is expected at the conclusion of the test sequence, onto the upper end of a solid, vertical, cylindrical, mild steel probe mounted on an essentially unyielding, horizontal surface. The probe must be 20 cm (7.9 in) in diameter, with the striking end forming the frustum of a right circular cone with the dimensions of 30 cm height, 2.5 cm top diameter, and a top edge rounded to a radius of not more than 6 mm (0.25 in). For packages less than 250 kg, the puncture test must be the same, except that a 250 kg probe must be dropped onto the specimen which must be placed on the surface; and
            (iv) The thermal test in § 71.73(c)(4), except that the duration of the test must be 60 minutes.
            (2) The package must be designed and constructed, and its contents limited, so that it would be subcritical, assuming reflection by 20 cm (7.9 in) of water but no water inleakage, when subjected to an impact on an unyielding surface at a velocity of 90 m/s normal to the surface, at such orientation so as to result in maximum damage. A separate, undamaged specimen can be used for this evaluation.

            (3) Allowance may not be made for the special design features in paragraph (c) of this section, unless water leakage into or out of void spaces is prevented following application of the tests in paragraphs (f)(1) and (f)(2) of this section, and subsequent application of the immersion test in § 71.73(c)(5).
            (g) Packages containing uranium hexafluoride only are excepted from the requirements of paragraph (b) of this section provided that:
            (1) Following the tests specified in § 71.73 (“Hypothetical accident conditions”), there is no physical contact between the valve body and any other component of the packaging, other than at its original point of attachment, and the valve remains leak tight;
            (2) There is an adequate quality control in the manufacture, maintenance, and repair of packagings;
            (3) Each package is tested to demonstrate closure before each shipment; and
            (4) The uranium is enriched to not more than 5 weight percent uranium-235.
            [60 FR 50264, Sept. 28, 1995; 61 FR 28724, June 6, 1996, as amended at 69 FR 3794, Jan. 26, 2004]

            § 71.57
            [Reserved]

            § 71.59
            Standards for arrays of fissile material packages.
            (a) A fissile material package must be controlled by either the shipper or the carrier during transport to assure that an array of such packages remains subcritical. To enable this control, the designer of a fissile material package shall derive a number “N” based on all the following conditions being satisfied, assuming packages are stacked together in any arrangement and with close full reflection on all sides of the stack by water:
            (1) Five times “N” undamaged packages with nothing between the packages would be subcritical;
            (2) Two times “N” damaged packages, if each package were subjected to the tests specified in § 71.73 (“Hypothetical accident conditions”) would be subcritical with optimum interspersed hydrogenous moderation; and
            (3) The value of “N” cannot be less than 0.5.
            (b) The CSI must be determined by dividing the number 50 by the value of “N” derived using the procedures specified in paragraph (a) of this section. The value of the CSI may be zero provided that an unlimited number of packages are subcritical, such that the value of “N” is effectively equal to infinity under the procedures specified in paragraph (a) of this section. Any CSI greater than zero must be rounded up to the first decimal place.
            (c) For a fissile material package which is assigned a CSI value—
            (1) Less than or equal to 50, that package may be shipped by a carrier in a nonexclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 50.
            (2) Less than or equal to 50, that package may be shipped by a carrier in an exclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 100.
            (3) Greater than 50, that package must be shipped by a carrier in an exclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 100.
            [60 FR 50264, Sept. 28, 1995, as amended at 69 FR 3795, Jan. 26, 2004]

            § 71.61
            Special requirements for Type B packages containing more than 10 5 A2.
            A Type B package containing more than 105A2 must be designed so that its undamaged containment system can withstand an external water pressure of 2 MPa (290 psi) for a period of not less than 1 hour without collapse, buckling, or inleakage of water.
            [69 FR 3795, Jan. 26, 2004]

            § 71.63
            Special requirement for plutonium shipments.
            Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than 0.74 TBq (20 Ci) of plutonium.
            [69 FR 3795, Jan. 26, 2004]

            § 71.64
            Special requirements for plutonium air shipments.

            (a) A package for the shipment of plutonium by air subject to § 71.88(a)(4), in addition to satisfying the requirements of §§ 71.41 through 71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in—

            (1) Section 71.74 (“Accident conditions for air transport of plutonium”)—

            (i) The containment vessel would not be ruptured in its post-tested condition, and the package must provide a sufficient degree of containment to restrict accumulated loss of plutonium contents to not more than an A2 quantity in a period of 1 week;
            (ii) The external radiation level would not exceed 10 mSv/h (1 rem/h) at a distance of 1 m (40 in) from the surface of the package in its post-tested condition in air; and
            (iii) A single package and an array of packages are demonstrated to be subcritical in accordance with this part, except that the damaged condition of the package must be considered to be that which results from the plutonium accident tests in § 71.74, rather than the hypothetical accident tests in § 71.73; and
            (2) Section 71.74(c), there would be no detectable leakage of water into the containment vessel of the package.
            (b) With respect to the package requirements of paragraph (a), there must be a demonstration or analytical assessment showing that—
            (1) The results of the physical testing for package qualification would not be adversely affected to a significant extent by—
            (i) The presence, during the tests, of the actual contents that will be transported in the package; and
            (ii) Ambient water temperatures ranging from 0.6 °C ( + 33 °F) to 38 °C ( + 100 °F) for those qualification tests involving water, and ambient atmospheric temperatures ranging from −40 °C (−40 °F) to + 54 °C ( + 130°F) for the other qualification tests.
            (2) The ability of the package to meet the acceptance standards prescribed for the accident condition sequential tests would not be adversely affected if one or more tests in the sequence were deleted.

            § 71.65
            Additional requirements.
            The Commission may, by rule, regulation, or order, impose requirements on any licensee, in addition to those established in this part, as it deems necessary or appropriate to protect public health or to minimize danger to life or property.

          Subpart F—Package, Special Form, and LSA-III Tests 2

              2 The package standards related to the tests in this subpart are contained in subpart E of this part.

            § 71.70
            Incorporations by reference.

            (a) The materials listed in this section are incorporated by reference in the corresponding sections noted and made a part of the regulations in part 71. These incorporations by reference were approved by the Director of the Federal Register under 5 U.S.C. 552(a) and 1 CFR part 51. These materials are incorporated as they exist on the date of the approval. A notice of any changes made to the material incorporated by reference will be published in the Federal Register, and the material must be available to the public. The materials can be examined, by appointment, at the NRC's Technical Library, which is located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-7000; email: Library.Resource@nrc.gov. The materials are also available from the sources listed below. All approved material is available for inspection at the National Archives and Records Administration (NARA). For information on the availability of this material at NARA, call 1-202-741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.

            (b) International Organization for Standardization, ISO Central Secretariat, Chemin de Blandonnet 8 CP 401, 1214 Vernier, Geneva, Switzerland; email: central@iso.org; phone: + 41 22 749 01 11; Web site: http://www.iso.org.

            (1) ISO 9978:1992(E), “Radiation protection—Sealed radioactive sources—Leakage test methods,” First Edition (February 15, 1992), incorporation by reference approved for § 71.75(a), is available for purchase from the American National Standards Institute, 25 West 43rd Street, 4th Floor, New York, NY 10036, 212-642-4900, http://www.ansi.org, or info@ansi.org.

            (2) ISO 2919:1999(E), “Radiation protection—Sealed radioactive sources—General requirements and classification,” Second Edition (February 15, 1999), incorporation by reference approved for § 71.75(d), is available on http://www.amazon.com.

            [80 FR 34013, June 12, 2015, as amended at 80 FR 48684, Aug. 14, 2015]

            § 71.71
            Normal conditions of transport.
            (a) Evaluation. Evaluation of each package design under normal conditions of transport must include a determination of the effect on that design of the conditions and tests specified in this section. Separate specimens may be used for the free drop test, the compression test, and the penetration test, if each specimen is subjected to the water spray test before being subjected to any of the other tests.
            (b) Initial conditions. With respect to the initial conditions for the tests in this section, the demonstration of compliance with the requirements of this part must be based on the ambient temperature preceding and following the tests remaining constant at that value between −29 °C (−20 °F) and + 38 °C ( + 100 °F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be considered to be the maximum normal operating pressure, unless a lower internal pressure consistent with the ambient temperature considered to precede and follow the tests is more unfavorable.
            (c) Conditions and tests—(1) Heat. An ambient temperature of 38 °C (100 °F) in still air, and insolation according to the following table:

              Insolation Data

                Form and location of surface
                Total insolation for a 12-hour period(g cal/cm2)

                Flat surfaces transported horizontally:

                Base
                None

                Other surfaces
                800

                Flat surfaces not transported horizontally
                200

                Curved surfaces
                400

            (2) Cold. An ambient temperature of −40 °C (−40 °F) in still air and shade.
            (3) Reduced external pressure. An external pressure of 25 kPa (3.5 lbf/in2) absolute.
            (4) Increased external pressure. An external pressure of 140 kPa (20 lbf/in2) absolute.
            (5) Vibration. Vibration normally incident to transport.
            (6) Water spray. A water spray that simulates exposure to rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour.
            (7) Free drop. Between 1.5 and 2.5 hours after the conclusion of the water spray test, a free drop through the distance specified below onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.

              Criteria for Free Drop Test (Weight/Distance)

                Package weight
                Kilograms
                (Pounds)
                Free drop distance
                Meters
                (Feet)

                Less than 5,000
                (Less than 11,000)
                1.2
                (4)

                5,000 to 10,000
                (11,000 to 22,000)
                0.9
                (3)

                10,000 to 15,000
                (22,000 to 33,100)
                0.6
                (2)

                More than 15,000
                (More than 33,100)
                0.3
                (1)

            (8) Corner drop. A free drop onto each corner of the package in succession, or in the case of a cylindrical package onto each quarter of each rim, from a height of 0.3 m (1 ft) onto a flat, essentially unyielding, horizontal surface. This test applies only to fiberboard, wood, or fissile material rectangular packages not exceeding 50 kg (110 lbs) and fiberboard, wood, or fissile material cylindrical packages not exceeding 100 kg (220 lbs).
            (9) Compression. For packages weighing up to 5000 kg (11,000 lbs), the package must be subjected, for a period of 24 hours, to a compressive load applied uniformly to the top and bottom of the package in the position in which the package would normally be transported. The compressive load must be the greater of the following:
            (i) The equivalent of 5 times the weight of the package; or
            (ii) The equivalent of 13 kPa (2 lbf/in2) multiplied by the vertically projected area of the package.
            (10) Penetration. Impact of the hemispherical end of a vertical steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13 lbs) mass, dropped from a height of 1 m (40 in) onto the exposed surface of the package that is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface.
            [60 FR 50264, Sept. 28, 1995, as amended at 81 FR 86910, Dec. 2, 2016]

            § 71.73
            Hypothetical accident conditions.
            (a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the tests specified in this section, in the order indicated, to determine their cumulative effect on a package or array of packages. An undamaged specimen may be used for the water immersion tests specified in paragraph (c)(6) of this section.
            (b) Test conditions. With respect to the initial conditions for the tests, except for the water immersion tests, to demonstrate compliance with the requirements of this part during testing, the ambient air temperature before and after the tests must remain constant at that value between −29 °C (−20 °F) and + 38 °C ( + 100 °F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be the maximum normal operating pressure, unless a lower internal pressure, consistent with the ambient temperature assumed to precede and follow the tests, is more unfavorable.
            (c) Tests. Tests for hypothetical accident conditions must be conducted as follows:
            (1) Free drop. A free drop of the specimen through a distance of 9 m (30 ft) onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.
            (2) Crush. Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 500-kg (1100-lb) mass from 9 m (30 ft) onto the specimen. The mass must consist of a solid mild steel plate 1 m (40 in) by 1 m (40 in) and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 500 kg (1100 lb), an overall density not greater than 1000 kg/m 3 (62.4 lb/ft 3) based on external dimension, and radioactive contents greater than 1000 A2 not as special form radioactive material. For packages containing fissile material, the radioactive contents greater than 1000 A2 criterion does not apply.
            (3) Puncture. A free drop of the specimen through a distance of 1 m (40 in) in a position for which maximum damage is expected, onto the upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizontal surface. The bar must be 15 cm (6 in) in diameter, with the top horizontal and its edge rounded to a radius of not more than 6 mm (0.25 in), and of a length as to cause maximum damage to the package, but not less than 20 cm (8 in) long. The long axis of the bar must be vertical.
            (4) Thermal. Exposure of the specimen fully engulfed, except for a simple support system, in a hydrocarbon fuel/air fire of sufficient extent, and in sufficiently quiescent ambient conditions, to provide an average emissivity coefficient of at least 0.9, with an average flame temperature of at least 800 °C (1475 °F) for a period of 30 minutes, or any other thermal test that provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 800 °C. The fuel source must extend horizontally at least 1 m (40 in), but may not extend more than 3 m (10 ft), beyond any external surface of the specimen, and the specimen must be positioned 1 m (40 in) above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the convective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified. Artificial cooling may not be applied after cessation of external heat input, and any combustion of materials of construction, must be allowed to proceed until it terminates naturally.
            (5) Immersion—fissile material. For fissile material subject to § 71.55, in those cases where water inleakage has not been assumed for criticality analysis, immersion under a head of water of at least 0.9 m (3 ft) in the attitude for which maximum leakage is expected.

            (6) Immersion—all packages. A separate, undamaged specimen must be subjected to water pressure equivalent to immersion under a head of water of at least 15 m (50 ft). For test purposes, an external pressure of water of 150 kPa (21.7 lbf/in2) gauge is considered to meet these conditions.
            [60 FR 50264, Sept. 28, 1995, as amended at 69 FR 3795, Jan. 26, 2004]

            § 71.74
            Accident conditions for air transport of plutonium.
            (a) Test conditions—Sequence of tests. A package must be physically tested to the following conditions in the order indicated to determine their cumulative effect.
            (1) Impact at a velocity of not less than 129 m/sec (422 ft/sec) at a right angle onto a flat, essentially unyielding, horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage at the conclusion of the test sequence.
            (2) A static compressive load of 31,800 kg (70,000 lbs) applied in the orientation expected to result in maximum damage at the conclusion of the test sequence. The force on the package must be developed between a flat steel surface and a 5 cm (2 in) wide, straight, solid, steel bar. The length of the bar must be at least as long as the diameter of the package, and the longitudinal axis of the bar must be parallel to the plane of the flat surface. The load must be applied to the bar in a manner that prevents any members or devices used to support the bar from contacting the package.
            (3) Packages weighing less than 227 kg (500 lbs) must be placed on a flat, essentially unyielding, horizontal surface, and subjected to a weight of 227 kg (500 lbs) falling from a height of 3 m (10 ft) and striking in the position expected to result in maximum damage at the conclusion of the test sequence. The end of the weight contacting the package must be a solid probe made of mild steel. The probe must be the shape of the frustum of a right circular cone, 30 cm (12 in) long, 20 cm (8 in) in diameter at the base, and 2.5 cm (1 in) in diameter at the end. The longitudinal axis of the probe must be perpendicular to the horizontal surface. For packages weighing 227 kg (500 lbs) or more, the base of the probe must be placed on a flat, essentially unyielding horizontal surface, and the package dropped from a height of 3 m (10 ft) onto the probe, striking in the position expected to result in maximum damage at the conclusion of the test sequence.
            (4) The package must be firmly restrained and supported such that its longitudinal axis is inclined approximately 45° to the horizontal. The area of the package that made first contact with the impact surface in paragraph (a)(1) of this section must be in the lowermost position. The package must be struck at approximately the center of its vertical projection by the end of a structural steel angle section falling from a height of at least 46 m (150 ft). The angle section must be at least 1.8 m (6 ft) in length with equal legs at least 13 cm (5 in) long and 1.3 cm (0.5 in) thick. The angle section must be guided in such a way as to fall end-on, without tumbling. The package must be rotated approximately 90° about its longitudinal axis and struck by the steel angle section falling as before.
            (5) The package must be exposed to luminous flames from a pool fire of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes. The luminous flames must extend an average of at least 0.9 m (3 ft) and no more than 3 m (10 ft) beyond the package in all horizontal directions. The position and orientation of the package in relation to the fuel must be that which is expected to result in maximum damage at the conclusion of the test sequence. An alternate method of thermal testing may be substituted for this fire test, provided that the alternate test is not of shorter duration and would not result in a lower heating rate to the package. At the conclusion of the thermal test, the package must be allowed to cool naturally or must be cooled by water sprinkling, whichever is expected to result in maximum damage at the conclusion of the test sequence.
            (6) Immersion under at least 0.9 m (3 ft) of water.
            (b) Individual free-fall impact test. (1) An undamaged package must be physically subjected to an impact at a velocity not less than the calculated terminal free-fall velocity, at mean sea level, at a right angle onto a flat, essentially unyielding, horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage.
            (2) This test is not required if the calculated terminal free-fall velocity of the package is less than 129 m/sec (422 ft/sec), or if a velocity not less than either 129 m/sec (422 ft/sec) or the calculated terminal free-fall velocity of the package is used in the sequential test of paragraph (a)(1) of this section.
            (c) Individual deep submersion test. An undamaged package must be physically submerged and physically subjected to an external water pressure of at least 4 MPa (600 lbs/in 2).

            § 71.75
            Qualification of special form radioactive material.
            (a) Special form radioactive materials must meet the test requirements of paragraph (b) of this section. Each solid radioactive material or capsule specimen to be tested must be manufactured or fabricated so that it is representative of the actual solid material or capsule that will be transported, with the proposed radioactive content duplicated as closely as practicable. Any differences between the material to be transported and the test material, such as the use of non-radioactive contents, must be taken into account in determining whether the test requirements have been met. In addition:
            (1) A different specimen may be used for each of the tests;
            (2) The specimen may not break or shatter when subjected to the impact, percussion, or bending tests;
            (3) The specimen may not melt or disperse when subjected to the heat test;

            (4) After each test, leaktightness or indispersibility of the specimen must be determined by a method no less sensitive than the leaching assessment procedure prescribed in paragraph (c) of this section. For a capsule resistant to corrosion by water, and which has an internal void volume greater than 0.1 milliliter, an alternative to the leaching assessment is a demonstration of leaktightness of × 10−4 torr-liter/s (1.3 × × 10−4 atm-cm3/s) based on air at 25 °C (77 °F) and one atmosphere differential pressure for solid radioactive content, or × 10−6 torr-liter/s (1.3 × × 10−6 atm−cm3/s) for liquid or gaseous radioactive content; and
            (5) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to the leaktightness procedure specified in this section, provided it is alternatively subjected to any of the tests prescribed in ISO 9978:1992(E), “Radiation protection—Sealed radioactive sources—Leakage test methods” (incorporated by reference, see § 71.70).
            (b) Test methods—(1) Impact Test. The specimen must fall onto the target from a height of 9 m (30 ft) or greater in the orientation expected to result in maximum damage. The target must be a flat, horizontal surface of such mass and rigidity that any increase in its resistance to displacement or deformation, on impact by the specimen, would not significantly increase the damage to the specimen.
            (2) Percussion Test. (i) The specimen must be placed on a sheet of lead that is supported by a smooth solid surface, and struck by the flat face of a steel billet so as to produce an impact equivalent to that resulting from a free drop of 1.4 kg (3 lbs) through 1 m (40 in);
            (ii) The flat face of the billet must be 25 millimeters (mm) (1 inch) in diameter with the edge rounded off to a radius of 3 mm ±0.3 mm (0.12 in ±0.012 in);
            (iii) The lead must be hardness number 3.5 to 4.5 on the Vickers scale and not more than 25 mm (1 inch) thick, and must cover an area greater than that covered by the specimen;
            (iv) A fresh surface of lead must be used for each impact; and
            (v) The billet must strike the specimen so as to cause maximum damage.
            (3) Bending test. (i) This test applies only to long, slender sources with a length of 10 cm (4 inches) or greater and a length to width ratio of 10 or greater;
            (ii) The specimen must be rigidly clamped in a horizontal position so that one half of its length protrudes from the face of the clamp;

            (iii) The orientation of the specimen must be such that the specimen will suffer maximum damage when its free end is struck by the flat face of a steel billet;

            (iv) The billet must strike the specimen so as to produce an impact equivalent to that resulting from a free vertical drop of 1.4 kg (3 lbs) through 1 m (40 in); and
            (v) The flat face of the billet must be 25 mm (1 inch) in diameter with the edges rounded off to a radius of 3 mm ±0.3 mm (.12 in ±0.012 in).
            (4) Heat test. The specimen must be heated in air to a temperature of not less than 800 °C (1475 °F), held at that temperature for a period of 10 minutes, and then allowed to cool.
            (c) Leaching assessment methods. (1) For indispersible solid material—
            (i) The specimen must be immersed for 7 days in water at ambient temperature. The water must have a pH of 6-8 and a maximum conductivity of 10 micromho per centimeter at 20° (68 °F);
            (ii) The water with specimen must then be heated to a temperature of 50 °C ±5 °C (122 °F ±9 °F) and maintained at this temperature for 4 hours.
            (iii) The activity of the water must then be determined;
            (iv) The specimen must then be stored for at least 7 days in still air of relative humidity not less than 90 percent at 30 °C (86 °F);
            (v) The specimen must then be immersed in water under the same conditions as in paragraph (c)(1)(i) of this section, and the water with specimen must be heated to 50 °C ±5 °C (122 °F ±9 °F) and maintained at that temperature for 4 hours;
            (vi) The activity of the water must then be determined. The sum of the activities determined here and in paragraph (c)(1)(iii) of this section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie (µCi)).
            (2) For encapsulated material—
            (i) The specimen must be immersed in water at ambient temperature. The water must have a pH of 6-8 and a maximum conductivity of 10 micromho per centimeter;
            (ii) The water and specimen must be heated to a temperature of 50 °C ±5 °C (122 °F ±9 °F) and maintained at this temperature for 4 hours;
            (iii) The activity of the water must then be determined;
            (iv) The specimen must then be stored for at least 7 days in still air at a temperature of 30 °C (86 °F) or greater;
            (v) The process in paragraph (c)(2)(i), (ii), and (iii) of this section must be repeated; and
            (vi) The activity of the water must then be determined. The sum of the activities determined here and in paragraph (c)(2)(iii) of this section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie (µCi)).
            (d) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to—
            (1) The impact test and the percussion test of this section, provided that the specimen is:
            (i) Less than 200 grams and alternatively subjected to the Class 4 impact test prescribed in ISO 2919:1999(E), “Radiation protection—Sealed radioactive sources—General requirements and classification” (incorporated by reference, see § 71.70); or
            (ii) Less than 500 grams and alternatively subjected to the Class 5 impact test prescribed in ISO 2919:1999(E), “Radioactive protection—Sealed radioactive sources—General requirements and classification” (incorporated by reference, see § 71.70); and
            (2) The heat test of this section, provided the specimen is alternatively subjected to the Class 6 temperature test specified in ISO 2919:1999(E), “Radioactive protection—Sealed radioactive sources—General requirements and classification” (incorporated by reference, see § 71.70).
            [60 FR 50264, Sept. 28, 1995, as amended at 80 FR 34013, June 12, 2015]

            § 71.77
            Qualification of LSA-III Material.
            (a) LSA-III material must meet the test requirements of paragraph (b) of this section. Any differences between the specimen to be tested and the material to be transported must be taken into account in determining whether the test requirements have been met.
            (b) Leaching test. (1) The specimen, representing no less than the entire contents of the package, must be immersed for 7 days in water at ambient temperature;

            (2) The volume of water to be used in the test must be sufficient to ensure that at the end of the test period the free volume of the unabsorbed and unreacted water remaining will be at least 10% of the volume of the specimen itself;
            (3) The water must have an initial pH of 6-8 and a maximum conductivity 10 micromho/cm at 20 °C (68 °F); and

            (4) The total activity of the free volume of water must be measured following the 7 day immersion test and must not exceed 0.1 A2.

          Subpart G—Operating Controls and Procedures

            § 71.81
            Applicability of operating controls and procedures.
            A licensee subject to this part, who, under a general or specific license, transports licensed material or delivers licensed material to a carrier for transport, shall comply with the requirements of this subpart G, with the quality assurance requirements of subpart H of this part, and with the general provisions of subpart A of this part.

            § 71.83
            Assumptions as to unknown properties.
            When the isotopic abundance, mass, concentration, degree of irradiation, degree of moderation, or other pertinent property of fissile material in any package is not known, the licensee shall package the fissile material as if the unknown properties have credible values that will cause the maximum neutron multiplication.

            § 71.85
            Preliminary determinations.
            Before the first use of any packaging for the shipment of licensed material—
            (a) The certificate holder shall ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could significantly reduce the effectiveness of the packaging;
            (b) Where the maximum normal operating pressure will exceed 35 kPa (5 lbf/in2) gauge, the certificate holder shall test the containment system at an internal pressure at least 50 percent higher than the maximum normal operating pressure, to verify the capability of that system to maintain its structural integrity at that pressure;
            (c) The certificate holder shall conspicuously and durably mark the packaging with its model number, serial number, gross weight, and a package identification number assigned by the NRC. Before applying the model number, the certificate holder shall determine that the packaging has been fabricated in accordance with the design approved by the Commission; and
            (d) The licensee shall ascertain that the determinations in paragraphs (a) through (c) of this section have been made.
            [60 FR 50264, Sept. 28, 1995, as amended at 80 FR 34013, June 12, 2015]

            § 71.87
            Routine determinations.
            Before each shipment of licensed material, the licensee shall ensure that the package with its contents satisfies the applicable requirements of this part and of the license. The licensee shall determine that—
            (a) The package is proper for the contents to be shipped;
            (b) The package is in unimpaired physical condition except for superficial defects such as marks or dents;
            (c) Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects;
            (d) Any system for containing liquid is adequately sealed and has adequate space or other specified provision for expansion of the liquid;
            (e) Any pressure relief device is operable and set in accordance with written procedures;
            (f) The package has been loaded and closed in accordance with written procedures;
            (g) For fissile material, any moderator or neutron absorber, if required, is present and in proper condition;
            (h) Any structural part of the package that could be used to lift or tie down the package during transport is rendered inoperable for that purpose, unless it satisfies the design requirements of § 71.45;
            (i) The level of non-fixed (removable) radioactive contamination on the external surfaces of each package offered for shipment is as low as reasonably achievable, and within the limits specified in DOT regulations in 49 CFR 173.443;

            (j) External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in § 71.47 at any time during transportation; and
            (k) Accessible package surface temperatures will not exceed the limits specified in § 71.43(g) at any time during transportation.

            § 71.88
            Air transport of plutonium.
            (a) Notwithstanding the provisions of any general licenses and notwithstanding any exemptions stated directly in this part or included indirectly by citation of 49 CFR chapter I, as may be applicable, the licensee shall assure that plutonium in any form, whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for air transport unless:
            (1) The plutonium is contained in a medical device designed for individual human application; or
            (2) The plutonium is contained in a material in which the specific activity is less than or equal to the activity concentration values for plutonium specified in Appendix A, Table A-2, of this part, and in which the radioactivity is essentially uniformly distributed; or

            (3) The plutonium is shipped in a single package containing no more than an A2 quantity of plutonium in any isotope or form, and is shipped in accordance with § 71.5; or
            (4) The plutonium is shipped in a package specifically authorized for the shipment of plutonium by air in the Certificate of Compliance for that package issued by the Commission.
            (b) Nothing in paragraph (a) of this section is to be interpreted as removing or diminishing the requirements of § 73.24 of this chapter.
            (c) For a shipment of plutonium by air which is subject to paragraph (a)(4) of this section, the licensee shall, through special arrangement with the carrier, require compliance with 49 CFR 175.704, U.S. Department of Transportation regulations applicable to the air transport of plutonium.
            [60 FR 50264, Sept. 28, 1995, as amended at 69 FR 3795, Jan. 26, 2004]

            § 71.89
            Opening instructions.
            Before delivery of a package to a carrier for transport, the licensee shall ensure that any special instructions needed to safely open the package have been sent to, or otherwise made available to, the consignee for the consignee's use in accordance with 10 CFR 20.1906(e).

            § 71.91
            Records.
            (a) Each licensee shall maintain, for a period of 3 years after shipment, a record of each shipment of licensed material not exempt under § 71.14, showing where applicable—
            (1) Identification of the packaging by model number and serial number;
            (2) Verification that there are no significant defects in the packaging, as shipped;
            (3) Volume and identification of coolant;
            (4) Type and quantity of licensed material in each package, and the total quantity of each shipment;
            (5) For each item of irradiated fissile material—
            (i) Identification by model number and serial number;
            (ii) Irradiation and decay history to the extent appropriate to demonstrate that its nuclear and thermal characteristics comply with license conditions; and
            (iii) Any abnormal or unusual condition relevant to radiation safety;
            (6) Date of the shipment;
            (7) For fissile packages and for Type B packages, any special controls exercised;
            (8) Name and address of the transferee;
            (9) Address to which the shipment was made; and
            (10) Results of the determinations required by § 71.87 and by the conditions of the package approval.
            (b) Each certificate holder shall maintain, for a period of 3 years after the life of the packaging to which they apply, records identifying the packaging by model number, serial number, and date of manufacture.

            (c) The licensee, certificate holder, and an applicant for a CoC, shall make available to the Commission for inspection, upon reasonable notice, all records required by this part. Records are only valid if stamped, initialed, or signed and dated by authorized personnel, or otherwise authenticated.

            (d) The licensee, certificate holder, and an applicant for a CoC shall maintain sufficient written records to furnish evidence of the quality of packaging. The records to be maintained include results of the determinations required by § 71.85; design, fabrication, and assembly records; results of reviews, inspections, tests, and audits; results of monitoring work performance and materials analyses; and results of maintenance, modification, and repair activities. Inspection, test, and audit records must identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted. These records must be retained for 3 years after the life of the packaging to which they apply.
            [60 FR 50264, Sept. 28, 1995, as amended at 69 FR 3795, Jan. 26, 2004; 80 FR 34013, June 12, 2015]

            § 71.93
            Inspection and tests.
            (a) The licensee, certificate holder, and applicant for a CoC shall permit the Commission, at all reasonable times, to inspect the licensed material, packaging, premises, and facilities in which the licensed material or packaging is used, provided, constructed, fabricated, tested, stored, or shipped.
            (b) The licensee, certificate holder, and applicant for a CoC shall perform, and permit the Commission to perform, any tests the Commission deems necessary or appropriate for the administration of the regulations in this chapter.
            (c) The certificate holder and applicant for a CoC shall notify the NRC, in accordance with § 71.1, 45 days in advance of starting fabrication of the first packaging under a CoC. This paragraph applies to any packaging used for the shipment of licensed material which has either—
            (1) A decay heat load in excess of 5 kW; or
            (2) A maximum normal operating pressure in excess of 103 kPa (15 lbf/in 2) gauge.
            [69 FR 3796, Jan. 26, 2004]

            § 71.95
            Reports.
            (a) The licensee, after requesting the certificate holder's input, shall submit a written report to the Commission of—
            (1) Instances in which there is a significant reduction in the effectiveness of any NRC-approved Type B or Type AF packaging during use; or
            (2) Details of any defects with safety significance in any NRC-approved Type B or fissile material packaging, after first use.
            (3) Instances in which the conditions of approval in the Certificate of Compliance were not observed in making a shipment.
            (b) The licensee shall submit a written report to the Commission of instances in which the conditions in the certificate of compliance were not followed during a shipment.
            (c) Each licensee shall submit, in accordance with § 71.1, a written report required by paragraph (a) or (b) of this section within 60 days of the event or discovery of the event. The licensee shall also provide a copy of each report submitted to the NRC to the applicable certificate holder. Written reports prepared under other regulations may be submitted to fulfill this requirement if the reports contain all the necessary information, and the appropriate distribution is made. Using an appropriate method listed in § 71.1(a), the licensee shall report to: ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards. These written reports must include the following:
            (1) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed to the event and significant corrective action taken or planned to prevent recurrence.
            (2) A clear, specific, narrative description of the event that occurred so that knowledgeable readers conversant with the requirements of part 71, but not familiar with the design of the packaging, can understand the complete event. The narrative description must include the following specific information as appropriate for the particular event.

            (i) Status of components or systems that were inoperable at the start of the event and that contributed to the event;
            (ii) Dates and approximate times of occurrences;
            (iii) The cause of each component or system failure or personnel error, if known;
            (iv) The failure mode, mechanism, and effect of each failed component, if known;
            (v) A list of systems or secondary functions that were also affected for failures of components with multiple functions;
            (vi) The method of discovery of each component or system failure or procedural error;
            (vii) For each human performance-related root cause, a discussion of the cause(s) and circumstances;
            (viii) The manufacturer and model number (or other identification) of each component that failed during the event; and
            (ix) For events occurring during use of a packaging, the quantities and chemical and physical form(s) of the package contents.
            (3) An assessment of the safety consequences and implications of the event. This assessment must include the availability of other systems or components that could have performed the same function as the components and systems that failed during the event.
            (4) A description of any corrective actions planned as a result of the event, including the means employed to repair any defects, and actions taken to reduce the probability of similar events occurring in the future.
            (5) Reference to any previous similar events involving the same packaging that are known to the licensee or certificate holder.
            (6) The name and telephone number of a person within the licensee's organization who is knowledgeable about the event and can provide additional information.
            (7) The extent of exposure of individuals to radiation or to radioactive materials without identification of individuals by name.
            (d) Report legibility. The reports submitted by licensees and/or certificate holders under this section must be of sufficient quality to permit reproduction and micrographic processing.
            [69 FR 3796, Jan. 26, 2004, as amended at 75 FR 73945, Nov. 30, 2010; 79 FR 75740, Dec. 19, 2014; 84 FR 65645, Nov. 29, 2019]

            § 71.97
            Advance notification of shipment of irradiated reactor fuel and nuclear waste.
            (a)(1) As specified in paragraphs (b), (c), and (d) of this section, each licensee shall provide advance notification to the governor of a State, or the governor's designee, of the shipment of licensed material, within or across the boundary of the State, before the transport, or delivery to a carrier, for transport, of licensed material outside the confines of the licensee's plant or other place of use or storage.
            (2) As specified in paragraphs (b), (c), and (d) of this section, after June 11, 2013, each licensee shall provide advance notification to the Tribal official of participating Tribes referenced in paragraph (c)(3)(iii) of this section, or the official's designee, of the shipment of licensed material, within or across the boundary of the Tribe's reservation, before the transport, or delivery to a carrier, for transport, of licensed material outside the confines of the licensee's plant or other place of use or storage.
            (b) Advance notification is also required under this section for the shipment of licensed material, other than irradiated fuel, meeting the following three conditions:
            (1) The licensed material is required by this part to be in Type B packaging for transportation;
            (2) The licensed material is being transported to or across a State boundary en route to a disposal facility or to a collection point for transport to a disposal facility; and
            (3) The quantity of licensed material in a single package exceeds the least of the following:
            (i) 3000 times the A1 value of the radionuclides as specified in appendix A, Table A-1 for special form radioactive material;
            (ii) 3000 times the A2 value of the radionuclides as specified in appendix A, Table A-1 for normal form radioactive material; or
            (iii) 1000 TBq (27,000 Ci).

            (c) Procedures for submitting advance notification. (1) The notification must be made in writing to:
            (i) The office of each appropriate governor or governor's designee;
            (ii) The office of each appropriate Tribal official or Tribal official's designee; and
            (iii) The Director, Office of Nuclear Security and Incident Response.
            (2) A notification delivered by mail must be postmarked at least 7 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.
            (3) A notification delivered by any other means than mail must reach the office of the governor or of the governor's designee or the Tribal official or Tribal official's designee at least 4 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.

            (i) A list of the names and mailing addresses of the governors' designees receiving advance notification of transportation of nuclear waste was published in the Federal Register on June 30, 1995 (60 FR 34306).

            (ii) Contact information for each State, including telephone and mailing addresses of governors and governors' designees, and participating Tribes, including telephone and mailing addresses of Tribal officials and Tribal official's designees, is available on the NRC Web site at: https://scp.nrc.gov/special/designee.pdf.

            (iii) A list of the names and mailing addresses of the governors' designees and Tribal officials' designees of participating Tribes is available on request from the Director, Division of Materials Safety, Security, State, and Tribal Programs, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
            (4) The licensee shall retain a copy of the notification as a record for 3 years.
            (d) Information to be furnished in advance notification of shipment. Each advance notification of shipment of irradiated reactor fuel or nuclear waste must contain the following information:
            (1) The name, address, and telephone number of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment;
            (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the regulations of DOT in 49 CFR 172.202 and 172.203(d);
            (3) The point of origin of the shipment and the 7-day period during which departure of the shipment is estimated to occur;
            (4) The 7-day period during which arrival of the shipment at State boundaries or Tribal reservation boundaries is estimated to occur;
            (5) The destination of the shipment, and the 7-day period during which arrival of the shipment is estimated to occur; and
            (6) A point of contact, with a telephone number, for current shipment information.
            (e) Revision notice. A licensee who finds that schedule information previously furnished to a governor or governor's designee or a Tribal official or Tribal official's designee, in accordance with this section, will not be met, shall telephone a responsible individual in the office of the governor of the State or of the governor's designee or the Tribal official or the Tribal official's designee and inform that individual of the extent of the delay beyond the schedule originally reported. The licensee shall maintain a record of the name of the individual contacted for 3 years.
            (f) Cancellation notice. (1) Each licensee who cancels an irradiated reactor fuel or nuclear waste shipment for which advance notification has been sent shall send a cancellation notice to the governor of each State or to the governor's designee previously notified, each Tribal official or to the Tribal official's designee previously notified, and to the Director, Office of Nuclear Security and Incident Response.

            (2) The licensee shall state in the notice that it is a cancellation and identify the advance notification that is being canceled. The licensee shall retain a copy of the notice as a record for 3 years.
            [60 FR 50264, Sept. 28, 1995, as amended at 67 FR 3586, Jan. 25, 2002; 68 FR 14529, Mar. 26, 2003; 68 FR 23575, May 5, 2003; 68 FR 58818, Oct. 10, 2003; 74 FR 62683, Dec. 1, 2009; 75 FR 73945, Nov. 30, 2010; 77 FR 34204, June 11, 2012; 78 FR 17021, Mar. 19, 2013; 79 FR 75741, Dec. 19, 2014; 80 FR 74981, Dec. 1, 2015; 83 FR 30288, June 28, 2018; 83 FR 58723, Nov. 21, 2018]

            § 71.99
            Violations.
            (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of—
            (1) The Atomic Energy Act of 1954, as amended;
            (2) Title II of the Energy Reorganization Act of 1974, as amended; or (3) A regulation or order issued pursuant to those Acts.
            (b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:
            (1) For violations of—
            (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;
            (ii) Section 206 of the Energy Reorganization Act;
            (iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section; or
            (iv) Any term , condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.
            (2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.

            § 71.100
            Criminal penalties.
            (a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 71 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.
            (b) The regulations in part 71 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 71.0, 71.2, 71.4, 71.6, 71.7, 71.10, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.40, 71.41, 71.43, 71.45, 71.47, 71.51, 71.55, 71.59, 71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, and 71.100.
            [60 FR 50264, Sept. 28, 1995, as amended at 69 FR 3796, Jan. 26, 2004]

          Subpart H—Quality Assurance

            Source:
            69 FR 3796, Jan. 26, 2004, unless otherwise noted.

            § 71.101
            Quality assurance requirements.
            (a) Purpose. This subpart describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging that are important to safety. As used in this subpart, “quality assurance” comprises all those planned and systematic actions necessary to provide adequate confidence that a system or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to control of the physical characteristics and quality of the material or component to predetermined requirements. Each certificate holder and applicant for a package approval is responsible for satisfying the quality assurance requirements that apply to design, fabrication, testing, and modification of packaging subject to this subpart. Each licensee is responsible for satisfying the quality assurance requirements that apply to its use of a packaging for the shipment of licensed material subject to this subpart.
            (b) Establishment of program. Each licensee, certificate holder, and applicant for a CoC shall establish, maintain, and execute a quality assurance program satisfying each of the applicable criteria of §§ 71.101 through 71.137 and satisfying any specific provisions that are applicable to the licensee's activities including procurement of packaging. The licensee, certificate holder, and applicant for a CoC shall execute the applicable criteria in a graded approach to an extent that is commensurate with the quality assurance requirement's importance to safety.
            (c) Approval of program. (1) Before the use of any package for the shipment of licensed material subject to this subpart, each licensee shall obtain Commission approval of its quality assurance program. Using an appropriate method listed in § 71.1(a), each licensee shall file a description of its quality assurance program, including a discussion of which requirements of this subpart are applicable and how they will be satisfied, by submitting the description to: ATTN: Document Control Desk, Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards.
            (2) Before the fabrication, testing, or modification of any package for the shipment of licensed material subject to this subpart, each certificate holder, or applicant for a Certificate of Compliance shall obtain Commission approval of its quality assurance program. Each certificate holder or applicant for a CoC shall, in accordance with § 71.1, file a description of its quality assurance program, including a discussion of which requirements of this subpart are applicable and how they will be satisfied.
            (d) Existing package designs. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979, and which have been designed in accordance with the provisions of this part in effect at the time of application for package approval. Those packages will be accepted as having been designed in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section.
            (e) Existing packages. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979, have been at least partially fabricated before that date, and for which the fabrication is in accordance with the provisions of this part in effect at the time of application for approval of package design. These packages will be accepted as having been fabricated and assembled in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section.
            (f) Previously approved programs. A Commission-approved quality assurance program that satisfies the applicable criteria of subpart H of this part, Appendix B of part 50 of this chapter, or subpart G of part 72 of this chapter, and that is established, maintained, and executed regarding transport packages, will be accepted as satisfying the requirements of paragraph (b) of this section. Before first use, the licensee, certificate holder, and applicant for a CoC shall notify the NRC, in accordance with § 71.1, of its intent to apply its previously approved subpart H, Appendix B, or subpart G quality assurance program to transportation activities. The licensee, certificate holder, and applicant for a CoC shall identify the program by date of submittal to the Commission, Docket Number, and date of Commission approval.
            (g) Radiography containers. A program for transport container inspection and maintenance limited to radiographic exposure devices, source changers, or packages transporting these devices and meeting the requirements of § 34.31(b) of this chapter or equivalent Agreement State requirement, is deemed to satisfy the requirements of §§ 71.17(b) and 71.101(b).
            [69 FR 3796, Jan. 26, 2004, as amended at 75 FR 73945, Nov. 30, 2010; 79 FR 75741, Dec. 19, 2014; 80 FR 34013, June 12, 2015; 84 FR 65645, Nov. 29, 2019]

            § 71.103
            Quality assurance organization.

            (a) The licensee, certificate holder, and applicant for a Certificate of Compliance shall be responsible for the establishment and execution of the quality assurance program. The licensee, certificate holder, and applicant for a Certificate of Compliance may delegate to others, such as contractors, agents, or consultants, the work of establishing and executing the quality assurance program, or any part of the quality assurance program, but shall retain responsibility for the program. These activities include performing the functions associated with attaining quality objectives and the quality assurance functions.

            (b) The quality assurance functions are—
            (1) Assuring that an appropriate quality assurance program is established and effectively executed; and
            (2) Verifying, by procedures such as checking, auditing, and inspection, that activities affecting the functions that are important to safety have been correctly performed.
            (c) The persons and organizations performing quality assurance functions must have sufficient authority and organizational freedom to—
            (1) Identify quality problems;
            (2) Initiate, recommend, or provide solutions; and
            (3) Verify implementation of solutions.
            (d) The persons and organizations performing quality assurance functions shall report to a management level that assures that the required authority and organizational freedom, including sufficient independence from cost and schedule, when opposed to safety considerations, are provided.
            (e) Because of the many variables involved, such as the number of personnel, the type of activity being performed, and the location or locations where activities are performed, the organizational structure for executing the quality assurance program may take various forms, provided that the persons and organizations assigned the quality assurance functions have the required authority and organizational freedom.
            (f) Irrespective of the organizational structure, the individual(s) assigned the responsibility for assuring effective execution of any portion of the quality assurance program, at any location where activities subject to this section are being performed, must have direct access to the levels of management necessary to perform this function.
            [69 FR 3796, Jan. 26, 2004, as amended at 80 FR 34014, June 12, 2015]

            § 71.105
            Quality assurance program.
            (a) The licensee, certificate holder, and applicant for a CoC shall establish, at the earliest practicable time consistent with the schedule for accomplishing the activities, a quality assurance program that complies with the requirements of §§ 71.101 through 71.137. The licensee, certificate holder, and applicant for a CoC shall document the quality assurance program by written procedures or instructions and shall carry out the program in accordance with those procedures throughout the period during which the packaging is used. The licensee, certificate holder, and applicant for a CoC shall identify the material and components to be covered by the quality assurance program, the major organizations participating in the program, and the designated functions of these organizations.
            (b) The licensee, certificate holder, and applicant for a CoC, through its quality assurance program, shall provide control over activities affecting the quality of the identified materials and components to an extent consistent with their importance to safety, and as necessary to assure conformance to the approved design of each individual package used for the shipment of radioactive material. The licensee, certificate holder, and applicant for a CoC shall assure that activities affecting quality are accomplished under suitably controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanliness; and assurance that all prerequisites for the given activity have been satisfied. The licensee, certificate holder, and applicant for a CoC shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test.
            (c) The licensee, certificate holder, and applicant for a CoC shall base the requirements and procedures of its quality assurance program on the following considerations concerning the complexity and proposed use of the package and its components:
            (1) The impact of malfunction or failure of the item to safety;
            (2) The design and fabrication complexity or uniqueness of the item;
            (3) The need for special controls and surveillance over processes and equipment;

            (4) The degree to which functional compliance can be demonstrated by inspection or test; and

            (5) The quality history and degree of standardization of the item.
            (d) The licensee, certificate holder, and applicant for a CoC shall provide for indoctrination and training of personnel performing activities affecting quality, as necessary to assure that suitable proficiency is achieved and maintained. The licensee, certificate holder, and applicant for a CoC shall review the status and adequacy of the quality assurance program at established intervals. Management of other organizations participating in the quality assurance program shall review regularly the status and adequacy of that part of the quality assurance program they are executing.

            § 71.106
            Changes to quality assurance program.
            (a) Each quality assurance program approval holder shall submit, in accordance with § 71.1(a), a description of a proposed change to its NRC-approved quality assurance program that will reduce commitments in the program description as approved by the NRC. The quality assurance program approval holder shall not implement the change before receiving NRC approval.
            (1) The description of a proposed change to the NRC-approved quality assurance program must identify the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the applicable requirements of subpart H of this part.
            (2) [Reserved]
            (b) Each quality assurance program approval holder may change a previously approved quality assurance program without prior NRC approval, if the change does not reduce the commitments in the quality assurance program previously approved by the NRC. Changes to the quality assurance program that do not reduce the commitments shall be submitted to the NRC every 24 months, in accordance with § 71.1(a). In addition to quality assurance program changes involving administrative improvements and clarifications, spelling corrections, and non-substantive changes to punctuation or editorial items, the following changes are not considered reductions in commitment:
            (1) The use of a quality assurance standard approved by the NRC that is more recent than the quality assurance standard in the certificate holder's or applicant's current quality assurance program at the time of the change;
            (2) The use of generic organizational position titles that clearly denote the position function, supplemented as necessary by descriptive text, rather than specific titles, provided that there is no substantive change to either the functions of the position or reporting responsibilities;
            (3) The use of generic organizational charts to indicate functional relationships, authorities, and responsibilities, or alternatively, the use of descriptive text, provided that there is no substantive change to the functional relationships, authorities, or responsibilities;
            (4) The elimination of quality assurance program information that duplicates language in quality assurance regulatory guides and quality assurance standards to which the quality assurance program approval holder has committed to on record; and
            (5) Organizational revisions that ensure that persons and organizations performing quality assurance functions continue to have the requisite authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations.
            (c) Each quality assurance program approval holder shall maintain records of quality assurance program changes.
            [80 FR 34014, June 12, 2015]

            § 71.107
            Package design control.

            (a) The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that applicable regulatory requirements and the package design, as specified in the license or CoC for those materials and components to which this section applies, are correctly translated into specifications, drawings, procedures, and instructions. These measures must include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from standards are controlled. Measures must be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the functions of the materials, parts, and components of the packaging that are important to safety.
            (b) The licensee, certificate holder, and applicant for a CoC shall establish measures for the identification and control of design interfaces and for coordination among participating design organizations. These measures must include the establishment of written procedures, among participating design organizations, for the review, approval, release, distribution, and revision of documents involving design interfaces. The design control measures must provide for verifying or checking the adequacy of design, by methods such as design reviews, alternate or simplified calculational methods, or by a suitable testing program. For the verifying or checking process, the licensee shall designate individuals or groups other than those who were responsible for the original design, but who may be from the same organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, the licensee, certificate holder, and applicant for a CoC shall include suitable qualification testing of a prototype or sample unit under the most adverse design conditions. The licensee, certificate holder, and applicant for a CoC shall apply design control measures to the following:
            (1) Criticality physics, radiation shielding, stress, thermal, hydraulic, and accident analyses;
            (2) Compatibility of materials;
            (3) Accessibility for inservice inspection, maintenance, and repair;
            (4) Features to facilitate decontamination; and
            (5) Delineation of acceptance criteria for inspections and tests.
            (c) The licensee, certificate holder, and applicant for a CoC shall subject design changes, including field changes, to design control measures commensurate with those applied to the original design. Changes in the conditions specified in the CoC require prior NRC approval.

            § 71.109
            Procurement document control.
            The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that adequate quality is required in the documents for procurement of material, equipment, and services, whether purchased by the licensee, certificate holder, and applicant for a CoC or by its contractors or subcontractors. To the extent necessary, the licensee, certificate holder, and applicant for a CoC shall require contractors or subcontractors to provide a quality assurance program consistent with the applicable provisions of this part.

            § 71.111
            Instructions, procedures, and drawings.
            The licensee, certificate holder, and applicant for a CoC shall prescribe activities affecting quality by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall require that these instructions, procedures, and drawings be followed. The instructions, procedures, and drawings must include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

            § 71.113
            Document control.
            The licensee, certificate holder, and applicant for a CoC shall establish measures to control the issuance of documents such as instructions, procedures, and drawings, including changes, that prescribe all activities affecting quality. These measures must assure that documents, including changes, are reviewed for adequacy, approved for release by authorized personnel, and distributed and used at the location where the prescribed activity is performed.

            § 71.115
            Control of purchased material, equipment, and services.

            (a) The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents. These measures must include provisions, as appropriate, for source evaluation and selection, objective evidence of quality furnished by the contractor or subcontractor, inspection at the contractor or subcontractor source, and examination of products on delivery.
            (b) The licensee, certificate holder, and applicant for a CoC shall have available documentary evidence that material and equipment conform to the procurement specifications before installation or use of the material and equipment. The licensee, certificate holder, and applicant for a CoC shall retain, or have available, this documentary evidence for the life of the package to which it applies. The licensee, certificate holder, and applicant for a CoC shall assure that the evidence is sufficient to identify the specific requirements met by the purchased material and equipment.
            (c) The licensee, certificate holder, and applicant for a CoC shall assess the effectiveness of the control of quality by contractors and subcontractors at intervals consistent with the importance, complexity, and quantity of the product or services.

            § 71.117
            Identification and control of materials, parts, and components.
            The licensee, certificate holder, and applicant for a CoC shall establish measures for the identification and control of materials, parts, and components. These measures must assure that identification of the item is maintained by heat number, part number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, installation, and use of the item. These identification and control measures must be designed to prevent the use of incorrect or defective materials, parts, and components.

            § 71.119
            Control of special processes.
            The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that special processes, including welding, heat treating, and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

            § 71.121
            Internal inspection.
            The licensee, certificate holder, and applicant for a CoC shall establish and execute a program for inspection of activities affecting quality by or for the organization performing the activity, to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. The inspection must be performed by individuals other than those who performed the activity being inspected. Examination, measurements, or tests of material or products processed must be performed for each work operation where necessary to assure quality. If direct inspection of processed material or products is not carried out, indirect control by monitoring processing methods, equipment, and personnel must be provided. Both inspection and process monitoring must be provided when quality control is inadequate without both. If mandatory inspection hold points, which require witnessing or inspecting by the licensee's designated representative and beyond which work should not proceed without the consent of its designated representative, are required, the specific hold points must be indicated in appropriate documents.

            § 71.123
            Test control.
            The licensee, certificate holder, and applicant for a CoC shall establish a test program to assure that all testing required to demonstrate that the packaging components will perform satisfactorily in service is identified and performed in accordance with written test procedures that incorporate the requirements of this part and the requirements and acceptance limits contained in the package approval. The test procedures must include provisions for assuring that all prerequisites for the given test are met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions. The licensee, certificate holder, and applicant for a CoC shall document and evaluate the test results to assure that test requirements have been satisfied.

            § 71.125
            Control of measuring and test equipment.
            The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that tools, gauges, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified times to maintain accuracy within necessary limits.

            § 71.127
            Handling, storage, and shipping control.
            The licensee, certificate holder, and applicant for a CoC shall establish measures to control, in accordance with instructions, the handling, storage, shipping, cleaning, and preservation of materials and equipment to be used in packaging to prevent damage or deterioration. When necessary for particular products, special protective environments, such as inert gas atmosphere, and specific moisture content and temperature levels must be specified and provided.

            § 71.129
            Inspection, test, and operating status.
            (a) The licensee, certificate holder, and applicant for a CoC shall establish measures to indicate, by the use of markings such as stamps, tags, labels, routing cards, or other suitable means, the status of inspections and tests performed upon individual items of the packaging. These measures must provide for the identification of items that have satisfactorily passed required inspections and tests, where necessary to preclude inadvertent bypassing of the inspections and tests.
            (b) The licensee shall establish measures to identify the operating status of components of the packaging, such as tagging valves and switches, to prevent inadvertent operation.

            § 71.131
            Nonconforming materials, parts, or components.
            The licensee, certificate holder, and applicant for a CoC shall establish measures to control materials, parts, or components that do not conform to the licensee's requirements to prevent their inadvertent use or installation. These measures must include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Nonconforming items must be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures.

            § 71.133
            Corrective action.
            The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that conditions adverse to quality, such as deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. In the case of a significant condition adverse to quality, the measures must assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken must be documented and reported to appropriate levels of management.

            § 71.135
            Quality assurance records.

            The licensee, certificate holder, and applicant for a Certificate of Compliance shall maintain sufficient written records to describe the activities affecting quality. These records must include changes to the quality assurance program as required by § 71.106, the instructions, procedures, and drawings required by § 71.111 to prescribe quality assurance activities, and closely related specifications such as required qualifications of personnel, procedures, and equipment. The records must include the instructions or procedures that establish a records retention program that is consistent with applicable regulations and designates factors such as duration, location, and assigned responsibility. The licensee, certificate holder, and applicant for a Certificate of Compliance shall retain these records for 3 years beyond the date when the licensee, certificate holder, and applicant for a Certificate of Compliance last engage in the activity for which the quality assurance program was developed. If any portion of the quality assurance program, written procedures or instructions is superseded, the licensee, certificate holder, and applicant for a Certificate of Compliance shall retain the superseded material for 3 years after it is superseded.
            [80 FR 34014, June 12, 2015]

            § 71.137
            Audits.
            The licensee, certificate holder, and applicant for a CoC shall carry out a comprehensive system of planned and periodic audits to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program. The audits must be performed in accordance with written procedures or checklists by appropriately trained personnel not having direct responsibilities in the areas being audited. Audited results must be documented and reviewed by management having responsibility in the area audited. Followup action, including reaudit of deficient areas, must be taken where indicated.

            Pt. 71, App. A
            Appendix A to Part 71—Determination of A1 and A2

            I. Values of A1 and A2 for individual radionuclides, which are the bases for many activity limits elsewhere in these regulations, are given in Table A-1. The curie (Ci) values specified are obtained by converting from the Terabecquerel (TBq) value. The Terabecquerel values are the regulatory standard. The curie values are for information only and are not intended to be the regulatory standard. Where values of A1 and A2 are unlimited, it is for radiation control purposes only. For nuclear criticality safety, some materials are subject to controls placed on fissile material.

            II. a. For individual radionuclides whose identities are known, but which are not listed in Table A-1, the A1 and A2 values contained in Table A-3 may be used. Otherwise, the licensee shall obtain prior Commission approval of the A1 and A2 values for radionuclides not listed in Table A-1, before shipping the material.
            b. For individual radionuclides whose identities are known, but which are not listed in Table A-2, the exempt material activity concentration and exempt consignment activity values contained in Table A-3 may be used. Otherwise, the licensee shall obtain prior Commission approval of the exempt material activity concentration and exempt consignment activity values for radionuclides not listed in Table A-2, before shipping the material.
            c. The licensee shall submit requests for prior approval, described under paragraphs II.a. and II.b. of this Appendix, to the Commission, in accordance with § 71.1 of this part.
            III. In the calculations of A1 and A2 for a radionuclide not in Table A-1, a single radioactive decay chain, in which radionuclides are present in their naturally occurring proportions, and in which no daughter radionuclide has a half-life either longer than 10 days, or longer than that of the parent radionuclide, shall be considered as a single radionuclide, and the activity to be taken into account, and the A1 and A2 value to be applied, shall be those corresponding to the parent radionuclide of that chain. In the case of radioactive decay chains in which any daughter radionuclide has a half-life either longer than 10 days, or greater than that of the parent radionuclide, the parent and those daughter radionuclides shall be considered as mixtures of different radionuclides.
            IV. For mixtures of radionuclides whose identities and respective activities are known, the following conditions apply:
            a. For special form radioactive material, the maximum quantity transported in a Type A package is as follows:

              ER12JN15.094

            where B(i) is the activity of radionuclide i in special form, and A1(i) is the A1 value for radionuclide i.
            b. For normal form radioactive material, the maximum quantity transported in a Type A package is as follows:

              ER12JN15.095

            where B(i) is the activity of radionuclide i in normal form, and A2(i) is the A2 value for radionuclide i.
            c. If the package contains both special and normal form radioactive material, the activity that may be transported in a Type A package is as follows:

              ER12JN15.096

            where B(i) is the activity of radionuclide i as special form radioactive material, A1(i) is the A1 value for radionuclide i, C(j) is the activity of radionuclide j as normal form radioactive material, and A2(j) is the A2 value for radionuclide j.
            d. Alternatively, the A1 value for mixtures of special form material may be determined as follows:

              ER12JN15.097

            where f(i) is the fraction of activity for radionuclide i in the mixture and A1(i) is the appropriate A1 value for radionuclide i.
            e. Alternatively, the A2 value for mixtures of normal form material may be determined as follows:

              ER12JN15.098

            where f(i) is the fraction of activity for radionuclide i in the mixture and A2(i) is the appropriate A2 value for radionuclide i.
            f. The exempt activity concentration for mixtures of nuclides may be determined as follows:

              ER12JN15.099

            where f(i) is the fraction of activity concentration of radionuclide i in the mixture and [A](i) is the activity concentration for exempt material containing radionuclide i.
            g. The activity limit for an exempt consignment for mixtures of radionuclides may be determined as follows:

              ER12JN15.100

            where f(i) is the fraction of activity of radionuclide i in the mixture and A(i) is the activity limit for exempt consignments for radionuclide i.

            V.a. When the identity of each radionuclide is known, but the individual activities of some of the radionuclides are not known, the radionuclides may be grouped, and the lowest A1 or A2 value, as appropriate, for the radionuclides in each group may be used in applying the formulas in paragraph IV. Groups may be based on the total alpha activity and the total beta/gamma activity when these are known, using the lowest A1 or A2 values for the alpha emitters and beta/gamma emitters.
            b. When the identity of each radionuclide is known but the individual activities of some of the radionuclides are not known, the radionuclides may be grouped and the lowest [A] (activity concentration for exempt material) or A (activity limit for exempt consignment) value, as appropriate, for the radionuclides in each group may be used in applying the formulas in paragraph IV of this appendix. Groups may be based on the total alpha activity and the total beta/gamma activity when these are known, using the lowest [A] or A values for the alpha emitters and beta/gamma emitters, respectively.

              Table A-1—A1 and A2 Values For Radionuclides

                Symbol ofradionuclide

                Element andatomic number

                A1 (TBq)
                A1 (Ci)b

                A2 (TBq)
                A2 (Ci)b

                Specific activity
                (TBq/g)
                (Ci/g)

                Ac-225 (a)
                Actinium (89)
                8.0 × 10−1

                2.2 × 101

                6.0 × 10−3

                1.6 × 10−1

                2.1 × 103

                5.8 × 104

                Ac-227 (a)

                9.0 × 10−1

                2.4 × 101

                9.0 × 10−5

                2.4 × 10−3

                2.7
                7.2 × 101

                Ac-228

                6.0 × 10−1

                1.6 × 101

                5.0 × 10−1

                1.4 × 101

                8.4 × 104

                2.2 × 106

                Ag-105
                Silver (47)
                2.0
                5.4 × 101

                2.0
                5.4 × 101

                1.1 × 103

                3.0 × 104

                Ag-108m (a)

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                9.7 × 10−1

                2.6 × 101

                Ag-110m (a)

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                1.8 × 102

                4.7 × 103

                Ag-111

                2.0
                5.4 × 101

                6.0 × 10−1

                1.6 × 101

                5.8 × 103

                1.6 × 105

                Al-26
                Aluminum (13)
                1.0 × 10−1

                2.7
                1.0 × 10−1

                2.7
                7.0 × 10−4

                1.9 × 10−2

                Am-241
                Americium (95)
                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                1.3 × 10−1

                3.4

                Am-242m (a)

                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                3.6 × 10−1

                1.0 × 101

                Am-243 (a)

                5.0
                1.4 × 102

                1.0 × 10−3

                2.7 × 10−2

                7.4 × 10−3

                2.0 × 10−1

                Ar-37
                Argon (18)
                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                3.7 × 103

                9.9 × 104

                Ar-39

                4.0 × 101

                1.1 × 103

                2.0 × 101

                5.4 × 102

                1.3
                3.4 × 101

                Ar-41

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                1.5 × 106

                4.2 × 107

                As-72
                Arsenic (33)
                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                6.2 × 104

                1.7 × 106

                As-73

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                8.2 × 102

                2.2 × 104

                As-74

                1.0
                2.7 × 101

                9.0 × 10−1

                2.4 × 101

                3.7 × 103

                9.9 × 104

                As-76

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                5.8 × 104

                1.6 × 106

                As-77

                2.0 × 101

                5.4 × 102

                7.0 × 10−1

                1.9 × 101

                3.9 × 104

                1.0 × 106

                At-211 (a)
                Astatine (85)
                2.0 × 101

                5.4 × 102

                5.0 × 10−1

                1.4 × 101

                7.6 × 104

                2.1 × 106

                Au-193
                Gold (79)
                7.0
                1.9 × 102

                2.0
                5.4 × 101

                3.4 × 104

                9.2 × 105

                Au-194

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                1.5 × 104

                4.1 × 105

                Au-195

                1.0 × 101

                2.7 × 102

                6.0
                1.6 × 102

                1.4 × 102

                3.7 × 103

                Au-198

                1.0
                2.7 × 101

                6.0 × 10−1

                1.6 × 101

                9.0 × 103

                2.4 × 105

                Au-199

                1.0 × 101

                2.7 × 102

                6.0 × 10−1

                1.6 × 101

                7.7 × 103

                2.1 × 105

                Ba-131 (a)
                Barium (56)
                2.0
                5.4 × 101

                2.0
                5.4 × 101

                3.1 × 103

                8.4 × 104

                Ba-133

                3.0
                8.1 × 101

                3.0
                8.1 × 101

                9.4
                2.6 × 102

                Ba-133m

                2.0 × 101

                5.4 × 102

                6.0 × 10−1

                1.6 × 101

                2.2 × 104

                6.1 × 105

                Ba-140 (a)

                5.0 × 10−1

                1.4 × 101

                3.0 × 10−1

                8.1
                2.7 × 103

                7.3 × 104

                Be-7
                Beryllium (4)
                2.0 × 101

                5.4 × 102

                2.0 × 101

                5.4 × 102

                1.3 × 104

                3.5 × 105

                Be-10

                4.0 × 101

                1.1 × 103

                6.0 × 10−1

                1.6 × 101

                8.3 × 10−4

                2.2 × 10−2

                Bi-205
                Bismuth (83)
                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                1.5 × 103

                4.2 × 104

                Bi-206

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                3.8 × 103

                1.0 × 105

                Bi-207

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                1.9
                5.2 × 101

                Bi-210

                1.0
                2.7 × 101

                6.0 × 10−1

                1.6 × 101

                4.6 × 103

                1.2 × 105

                Bi-210m (a)

                6.0 × 10−1

                1.6 × 101

                2.0 × 10−2

                5.4 × 10−1

                2.1 × 10−5

                5.7 × 10−4

                Bi-212 (a)

                7.0 × 10−1

                1.9 × 101

                6.0 × 10−1

                1.6 × 101

                5.4 × 105

                1.5 × 107

                Bk-247
                Berkelium (97)
                8.0
                2.2 × 102

                8.0 × 10−4

                2.2 × 10−2

                3.8 × 10−2

                1.0

                Bk-249 (a)

                4.0 × 101

                1.1 × 103

                3.0 × 10−1

                8.1
                6.1 × 101

                1.6 × 103

                Br-76
                Bromine (35)
                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                9.4 × 104

                2.5 × 106

                Br-77

                3.0
                8.1 × 101

                3.0
                8.1 × 101

                2.6 × 104

                7.1 × 105

                Br-82

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                4.0 × 104

                1.1 × 106

                C-11
                Carbon (6)
                1.0
                2.7 × 101

                6.0 × 10−1

                1.6 × 101

                3.1 × 107

                8.4 × 108

                C-14

                4.0 × 101

                1.1 × 103

                3.0
                8.1 × 101

                1.6 × 10−1

                4.5

                Ca-41
                Calcium (20)
                Unlimited
                Unlimited
                Unlimited
                Unlimited
                3.1 × 10−3

                8.5 × 10−2

                Ca-45

                4.0 × 101

                1.1 × 103

                1.0
                2.7 × 101

                6.6 × 102

                1.8 × 104

                Ca-47 (a)

                3.0
                8.1 × 101

                3.0 × 10−1

                8.1
                2.3 × 104

                6.1 × 105

                Cd-109
                Cadmium (48)
                3.0 × 101

                8.1 × 102

                2.0
                5.4 × 101

                9.6 × 101

                2.6 × 103

                Cd-113m

                4.0 × 101

                1.1 × 103

                5.0 × 10−1

                1.4 × 101

                8.3
                2.2 × 102

                Cd-115 (a)

                3.0
                8.1 × 101

                4.0 × 10−1

                1.1 × 101

                1.9 × 104

                5.1 × 105

                Cd-115m

                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                9.4 × 102

                2.5 × 104

                Ce-139
                Cerium (58)
                7.0
                1.9 × 102

                2.0
                5.4 × 101

                2.5 × 102

                6.8 × 103

                Ce-141

                2.0 × 101

                5.4 × 102

                6.0 × 10−1

                1.6 × 101

                1.1 × 103

                2.8 × 104

                Ce-143

                9.0 × 10−1

                2.4 × 101

                6.0 × 10−1

                1.6 × 101

                2.5 × 104

                6.6 × 105

                Ce-144 (a)

                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                1.2 × 102

                3.2 × 103

                Cf-248
                Californium (98)
                4.0 × 101

                1.1 × 103

                6.0 × 10−3

                1.6 × 10−1

                5.8 × 101

                1.6 × 103

                Cf-249

                3.0
                8.1 × 101

                8.0 × 10−4

                2.2 × 10−2

                1.5 × 10−1

                4.1

                Cf-250

                2.0 × 101

                5.4 × 102

                2.0 × 10−3

                5.4 × 10−2

                4.0
                1.1 × 102

                Cf-251

                7.0
                1.9 × 102

                7.0 × 10−4

                1.9 × 10−2

                5.9 × 10−2

                1.6

                Cf-252

                1.0 × 10−1

                2.7
                3.0 × 10−3

                8.1 × 10−2

                2.0 × 101

                5.4 × 102

                Cf-253 (a)

                4.0 × 101

                1.1 × 103

                4.0 × 10−2

                1.1
                1.1 × 103

                2.9 × 104

                Cf-254

                1.0 × 10−3

                2.7 × 10−2

                1.0 × 10−3

                2.7 × 10−2

                3.1 × 102

                8.5 × 103

                Cl-36
                Chlorine (17)
                1.0 × 101

                2.7 × 102

                6.0 × 10−1

                1.6 × 101

                1.2 × 10−3

                3.3 × 10−2

                Cl-38

                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                4.9 × 106

                1.3 × 108

                Cm-240
                Curium (96)
                4.0 × 101

                1.1 × 103

                2.0 × 10−2

                5.4 × 10−1

                7.5 × 102

                2.0 × 104

                Cm-241

                2.0
                5.4 × 101

                1.0
                2.7 × 101

                6.1 × 102

                1.7 × 104

                Cm-242

                4.0 × 101

                1.1 × 103

                1.0 × 10−2

                2.7 × 10−1

                1.2 × 102

                3.3 × 103

                Cm-243

                9.0
                2.4 × 102

                1.0 × 10−3

                2.7 × 10−2

                1.9 × 10−3

                5.2 × 101

                Cm-244

                2.0 × 101

                5.4 × 102

                2.0 × 10−3

                5.4 × 10−2

                3.0
                8.1 × 101

                Cm-245

                9.0
                2.4 × 102

                9.0 × 10−4

                2.4 × 10−2

                6.4 × 10−3

                1.7 × 10−1

                Cm-246

                9.0
                2.4 × 102

                9.0 × 10−4

                2.4 × 10−2

                1.1 × 10−2

                3.1 × 10−1

                Cm-247 (a)

                3.0
                8.1 × 101

                1.0 × 10−3

                2.7 × 10−2

                3.4 × 10−6

                9.3 × 10−5

                Cm-248

                2.0 × 10−2

                5.4 × 10−1

                3.0 × 10−4

                8.1 × 10−3

                1.6 × 10−4

                4.2 × 10−3

                Co-55
                Cobalt (27)
                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                1.1 × 105

                3.1 × 106

                Co-56

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                1.1 × 103

                3.0 × 104

                Co-57

                1.0 × 101

                2.7 × 102

                1.0 × 101

                2.7 × 102

                3.1 × 102

                8.4 × 103

                Co-58

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                1.2 × 103

                3.2 × 104

                Co-58m

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                2.2 × 105

                5.9 × 106

                Co-60

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                4.2 × 101

                1.1 × 103

                Cr-51
                Chromium (24)
                3.0 × 101

                8.1 × 102

                3.0 × 101

                8.1 × 102

                3.4 × 103

                9.2 × 104

                Cs-129
                Cesium (55)
                4.0
                1.1 × 102

                4.0
                1.1 × 102

                2.8 × 104

                7.6 × 105

                Cs-131

                3.0 × 101

                8.1 × 102

                3.0 × 101

                8.1 × 102

                3.8 × 103

                1.0 × 105

                Cs-132

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                5.7 × 103

                1.5 × 105

                Cs-134

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                4.8 × 101

                1.3 × 103

                Cs-134m

                4.0 × 101

                1.1 × 103

                6.0 × 10−1

                1.6 × 101

                3.0 × 105

                8.0 × 106

                Cs-135

                4.0 × 101

                1.1 × 103

                1.0
                2.7 × 101

                4.3 × 10−5

                1.2 × 10−3

                Cs-136

                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                2.7 × 103

                7.3 × 104

                Cs-137 (a)

                2.0
                5.4 × 101

                6.0 × 10−1

                1.6 × 101

                3.2
                8.7 × 101

                Cu-64
                Copper (29)
                6.0
                1.6 × 102

                1.0
                2.7 × 101

                1.4 × 105

                3.9 × 106

                Cu-67

                1.0 × 101

                2.7 × 102

                7.0 × 10−1

                1.9 × 101

                2.8 × 104

                7.6 × 105

                Dy-159
                Dysprosium (66)
                2.0 × 101

                5.4 × 102

                2.0 × 101

                5.4 × 102

                2.1 × 102

                5.7 × 103

                Dy-165

                9.0 × 10−1

                2.4 × 101

                6.0 × 10−1

                1.6 × 101

                3.0 × 105

                8.2 × 106

                Dy-166 (a)

                9.0 × 10−1

                2.4 × 101

                3.0 × 10−1

                8.1
                8.6 × 103

                2.3 × 105

                Er-169
                Erbium (68)
                4.0 × 101

                1.1 × 103

                1.0
                2.7 × 101

                3.1 × 103

                8.3 × 104

                Er-171

                8.0 × 10−1

                2.2 × 101

                5.0 × 10−1

                1.4 × 101

                9.0 × 104

                2.4 × 106

                Eu-147
                Europium (63)
                2.0
                5.4 × 101

                2.0
                5.4 × 101

                1.4 × 103

                3.7 × 104

                Eu-148

                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                6.0 × 102

                1.6 × 104

                Eu-149

                2.0 × 101

                5.4 × 102

                2.0 × 101

                5.4 × 102

                3.5 × 102

                9.4 × 103

                Eu-150 (short lived)

                2.0
                5.4 × 101

                7.0 × 10−1

                1.9 × 101

                6.1 × 104

                1.6 × 106

                Eu-150 (long lived)

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                6.1 × 104

                1.6 × 106

                Eu-152

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                6.5
                1.8 × 102

                Eu-152m

                8.0 × 10−1

                2.2 × 101

                8.0 × 10−1

                2.2 × 101

                8.2 × 104

                2.2 × 106

                Eu-154

                9.0 × 10−1

                2.4 × 101

                6.0 × 10−1

                1.6 × 101

                9.8
                2.6 × 102

                Eu-155

                2.0 × 101

                5.4 × 102

                3.0
                8.1 × 101

                1.8 × 101

                4.9 × 102

                Eu-156

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                2.0 × 103

                5.5 × 104

                F-18
                Fluorine (9)
                1.0
                2.7 × 101

                6.0 × 10−1

                1.6 × 101

                3.5 × 106

                9.5 × 107

                Fe-52 (a)
                Iron (26)
                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                2.7 × 105

                7.3 × 106

                Fe-55

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                8.8 × 101

                2.4 × 103

                Fe-59

                9.0 × 10−1

                2.4 × 101

                9.0 × 10−1

                2.4 × 101

                1.8 × 103

                5.0 × 104

                Fe-60 (a)

                4.0 × 101

                1.1 × 103

                2.0 × 10−1

                5.4
                7.4 × 10−4

                2.0 × 10−2

                Ga-67
                Gallium (31)
                7.0
                1.9 × 102

                3.0
                8.1 × 101

                2.2 × 104

                6.0 × 105

                Ga-68

                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                1.5 × 106

                4.1 × 107

                Ga-72

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                1.1 × 105

                3.1 × 106

                Gd-146 (a)
                Gadolinium (64)
                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                6.9 × 102

                1.9 × 104

                Gd-148

                2.0 × 101

                5.4 × 102

                2.0 × 10−3

                5.4 × 10−2

                1.2
                3.2 × 101

                Gd-153

                1.0 × 101

                2.7 × 102

                9.0
                2.4 × 102

                1.3 × 102

                3.5 × 103

                Gd-159

                3.0
                8.1 × 101

                6.0 × 10−1

                1.6 × 101

                3.9 × 104

                1.1 × 106

                Ge-68 (a)
                Germanium (32)
                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                2.6 × 102

                7.1 × 103

                Ge-71

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                5.8 × 103

                1.6 × 105

                Ge-77

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                1.3 × 105

                3.6 × 106

                Hf-172 (a)
                Hafnium (72)
                6.0 × 10−1

                1.6 × 101

                6.0 × 10−1

                1.6 × 101

                4.1 × 101

                1.1 × 103

                Hf-175

                3.0
                8.1 × 101

                3.0
                8.1 × 101

                3.9 × 102

                1.1 × 104

                Hf-181

                2.0
                5.4 × 101

                5.0 × 10−1

                1.4 × 101

                6.3 × 102

                1.7 × 104

                Hf-182

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                8.1 × 10−6

                2.2 × 10−4

                Hg-194 (a)
                Mercury (80)
                1.0
                2.7 × 101

                1.0
                2.7 × 101

                1.3 × 10−1

                3.5

                Hg-195m (a)

                3.0
                8.1 × 101

                7.0 × 10−1

                1.9 × 101

                1.5 × 104

                4.0 × 105

                Hg-197

                2.0 × 101

                5.4 × 102

                1.0 × 101

                2.7 × 102

                9.2 × 103

                2.5 × 105

                Hg-197m

                1.0 × 101

                2.7 × 102

                4.0 × 10−1

                1.1 × 101

                2.5 × 104

                6.7 × 105

                Hg-203

                5.0
                1.4 × 102

                1.0
                2.7 × 101

                5.1 × 102

                1.4 × 104

                Ho-166
                Holmium (67)
                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                2.6 × 104

                7.0 × 105

                Ho-166m

                6.0 × 10−1

                1.6 × 101

                5.0 × 10−1

                1.4 × 101

                6.6 × 10−2

                1.8

                I-123
                Iodine (53)
                6.0
                1.6 × 102

                3.0
                8.1 × 101

                7.1 × 104

                1.9 × 106

                I-124

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                9.3 × 103

                2.5 × 105

                I-125

                2.0 × 101

                5.4 × 102

                3.0
                8.1 × 101

                6.4 × 102

                1.7 × 104

                I-126

                2.0
                5.4 × 101

                1.0
                2.7 × 101

                2.9 × 103

                8.0 × 104

                I-129

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                6.5 × 10−6

                1.8 × 10−4

                I-131

                3.0
                8.1 × 101

                7.0 × 10−1

                1.9 × 101

                4.6 × 103

                1.2 × 105

                I-132

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                3.8 × 105

                1.0 × 107

                I-133

                7.0 × 10−1

                1.9 × 101

                6.0 × 10−1

                1.6 × 101

                4.2 × 104

                1.1 × 106

                I-134

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                9.9 × 105

                2.7 × 107

                I-135 (a)

                6.0 × 10−1

                1.6 × 101

                6.0 × 10−1

                1.6 × 101

                1.3 × 105

                3.5 × 106

                In-111
                Indium (49)
                3.0
                8.1 × 101

                3.0
                8.1 × 101

                1.5 × 104

                4.2 × 105

                In-113m

                4.0
                1.1 × 102

                2.0
                5.4 × 101

                6.2 × 105

                1.7 × 107

                In-114m (a)

                1.0 × 101

                2.7 × 102

                5.0 × 10−1

                1.4 × 101

                8.6 × 102

                2.3 × 104

                In-115m

                7.0
                1.9 × 102

                1.0
                2.7 × 101

                2.2 × 105

                6.1 × 106

                Ir-189 (a)
                Iridium (77)
                1.0 × 101

                2.7 × 102

                1.0 × 101

                2.7 × 102

                1.9 × 103

                5.2 × 104

                Ir-190

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                2.3 × 103

                6.2 × 104

                Ir-192

                  c 1.0

                  c 2.7 × 101

                6.0 × 10−1

                1.6 × 101

                3.4 × 102

                9.2 × 103

                Ir-194

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                3.1 × 104

                8.4 × 105

                K-40
                Potassium (19)
                9.0 × 10−1

                2.4 × 101

                9.0 × 10−1

                2.4 × 101

                2.4 × 10−7

                6.4 × 10−6

                K-42

                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                2.2 × 105

                6.0 × 106

                K-43

                7.0 × 10−1

                1.9 × 101

                6.0 × 10−1

                1.6 × 101

                1.2 × 105

                3.3 × 106

                Kr-79
                Krypton (36)
                4.0
                1.1 × 102

                2.0
                5.4 × 101

                4.2 × 104

                1.1 × 106

                Kr-81

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                7.8 × 10−4

                2.1 × 10−2

                Kr-85

                1.0 × 101

                2.7 × 102

                1.0 × 101

                2.7 × 102

                1.5 × 101

                3.9 × 102

                Kr-85m

                8.0
                2.2 × 102

                3.0
                8.1 × 101

                3.0 × 105

                8.2 × 106

                Kr-87

                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                1.0 × 106

                2.8 × 107

                La-137
                Lanthanum (57)
                3.0 × 101

                8.1 × 102

                6.0
                1.6 × 102

                1.6 × 10−3

                4.4 × 10−2

                La-140

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                2.1 × 104

                5.6 × 105

                Lu-172
                Lutetium (71)
                6.0 × 10−1

                1.6 × 101

                6.0 × 10−1

                1.6 × 101

                4.2 × 103

                1.1 × 105

                Lu-173

                8.0
                2.2 × 102

                8.0
                2.2 × 102

                5.6 × 101

                1.5 × 103

                Lu-174

                9.0
                2.4 × 102

                9.0
                2.4 × 102

                2.3 × 101

                6.2 × 102

                Lu-174m

                2.0 × 101

                5.4 × 102

                1.0 × 101

                2.7 × 102

                2.0 × 102

                5.3 × 103

                Lu-177

                3.0 × 101

                8.1 × 102

                7.0 × 10−1

                1.9 × 101

                4.1 × 103

                1.1 × 105

                Mg-28 (a)
                Magnesium (12)
                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                2.0 × 105

                5.4 × 106

                Mn-52
                Manganese (25)
                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                1.6 × 104

                4.4 × 105

                Mn-53

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                6.8 × 10−5

                1.8 × 10−3

                Mn-54

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                2.9 × 102

                7.7 × 103

                Mn-56

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                8.0 × 105

                2.2 × 107

                Mo-93
                Molybdenum (42)
                4.0 × 101

                1.1 × 103

                2.0 × 101

                5.4 × 102

                4.1 × 10−2

                1.1

                Mo-99 a h

                1.0
                2.7 × 101

                6.0 × 10−1

                1.6 × 101

                1.8 × 104

                4.8 × 105

                N-13
                Nitrogen (7)
                9.0 × 10−1

                2.4 × 101

                6.0 × 10−1

                1.6 × 101

                5.4 × 107

                1.5 × 109

                Na-22
                Sodium (11)
                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                2.3 × 102

                6.3 × 103

                Na-24

                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                3.2 × 105

                8.7 × 106

                Nb-93m
                Niobium (41)
                4.0 × 101

                1.1 × 103

                3.0 × 101

                8.1 × 102

                8.8
                2.4 × 102

                Nb-94

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                6.9 × 10−3

                1.9 × 10−1

                Nb-95

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                1.5 × 103

                3.9 × 104

                Nb-97

                9.0 × 10−1

                2.4 × 101

                6.0 × 10−1

                1.6 × 101

                9.9 × 105

                2.7 × 107

                Nd-147
                Neodymium (60)
                6.0
                1.6 × 102

                6.0 × 10−1

                1.6 × 101

                3.0 × 103

                8.1 × 104

                Nd-149

                6.0 × 10−1

                1.6 × 101

                5.0 × 10−1

                1.4 × 101

                4.5 × 105

                1.2 × 107

                Ni-59
                Nickel (28)
                Unlimited
                Unlimited
                Unlimited
                Unlimited
                3.0 × 10−3

                8.0 × 10−2

                Ni-63

                4.0 × 101

                1.1 × 103

                3.0 × 101

                8.1 × 102

                2.1
                5.7 × 101

                Ni-65

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                7.1 × 105

                1.9 × 107

                Np-235
                Neptunium (93)
                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                5.2 × 101

                1.4 × 103

                Np-236 (short-lived)

                2.0 × 101

                5.4 × 102

                2.0
                5.4 × 101

                4.7 × 10−4

                1.3 × 10−2

                Np-236 (long-lived)

                9.0 × 100

                2.4 × 102

                2.0 × 10−2

                5.4 × 10−1

                4.7 × 10−4

                1.3 × 10−2

                Np-237

                2.0 × 101

                5.4 × 102

                2.0 × 10−3

                5.4 × 10−2

                2.6 × 10−5

                7.1 × 10−4

                Np-239

                7.0
                1.9 × 102

                4.0 × 10−1

                1.1 × 101

                8.6 × 103

                2.3 × 105

                Os-185
                Osmium (76)
                1.0
                2.7 × 101

                1.0
                2.7 × 101

                2.8 × 102

                7.5 × 103

                Os-191

                1.0 × 101

                2.7 × 102

                2.0
                5.4 × 101

                1.6 × 103

                4.4 × 104

                Os-191m

                4.0 × 101

                1.1 × 103

                3.0 × 101

                8.1 × 102

                4.6 × 104

                1.3 × 106

                Os-193

                2.0
                5.4 × 101

                6.0 × 10−1

                1.6 × 101

                2.0 × 104

                5.3 × 105

                Os-194 (a)

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                1.1 × 101

                3.1 × 102

                P-32
                Phosphorus (15)
                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                1.1 × 104

                2.9 × 105

                P-33

                4.0 × 101

                1.1 × 103

                1.0
                2.7 × 101

                5.8 × 103

                1.6 × 105

                Pa-230 (a)
                Protactinium (91)
                2.0
                5.4 × 101

                7.0 × 10−2

                1.9
                1.2 × 103

                3.3 × 104

                Pa-231

                4.0
                1.1 × 102

                4.0 × 10−4

                1.1 × 10−2

                1.7 × 10−3

                4.7 × 10−2

                Pa-233

                5.0
                1.4 × 102

                7.0 × 10−1

                1.9 × 101

                7.7 × 102

                2.1 × 104

                Pb-201
                Lead (82)
                1.0
                2.7 × 101

                1.0
                2.7 × 101

                6.2 × 104

                1.7 × 106

                Pb-202

                4.0 × 101

                1.1 × 103

                2.0 × 101

                5.4 × 102

                1.2 × 10−4

                3.4 × 10−3

                Pb-203

                4.0
                1.1 × 102

                3.0
                8.1 × 101

                1.1 × 104

                3.0 × 105

                Pb-205

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                4.5 × 10−6

                1.2 × 10−4

                Pb-210 (a)

                1.0
                2.7 × 101

                5.0 × 10−2

                1.4
                2.8
                7.6 × 101

                Pb-212 (a)

                7.0 × 10−1

                1.9 × 101

                2.0 × 10−1

                5.4
                5.1 × 104

                1.4 × 106

                Pd-103 (a)
                Palladium (46)
                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                2.8 × 103

                7.5 × 104

                Pd-107

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                1.9 × 10−5

                5.1 × 10−4

                Pd-109

                2.0
                5.4 × 101

                5.0 × 10−1

                1.4 × 101

                7.9 × 104

                2.1 × 106

                Pm-143
                Promethium (61)
                3.0
                8.1 × 101

                3.0
                8.1 × 101

                1.3 × 102

                3.4 × 103

                Pm-144

                7.0 × 10−1

                1.9 × 101

                7.0 × 10−1

                1.9 × 101

                9.2 × 101

                2.5 × 103

                Pm-145

                3.0 × 101

                8.1 × 102

                1.0 × 101

                2.7 × 102

                5.2
                1.4 × 102

                Pm-147

                4.0 × 101

                1.1 × 103

                2.0
                5.4 × 101

                3.4 × 101

                9.3 × 102

                Pm-148m (a)

                8.0 × 10−1

                2.2 × 101

                7.0 × 10−1

                1.9 × 101

                7.9 × 102

                2.1 × 104

                Pm-149

                2.0
                5.4 × 101

                6.0 × 10−1

                1.6 × 101

                1.5 × 104

                4.0 × 105

                Pm-151

                2.0
                5.4 × 101

                6.0 × 10−1

                1.6 × 101

                2.7 × 104

                7.3 × 105

                Po-210
                Polonium (84)
                4.0 × 101

                1.1 × 103

                2.0 × 10−2

                5.4 × 10−1

                1.7 × 102

                4.5 × 103

                Pr-142
                Praseodymium (59)
                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                4.3 × 104

                1.2 × 106

                Pr-143

                3.0
                8.1 × 101

                6.0 × 10−1

                1.6 × 101

                2.5 × 103

                6.7 × 104

                Pt-188 (a)
                Platinum (78)
                1.0
                2.7 × 101

                8.0 × 10−1

                2.2 × 101

                2.5 × 103

                6.8 × 104

                Pt-191

                4.0
                1.1 × 102

                3.0
                8.1 × 101

                8.7 × 103

                2.4 × 105

                Pt-193

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                1.4
                3.7 × 101

                Pt-193m

                4.0 × 101

                1.1 × 103

                5.0 × 10−1

                1.4 × 101

                5.8 × 103

                1.6 × 105

                Pt-195m

                1.0 × 101

                2.7 × 102

                5.0 × 10−1

                1.4 × 101

                6.2 × 103

                1.7 × 105

                Pt-197

                2.0 × 101

                5.4 × 102

                6.0 × 10−1

                1.6 × 101

                3.2 × 104

                8.7 × 105

                Pt-197m

                1.0 × 101

                2.7 × 102

                6.0 × 10−1

                1.6 × 101

                3.7 × 105

                1.0 × 107

                Pu-236
                Plutonium (94)
                3.0 × 101

                8.1 × 102

                3.0 × 10−3

                8.1 × 10−2

                2.0 × 101

                5.3 × 102

                Pu-237

                2.0 × 101

                5.4 × 102

                2.0 × 101

                5.4 × 102

                4.5 × 102

                1.2 × 104

                Pu-238

                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                6.3 × 10−1

                1.7 × 101

                Pu-239

                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                2.3 × 10−3

                6.2 × 10−2

                Pu-240

                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                8.4 × 10−3

                2.3 × 10−1

                Pu-241 (a)

                4.0 × 101

                1.1 × 103

                6.0 × 10−2

                1.6
                3.8
                1.0 × 102

                Pu-242

                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                1.5 × 10−4

                3.9 × 10−3

                Pu-244 (a)

                4.0 × 10−1

                1.1 × 101

                1.0 × 10−3

                2.7 × 10−2

                6.7 × 10−7

                1.8 × 10−5

                Ra-223 (a)
                Radium (88)
                4.0 × 10−1

                1.1 × 101

                7.0 × 10−3

                1.9 × 10−1

                1.9 × 103

                5.1 × 104

                Ra-224 (a)

                4.0 × 10−1

                1.1 × 101

                2.0 × 10−2

                5.4 × 10−1

                5.9 × 103

                1.6 × 105

                Ra-225 (a)

                2.0 × 10−1

                5.4
                4.0 × 10−3

                1.1 × 10−1

                1.5 × 103

                3.9 × 104

                Ra-226 (a)

                2.0 × 10−1

                5.4
                3.0 × 10−3

                8.1 × 10−2

                3.7 × 10−2

                1.0

                Ra-228 (a)

                6.0 × 10−1

                1.6 × 101

                2.0 × 10−2

                5.4 × 10−1

                1.0 × 101

                2.7 × 102

                Rb-81
                Rubidium (37)
                2.0
                5.4 × 101

                8.0 × 10−1

                2.2 × 101

                3.1 × 105

                8.4 × 106

                Rb-83 (a)

                2.0
                5.4 × 101

                2.0
                5.4 × 101

                6.8 × 102

                1.8 × 104

                Rb-84

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                1.8 × 103

                4.7 × 104

                Rb-86

                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                3.0 × 103

                8.1 × 104

                Rb-87

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                3.2 × 10−9

                8.6 × 10−8

                Rb(nat)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                6.7 × 106

                1.8 × 108

                Re-184
                Rhenium (75)
                1.0
                2.7 × 101

                1.0
                2.7 × 101

                6.9 × 102

                1.9 × 104

                Re-184m

                3.0
                8.1 × 101

                1.0
                2.7 × 101

                1.6 × 102

                4.3 × 103

                Re-186

                2.0
                5.4 × 101

                6.0 × 10−1

                1.6 × 101

                6.9 × 103

                1.9 × 105

                Re-187

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                1.4 × 10−9

                3.8 × 10−8

                Re-188

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                3.6 × 104

                9.8 × 105

                Re-189 (a)

                3.0
                8.1 × 101

                6.0 × 10−1

                1.6 × 101

                2.5 × 104

                6.8 × 105

                Re(nat)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                0.0
                2.4 × 10−8

                Rh-99
                Rhodium (45)
                2.0
                5.4 × 101

                2.0
                5.4 × 101

                3.0 × 103

                8.2 × 104

                Rh-101

                4.0
                1.1 × 102

                3.0
                8.1 × 101

                4.1 × 101

                1.1 × 103

                Rh-102

                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                4.5 × 101

                1.2 × 103

                Rh-102m

                2.0
                5.4 × 101

                2.0
                5.4 × 101

                2.3 × 102

                6.2 × 103

                Rh-103m

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                1.2 × 106

                3.3 × 107

                Rh-105

                1.0 × 101

                2.7 × 102

                8.0 × 10−1

                2.2 × 101

                3.1 × 104

                8.4 × 105

                Rn-222 (a)
                Radon (86)
                3.0 × 10−1

                8.1
                4.0 × 10−3

                1.1 × 10−1

                5.7 × 103

                1.5 × 105

                Ru-97
                Ruthenium (44)
                5.0
                1.4 × 102

                5.0
                1.4 × 102

                1.7 × 104

                4.6 × 105

                Ru-103 (a)

                2.0
                5.4 × 101

                2.0
                5.4 × 101

                1.2 × 103

                3.2 × 104

                Ru-105

                1.0
                2.7 × 101

                6.0 × 10−1

                1.6 × 101

                2.5 × 105

                6.7 × 106

                Ru-106 (a)

                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                1.2 × 102

                3.3 × 103

                S-35
                Sulphur (16)
                4.0 × 101

                1.1 × 103

                3.0
                8.1 × 101

                1.6 × 103

                4.3 × 104

                Sb-122
                Antimony (51)
                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                1.5 × 104

                4.0 × 105

                Sb-124

                6.0 × 10−1

                1.6 × 101

                6.0 × 10−1

                1.6 × 101

                6.5 × 102

                1.7 × 104

                Sb-125

                2.0
                5.4 × 101

                1.0
                2.7 × 101

                3.9 × 101

                1.0 × 103

                Sb-126

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                3.1 × 103

                8.4 × 104

                Sc-44
                Scandium (21)
                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                6.7 × 105

                1.8 × 107

                Sc-46

                5.0 × 10−1

                1.4 × 101

                5.0 × 10−1

                1.4 × 101

                1.3 × 103

                3.4 × 104

                Sc-47

                1.0 × 101

                2.7 × 102

                7.0 × 10−1

                1.9 × 101

                3.1 × 104

                8.3 × 105

                Sc-48

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                5.5 × 104

                1.5 × 106

                Se-75
                Selenium (34)
                3.0
                8.1 × 101

                3.0
                8.1 × 101

                5.4 × 102

                1.5 × 104

                Se-79

                4.0 × 101

                1.1 × 103

                2.0
                5.4 × 101

                2.6 × 10−3

                7.0 × 10−2

                Si-31
                Silicon (14)
                6.0 × 10−1

                1.6 × 101

                6.0 × 10−1

                1.6 × 101

                1.4 × 106

                3.9 × 107

                Si-32

                4.0 × 101

                1.1 × 103

                5.0 × 10−1

                1.4 × 101

                3.9
                1.1 × 102

                Sm-145
                Samarium (62)
                1.0 × 101

                2.7 × 102

                1.0 × 101

                2.7 × 102

                9.8 × 101

                2.6 × 103

                Sm-147

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                8.5 × 10−1

                2.3 × 10−8

                Sm-151

                4.0 × 101

                1.1 × 103

                1.0 × 101

                2.7 × 102

                9.7 × 10−1

                2.6 × 101

                Sm-153

                9.0
                2.4 × 102

                6.0 × 10−1

                1.6 × 101

                1.6 × 104

                4.4 × 105

                Sn-113 (a)
                Tin (50)
                4.0
                1.1 × 102

                2.0
                5.4 × 101

                3.7 × 102

                1.0 × 104

                Sn-117m

                7.0
                1.9 × 102

                4.0 × 10−1

                1.1 × 101

                3.0 × 103

                8.2 × 104

                Sn-119m

                4.0 × 101

                1.1 × 103

                3.0 × 101

                8.1 × 102

                1.4 × 102

                3.7 × 103

                Sn-121m (a)

                4.0 × 101

                1.1 × 103

                9.0 × 10−1

                2.4 × 101

                2.0
                5.4 × 101

                Sn-123

                8.0 × 10−1

                2.2 × 101

                6.0 × 10−1

                1.6 × 101

                3.0 × 102

                8.2 × 103

                Sn-125

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                4.0 × 103

                1.1 × 105

                Sn-126 (a)

                6.0 × 10−1

                1.6 × 101

                4.0 × 10−1

                1.1 × 101

                1.0 × 10−3

                2.8 × 10−2

                Sr-82 (a)
                Strontium (38)
                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                2.3 × 103

                6.2 × 104

                Sr-85

                2.0
                5.4 × 101

                2.0
                5.4 × 101

                8.8 × 102

                2.4 × 104

                Sr-85m

                5.0
                1.4 × 102

                5.0
                1.4 × 102

                1.2 × 106

                3.3 × 107

                Sr-87m

                3.0
                8.1 × 101

                3.0
                8.1 × 101

                4.8 × 105

                1.3 × 107

                Sr-89

                6.0 × 10−1

                1.6 × 101

                6.0 × 10−1

                1.6 × 101

                1.1 × 103

                2.9 × 104

                Sr-90 (a)

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                5.1
                1.4 × 102

                Sr-91 (a)

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                1.3 × 105

                3.6 × 106

                Sr-92 (a)

                1.0
                2.7 × 101

                3.0 × 10−1

                8.1
                4.7 × 105

                1.3 × 107

                T(H-3)
                Tritium (1)
                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                3.6 × 102

                9.7 × 103

                Ta-178 (long-lived)
                Tantalum (73)
                1.0
                2.7 × 101

                8.0 × 10−1

                2.2 × 101

                4.2 × 106

                1.1 × 108

                Ta-179

                3.0 × 101

                8.1 × 102

                3.0 × 101

                8.1 × 102

                4.1 × 101

                1.1 × 103

                Ta-182

                9.0 × 10−1

                2.4 × 101

                5.0 × 10−1

                1.4 × 101

                2.3 × 102

                6.2 × 103

                Tb-157
                Terbium (65)
                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                5.6 × 10−1

                1.5 × 101

                Tb-158

                1.0
                2.7 × 101

                1.0
                2.7 × 101

                5.6 × 10−1

                1.5 × 101

                Tb-160

                1.0
                2.7 × 101

                6.0 × 10−1

                1.6 × 101

                4.2 × 102

                1.1 × 104

                Tc-95m (a)
                Technetium (43)
                2.0
                5.4 × 101

                2.0
                5.4 × 101

                8.3 × 102

                2.2 × 104

                Tc-96

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                1.2 × 104

                3.2 × 105

                Tc-96m (a)

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                1.4 × 106

                3.8 × 107

                Tc-97

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                5.2 × 10−5

                1.4 × 10−3

                Tc-97m

                4.0 × 101

                1.1 × 103

                1.0
                2.7 × 101

                5.6 × 102

                1.5 × 104

                Tc-98

                8.0 × 10−1

                2.2 × 101

                7.0 × 10−1

                1.9 × 101

                3.2 × 10−5

                8.7 × 10−4

                Tc-99

                4.0 × 101

                1.1 × 103

                9.0 × 10−1

                2.4 × 101

                6.3 × 10−4

                1.7 × 10−2

                Tc-99m

                1.0 × 101

                2.7 × 102

                4.0
                1.1 × 102

                1.9 × 105

                5.3 × 106

                Te-121
                Tellurium (52)
                2.0
                5.4 × 101

                2.0
                5.4 × 101

                2.4 × 103

                6.4 × 104

                Te-121m

                5.0
                1.4 × 102

                3.0
                8.1 × 101

                2.6 × 102

                7.0 × 103

                Te-123m

                8.0
                2.2 × 102

                1.0
                2.7 × 101

                3.3 × 102

                8.9 × 103

                Te-125m

                2.0 × 101

                5.4 × 102

                9.0 × 10−1

                2.4 × 101

                6.7 × 102

                1.8 × 104

                Te-127

                2.0 × 101

                5.4 × 102

                7.0 × 10−1

                1.9 × 101

                9.8 × 104

                2.6 × 106

                Te-127m (a)

                2.0 × 101

                5.4 × 102

                5.0 × 10−1

                1.4 × 101

                3.5 × 102

                9.4 × 103

                Te-129

                7.0 × 10−1

                1.9 × 101

                6.0 × 10−1

                1.6 × 101

                7.7 × 105

                2.1 × 107

                Te-129m (a)

                8.0 × 10−1

                2.2 × 101

                4.0 × 10−1

                1.1 × 101

                1.1 × 103

                3.0 × 104

                Te-131m (a)

                7.0 × 10−1

                1.9 × 101

                5.0 × 10−1

                1.4 × 101

                3.0 × 104

                8.0 × 105

                Te-132 (a)

                5.0 × 10−1

                1.4 × 101

                4.0 × 10−1

                1.1 × 101

                1.1 × 104

                3.0 × 105

                Th-227
                Thorium (90)
                1.0 × 101

                2.7 × 102

                5.0 × 10−3

                1.4 × 10−1

                1.1 × 103

                3.1 × 104

                Th-228 (a)

                5.0 × 10−1

                1.4 × 101

                1.0 × 10−3

                2.7 × 10−2

                3.0 × 101

                8.2 × 102

                Th-229

                5.0
                1.4 × 102

                5.0 × 10−4

                1.4 × 10−2

                7.9 × 10−3

                2.1 × 10−1

                Th-230

                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                7.6 × 10−4

                2.1 × 10−2

                Th-231

                4.0 × 101

                1.1 × 103

                2.0 × 10−2

                5.4 × 10−1

                2.0 × 104

                5.3 × 105

                Th-232

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                4.0 × 10−9

                1.1 × 10−7

                Th-234 (a)

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                8.6 × 102

                2.3 × 104

                Th(nat)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                8.1 × 10−9

                2.2 × 10−7

                Ti-44 (a)
                Titanium (22)
                5.0 × 10−1

                1.4 × 101

                4.0 × 10−1

                1.1 × 101

                6.4
                1.7 × 102

                Tl-200
                Thallium (81)
                9.0 × 10−1

                2.4 × 101

                9.0 × 10−1

                2.4 × 101

                2.2 × 104

                6.0 × 105

                Tl-201

                1.0 × 101

                2.7 × 102

                4.0
                1.1 × 102

                7.9 × 103

                2.1 × 105

                Tl-202

                2.0
                5.4 × 101

                2.0
                5.4 × 101

                2.0 × 103

                5.3 × 104

                Tl-204

                1.0 × 101

                2.7 × 102

                7.0 × 10−1

                1.9 × 101

                1.7 × 101

                4.6 × 102

                Tm-167
                Thulium (69)
                7.0
                1.9 × 102

                8.0 × 10−1

                2.2 × 101

                3.1 × 103

                8.5 × 104

                Tm-170

                3.0
                8.1 × 101

                6.0 × 10−1

                1.6 × 101

                2.2 × 102

                6.0 × 103

                Tm-171

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                U-230 (fast lung absorption) (a)(d)
                Uranium (92)
                4.0 × 101

                1.1 × 103

                1.0 × 10−1

                2.7
                1.0 × 103

                2.7 × 104

                U-230 (medium lung absorption) (a)(e)

                4.0 × 101

                1.1 × 103

                4.0 × 10−3

                1.1 × 10−1

                1.0 × 103

                2.7 × 104

                U-230 (slow lung absorption) (a)(f)

                3.0 × 101

                8.1 × 102

                3.0 × 10−3

                8.1 × 10−2

                1.0 × 103

                2.7 × 104

                U-232 (fast lung absorption) (d)

                4.0 × 101

                1.1 × 103

                1.0 × 10−2

                2.7 × 10−1

                8.3 × 10−1

                2.2 × 101

                U-232 (medium lung absorption) (e)

                4.0 × 101

                1.1 × 103

                7.0 × 10−3

                1.9 × 10−1

                8.3 × 10−1

                2.2 × 101

                U-232 (slow lung absorption) (f)

                1.0 × 101

                2.7 × 102

                1.0 × 10−3

                2.7 × 10−2

                8.3 × 10−1

                2.2 × 101

                U-233 (fast lung absorption) (d)

                4.0 × 101

                1.1 × 103

                9.0 × 10−2

                2.4
                3.6 × 10−4

                9.7 × 10−3

                U-233 (medium lung absorption) (e)

                4.0 × 101

                1.1 × 103

                2.0 × 10−2

                5.4 × 10−1

                3.6 × 10−4

                9.7 × 10−3

                U-233 (slow lung absorption) (f)

                4.0 × 101

                1.1 × 103

                6.0 × 10−3

                1.6 × 10−1

                3.6 × 10−4

                9.7 × 10−3

                U-234 (fast lung absorption) (d)

                4.0 × 101

                1.1 × 103

                9.0 × 10−2

                2.4
                2.3 × 10−4

                6.2 × 10−3

                U-234 (medium lung absorption) (e)

                4.0 × 101

                1.1 × 103

                2.0 × 10−2

                5.4 × 10−1

                2.3 × 10−4

                6.2 × 10−3

                U-234 (slow lung absorption) (f)

                4.0 × 101

                1.1 × 103

                6.0 × 10−3

                1.6 × 10−1

                2.3 × 10−4

                6.2 × 10−3

                U-235 (all lung absorption types) (a),(d),(e),(f)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                8.0 × 10−8

                2.2 × 10−6

                U-236 (fast lung absorption) (d)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                2.4 × 10−6

                6.5 × 10−5

                U-236 (medium lung absorption) (e)

                4.0 × 101

                1.1 × 103

                2.0 × 10−2

                5.4 × 10−1

                2.4 × 10−6

                6.5 × 10−5

                U-236 (slow lung absorption) (f)

                4.0 × 101

                1.1 × 103

                6.0 × 10−3

                1.6 × 10−1

                2.4 × 10−6

                6.5 × 10−5

                U-238 (all lung absorption types) (d),(e),(f)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                1.2 × 10−8

                3.4 × 10−7

                U (nat)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                2.6 × 10−8

                7.1 × 10−7

                U (enriched to 20% or less)(g)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                See Table A-4
                See Table A-4

                U (dep)

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                See Table A-4
                See Table A-4

                V-48
                Vanadium (23)
                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                6.3 × 103

                1.7 × 105

                V-49

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                3.0 × 102

                8.1 × 103

                W-178 (a)
                Tungsten (74)
                9.0
                2.4 × 102

                5.0
                1.4 × 102

                1.3 × 103

                3.4 × 104

                W-181

                3.0 × 101

                8.1 × 102

                3.0 × 101

                8.1 × 102

                2.2 × 102

                6.0 × 103

                W-185

                4.0 × 101

                1.1 × 103

                8.0 × 10−1

                2.2 × 101

                3.5 × 102

                9.4 × 103

                W-187

                2.0
                5.4 × 101

                6.0 × 10−1

                1.6 × 101

                2.6 × 104

                7.0 × 105

                W-188 (a)

                4.0 × 10−1

                1.1 × 101

                3.0 × 10−1

                8.1
                3.7 × 102

                1.0 × 104

                Xe-122 (a)
                Xenon (54)
                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                4.8 × 104

                1.3 × 106

                Xe-123

                2.0
                5.4 × 101

                7.0 × 10−1

                1.9 × 101

                4.4 × 105

                1.2 × 107

                Xe-127

                4.0
                1.1 × 102

                2.0
                5.4 × 101

                1.0 × 103

                2.8 × 104

                Xe-131m

                4.0 × 101

                1.1 × 103

                4.0 × 101

                1.1 × 103

                3.1 × 103

                8.4 × 104

                Xe-133

                2.0 × 101

                5.4 × 102

                1.0 × 101

                2.7 × 102

                6.9 × 103

                1.9 × 105

                Xe-135

                3.0
                8.1 × 101

                2.0
                5.4 × 101

                9.5 × 104

                2.6 × 106

                Y-87 (a)
                Yttrium (39)
                1.0
                2.7 × 101

                1.0
                2.7 × 101

                1.7 × 104

                4.5 × 105

                Y-88

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                5.2 × 102

                1.4 × 104

                Y-90

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                2.0 × 104

                5.4 × 105

                Y-91

                6.0 × 10−1

                1.6 × 101

                6.0 × 10−1

                1.6 × 101

                9.1 × 102

                2.5 × 104

                Y-91m

                2.0
                5.4 × 101

                2.0
                5.4 × 101

                1.5 × 106

                4.2 × 107

                Y-92

                2.0 × 10−1

                5.4
                2.0 × 10−1

                5.4
                3.6 × 105

                9.6 × 106

                Y-93

                3.0 × 10−1

                8.1
                3.0 × 10−1

                8.1
                1.2 × 105

                3.3 × 106

                Yb-169
                Ytterbium (70)
                4.0
                1.1 × 102

                1.0
                2.7 × 101

                8.9 × 102

                2.4 × 104

                Yb-175

                3.0 × 101

                8.1 × 102

                9.0 × 10−1

                2.4 × 101

                6.6 × 103

                1.8 × 105

                Zn-65
                Zinc (30)
                2.0
                5.4 × 101

                2.0
                5.4 × 101

                3.0 × 102

                8.2 × 103

                Zn-69

                3.0
                8.1 × 101

                6.0 × 10−1

                1.6 × 101

                1.8 × 106

                4.9 × 107

                Zn-69m (a)

                3.0
                8.1 × 101

                6.0 × 10−1

                1.6 × 101

                1.2 × 105

                3.3 × 106

                Zr-88
                Zirconium (40)
                3.0
                8.1 × 101

                3.0
                8.1 × 101

                6.6 × 102

                1.8 × 104

                Zr-93

                Unlimited
                Unlimited
                Unlimited
                Unlimited
                9.3 × 10−5

                2.5 × 10−3

                Zr-95 (a)

                2.0
                5.4 × 101

                8.0 × 10−1

                2.2 × 101

                7.9 × 102

                2.1 × 104

                Zr-97 (a)

                4.0 × 10−1

                1.1 × 101

                4.0 × 10−1

                1.1 × 101

                7.1 × 104

                1.9 × 106

                a A1 and/or A2 values include contributions from daughter nuclides with half-lives less than 10 days, as listed in the following:

                Mg-28
                Al-28

                Ca-47
                Sc-47

                Ti-44
                Sc-44

                Fe-52
                Mn-52m

                Fe-60
                Co-60m

                Zn-69m
                Zn-69

                Ge-68
                Ga-68

                Rb-83
                Kr-83m

                Sr-82
                Rb-82

                Sr-90
                Y-90

                Sr-91
                Y-91m

                Sr-92
                Y-92

                Y-87
                Sr-87m

                Zr-95
                Nb-95m

                Zr-97
                Nb-97m, Nb-97

                Mo-99
                Tc-99m

                Tc-95m
                Tc-95

                Tc-96m
                Tc-96

                Ru-103
                Rh-103m

                Ru-106
                Rh-106

                Pd-103
                Rh-103m

                Ag-108m
                Ag-108

                Ag-110m
                Ag-110

                Cd-115
                In-115m

                In-114m
                In-114

                Sn-113
                In-113m

                Sn-121m
                Sn-121

                Sn-126
                Sb-126m

                Te-127m
                Te-127

                Te-129m
                Te-129

                Te-131m
                Te-131

                Te-132
                I-132

                I-135
                Xe-135m

                Xe-122
                I-122

                Cs-137
                Ba-137m

                Ba-131
                Cs-131

                Ba-140
                La-140

                Ce-144
                Pr-144m, Pr-144

                Pm-148m
                Pm-148

                Gd-146
                Eu-146

                Dy-166
                Ho-166

                Hf-172
                Lu-172

                W-178
                Ta-178

                W-188
                Re-188

                Re-189
                Os-189m

                Os-194
                Ir-194

                Ir-189
                Os-189m

                Pt-188
                Ir-188

                Hg-194
                Au-194

                Hg-195m
                Hg-195

                Pb-210
                Bi-210

                Pb-212
                Bi-212, Tl-208, Po-212

                Bi-210m
                Tl-206

                Bi-212
                Tl-208, Po-212

                At-211
                Po-211

                Rn-222
                Po-218, Pb-214, At-218, Bi-214, Po-214

                Ra-223
                Rn-219, Po-215, Pb-211, Bi-211, Po-211, Tl-207

                Ra-224
                Rn-220, Po-216, Pb-212, Bi-212, Tl-208, Po-212

                Ra-225
                Ac-225, Fr-221, At-217, Bi-213, Tl-209, Po-213, Pb-209

                Ra-226
                Rn-222, Po-218, Pb-214, At-218, Bi-214, Po-214

                Ra-228
                Ac-228

                Ac-225
                Fr-221, At-217, Bi-213, Tl-209, Po-213, Pb-209

                Ac-227
                Fr-223

                Th-228
                Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208, Po-212

                Th-234
                Pa-234m, Pa-234

                Pa-230
                Ac-226, Th-226, Fr-222, Ra-222, Rn-218, Po-214

                U-230
                Th-226, Ra-222, Rn-218, Po-214

                U-235
                Th-231

                Pu-241
                U-237

                Pu-244
                U-240, Np-240m

                Am-242m
                Am-242, Np-238

                Am-243
                Np-239

                Cm-247
                Pu-243

                Bk-249
                Am-245

                Cf-253
                Cm-249

                b The values of A1 and A2 in Curies (Ci) are approximate and for information only; the regulatory standard units are Terabecquerels (TBq) (see Appendix A to Part 71—Determination of A1 and A2, Section I).

                c The activity of Ir-192 in special form may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the source.

                d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.

                e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.

                f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.

                g These values apply to unirradiated uranium only.

                h A2 = 0.74 TBq (20 Ci) for Mo-99 for domestic use.

              Table A-2—Exempt Material Activity Concentrations and Exempt Consignment Activity Limits for Radionuclides

                Symbol ofradionuclide

                Element and atomic number
                Activity concentration for exempt material(Bq/g)

                Activity concentration for exempt material(Ci/g)

                Activity limit for exempt consignment(Bq)

                Activity limit for exempt consignment(Ci)

                Ac-225
                Actinium (89)
                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Ac-227

                1.0 × 10−1

                2.7 × 10−12

                1.0 × 103

                2.7 × 10−8

                Ac-228

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Ag-105
                Silver (47)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ag-108m (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Ag-110m

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Ag-111

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Al-26
                Aluminum (13)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Am-241
                Americium (95)
                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Am-242m (b)

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Am-243 (b)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Ar-37
                Argon (18)
                1.0 × 106

                2.7 × 10−5

                1.0 × 108

                2.7 × 10−3

                Ar-39

                1.0 × 107

                2.7 × 10−4

                1.0 × 104

                2.7 × 10−7

                Ar-41

                1.0 × 102

                2.7 × 10−9

                1.0 × 109

                2.7 × 10−2

                As-72
                Arsenic (33)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                As-73

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                As-74

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                As-76

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                As-77

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                At-211
                Astatine (85)
                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Au-193
                Gold (79)
                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Au-194

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Au-195

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Au-198

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Au-199

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ba-131
                Barium (56)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ba-133

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ba-133m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ba-140 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Be-7
                Beryllium (4)
                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Be-10

                1.0 × 104

                2.7 × 10−7

                1.0 × 106

                2.7 × 10−5

                Bi-205
                Bismuth (83)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Bi-206

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Bi-207

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Bi-210

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Bi-210m

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Bi-212 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Bk-247
                Berkelium (97)
                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Bk-249

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Br-76
                Bromine (35)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Br-77

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Br-82

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                C-11
                Carbon (6)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                C-14

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Ca-41
                Calcium (20)
                1.0 × 105

                2.7 × 10−6

                1.0 × 107

                2.7 × 10−4

                Ca-45

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Ca-47

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Cd-109
                Cadmium (48)
                1.0 × 104

                2.7 × 10−7

                1.0 × 106

                2.7 × 10−5

                Cd-113m

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Cd-115

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Cd-115m

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Ce-139
                Cerium (58)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ce-141

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Ce-143

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ce-144 (b)

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Cf-248
                Californium (98)
                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Cf-249

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Cf-250

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Cf-251

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Cf-252

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Cf-253

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Cf-254

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Cl-36
                Chlorine (17)
                1.0 × 104

                2.7 × 10−7

                1.0 × 106

                2.7 × 10−5

                Cl-38

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Cm-240
                Curium (96)
                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Cm-241

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Cm-242

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Cm-243

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Cm-244

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Cm-245

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Cm-246

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Cm-247

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Cm-248

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Co-55
                Cobalt (27)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Co-56

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Co-57

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Co-58

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Co-58m

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Co-60

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Cr-51
                Chromium (24)
                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Cs-129
                Cesium (55)
                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Cs-131

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Cs-132

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Cs-134

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Cs-134m

                1.0 × 103

                2.7 × 10−8

                1.0 × 105

                2.7 × 10−6

                Cs-135

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Cs-136

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Cs-137 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Cu-64
                Copper (29)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Cu-67

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Dy-159
                Dysprosium (66)
                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Dy-165

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Dy-166

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Er-169
                Erbium (68)
                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Er-171

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Eu-147
                Europium (63)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Eu-148

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Eu-149

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Eu-150 (short lived)

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Eu-150 (long lived)

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Eu-152

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Eu-152m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Eu-154

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Eu-155

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Eu-156

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                F-18
                Fluorine (9)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Fe-52
                Iron (26)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Fe-55

                1.0 × 104

                2.7 × 10−7

                1.0 × 106

                2.7 × 10−5

                Fe-59

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Fe-60

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Ga-67
                Gallium (31)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ga-68

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Ga-72

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Gd-146
                Gadolinium (64)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Gd-148

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Gd-153

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Gd-159

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Ge-68
                Germanium (32)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Ge-71

                1.0 × 104

                2.7 × 10−7

                1.0 × 108

                2.7 × 10−3

                Ge-77

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Hf-172
                Hafnium (72)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Hf-175

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Hf-181

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Hf-182

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Hg-194
                Mercury (80)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Hg-195m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Hg-197

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Hg-197m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Hg-203

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Ho-166
                Holmium (67)
                1.0 × 103

                2.7 × 10−8

                1.0 × 105

                2.7 × 10−6

                Ho-166m

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                I-123
                Iodine (53)
                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                I-124

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                I-125

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                I-126

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                I-129

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                I-131

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                I-132

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                I-133

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                I-134

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                I-135

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                In-111
                Indium (49)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                In-113m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                In-114m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                In-115m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ir-189
                Iridium (77)
                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Ir-190

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Ir-192

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Ir-194

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                K-40
                Potassium (19)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                K-42

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                K-43

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Kr-79
                Krypton (36)
                1.0 × 103

                2.7 × 10−8

                1.0 × 105

                2.7 × 10−6

                Kr-81

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Kr-85

                1.0 × 105

                2.7 × 10−6

                1.0 × 104

                2.7 × 10−7

                Kr-85m

                1.0 × 103

                2.7 × 10−8

                1.0 × 1010

                2.7 × 10−1

                Kr-87

                1.0 × 102

                2.7 × 10−9

                1.0 × 109

                2.7 × 10−2

                La-137
                Lanthanum (57)
                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                La-140

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Lu-172
                Lutetium (71)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Lu-173

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Lu-174

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Lu-174m

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Lu-177

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Mg-28
                Magnesium (12)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Mn-52
                Manganese (25)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Mn-53

                1.0 × 104

                2.7 × 10−7

                1.0 × 109

                2.7 × 10−2

                Mn-54

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Mn-56

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Mo-93
                Molybdenum (42)
                1.0 × 103

                2.7 × 10−8

                1.0 × 108

                2.7 × 10−3

                Mo-99

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                N-13
                Nitrogen (7)
                1.0 × 102

                2.7 × 10−9

                1.0 × 109

                2.7 × 10−2

                Na-22
                Sodium (11)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Na-24

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Nb-93m
                Niobium (41)
                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Nb-94

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Nb-95

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Nb-97

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Nd-147
                Neodymium (60)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Nd-149

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ni-59
                Nickel (28)
                1.0 × 104

                2.7 × 10−7

                1.0 × 108

                2.7 × 10−3

                Ni-63

                1.0 × 105

                2.7 × 10−6

                1.0 × 108

                2.7 × 10−3

                Ni-65

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Np-235
                Neptunium (93)
                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Np-236 (short-lived)

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Np-236 (long-lived)

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Np-237 (b)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Np-239

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Os-185
                Osmium (76)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Os-191

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Os-191m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Os-193

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Os-194

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                P-32
                Phosphorus (15)
                1.0 × 103

                2.7 × 10−8

                1.0 × 105

                2.7 × 10−6

                P-33

                1.0 × 105

                2.7 × 10−6

                1.0 × 108

                2.7 × 10−3

                Pa-230
                Protactinium (91)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Pa-231

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Pa-233

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Pb-201
                Lead (82)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Pb-202

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Pb-203

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Pb-205

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Pb-210 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Pb-212 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Pd-103
                Palladium (46)
                1.0 × 103

                2.7 × 10−8

                1.0 × 108

                2.7 × 10−3

                Pd-107

                1.0 × 105

                2.7 × 10−6

                1.0 × 108

                2.7 × 10−3

                Pd-109

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Pm-143
                Promethium (61)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Pm-144

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Pm-145

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Pm-147

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Pm-148m

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Pm-149

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Pm-151

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Po-210
                Polonium (84)
                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Pr-142
                Praseodymium (59)
                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Pr-143

                1.0 × 104

                2.7 × 10−7

                1.0 × 106

                2.7 × 10−5

                Pt-188
                Platinum (78)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Pt-191

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Pt-193

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Pt-193m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Pt-195m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Pt-197

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Pt-197m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Pu-236
                Plutonium (94)
                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Pu-237

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Pu-238

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Pu-239

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Pu-240

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Pu-241

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Pu-242

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Pu-244

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Ra-223 (b)
                Radium (88)
                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Ra-224 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Ra-225

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Ra-226 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Ra-228 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Rb-81
                Rubidium (37)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Rb-83

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Rb-84

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Rb-86

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Rb-87

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Rb(nat)

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Re-184
                Rhenium (75)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Re-184m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Re-186

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Re-187

                1.0 × 106

                2.7 × 10−5

                1.0 × 109

                2.7 × 10−2

                Re-188

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Re-189

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Re(nat)

                1.0 × 106

                2.7 × 10−5

                1.0 × 109

                2.7 × 10−2

                Rh-99
                Rhodium (45)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Rh-101

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Rh-102

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Rh-102m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Rh-103m

                1.0 × 104

                2.7 × 10−7

                1.0 × 108

                2.7 × 10−3

                Rh-105

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Rn-222 (b)
                Radon (86)
                1.0 × 101

                2.7 × 10−10

                1.0 × 108

                2.7 × 10−3

                Ru-97
                Ruthenium (44)
                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Ru-103

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Ru-105

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Ru-106 (b)

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                S-35
                Sulphur (16)
                1.0 × 105

                2.7 × 10−6

                1.0 × 108

                2.7 × 10−3

                Sb-122
                Antimony (51)
                1.0 × 102

                2.7 × 10−9

                1.0 × 104

                2.7 × 10−7

                Sb-124

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Sb-125

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Sb-126

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Sc-44
                Scandium (21)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Sc-46

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Sc-47

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Sc-48

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Se-75
                Selenium (34)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Se-79

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Si-31
                Silicon (14)
                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Si-32

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Sm-145
                Samarium (62)
                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Sm-147

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Sm-151

                1.0 × 104

                2.7 × 10−7

                1.0 × 108

                2.7 × 10−3

                Sm-153

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Sn-113
                Tin (50)
                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Sn-117m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Sn-119m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Sn-121m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Sn-123

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Sn-125

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Sn-126

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Sr-82
                Strontium (38)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Sr-85

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Sr-85m

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Sr-87m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Sr-89

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Sr-90 (b)

                1.0 × 102

                2.7 × 10−9

                1.0 × 104

                2.7 × 10−7

                Sr-91

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Sr-92

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                T(H-3)
                Tritium (1)
                1.0 × 106

                2.7 × 10−5

                1.0 × 109

                2.7 × 10−2

                Ta-178 (long-lived)
                Tantalum (73)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Ta-179

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Ta-182

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Tb-157
                Terbium (65)
                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Tb-158

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Tb-160

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Tc-95m
                Technetium (43)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Tc-96

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Tc-96m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Tc-97

                1.0 × 103

                2.7 × 10−8

                1.0 × 108

                2.7 × 10−3

                Tc-97m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Tc-98

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Tc-99

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                Tc-99m

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Te-121
                Tellurium (52)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Te-121m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Te-123m

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Te-125m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Te-127

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Te-127m

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Te-129

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Te-129m

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Te-131m

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Te-132

                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Th-227
                Thorium (90)
                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Th-228 (b)

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Th-229 (b)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Th-230

                1.0
                2.7 × 10−11

                1.0 × 104

                2.7 × 10−7

                Th-231

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Th-232

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                Th-234 (b)

                1.0 × 103

                2.7 × 10−8

                1.0 × 105

                2.7 × 10−6

                Th (nat) (b)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                Ti-44
                Titanium (22)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                Tl-200
                Thallium (81)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Tl-201

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Tl-202

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Tl-204

                1.0 × 104

                2.7 × 10−7

                1.0 × 104

                2.7 × 10−7

                Tm-167
                Thulium (69)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Tm-170

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Tm-171

                1.0 × 104

                2.7 × 10−7

                1.0 × 108

                2.7 × 10−3

                U-230 (fast lung absorption) (b),(d)
                Uranium (92)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                U-230 (medium lung absorption) (e)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-230 (slow lung absorption) (f)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-232 (fast lung absorption) (b),(d)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                U-232 (medium lung absorption) (e)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-232 (slow lung absorption) (f)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-233 (fast lung absorption) (d)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-233 (medium lung absorption) (e)

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                U-233 (slow lung absorption) (f)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                U-234 (fast lung absorption) (d)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-234 (medium lung absorption) (e)

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                U-234 (slow lung absorption) (f)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                U-235 (all lung absorption types) (b),(d),(e),(f)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-236 (fast lung absorption) (d)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-236 (medium lung absorption) (e)

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                U-236 (slow lung absorption) (f)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U-238 (all lung absorption types) (b),(d),(e),(f)

                1.0 × 101

                2.7 × 10−10

                1.0 × 104

                2.7 × 10−7

                U (nat) (b)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                U (enriched to 20% or less)(g)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                U (dep)

                1.0
                2.7 × 10−11

                1.0 × 103

                2.7 × 10−8

                V-48
                Vanadium (23)
                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                V-49

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                W-178
                Tungsten (74)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                W-181

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                W-185

                1.0 × 104

                2.7 × 10−7

                1.0 × 107

                2.7 × 10−4

                W-187

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                W-188

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Xe-122
                Xenon (54)
                1.0 × 102

                2.7 × 10−9

                1.0 × 109

                2.7 × 10−2

                Xe-123

                1.0 × 102

                2.7 × 10−9

                1.0 × 109

                2.7 × 10−2

                Xe-127

                1.0 × 103

                2.7 × 10−8

                1.0 × 105

                2.7 × 10−6

                Xe-131m

                1.0 × 104

                2.7 × 10−7

                1.0 × 104

                2.7 × 10−7

                Xe-133

                1.0 × 103

                2.7 × 10−8

                1.0 × 104

                2.7 × 10−7

                Xe-135

                1.0 × 103

                2.7 × 10−8

                1.0 × 1010

                2.7 × 10−1

                Y-87
                Yttrium (39)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Y-88

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Y-90

                1.0 × 103

                2.7 × 10−8

                1.0 × 105

                2.7 × 10−6

                Y-91

                1.0 × 103

                2.7 × 10−8

                1.0 × 106

                2.7 × 10−5

                Y-91m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Y-92

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Y-93

                1.0 × 102

                2.7 × 10−9

                1.0 × 105

                2.7 × 10−6

                Yb-169
                Ytterbium (70)
                1.0 × 102

                2.7 × 10−9

                1.0 × 107

                2.7 × 10−4

                Yb-175

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Zn-65
                Zinc (30)
                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Zn-69

                1.0 × 104

                2.7 × 10−7

                1.0 × 106

                2.7 × 10−5

                Zn-69m

                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Zr-88
                Zirconium (40)
                1.0 × 102

                2.7 × 10−9

                1.0 × 106

                2.7 × 10−5

                Zr-93 (b)

                1.0 × 103

                2.7 × 10−8

                1.0 × 107

                2.7 × 10−4

                Zr-95

                1.0 × 101

                2.7 × 10−10

                1.0 × 106

                2.7 × 10−5

                Zr-97 (b)

                1.0 × 101

                2.7 × 10−10

                1.0 × 105

                2.7 × 10−6

                a [Reserved]

                b Parent nuclides and their progeny included in secular equilibrium are listed as follows:

                Sr-90
                Y-90

                Zr-93
                Nb-93m

                Zr-97
                Nb-97

                Ru-106
                Rh-106

                Ag-108m
                Ag-108

                Cs-137
                Ba-137m

                Ce-144
                Pr-144

                Ba-140
                La-140

                Bi-212
                Tl-208 (0.36), Po-212 (0.64)

                Pb-210
                Bi-210, Po-210

                Pb-212
                Bi-212, Tl-208 (0.36), Po-212 (0.64)

                Rn-222
                Po-218, Pb-214, Bi-214, Po-214

                Ra-223
                Rn-219, Po-215, Pb-211, Bi-211, Tl-207

                Ra-224
                Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)

                Ra-226
                Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210

                Ra-228
                Ac-228

                Th-228
                Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212(0.64)

                Th-229
                Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209

                Th-nat
                Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)

                Th-234
                Pa-234m

                U-230
                Th-226, Ra-222, Rn-218, Po-214

                U-232
                Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)

                U-235
                Th-231

                U-238
                Th-234, Pa-234m

                U-nat
                Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210

                Np-237
                Pa-233

                Am-242m
                Am-242

                Am-243
                Np-239

                c [Reserved]

                d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.

                e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.

                f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.

                g These values apply to unirradiated uranium only.

              Table A-3—General Values for A1 and A2

                Contents
                A1

                (TBq)
                (Ci)
                A2

                (TBq)
                (Ci)
                Activity concentration for exempt material(Bq/g)

                Activity concentration for exempt material(Ci/g)

                Activity limits for exempt consignments(Bq)

                Activity limits for exempt consignments(Ci)

                Only beta or gamma emitting radionuclides are known to be present
                1 × 10−1

                2.7 × 100

                2 × 10−2

                5.4 × 10−1

                1 × 101

                2.7 × 10−10

                1 × 104

                2.7 × 10−7

                Alpha emitting nuclides, but no neutron emitters, are known to be present (a)
                2 × 10−1

                5.4 × 100

                9 × 10−5

                2.4 × 10−3

                1 × 10−1

                2.7 × 10−12

                1 × 103

                2.7 × 10−8

                Neutron emitting nuclides are known to be present or no relevant data are available
                1 × 10−3

                2.7 × 10−2

                9 × 10−5

                2.4 × 10−3

                1 × 10−1

                2.7 × 10−12

                1 × 103

                2.7 × 10−8

                a If beta or gamma emitting nuclides are known to be present, the A1 value of 0.1 TBq (2.7 Ci) should be used.

              Table A-4—Activity-mass Relationships for Uranium

                Uranium Enrichment 1 wt % U-235 present
                Specific Activity
                TBq/g
                Ci/g

                0.45
                1.8 × 10−8

                5.0 × 10−7

                0.72
                2.6 × 10−8

                7.1 × 10−7

                1
                2.8 × 10−8

                7.6 × 10−7

                1.5
                3.7 × 10−8

                1.0 × 10−6

                5
                1.0 × 10−7

                2.7 × 10−6

                10
                1.8 × 10−7

                4.8 × 10−6

                20
                3.7 × 10−7

                1.0 × 10−5

                35
                7.4 × 10−7

                2.0 × 10−5

                50
                9.3 × 10−7

                2.5 × 10−5

                90
                2.2 × 10−6

                5.8 × 10−5

                93
                2.6 × 10−6

                7.0 × 10−5

                95
                3.4 × 10−6

                9.1 × 10−5

                1 The figures for uranium include representative values for the activity of the uranium-234 that is concentrated during the enrichment process.

            [69 FR 3800, Jan. 26, 2004; 69 FR 58039, Sept. 29, 2004, as amended at 77 FR 39908, July 6, 2012; 80 FR 34014, June 12, 2015]

        Pt. 72
        PART 72—LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE

            Subpart A—General Provisions
            Sec.
            72.1
            Purpose.
            72.2
            Scope.
            72.3
            Definitions.
            72.4
            Communications.
            72.5
            Interpretations.
            72.6
            License required; types of licenses.
            72.7
            Specific exemptions.
            72.8
            Denial of licensing by Agreement States.
            72.9
            Information collection requirements: OMB approval.
            72.10
            Employee protection.
            72.11
            Completeness and accuracy of information.
            72.12
            Deliberate misconduct.
            72.13
            Applicability.

            Subpart B—License Application, Form, and Contents
            72.16
            Filing of application for specific license.
            72.18
            Elimination of repetition.
            72.20
            Public inspection of application.
            72.22
            Contents of application: General and financial information.
            72.24
            Contents of application: Technical information.
            72.26
            Contents of application: Technical specifications.
            72.28
            Contents of application: Applicant's technical qualifications.
            72.30
            Financial assurance and recordkeeping for decommissioning.
            72.32
            Emergency Plan.
            72.34
            Environmental report.

            Subpart C—Issuance and Conditions of License
            72.40
            Issuance of license.
            72.42
            Duration of license; renewal.
            72.44
            License conditions.
            72.46
            Public hearings.
            72.48
            Changes, tests, and experiments.
            72.50
            Transfer of license.
            72.52
            Creditor regulations.
            72.54
            Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas.
            72.56
            Application for amendment of license.
            72.58
            Issuance of amendment.
            72.60
            Modification, revocation, and suspension of license.
            72.62
            Backfitting.

            Subpart D—Records, Reports, Inspections, and Enforcement
            72.70
            Safety analysis report updating.

            72.72
            Material balance, inventory, and records requirements for stored materials.
            72.74
            Reports of accidental criticality or loss of special nuclear material.
            72.75
            Reporting requirements for specific events and conditions.
            72.76
            Material status reports.
            72.78
            Nuclear material transaction reports.
            72.79
            Facility information and verification.
            72.80
            Other records and reports.
            72.82
            Inspections and tests.
            72.84
            Violations.
            72.86
            Criminal penalties.

            Subpart E—Siting Evaluation Factors
            72.90
            General considerations.
            72.92
            Design basis external natural events.
            72.94
            Design basis external man-induced events.
            72.96
            Siting limitations.
            72.98
            Identifying regions around an ISFSI or MRS site.
            72.100
            Defining potential effects of the ISFSI or MRS on the region.
            72.102
            Geological and seismological characteristics for applications before October 16, 2003 and applications for other than dry cask modes of storage.
            72.103
            Geological and seismological characteristics for applications for dry cask modes of storage on or after October 16, 2003.
            72.104
            Criteria for radioactive materials in effluents and direct radiation from an ISFSI or MRS.
            72.106
            Controlled area of an ISFSI or MRS.
            72.108
            Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation.

            Subpart F—General Design Criteria
            72.120
            General considerations.
            72.122
            Overall requirements.
            72.124
            Criteria for nuclear criticality safety.
            72.126
            Criteria for radiological protection.
            72.128
            Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling.
            72.130
            Criteria for decommissioning.

            Subpart G—Quality Assurance
            72.140
            Quality assurance requirements.
            72.142
            Quality assurance organization.
            72.144
            Quality assurance program.
            72.146
            Design control.
            72.148
            Procurement document control.
            72.150
            Instructions, procedures, and drawings.
            72.152
            Document control.
            72.154
            Control of purchased material, equipment, and services.
            72.156
            Identification and control of materials, parts, and components.
            72.158
            Control of special processes.
            72.160
            Licensee and certificate holder inspection.
            72.162
            Test control.
            72.164
            Control of measuring and test equipment.
            72.166
            Handling, storage, and shipping control.
            72.168
            Inspection, test, and operating status.
            72.170
            Nonconforming materials, parts, or components.
            72.172
            Corrective action.
            72.174
            Quality assurance records.
            72.176
            Audits.

            Subpart H—Physical Protection
            72.180
            Physical protection plan.
            72.182
            Design for physical protection.
            72.184
            Safeguards contingency plan.
            72.186
            Change to physical security and safeguards contingency plans.

            Subpart I—Training and Certification of Personnel
            72.190
            Operator requirements.
            72.192
            Operator training and certification program.
            72.194
            Physical requirements.

            Subpart J—Provision of MRS Information to State Governments and Indian Tribes
            72.200
            Provision of MRS information.
            72.202
            Participation in license reviews.
            72.204
            Notice to States.
            72.206
            Representation.

            Subpart K—General License for Storage of Spent Fuel at Power Reactor Sites
            72.210
            General license issued.
            72.212
            Conditions of general license issued under § 72.210.
            72.214
            List of approved spent fuel storage casks.
            72.216
            [Reserved]
            72.218
            Termination of licenses.
            72.220
            Violations.

            Subpart L—Approval of Spent Fuel Storage Casks
            72.230
            Procedures for spent fuel storage cask submittals.
            72.232
            Inspection and tests.
            72.234
            Conditions of approval.
            72.236
            Specific requirements for spent fuel storage cask approval and fabrication.
            72.238
            Issuance of an NRC Certificate of Compliance.
            72.240
            Conditions for spent fuel storage cask renewal.
            72.242
            Recordkeeping and reports.
            72.244

            Application for amendment of a certificate of compliance.

            72.246
            Issuance of amendment to a certificate of compliance.
            72.248
            Safety analysis report updating.

          Authority:
          Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42 U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e, 2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); National Environmental Policy Act of 1969 (42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 117(a), 132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 U.S.C. 10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 10168, 10198(a)); 44 U.S.C. 3504 note.

          Source:
          53 FR 31658, Aug. 19, 1988, unless otherwise noted.

          Editorial Note:
          Nomenclature changes to part 72 appear at 80 FR 74981, Dec. 1, 2015, and at 81 FR 86910, Dec. 2, 2016.

          Subpart A—General Provisions

            § 72.1
            Purpose.
            The regulations in this part establish requirements, procedures, and criteria for the issuance of licenses to receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC) waste, and other radioactive materials associated with spent fuel storage in an independent spent fuel storage installation (ISFSI) and the terms and conditions under which the Commission will issue these licenses. The regulations in this part also establish requirements, procedures, and criteria for the issuance of licenses to the Department of Energy (DOE) to receive, transfer, package, and possess power reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive materials associated with the storage of these materials in a monitored retrievable storage installation (MRS). The term Monitored Retrievable Storage Installation or MRS, as defined in § 72.3, is derived from the Nuclear Waste Policy Act (NWPA) and includes any installation that meets this definition. The regulations in this part also establish requirements, procedures, and criteria for the issuance of Certificates of Compliance approving spent fuel storage cask designs.
            [66 FR 51838, Oct. 11, 2001]

            § 72.2
            Scope.
            (a) Except as provided in § 72.6(b), licenses issued under this part are limited to the receipt, transfer, packaging, and possession of:
            (1) Power reactor spent fuel to be stored in a complex that is designed and constructed specifically for storage of power reactor spent fuel aged for at least one year, other radioactive materials associated with spent fuel storage, and power reactor-related GTCC waste in a solid form in an independent spent fuel storage installation (ISFSI); or
            (2) Power reactor spent fuel to be stored in a monitored retrievable storage installation (MRS) owned by DOE that is designed and constructed specifically for the storage of spent fuel aged for at least one year, high-level radioactive waste that is in a solid form, other radioactive materials associated with storage of these materials, and power reactor-related GTCC waste that is in a solid form.
            (b) The regulations in this part pertaining to an independent spent fuel storage installation (ISFSI) and a spent fuel storage cask apply to all persons in the United States, including persons in Agreement States. The regulations in this part pertaining to a monitored retrievable storage installation (MRS) apply only to DOE.
            (c) The requirements of this regulation are applicable, as appropriate, to both wet and dry modes of storage of—
            (1) Spent fuel and solid reactor-related GTCC waste in an independent spent fuel storage installation (ISFSI); and
            (2) Spent fuel, solid high-level radioactive waste, and solid reactor-related GTCC waste in a monitored retrievable storage installation (MRS).
            (d) Licenses covering the storage of spent fuel in an existing spent fuel storage installation shall be issued in accordance with the requirements of this part as stated in § 72.40, as applicable.

            (e) This part also gives notice to all persons who knowingly provide to any licensee, certificate holder, applicant for a license or certificate, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licensee's, certificate holder's, or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 72.12.
            (f) Certificates of Compliance approving spent fuel storage cask designs shall be issued in accordance with the requirements of subpart L of this part.
            [53 FR 31658, Aug. 19, 1988, as amended at 56 FR 40692, Aug. 15, 1991; 63 FR 1900, Jan. 13, 1998; 64 FR 33183, June 22, 1999; 64 FR 56121, Oct. 15, 1999; 66 FR 51838, Oct. 11, 2001]

            § 72.3
            Definitions.
            As used in this part:

              Act means the Atomic Energy Act of 1954 (68 Stat. 919) including any amendments thereto.

              Affected Indian Tribe means any Indian Tribe—
            (1) Within whose reservation boundaries a monitored retrievable storage facility is proposed to be located;

            (2) Whose federally defined possessory or usage rights to other lands outside of the reservation's boundaries arising out of congressionally ratified treaties may be substantially and adversely affected by the locating of such a facility: Provided, That the Secretary of the Interior finds, upon the petition of the appropriate governmental officials of the Tribe, that such effects are both substantial and adverse to the Tribe.

              Affected unit of local government means any unit of local government with jurisdiction over the site where an MRS is proposed to be located.

              AMP, for the purposes of this part, means a program for addressing aging effects that may include prevention, mitigation, condition monitoring, and performance monitoring.

              As low as is reasonably achievable (ALARA) means as low as is reasonably achievable taking into account the state of technology, and the economics of improvement in relation to—
            (1) Benefits to the public health and safety,
            (2) Other societal and socioeconomic considerations, and
            (3) The utilization of atomic energy in the public interest.

              Atomic energy means all forms of energy released in the course of nuclear fission or nuclear transformation.

              Byproduct material means—
            (1) Any radioactive material (except special nuclear material) yielded in, or made radioactive by, exposure to the radiation incident to the process of producing or using special nuclear material;
            (2)(i) Any discrete source of radium-226 that is produced, extracted, or converted after extraction, before, on, or after August 8, 2005, for use for a commercial, medical, or research activity; or
            (ii) Any material that—
            (A) Has been made radioactive by use of a particle accelerator; and
            (B) Is produced, extracted, or converted after extraction, before, on, or after August 8, 2005, for use for a commercial, medical, or research activity; and
            (3) Any discrete source of naturally occurring radioactive material, other than source material, that—
            (i) The Commission, in consultation with the Administrator of the Environmental Protection Agency, the Secretary of Energy, the Secretary of Homeland Security, and the head of any other appropriate Federal agency, determines would pose a threat similar to the threat posed by a discrete source of radium-226 to the public health and safety or the common defense and security; and
            (ii) Before, on, or after August 8, 2005, is extracted or converted after extraction for use in a commercial,
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