Source: https://regulations.justia.com/regulations/fedreg/2008/10/07/E8-23342.html
Timestamp: 2020-08-07 22:34:11
Document Index: 424824461

Matched Legal Cases: ['art 2', 'art 50', 'art 20', 'art 26', 'art 26', 'art 26', 'art 26', 'art 26', 'art 50', 'art 1', 'art 2', 'art 20', 'art 26', 'art 26', 'art 26', 'art 26', 'art 26']

Biweekly Notice; Applications and Amendments to Facility Operating Licenses; Involving No Significant Hazards Considerations, 58669-58684 [E8-23342] :: Nuclear Regulatory Commission :: Agencies And Commissions :: Regulation Tracker :: Justia
Justia Regulation Tracker Agencies And Commissions Nuclear Regulatory Commission Biweekly Notice; Applications and Amendments to Facility Operating Licenses; Involving No Significant Hazards Considerations, 58669-58684 [E8-23342]
Biweekly Notice; Applications and Amendments to Facility Operating Licenses; Involving No Significant Hazards Considerations, 58669-58684 [E8-23342]
Download as PDF Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices This meeting, from 9 a.m. to 11:30 a.m. (ending time is approximate), will be open to the public on a space available basis. The meeting will begin with opening remarks and will include a poetry reading by David Lehman, a performance from Shakespeare by Aquila Theater Company, and a jazz performance by pianist Helen Sung. After the presentations the Council will review and vote on applications and guidelines, and the meeting will end with remarks and Council members’ farewell to the Chairman. If, in the course of the open session discussion, it becomes necessary for the Council to discuss non-public commercial or financial information of intrinsic value, the Council will go into closed session pursuant to subsection (c)(4) of the Government in the Sunshine Act, 5 U.S.C. 552b. Additionally, discussion concerning purely personal information about individuals, submitted with grant applications, such as personal biographical and salary data or medical information, may be conducted by the Council in closed session in accordance with subsection (c) (6) of 5 U.S.C. 552b. Any interested persons may attend, as observers, Council discussions and reviews that are open to the public. If you need special accommodations due to a disability, please contact the Office of AccessAbility, National Endowment for the Arts, 1100 Pennsylvania Avenue, NW., Washington, DC 20506, 202/682– 5532, TTY–TDD 202/682–5429, at least seven (7) days prior to the meeting. Further information with reference to this meeting can be obtained from the Office of Communications, National Endowment for the Arts, Washington, DC 20506, at 202/682–5570. Dated: October 2, 2008. Kathy Plowitz-Worden, Panel Coordinator, Office of Guidelines and Panel Operations. [FR Doc. E8–23705 Filed 10–6–08; 8:45 am] BILLING CODE 7537–01–P NUCLEAR REGULATORY COMMISSION mstockstill on PROD1PC66 with NOTICES Biweekly Notice; Applications and Amendments to Facility Operating Licenses; Involving No Significant Hazards Considerations I. Background Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. This biweekly notice includes all notices of amendments issued, or proposed to be issued from September 11, 2008 to September 24, 2008. The last biweekly notice was published on September 23, 2008 (73 FR 54862). Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission’s regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below. The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a PO 00000 Frm 00141 Fmt 4703 Sfmt 4703 58669 notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently. Written comments may be submitted by mail to the Chief, Rulemaking, Directives and Editing Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below. Within 60 days after the date of publication of this notice, person(s) may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request via electronic submission through the NRC E-Filing system for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission’s ‘‘Rules of Practice for Domestic Licensing Proceedings’’ in 10 CFR Part 2. Interested person(s) should consult a current copy of 10 CFR 2.309, which is available at the Commission’s PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System’s (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a E:\FR\FM\07OCN1.SGM 07OCN1 mstockstill on PROD1PC66 with NOTICES 58670 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor’s/petitioner’s right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor’s/petitioner’s property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor’s/petitioner’s interest. The petition must also set forth the specific contentions which the petitioner/ requestor seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/ requestor to relief. A petitioner/ requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment. A request for hearing or a petition for leave to intervene must be filed in accordance with the NRC E-Filing rule, which the NRC promulgated on August 28, 2007 (72 FR 49139). The E-Filing process requires participants to submit and serve documents over the Internet or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek a waiver in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least five (5) days prior to the filing deadline, the petitioner/requestor must contact the Office of the Secretary by e-mail at hearingdocket@nrc.gov, or by calling (301) 415–1677, to request (1) a digital ID certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and/or (2) creation of an electronic docket for the proceeding (even in instances in which the petitioner/requestor (or its counsel or representative) already holds an NRCissued digital ID certificate). Each petitioner/requestor will need to download the Workplace Forms ViewerTM to access the Electronic Information Exchange (EIE), a component of the E-Filing system. The Workplace Forms ViewerTM is free and is available at http://www.nrc.gov/sitehelp/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is available on NRC’s public Web site at http://www.nrc.gov/ site-help/e-submittals/applycertificates.html. Once a petitioner/requestor has obtained a digital ID certificate, had a docket created, and downloaded the EIE viewer, it can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at PO 00000 Frm 00142 Fmt 4703 Sfmt 4703 http://www.nrc.gov/site-help/esubmittals.html. A filing is considered complete at the time the filer submits its documents through EIE. To be timely, an electronic filing must be submitted to the EIE system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The EIE system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/ petition to intervene is filed so that they can obtain access to the document via the E-Filing system. A person filing electronically may seek assistance through the ‘‘Contact Us’’ link located on the NRC Web site at http://www.nrc.gov/site-help/esubmittals.html or by calling the NRC technical help line, which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, Monday through Friday. The help line number is (800) 397–4209 or locally, (301) 415–4737. Participants who believe that they have a good cause for not submitting documents electronically must file a motion, in accordance with 10 CFR 2.302(g), with their initial paper filing requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered complete by firstclass mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. Non-timely requests and/or petitions and contentions will not be entertained absent a determination by the Commission, the presiding officer, or E:\FR\FM\07OCN1.SGM 07OCN1 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices mstockstill on PROD1PC66 with NOTICES the Atomic Safety and Licensing Board that the petition and/or request should be granted and/or the contentions should be admitted, based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)–(viii). To be timely, filings must be submitted no later than 11:59 p.m. Eastern Time on the due date. Documents submitted in adjudicatory proceedings will appear in NRC’s electronic hearing docket which is available to the public at http:// ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant to an order of the Commission, an Atomic Safety and Licensing Board, or a Presiding Officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. For further details with respect to this amendment action, see the application for amendment which is available for public inspection at the Commission’s PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http:// www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397– 4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50–317 and 50–318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland Date of amendments request: August 28, 2008. Description of amendments request: The amendment would relocate the main steam isolation valve times in Technical Specification (TS) section 3.7.2, ‘‘Main Steam Isolation Valves (MSIVs)’’ to the licensee controlled document that is referenced in the Bases. In addition, the valve isolation times in the TS are replaced with the phrase ‘‘within limits.’’ The changes are consistent with the Nuclear Regulatory Commission approved Technical Specification Task Force (TSTF)–491, Revision 2, ‘‘Removal of Main Steam and Main Feedwater Valve Isolation Times From Technical Specifications.’’ VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 The availability of the TS improvement was published in the Federal Register on December 29, 2006 (71 FR 250) as part of the consolidated item improvement process (CLIIP). Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed change allows relocating main steam and main feedwater valve isolation times to the Licensee Controlled Document that is referenced in the Bases. The proposed change is described in Technical Specification Task Force (TSTF) Standard TS Change Traveler TSTF–491 related to relocating the main steam and main feedwater valves isolation times to the Licensee Controlled Document that is referenced in the Bases and replacing the isolation time with the phrase, ‘‘within limits.’’ The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). The proposed changes relocate the main steam and main feedwater isolation valve times to the Licensee Controlled Document that is referenced in the Bases. The requirements to perform the testing of these isolation valves are retained in the TS. Future changes to the Bases or licensee-controlled document will be evaluated pursuant to the requirements of 10 CFR 50.59, ‘‘Changes, tests and experiments’’, to ensure that such changes do not result in more than minimal increase in the probability or consequences of an accident previously evaluated. The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely affect the ability of structures, systems and components (SSCs) to perform their intended safety function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological consequences of any accident previously evaluated. Further, the proposed changes do not increase the types and the amounts of radioactive effluent that may be released, nor significantly increase individual or cumulative occupation/public radiation exposures. Therefore, the changes do not involve a significant increase in the probability or consequences of any accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated The proposed changes relocate the main steam and main feedwater valve isolation PO 00000 Frm 00143 Fmt 4703 Sfmt 4703 58671 times to the Licensee Controlled Document that is referenced in the Bases. In addition, the valve isolation times are replaced in the TS with the phrase ‘‘within limits’’. The changes do not involve a physical altering of the plant (i.e., no new or different type of equipment will be installed) or a change in methods governing normal pant operation. The requirements in the TS continue to require testing of the main steam and main feedwater isolation valves to ensure the proper functioning of these isolation valves. Therefore, the changes do not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The proposed changes relocate the main steam and main feedwater valve isolation times to the Licensee Controlled Document that is referenced in the Bases. In addition, the valve isolation times are replaced in the TS with the phrase ‘‘within limits.’’ Instituting the proposed changes will continue to ensure the testing of main steam and main feedwater isolation valves. Changes to the Bases are license controlled document are performed in accordance with 10 CFR 50.59. This approach provides an effective level of regulatory control and ensures that main steam and feedwater isolation valve testing is conducted such that there is no significant reduction in the margin of safety. The margin of safety provided by the isolation valves is unaffected by the proposed changes since there continue to be TS requirements to ensure the testing of main steam and main feedwater isolation valves. The proposed changes maintain sufficient controls to preserve the current margins of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration. Attorney for licensee: Carey Fleming, Sr. Counsel–Nuclear Generation, Constellation Generation Group, LLC, 750 East Pratt Street, 17th floor, Baltimore, MD 21202. NRC Branch Chief: Mark G. Kowal. Carolina Power & Light Company, Docket Nos. 50–325 and 50–324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of amendments request: June 19, 2008. Description of amendments request: The proposed change would: (1) Revise Technical Specifications (TS) control rod notch surveillance requirement (SR) frequency in TS 3.1.3, ‘‘Control Rod Operability,’’ and (2) revise Example 1.4–3 in Section 1.4, ‘‘Frequency,’’ to clarify the applicability of the 1.25 surveillance test extension. The licensee E:\FR\FM\07OCN1.SGM 07OCN1 58672 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices mstockstill on PROD1PC66 with NOTICES is proposing to adopt the approved Technical Specification Task Force (TSTF) change traveler TSTF–475, Revision 1, ‘‘Control Rod Notch Testing Frequency.’’ A notice of availability of TSTF–475, Revision 1, was published in the Federal Register on November 13, 2007 (72 FR 63935). In addition, the proposed amendment would remove Note 2 associated with SR 3.1.3.3 for Unit 1, which is a cyclespecific note and has expired. This change is administrative in nature and does not affect the no significant hazards consideration (NSHC) determination. Basis for proposed NSHC determination: As required by 10 CFR 50.91(a), the licensee, in its application dated June 19, 2008, affirmed the applicability of the published model NSHC determination, which is presented below: Report, ‘‘CRD Notching Surveillance Testing for Limerick Generating Station,’’ dated November 2006, concludes that extending the control rod notch test interval from weekly to monthly is not expected to impact the reliability of the scram system and that the analysis supports the decision to change the surveillance frequency. Therefore, the proposed changes in TSTF–475, Revision 1 [. . .] do not involve a significant reduction in a margin of safety. Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed change generically implements TSTF–475, Revision 1, ‘‘Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.’’ TSTF–475, Revision 1 modifies NUREG–1433 (BWR/4) and NUREG–1434 (BWR/6) STS. The changes: (1) revise TS testing frequency for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ‘‘Control Rod OPERABILITY’’, (2) [not applicable to BSEP], and (3) revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. The consequences of an accident after adopting TSTF–475, Revision 1 are no different than the consequences of an accident prior to adoption. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously analyzed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety TSTF–475, Revision 1 will: (1) revise the TS SR 3.1.3.2 frequency in TS 3.1.3, ‘‘Control Rod OPERABILITY’’, (2) [not applicable to BSEP], and (3) revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. The GE Nuclear Energy Carolina Power & Light Company, Docket Nos. 50–325 and 50–324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 Based on the review of the above analysis, the NRC staff finds that the three standards in 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC. Attorney for licensee: David T. Conley, Associate General Counsel II– Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, NC 27602. NRC Branch Chief: Thomas H. Boyce. Date of amendments request: June 19, 2008. Description of amendments request: The proposed change would revise Limiting Condition for Operation (LCO) 3.10.1, and the associated Bases, to expand its scope to include provisions for temperature excursions greater than 212 degrees Fahrenheit (°F) as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4. The NRC issued a ‘‘Notice of Availability of Model Application on Technical Specification Improvement to Modify Requirements Regarding LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation Using the Consolidated Line Item Improvement Process,’’ associated with Technical Specification Task Force (TSTF) Improved Standard Technical Specification Change Traveler, TSTF– 484, Revision 0, in the Federal Register on October 27, 2006 (71 FR 63050). The NRC also issued a Federal Register notice on August 21, 2006 (71 FR 48561) that provided a model safety evaluation and a model no significant hazards consideration (NSHC) determination relating to modification of requirements regarding LCO 3.10.1, ‘‘Inservice Leak and Hydrostatic Testing Operation.’’ In its application dated June 19, 2008, the licensee affirmed the applicability of the model NSHC determination. PO 00000 Frm 00144 Fmt 4703 Sfmt 4703 Basis for proposed NSHC determination: As required by 10 CFR Part 50.91(a), an analysis of the issue of NSHC determination is presented below: Criterion 1: The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact the probability or consequences of an accident previously evaluated. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Criterion 2: The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. No new operational conditions beyond those currently allowed by LCO 3.10.1 are introduced. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements or eliminate any existing requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3: The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety Technical Specifications currently allow for operation at greater than 212 °F while imposing MODE 4 requirements in addition to the secondary containment requirements required to be met. Extending the activities that can apply this allowance will not adversely impact any margin of safety. Allowing completion of inspections and testing and supporting completion of scram time testing initiated in conjunction with an inservice leak or hydrostatic test prior to power operation results in enhanced safe operations by eliminating unnecessary maneuvers to control reactor temperature and pressure. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, the proposed change presents NSHCs under the standards set forth in 10 CFR 50.92(c). Therefore, the NRC staff proposes to determine that the amendment request involves NSHC. Attorney for licensee: David T. Conley, Associate General Counsel II– E:\FR\FM\07OCN1.SGM 07OCN1 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, NC 27602. NRC Branch Chief: Thomas H. Boyce. mstockstill on PROD1PC66 with NOTICES Entergy Nuclear Operations, Inc., Docket No. 50–3, Indian Point Energy Center, Unit 1 Westchester County, New York Date of amendment request: June 26, 2008. Description of amendment request: The proposed amendment would delete license conditions and Technical Specification (TS) requirements which relate to the storage of spent nuclear fuel in the Indian Point Unit 1 (IP1) Fuel Handling Building Spent Fuel Pool. The spent fuel is to be transferred to, and stored at, the existing Indian Point Independent Spent Fuel Storage Installation (ISFSI), Docket No. 72–51. The removal of the stored spent fuel and drain down of the spent fuel pools renders many of the license conditions and TS requirements unnecessary and burdensome. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? No. The proposed changes are all contingent on the prior removal of the stored spent fuel from the IP1 Spent Fuel Pool (SFP) to the Indian Point Energy Center (IPEC) ISFSI. The accidents previously evaluated in the IP1 Final Safety Analysis Report (FSAR), which consists of the IP1 Decommission Plan and Supplemental Environmental Information, are stored fuel related accidents. The removal of the stored fuel from the IP1 facility to the IPEC ISFSI precludes the possibility of these accidents. Consequently, the proposed changes to the license do not involve a significant increase in the probability or the consequences of an accident previously evaluated. 2. Does the proposed change create the probability of a new or different accident from any accident previously evaluated? The proposed changes are all contingent on the prior removal of the stored spent fuel from the IP1 SFP to the IPEC ISFSI. With the removal of the stored spent fuel from the IP1 facility, and considering the IP1 has been in a SAFESTOR mode for over thirty years, no significant source term remains which could result in any postulated radiological event that would impact the health and safety of the public. Therefore, the proposed changes to the IP1 license consequently do not create the possibility of a new or different kind of accident from any accident previously evaluated. VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 3. Does the proposed change involve a significant reduction in a margin of safety? No. The proposed changes are all contingent on the prior removal of the stored spent fuel from the IP1 SFP to the IPEC ISFSI. Upon the removal of spent fuel, the Technical Specifications being deleted no longer are required to protect the health and safety of the public or occupational workers from the potential adverse conditions, hazards or accidents as discussed in the FSAR. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant reduction in the margin of safety. Based upon the reasoning presented above it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601. NRC Branch Chief: Theodore Smith, Acting. Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50–416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Date of amendment request: September 11, 2008. Description of amendment request: The amendment would revise several surveillance requirements (SRs) and add SR 3.8.1.21 in Technical Specification (TS) 3.8.1, ‘‘AC [alternating current] Sources—Operating,’’ and TS 3.8.2, ‘‘AC Sources—Shutdown.’’ The amendment would allow the slow-start testing sequence of the diesel generators in order to reduce the stress and wear on the equipment. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change affects the surveillance requirements for the Diesel Generators (DGs). The DGs are onsite standby power sources intended to provide redundant and reliable power to ESF [Engineered Safety Feature] systems credited as accident mitigating features in design basis [accident] analyses. Per NRC Regulatory Guide (RG) 1.9, Revision 3, which is PO 00000 Frm 00145 Fmt 4703 Sfmt 4703 58673 referenced in Grand Gulf Nuclear Station (GGNS) UFSAR [Updated Final Safety Analysis Report Section] 8.3.1.2.1, the proposed change is intended to allow slower starts of the DGs during testing in order to reduce DG aging effects due to excessive testing conditions. As such, the proposed change will result in improved DG reliability and availability, thereby providing additional assurance that the DGs will be capable of performing their safety function. The method of starting the emergency diesel generators for testing purposes does not affect the probability of any previously evaluated accident. Although the change allows slower starts for the monthly tests, the more rapid start function, assumed in the accident analysis, is unchanged and will be verified on a 184 day frequency. Therefore the accident analysis consequences are not affected [by the proposed change]. Therefore, the proposed change does not involve a significant increase in the probability [or] consequences of any accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change affects the surveillance requirements for the onsite ac sources, i.e. the Diesel Generators. Accordingly, the proposed change does not involve any change to the configuration or method of operation of any plant equipment that could cause an accident. In addition, no new failure modes have been created nor has any new limiting failure been introduced as a result of the proposed surveillance changes. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change is intended to bring the existing GGNS TS requirements for the onsite AC sources in line with regulatory guidance. Under the proposed change, the DGs will remain capable of performing their safety function, and the effects of aging on the DGs will be reduced by eliminating unnecessary testing. The DG start times assumed in the current accident analyses are unchanged and will be verified on a 184 day frequency. Therefore, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Terence A. Burke, Associate General Counsel— Nuclear Entergy Services, Inc., 1340 E:\FR\FM\07OCN1.SGM 07OCN1 58674 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices mstockstill on PROD1PC66 with NOTICES Echelon Parkway, Jackson, Mississippi 39213. NRC Branch Chief: Michael T. Markley. GE Hitachi Nuclear Energy (GEH), License No. DR–10, Docket No. 50–183, ESADA Vallecitos Experimental Superheat Reactor (EVESR) Date of amendment request: June 23, 2008. Description of amendment request: The proposed license amendment would modify the Technical Specification (TS) requirements to revise the scope of dismantling activities that GEH can perform under The Vallecitos Nuclear Center Liabilities Reduction Project and specify radiological control requirements of 10 CFR Part 20. Two TS changes are proposed. The proposed changes to the TS: • Allow GEH to conduct dismantling activities below the 549-ft elevation level within the containment building; and • Revise the physical security requirements for access to areas below the 549-ft elevation level within the containment building. The application for license amendment is available electronically at the NRC’s Electronic Reading Room at http://www.nrc.gov/reading-rm/ adams.html. From this site, you can access the NRC’s Agencywide Document Access and Management System (ADAMS), which provides text and image files of NRC’s public documents. The ADAMS accession number for the June 23, 2008, request is ML081780099. If you do not have access to ADAMS, or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1–800–397–4209, 301– 415–4737, or by e-mail to pdr@nrc.gov. These documents may also be viewed electronically on the public computers located at the NRC’s PDR, 01F21, One White Flint North, 11555 Rockville Pike, Rockville, MD 20852. The PDR reproduction contractor will copy documents for a fee. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: (1) Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Proposed change one is an administrative change submitted to clarify the area where VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 dismantling activities will occur as authorized by the facility license. The majority of the component removal activities will occur in areas below the 549-ft. elevation. Proposed change two removes the specific shielding and covering requirements for the reactor vessel, shield plug storage pit and the empty spent fuel storage pit and modifies the access control requirements to be consistent with 10 CFR 20. The EVESR reactor was shutdown in 1967 and has remained in a ‘‘Possess Only’’ status. All fuel bundles were removed from the facility and the radiation and contamination levels have been reduced by the removal of radioactive material and natural decay. No aspect of the proposed changes will involve a significant increase in the probability or consequences of an accident previously evaluated. (2) Does the change create the possibility of a new or different kind of accident from any accident evaluated? Response: No. Proposed change one is an administrative, therefore there it cannot create a new or different kind of accident. Removal of the specific shielding and covering requirements for the reactor vessel, shield plug storage pit and the empty spent fuel storage pit and modification of the access control requirements as described in proposed change two will not impact the function or integrity of the reactor pressure vessel, which is the primary safety system required to be maintained by the license. The proposed changes do not create the possibility of a new or different kind of accident from any accident evaluated. (3) Does the change involve a significant reduction in a margin of safety? Response: No. Removal of the specific shielding, covering and access control requirements will not result in a reduction of the margin of the safety for the EVESR facility. These controls were implemented to provide shielding and access controls to High Radiation Areas. Since the reactor is no longer operating and the radiological conditions have been significantly reduced, the specific controls specified in the current technical specifications are not required. All areas in the EVESR containment will be controlled in accordance with 10 CFR 20. High Radiation areas will be controlled in a manner consistent with the requirements of 10 CFR 20.1601. The proposed changes do not affect the margins of safety. The NRC staff has reviewed the licensee’s analysis and, based upon the staff’s review of the licensee’s analysis, as well as the staff’s own evaluation, the staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. GEH, Manager, Regulatory Compliance & EHS: LaTonya L. Mahlahla. NRC Branch Chief: Andrew Persinko. PO 00000 Frm 00146 Fmt 4703 Sfmt 4703 Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket Nos. 50–220 and 50– 410, Nine Mile Point Nuclear Station Unit Nos. 1 and 2 (NMP 1 and 2), Oswego County, New York Date of amendment request: June 24, 2008. Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) by (1) replacing the references to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code with references to the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code); and (2) revising the allowance to extend Inservice Testing (IST) frequencies by 25 percent to clearly state that the allowance is applicable to IST frequencies of 2 years or less. The proposed changes are based on TS Task Force (TSTF) Standard Technical Specification Change Traveler 479–A, Revision 0, ‘‘Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less.’’ Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes revise the IST Program sections of the NMP1 and NMP2 TS to maintain consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the IST of pumps and valves that are classified as ASME Code Class 1, Class 2, and Class 3. The proposed changes incorporate revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The proposed changes also revise the allowance to extend IST frequencies by 25 percent to clearly state that this allowance is applicable to IST frequencies of 2 years or less. The proposed TS changes are administrative in nature. They do not impact any accident initiators, the ability to mitigate previously evaluated accidents, or the assumptions used in evaluating the radiological consequences of previously evaluated accidents. The proposed changes do not involve the addition or removal of any equipment, or any design changes to the facilities. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. E:\FR\FM\07OCN1.SGM 07OCN1 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices The proposed changes revise the IST Program sections of the NMP1 and NMP2 TS to maintain consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the IST of pumps and valves that are classified as ASME Code Class 1, Class 2, and Class 3. The proposed changes incorporate revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves. The proposed changes also revise the allowance to extend IST frequencies by 25 percent to clearly state that this allowance is applicable to IST frequencies of 2 years or less. The proposed TS changes are administrative in nature. They do not involve a modification to the physical configuration of the plants (i.e., no new equipment will be installed) or involve a change in the methods governing normal plant operation. The proposed changes will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or failure mechanism. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed TS changes are administrative in nature. They do not involve a modification to the physical configuration of the plants (i.e., no new equipment will be installed) or change the methods governing normal plant operation. The proposed changes do not modify the safety limits or setpoints at which protective actions are initiated, and do not change the requirements governing operation or availability of safety equipment assumed to operate to preserve margins of safety. The incorporation of revisions to the ASME Code results in a net improvement in the measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. mstockstill on PROD1PC66 with NOTICES The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & Strawn, 1700 K Street, NW., Washington, DC 20006. NRC Branch Chief: Mark G. Kowal. PPL Susquehanna, LLC, Docket Nos. 50–387 and 50–388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania Date of amendment request: July 7, 2008. Description of amendment request: The proposed amendment would revise VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 the Technical Specification (TS) testing frequency for the Surveillance Requirement (SR) in TS 3.1.4, ‘‘Control Rod Scram Times.’’ The proposed change revises the frequency of SR 3.1.4.2, control rod scram time testing, from ‘‘120 days cumulative operation in Mode 1’’ to ‘‘200 days cumulative operation in Mode 1.’’ These changes are based on TS Task Force (TSTF) change traveler TSTF–460 (Revision 0) that has been approved generically for the Boiling-water reactor (BWR) Standard TS, NUREG–1433 (BWR/4) and NUREG–1434 (BWR/6) by revising the frequency of SR 3.1.4.2, control rod scram time testing, from ‘‘120 days cumulative operation in MODE 1’’ to ‘‘200 days cumulative operation in MODE 1.’’ The NRC staff issued a notice of availability of a model no significant hazards consideration determination (NSHCD) for referencing in licensing amendment applications in the Federal Register on August 23, 2004 (69 FR 51864) using the consolidated line item improvement process (CLIIP). The licensee affirmed the applicability of the model NSHC determination and the model safety evaluation in its application dated July 7, 2008. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC, based on the model NSHCD published in the Federal Register on August 23, 2004 (69 FR 51864), is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident, previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The frequency of surveillance testing is not an initiator of any accident previously evaluated. The frequency of surveillance testing does not affect the ability to mitigate any accident previously evaluated, as the tested component is still required to be operable. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change extends the frequency for testing control rod scram time testing from every 120 days of cumulative Mode 1 operation to 200 days of cumulative Mode 1 operation. The proposed change does not result in any new or different modes of plant operation. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. PO 00000 Frm 00147 Fmt 4703 Sfmt 4703 58675 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change continues to test the control rod scram time to ensure the assumptions in the safety analysis are protected. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101–1179. NRC Branch Chief: Mark Kowal. PPL Susquehanna, LLC, Docket Nos. 50–387 and 50–388, Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania Date of amendment request: July 7, 2008. Description of amendment request: PPL Susquehanna, LLC (the licensee) requests adoption of the Nuclear Regulatory Commission (NRC) approved Technical Specification Task Force (TSTF) change traveler TSTF–475, (Revision 1), ‘‘Control Rod Notch Testing Frequency and SRM [Source Range Monitor] Insert Control Rod Action,’’ to change the Standard Technical Specifications (STS) for General Electric (GE) Plants (NUREG– 1433, BWR/4 to the plant specific TS, that allows: (1) Revising the frequency of Surveillance Requirement (SR) 3.1.3.2, notch testing of fully withdrawn control rod, from ‘‘7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM’’ to ‘‘31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP [Low Power Set Point] of the RWM [Rod With Minimizer]’’, and (2) revising Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the ‘‘SURVEILLANCE’’ column in addition to the time periods in the ‘‘FREQUENCY’’ column. The NRC staff issued a notice of availability in the Federal Register on November 13, 2007, (72 FR 63935), which included a model safety evaluation (SE) and model no significant hazards consideration determination (NSHCD), using the consolidated line-item improvement process (CLIIP), of possible amendments to revise the plant specific TS, to allow: (1) Revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn control rod, from ‘‘7 days after the control rod E:\FR\FM\07OCN1.SGM 07OCN1 mstockstill on PROD1PC66 with NOTICES 58676 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices is withdrawn and THERMAL POWER is greater than the LPSP of RWM’’ to ‘‘31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM’’, (2) adding the word ‘‘fully’’ to LCO 3.3.1.2 Required Action E.2 to clarify the requirement to fully insert all insertable control rods in core cells containing one or more fuel assemblies when the associated SRM instrument is inoperable, and (3) revising Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the ‘‘SURVEILLANCE’’ column in addition to the time periods in the ‘‘FREQUENCY’’ column. The licensee affirmed the applicability of the model SE and model NSHC determination in its application dated July 7, 2008. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of NSHC, based on the model NSHCD published in the Federal Register on November 13, 2007 (72 FR 63935), is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed change generically implements TSTF–475, Revision 1, ‘‘Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.’’ TSTF–475, Revision 1 modifies NUREG–1433 (BWR/4) and NUREG– 1434 (BWR/6) STS. The changes: (1) revise TS testing frequency for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ‘‘Control Rod OPERABILITY’’, (2) clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, Required Action E.2, ‘‘Source Range Monitoring Instrumentation’’ (NUREG–1434 only), and (3) revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. Implementing TSTF–475, Revision 1 does not change the control rod notch test method. Implementing TSTF–475, Revision 1 decreases the performance frequency of the control rod notch test. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated. The consequences of an accident after adopting TSTF–475, Revision 1 are no different than the consequences of an accident prior to adoption. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously analyzed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety The proposed amendment will: (1) Revise the TS SR 3.1.3.2 frequency in TS 3.1.3, ‘‘Control Rod OPERABILITY’’, and (2) revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. The GE Nuclear Energy Report, ‘‘CRD Notching Surveillance Testing for Limerick Generating Station,’’ dated November 2006, concludes that extending the control rod notch test interval from weekly to monthly is not expected to impact the reliability of the scram system and that the analysis supports the decision to change the surveillance frequency. Therefore, the proposed changes in TSTF–475, Revision 1 do not involve a significant reduction in a margin of safety. The NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101–1179. NRC Branch Chief: Mark Kowal. PSEG Nuclear LLC, Docket No. 50–354, Hope Creek Generating Station, Salem County, New Jersey PSEG Nuclear LLC, Docket Nos. 50–272 and 50–311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of amendment request: July 21, 2008. Description of amendment request: The proposed amendments would delete the requirements related to plant staff working hours from Section 6.0, ‘‘Administrative Controls’’ of the respective plants’ Technical Specifications (TSs). The current working hour requirements were PO 00000 Frm 00148 Fmt 4703 Sfmt 4703 incorporated into the TSs as a result of the guidance in Nuclear Regulatory Commission (NRC) Generic Letter (GL) 82–12, ‘‘Nuclear Power Plant Staff Working Hours.’’ The guidance in GL 82–12 has been superseded by the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 26, ‘‘Fitness for Duty Programs,’’ Subpart I, ‘‘Managing Fatigue’’ which was published in the Federal Register on March 31, 2008, as part of the final rulemaking for Part 26. As discussed in the Federal Register notice for the final rule (73 FR 16966), Subpart I must be implemented by licensees no later than October 1, 2009. The licensee stated that the proposed amendments would support implementation of the new requirements in Subpart I. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The removal of GL 82–12 administrative controls will not remove the requirement to control work hours and manage fatigue. Removal of TS controls required by GL 82– 12 will be performed concurrently with the implementation of the more conservative [10 CFR Part 26], Subpart I, requirements. The proposed changes do not impact the physical configuration or function of plant structures, systems, or components (SSCs) or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not impact the initiators or assumptions of analyzed events, nor do they impact the mitigation of accidents or transient events. Because these new requirements are more conservative with respect to work hour controls and fatigue management, this will not significantly increase the probability or consequence of an accident previously evaluated. 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed changes remove GL 82–12 administrative controls from [the] TS to support the implementation of Subpart I to [10 CFR Part 26]. The Subpart I regulations are more restrictive than the current guidance in [the] TS and would add conservatism to work hour controls and fatigue management. Work hours will continue to be controlled in accordance with NRC requirements. The new rule continues to allow for deviations from controls to mitigate or prevent a condition adverse to safety or necessary to maintain the security of the facility. This ensures that the new rule will not restrict work hours at the E:\FR\FM\07OCN1.SGM 07OCN1 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices expense of the health and safety of the public as well as plant personnel. The proposed changes do not alter plant configuration, require that new plant equipment be installed, alter assumptions made about accidents previously evaluated, add any initiators, or impact the function of plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected. Because the proposed changes do not remove the station’s requirement to control work hours and increases the conservatism of work hour controls by changing administrative scheduling requirements, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. Do the proposed changes involve a significant reduction in a margin of safety? Response: No. An input to maintaining the margin of safety is the control of work hours in managing fatigue. Salem and Hope Creek Generating Stations will continue their fitness-for-duty and behavioral observation programs, both of which will be strengthened by compliance with the new Part 26 regulation. The proposed changes add conservatism to fatigue management and contribute to the margin of safety. The proposed changes do not involve any physical changes to plant SSCs or the manner in which SSCs are operated, maintained, modified, tested, or inspected. The proposed changes do not involve a change to any safety limits, limiting safety system settings, limiting conditions of operation, or design parameters for any SSC. The proposed changes do not impact any safety analysis assumptions and do not involve a change in initial conditions, system response times, or other parameters affecting an accident analysis. Therefore, the proposed changes do not involve a significant reduction in the margin of safety. mstockstill on PROD1PC66 with NOTICES The NRC staff has reviewed the licensee’s analysis and, based on this review, with changes in the areas noted above, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. NRC Branch Chief: Harold K. Chernoff. PSEG Nuclear LLC, Docket No. 50–354, Hope Creek Generating Station, Salem County, New Jersey Date of amendment request: July 30, 2008. Description of amendment request: The proposed amendment would relocate Technical Specification (TS) 3/ 4.7.5, ‘‘Snubbers,’’ to the Hope Creek Generating Station (HCGS) Technical Requirements Manual (TRM). TS VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 6.10.3.l, which specifies retention requirements for records of snubber service life monitoring pursuant to TS 4.7.5, would also be relocated to the TRM. In addition, the amendment would add new TS Limiting Condition for Operation (LCO) 3.0.8, ‘‘Inoperability of Snubbers,’’ and would modify LCO 3.0.1 to reference LCO 3.0.8. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change to relocate TS 3/4.7.5 to the TRM is administrative in nature and does not involve the modification of any plant equipment or affect basic plant operation. Snubber operability and surveillance requirements will be contained in the TRM to ensure design assumptions for accident mitigation are maintained. The proposed change to add LCO 3.0.8 allows a delay time for entering a supported system technical specification (TS) when the inoperability is due solely to an inoperable snubber if risk is assessed and managed. Entrance into TS actions or delaying entrance into actions is not an initiator of any accident previously evaluated. Consequently, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on [the] allowance provided by proposed LCO 3.0.8 are no different than the consequences of an accident while relying on the current TS required actions in effect without the allowance provided by proposed LCO 3.0.8. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change to relocate TS 3/4.7.5 to the TRM is administrative and does not involve any physical alteration of plant equipment. The proposed change does not change the method by which any safetyrelated system performs its function. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change to add LCO 3.0.8 does not involve a physical alteration of the plant (no new or different type of equipment PO 00000 Frm 00149 Fmt 4703 Sfmt 4703 58677 will be installed). Allowing delay times for entering supported system TS when inoperability is due solely to inoperable snubbers, if risk is assessed and managed, will not introduce new failure modes or effects. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. The proposed change to relocate TS 3/4.7.5 to the TRM is administrative in nature, does not negate any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by requirements that are retained, but relocated from the TS to the TRM. The proposed change to add LCO 3.0.8 to [the] TS allows a delay time before declaring supported TS systems inoperable when the associated snubber(s) cannot perform the required safety function. The proposed change retains an allowance in the current HCGS TS while upgrading it to be more conservative for snubbers supporting multiple trains or sub-systems of an associated system. The updated TS will continue to provide an adequate margin of safety for plant operation upon incorporation of LCO 3.0.8. The station design and safety analysis assumptions provide margin in the form of redundancy to account for periods of time when system capability is reduced. Therefore, the proposed change does not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, with changes in the areas noted above, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038. NRC Branch Chief: Harold K. Chernoff. Southern Nuclear Operating Company, Inc., Docket Nos. 50–424 and 50–425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request: August 12, 2008. Description of amendment request: The proposed amendment deletes License Condition 2.H, which requires reporting of violations of operating license requirements found in license condition 2.C. E:\FR\FM\07OCN1.SGM 07OCN1 58678 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change involves the deletion of a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not significantly increase the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change is administrative in that it deletes a reporting requirement. The change does not add new plant equipment, change existing plant equipment, or affect the operating practices of the facility. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. The proposed change deletes a reporting requirement. The change does not affect plant equipment or operating practices and therefore does not involve a significant reduction in a margin of safety. mstockstill on PROD1PC66 with NOTICES The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308–2216. NRC Branch Chief: Melanie C. Wong. Union Electric Company, Docket No. 50–483, Callaway Plant, Unit 1, Callaway County, Missouri Date of amendment request: June 3, 2008. Description of amendment request: The proposed changes will revise Technical Specifications (TSs) 3.3.7, 3.3.8, 3.7.10, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and 3.8.10. This amendment will (1) delete MODES 5 and 6 from the Control Room Emergency Ventilation System and its actuation instrumentation in TS 3.7.10 and TS 3.3.7; (2) adopt U.S. Nuclear Regulatory Commission (NRC)approved traveler TSTF–36–A for TSs 3.3.8, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and 3.8.10; and (3) add a more restrictive change to the Limiting Condition for VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 Operation (LCO) Applicability for TSs 3.8.2, 3.8.5, 3.8.8, and 3.8.10 such that these LCOs apply not only during MODES 5 and 6, but also during the movement of irradiated fuel assemblies regardless of the MODE in which the plant is operating. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed changes to delete MODES 5 and 6 from the LCO Applicability of Technical Specifications (TSs) 3.3.7 and 3.7.10, adopt TSTF–36–A, and revise the LCO Applicability of the shutdown electrical specifications to be more restrictive does not alter plant design or operation; therefore, these changes will not increase the probability of any accident. Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no design changes. All design, material, and construction standards that were applicable prior to this amendment request will be maintained. There will be no changes to any design or operating limits. The proposed changes will not adversely affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not physically alter safety-related systems nor affect the way in which safety-related systems perform their functions. Deleting MODES 5 and 6 from the LCO Applicability of TSs 3.3.7 and 3.7.10 does not significantly increase the consequences of any accident since it has been demonstrated that the radiological consequences to control room occupants from a waste gas decay tank rupture will remain much less than the regulatory limits with no mitigation from the Control Room Emergency Ventilation System (CREVS) in MODES 5 and 6. The acceptance criteria for this event will continue to be met. The adoption of TSTF–36–A will not affect the equipment and LCOs needed to mitigate the consequences of a fuel handling accident in the fuel building; however, this change will reduce the chances of an unnecessary plant shutdown due to activities in the fuel building that have no bearing on the operation of the rest of the plant and the reactor core inside the containment building. The changes to the shutdown electrical specifications will add an additional restriction that is consistent with the PO 00000 Frm 00150 Fmt 4703 Sfmt 4703 objective of being able to mitigate a fuel handling accident during all situations, including a full core offload, in which such an accident could occur. All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. After a postulated release from a waste gas decay tank rapture no CREVS mitigation is required. The applicable radiological dose criteria will continue to be met. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. There are no proposed design changes nor are there any changes in the method by which any safety-related plant structure, system, or component (SSC) performs its specified safety function. The proposed changes will not affect the normal method of plant operation or change any operating parameters. Equipment performance necessary to fulfill safety analysis missions will be unaffected. The proposed changes will not alter any assumptions required to meet the safety analysis acceptance criteria. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safetyrelated system as a result of this amendment. The proposed amendment will not alter the design or performance of the 7300 Process Protection System, Nuclear Instrumentation System, or Solid State Protection System used in the plant protection systems. The proposed changes to delete MODES 5 and 6 from the LCO Applicability of TSs 3.3.7 and 3.7.10, adopt TSTF–36–A, and revise the LCO Applicability of the shutdown electrical specifications to be more restrictive do not, therefore, create the possibility of a new or different accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety? Response: No. There will be no effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor (FQ), nuclear enthalpy rise hot channel-factor (FAH), loss of coolant accident peak cladding temperature (LOCA PCT), peak local power density, or any other margin of safety. The applicable radiological dose consequence acceptance criteria will continue to be met. It has been demonstrated that the CREVS and its actuation instrumentation are not required to mitigate the control room radiological consequences of a waste gas decay tank rupture. The proposed changes do not eliminate any surveillances or alter the frequency of E:\FR\FM\07OCN1.SGM 07OCN1 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices surveillances required by the Technical Specifications. None of the acceptance criteria for any accident analysis will be changed. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration. Attorney for licensee: John O’Neill, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037. NRC Branch Chief: Michael T. Markley. Wolf Creek Nuclear Operating Corporation, Docket No. 50–482, Wolf Creek Generating Station, Coffey County, Kansas mstockstill on PROD1PC66 with NOTICES Date of amendment request: August 14, 2008. Description of amendment request: The proposed amendment would revise the Technical Specification (TS) 3.3.2, ‘‘Engineered Safety Feature Actuation System (ESFAS)’’ to extend the Surveillance Frequency on selected ESFAS slave relays from 92 days to 18 months. Justification for extending the slave relay Surveillance Frequency is based on information contained in the Westinghouse Electric Corporation reports WCAP–13878–P–A, Revision 2 (proprietary version), and WCAP– 14117–NP–A, Revision 2 (nonproprietary version), ‘‘Reliability Assessment of Potter & Brumfield MDR Series Relays,’’ dated August 2000. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration (NSHC), which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change will not result in a condition where the design, material, and construction standards that were applicable prior to the change are altered. The same Engineered Safety Feature Actuation System (ESFAS) instrumentation will be used and the same ESFAS system reliability is expected. Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no design changes. There will be no changes to any design or operating limits. VerDate Aug<31>2005 18:36 Oct 06, 2008 Jkt 217001 The proposed changes will not change accident initiators or precursors assumed or postulated in the Updated Safety Analysis Report (USAR) described accident analyses, nor will they alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not physically alter safety related systems, nor do they affect the way in which safety related systems perform their functions. All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the USAR. The applicable radiological dose acceptance criteria will continue to be met. Based on the above considerations, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. There are no proposed design changes, nor are there any changes in the method by which any safety-related plant SSC performs its specified safety function. Changing the interval for periodically verifying the ESFAS slave relays will not create any new accident initiators or scenarios. The proposed changes will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed changes will not alter any assumptions made in the safety analyses. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safetyrelated system as a result of this amendment. The proposed amendment will not alter the design or performance of the 7300 Process Protection System, Nuclear Instrumentation System, or Solid State Protection System used in the plant protection systems. Therefore, the proposed changes do not create the possibility of a new or different accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. The proposed change will not affect the total ESFAS response assumed in the safety analysis because the reliability of the slave relays will not be significantly affected by the increased surveillance interval. The relays PO 00000 Frm 00151 Fmt 4703 Sfmt 4703 58679 have demonstrated a high reliability and insensitivity to short term wear and aging effects. The overall reliability, redundancy, and diversity assumed available for the protection and mitigation of accident and transient conditions is unaffected by this proposed change. There will be no effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor (F2), nuclear enthalpy rise hot channel factor (F∆H), loss of coolant accident peak cladding temperature (LOCA PCT), peak local power density, or any other margin of safety. The applicable radiological dose consequence acceptance criteria for design-basis transients and accidents will continue to be met. None of the acceptance criteria for any accident analysis will be changed. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC. Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037. NRC Branch Chief: Michael T. Markley. Wolf Creek Nuclear Operating Corporation, Docket No. 50–482, Wolf Creek Generating Station, Coffey County, Kansas Date of amendment request: August 14, 2008. Description of amendment request: The proposed amendment would revise the Technical Specification (TS) 3.3.2, ‘‘Engineered Safety Feature Actuation System (ESFAS) Instrumentation,’’ TS 3.7.2, ‘‘Main Steam Isolation Valves (MSIVs),’’ and add New TS 3.7.19, ‘‘Secondary System Isolation Valves (SSIVs).’’ TS 3.7.2 is being revised to add MSIV bypass valves to the scope of TS 3.7.2. TS Table 3.3.2–1 is being revised to reflect the addition of the MSIV bypass valves to TS 3.7.2 and the associated applicability to be consistent with Westinghouse Standard Technical Specifications (NUREG–1431, Revision 31). TS 3.7.19 is being added to include a Limiting condition for Operation (LCO), Conditions/Required Actions and Surveillance Requirements for the steam generator blowdown isolation valves and steam generator blowdown sample isolation valves. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the E:\FR\FM\07OCN1.SGM 07OCN1 58680 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices mstockstill on PROD1PC66 with NOTICES licensee has provided its analysis of the issue of no significant hazards consideration (NSHC), which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change adds requirements to the TS to ensure that systems and components are maintained consistent with the safety analysis and licensing basis. Requirements are incorporated into the TS for secondary system isolation valves. These changes do not involve any design or physical changes to the facility, including the SSIVs themselves. The design and functional performance requirements, operational characteristics, and reliability of the SSIVs are unchanged. Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no design changes. All design, material, and construction standards that were applicable prior to this amendment request will be maintained. There will be no changes to any design or operating limits. The proposed changes will not change accident initiators or precursors assumed or postulated in the Updated Safety Analysis Report (USAR) described accident analyses, nor will they alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not physically alter safety related systems, nor do they affect the way in which safety related systems perform their functions. All accident analysis acceptance criteria will continue to be met with the proposed changes. The proposed changes will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. The proposed changes will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the USAR. The applicable radiological dose acceptance criteria will continue to be met. Based on the above considerations, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. There are no proposed design changes, nor are there any changes in the method by which any safety related plant SSC performs its specified safety function. The proposed changes will not affect the normal method of plant operation or change any operating parameters. No equipment performance VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 requirements will be affected. The proposed changes will not alter any assumptions made in the safety analyses. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety related system as a result of this amendment. The proposed amendment will not alter the design or performance of the [Analog Series] 7300 Process Protection System, Nuclear Instrumentation System, or Solid State Protection System used in the plant protection systems. Therefore, the proposed changes do not create the possibility of a new or different accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. There will be no effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio (DNBR) limits, heat flux hot channel factor (FQ), nuclear enthalpy rise hot channel factor (FAH), loss of coolant accident peak cladding temperature (LOCA PCT), peak local power density, or any other margin of safety. The applicable radiological dose consequence acceptance criteria for design-basis transients and accidents will continue to be met. The proposed changes do not eliminate any surveillances or alter the frequency of surveillances required by the Technical Specifications. None of the acceptance criteria for any accident analysis will be changed. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC. Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037. NRC Branch Chief: Michael T. Markley. Wolf Creek Nuclear Operating Corporation, Docket No. 50–482, Wolf Creek Generating Station, Coffey County, Kansas Date of amendment request: August 18, 2008. Description of amendment request: The proposed amendment would revise the Technical Specification (TS) 3.5.2, ‘‘ECCS [Emergency Core Cooling System]—Operating’’ requirements. The change is in accordance with Technical Specification Task Force (TSTF) TSTF– 325–A, Revision 0, ‘‘ECCS Conditions PO 00000 Frm 00152 Fmt 4703 Sfmt 4703 and Required Actions with <100% Equivalent ECCS Flow.’’ Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration (NSHC), which is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. The proposed change corrects the structure of the ACTIONS table to assure its correct application. There is no change or intent in the way the Conditions are actually applied. The literal interpretation of the existing Conditions structure could, under some circumstances, provide longer than intended Completion Times for restoration of OPERABILITY. Since the proposed change affects neither the Conditions intent nor its application, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated? Response: No. The proposed change corrects the structure of the ACTIONS table to assure its correct application. The proposed change does not result in any physical alterations to the plant configuration, no new equipment additions, no equipment interface modifications, and no changes to any equipment function or the method of operating the equipment are being made. As the proposed change would not change the design, configuration or operation of the plant, no new or different kinds of accident modes are created. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed amendment involve a significant reduction in the margin of safety? Response: No. The proposed change corrects the structure of the LCO [Limiting Condition for Operation] to assure its correct application. The proposed change is consistent with the requirements of the Technical Specifications. There is no change in intent or in the way the LCO is applied. Therefore, the proposed change does not involve a significant reduction in the margin of safety. The NRC staff has reviewed the licensee’s analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves NSHC. Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw Pittman LLP, 2300 N Street, NW., Washington, DC 20037. NRC Branch Chief: Michael T. Markley. E:\FR\FM\07OCN1.SGM 07OCN1 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices mstockstill on PROD1PC66 with NOTICES Notice of Issuance of Amendments to Facility Operating Licenses During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission’s rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing in connection with these actions was published in the Federal Register as indicated. Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated. For further details with respect to the action see (1) the applications for amendment, (2) the amendment, and (3) the Commission’s related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission’s Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/ reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397–4209, (301) 415–4737 or by e-mail to pdr@nrc.gov. VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 AmerGen Energy Company, LLC, Docket No. 50–461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois Date of application for amendment: December 12, 2006, as supplemented by letters dated November 16, 2007, and May 16 and June 27, 2008. Brief description of amendment: The amendment would increase the interval between the local power range monitor (LPRM) calibrations from 1000 megawatt-days/ton (MWD/T) to 2000 MWD/T as required by the Clinton Power Station technical specification surveillance requirements 3.3.1.1.8 and SR 3.3.1.2.2. Date of issuance: September 12, 2008. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment No.: 181. Facility Operating License No. NPF– 62: The amendment revised the Technical Specifications and License. Date of initial notice in Federal Register: May 22, 2007 (72 FR 28718). The November 16, 2007, and May 16 and June 27, 2008, supplements, contained clarifying information and did not change the NRC staff’s initial proposed finding of no significant hazards consideration. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 12, 2008. No significant hazards consideration comments received: No. Dominion Nuclear Connecticut, Inc., Docket Nos. 50–336 and 50–423 Millstone Power Station, Unit Nos. 2 and 3, New London County, Connecticut Date of application for amendments: July 13, 2007, as supplemented by letters dated December 7, 2007, March 5, March 25, April 28, June 9, June 26, and July 28, 2008. Brief description of amendments: The amendment changed the Millstone Power Station, Unit Nos. 2 and 3 Technical Specifications. This amendment established more effective and appropriate action, surveillance, and administrative requirements related to ensuring the habitability of the control room envelope in accordance with the Nuclear Regulatory Commission-approved Technical Specification Task Force (TSTF) Standard Technical Specification change traveler TSTF–448, Revision 3, ‘‘Control Room Habitability.’’ Additionally, the amendment changed the ‘‘irradiated fuel movement’’ terminology and adopted ‘‘movement of PO 00000 Frm 00153 Fmt 4703 Sfmt 4703 58681 recently irradiated fuel assemblies’’ terminology with TSTF–448, Revision 3. Date of issuance: September 18, 2008. Effective date: As of the date of issuance and shall be implemented within 180 days from the date of issuance. Amendment Nos.: 305 and 243. Renewed Facility Operating License No. DPR–65 and NPF–49: Amendments revised the License and Technical Specifications. Date of initial notice in Federal Register: May 16, 2008 and July 1, 2008 (73 FR 28534 and 73 FR 37506, respectively). The supplements dated June 9, June 26, and July 28, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 18, 2008. No significant hazards consideration comments received: No. Energy Northwest, Docket No. 50–397, Columbia Generating Station, Benton County, Washington Date of application for amendment: July 30, 2007, as supplemented August 28, 2008. Brief description of amendment: The amendment revises Technical Specifications 3.3.3.1, ‘‘Post Accident Monitoring (PAM) Instrumentation,’’ 3.3.6.1, ‘‘Primary Containment Isolation Instrumentation,’’ 3.6.1.3, ‘‘Primary Containment Isolation Valves (PCIVs),’’ and 3.6.4.2, ‘‘Secondary Containment Isolation Valves (SCIVs).’’ The proposed changes adopt the following TS Task Force (TSTF) Travelers that have been previously approved by the NRC: TSTF– 45–A, Revision 2, TSTF–46–A, Revision 1, TSTF–207–A, Revision 5, TSTF–269– A, Revision 2, TSTF–295–A, Revision 0, TSTF–306–A, Revision 2, and TSTF– 323–A, Revision 0. Date of issuance: September 15, 2008. Effective date: As of its date of issuance and shall be implemented within 90 days from the date of issuance. Amendment No.: 208. Facility Operating License No. NPF– 21: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: August 28, 2007 (72 FR 49573). The supplemental letter dated August 28, 2008, provided additional information that clarified the E:\FR\FM\07OCN1.SGM 07OCN1 58682 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices application, did not expand the scope of the application originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 2008, No significant hazards consideration comments received: No. mstockstill on PROD1PC66 with NOTICES Energy Northwest, Docket No. 50–397, Columbia Generating Station, Benton County, Washington Date of application for amendment: May 7, 2008 Brief description of amendment: The amendment revises Technical Specification Limiting Condition for Operation 3.10.1, and approves the associated Bases, to expand its scope to include provisions for temperature excursions greater than 200 degrees Fahrenheit as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4. Date of issuance: September 16, 2008. Effective date: As of its date of issuance and shall be implemented within 60 days from the date of issuance. Amendment No.: 209. Facility Operating License No. NPF– 21: The amendment revised the Facility Operating License and Technical Specifications. Date of initial notice in Federal Register: July 15, 2008 (73 FR 40630). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 16, 2008. No significant hazards consideration comments received: No. FPL Energy Duane Arnold, LLC, Docket No. 50–331, Duane Arnold Energy Center, Linn County, Iowa Date of application for amendment: February 19, 2008. Brief description of amendment: The amendment revises the Technical Specification Actions for the Emergency Diesel Generators (EDG) to remove the conditional surveillance requirement to test the alternate EDG whenever one EDG is taken out of service for preplanned preventive maintenance and testing. Date of issuance: September 9, 2008. Effective date: As of the date of issuance and shall be implemented within 30 days. Amendment No.: 270. Facility Operating License No. DPR– 49: The amendment revised the Technical Specifications. VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 Date of initial notice in Federal Register: June 13, 2008 (73 FR 33853). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 9, 2008. No significant hazards consideration comments received: No. Luminant Generation Company LLC, Docket Nos. 50–445 and 50–446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas Date of amendment request: November 29, 2007. Brief description of amendments: The amendments revised the Technical Specification (TS) 3.6.7, ‘‘Spray Additive System,’’ to allow modifications to the facility potentially required to address U.S. Nuclear Regulatory Commission (NRC) Generic Letter 2004–02, ‘‘Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accident at Pressurized-Water Reactors’’ and authorized changes to TS 3.6.7 to remove the current surveillances for sodium hydroxide and insert a surveillance to ensure equilibrium sump pH is greater than or equal to 7.1. Date of issuance: September 12, 2008. Effective date: As of the date of issuance and shall be implemented within 120 days from the date of issuance. Amendment Nos.: Unit 1–147, Unit 2–147. Facility Operating License Nos. NPF– 87 and NPF–89: The amendments revised the Facility Operating Licenses and Technical Specifications. Date of initial notice in Federal Register: December 31, 2007 (72 FR 74360). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 12, 2008. No significant hazards consideration comments received: No. Nine Mile Point Nuclear Station, LLC, Docket No. 50–220, Nine Mile Point Nuclear Station, Unit No. 1, Oswego County, New York Date of application for amendment: July 23, 2007, as supplemented by letter dated January 24, 2008. Brief description of amendment: The amendment revises Technical Specification (TS) Section 3.1.1, ‘‘Control Rod System,’’ to incorporate a provision that should the rod worth minimizer (RWM) become inoperable before a reactor startup is commenced or before the first 12 control rods have been withdrawn, startup will be allowed to continue. This provision will rely on the RWM function being performed PO 00000 Frm 00154 Fmt 4703 Sfmt 4703 manually and will require a double check of compliance with the control rod program by a second licensed operator or other qualified member of the technical staff. The use of this allowance will be limited to one startup in the last calendar year. Date of issuance: July 29, 2008. Effective date: As of the date of issuance to be implemented within 60 days. Amendment No.: 196. Renewed Facility Operating License No. DPR–63: Amendment revised the License and TSs. Date of initial notice in Federal Register: September 11, 2007 (72 FR 51863). The supplemental letter dated January 24, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff’s initial proposed no significant hazards consideration determination. The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated July 29, 2008. No significant hazards consideration comments received: No. Northern States Power Company, Docket No. 50–263, Monticello Nuclear Generating Plant, Wright County, Minnesota Date of application for amendment: April 16, 2008, as supplemented by letter dated August 6, 2008. Brief description of amendment: The amendment conforms Renewed Facility Operating License No. DPR–22 to reflect the fact that Northern States Power Company holds the operating authority of the unit as of the date of this amendment. This license transfer was previously approved by an Order dated September 15, 2008. Date of issuance: September 22, 2008. Effective date: As of the date of issuance and shall be implemented within 1 year. Amendment No.: 156. Facility Operating License No. DPR– 22: Amendment revised the Technical Specifications. Date of initial notice in Federal Register: June 5, 2008 (73 FR 32057). The Commission’s related evaluation of the amendment is contained in a Safety Evaluation dated September 15, 2008. No significant hazards consideration comments received: As provided in 10 CFR 2.1315, no public comments with respect to significant hazards considerations were solicited. E:\FR\FM\07OCN1.SGM 07OCN1 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices Northern States Power Company, Docket Nos. 50–282 and 50–306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota Date of application for amendments: April 16, 2008, as supplemented by letter dated August 6, 2008. Brief description of amendments: The amendments conform the Technical Specifications and Facility Operating License Nos. DPR–42 and DPR–60 to reflect the fact that Northern States Power Company holds the operating authority of the units as of the date of these amendments. This license transfer was previously approved by an Order dated September 15, 2008. Date of issuance: September 22, 2008. Effective date: As of the date of issuance and shall be implemented within 1 year. Amendment Nos.: 188 (for Unit 1) and 177 (for Unit 2). Facility Operating License Nos. DPR– 42 and DPR–60: Amendments revised the Facility Operating Licenses and the Technical Specifications. Date of initial notice in Federal Register: June 5, 2008 (73 FR 32055). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 15, 2008. No significant hazards consideration comments received: As provided in 10 CFR 2.1315, no public comments with respect to significant hazards considerations were solicited. mstockstill on PROD1PC66 with NOTICES PSEG Nuclear LLC, Docket Nos. 50–272 and 50–311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey Date of application for amendments: March 11, 2008, as supplemented on June 17, and July 23, 2008. Brief description of amendments: The amendments revise the Technical Specification (TS) requirements for fuel decay time prior to commencing movement of irradiated fuel in the reactor pressure vessel. Date of issuance: September 24, 2008. Effective date: As of the date of issuance, to be implemented within 30 days. Amendment Nos.: 289 and 273. Facility Operating License Nos. DPR– 70 and DPR–75: The amendments revise the TSs and the license. Date of initial notice in Federal Register: July 15, 2008 (73 FR 40631). The letters dated June 17, and July 23, 2008, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice. VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 24, 2008. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50–321 and 50–366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia Date of application for amendments: April 29, 2008. Brief description of amendments: The amendments revise Technical Specification Figure 3.1.7–1, ‘‘Sodium Penataborate Solution Volume Versus Concentration Requirements,’’ by implementing an editorial change to improve the readability of the figure. Date of issuance: September 23, 2008. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment Nos.: Unit 1–257, Unit 2–201. Renewed Facility Operating License Nos. DPR–57 and NPF–5: Amendments revised the licenses and the technical specifications. Date of initial notice in Federal Register: June 3, 2008 (73 FR 31723). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 23, 2008. No significant hazards consideration comments received: No. Southern Nuclear Operating Company, Inc., Docket Nos. 50–424 and 50–425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of application for amendments: June 27, 2008. Brief description of amendments: The amendments revised the combined Vogtle Electric Generating Plant, Units 1 and 2 Technical Specifications (TS) 5.5.9, ‘‘Steam Generator (SG) Program’’ and TS 5.6.10, ‘‘Steam Generator Tube Inspection Report,’’ to incorporate a one-cycle interim alternate repair criterion in the provisions for SG tube repair criteria for VEGP Unit 2 during refueling outage 2R13 and the subsequent operating cycle. Date of issuance: September 16, 2008. Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance. Amendment Nos.: Unit 1–152, Unit 2–133. PO 00000 Frm 00155 Fmt 4703 Sfmt 4703 58683 Facility Operating License Nos. NPF– 68 and NPF–81: Amendments revised the technical specifications. Date of initial notice in Federal Register: July 14, 2008 (73 FR 40394). The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 16, 2008. No significant hazards consideration comments received: No. STP Nuclear Operating Company, Docket Nos. 50–498 and 50–499, South Texas Project, Units 1 and 2, Matagorda County, Texas Date of amendment request: August 27, 2007, as supplemented by letters dated March 27 and September 5, 2008. Brief description of amendments: The amendments revised the South Texas Project, Units 1 and 2 fire protection program to allow the performance of operator manual actions to achieve and maintain safe shutdown in the event of a fire, in lieu of meeting circuit separation requirements specified in Title 10 of the Code of Federal Regulations, Part 50, Appendix R, Section III.G.2, for a fire in Fire Area 32 located in the Mechanical/Electrical Auxiliary Building. License Condition 2.E of the operating licenses is revised. Date of issuance: September 16, 2008. Effective date: As of the date of issuance and shall be implemented within 60 days of issuance. Amendment Nos.: Unit 1–186, Unit 2–173. Facility Operating License Nos. NPF– 76 and NPF–80: The amendments revised the Facility Operating Licenses. Date of initial notice in Federal Register: November 20, 2007 (72 FR 65373). The supplemental letters dated March 27 and September 5, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination as published in the Federal Register. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 16, 2008. No significant hazards consideration comments received: No. Virginia Electric and Power Company, et al., Docket Nos. 50–280 and 50–281, Surry Power Station, Units 1 and 2, Surry County, Virginia Date of application for amendments: September 19, 2007, as supplemented on April 11, 2008. Brief Description of amendments: These amendments revised various Technical Specification (TS) setting E:\FR\FM\07OCN1.SGM 07OCN1 58684 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices limits and the overtemperature DT/ overpower DT time constants in TS 2.3 and TS 3.7. The methodology for determining the revised setting limits and time constants is in agreement with methods 1 and 2 in ‘‘The Instrumentation, Systems, and Automation Society (ISA),’’ Standard ISA–R67.04, Part II, ‘‘Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation.’’ Date of issuance: September 17, 2008. Effective date: As of the date of issuance and shall be implemented within 60 days. Amendment Nos.: 261 and 261. Renewed Facility Operating License Nos. DPR–32 and DPR–37: Amendments changed the licenses and the technical specifications. Date of initial notice in Federal Register: October 23, 2007 (72 FR 60036). The supplement dated April 11, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff’s original proposed no significant hazards consideration determination published in the Federal Register on October 23, 2007. The Commission’s related evaluation of the amendments is contained in a Safety Evaluation dated September 17, 2008. No significant hazards consideration comments received: No. Dated at Rockville, Maryland, this 29th day of September 2008. For the Nuclear Regulatory Commission. Joseph G. Giitter, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. E8–23342 Filed 10–6–08; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION Sunshine Federal Register Notice mstockstill on PROD1PC66 with NOTICES AGENCY HOLDING THE MEETINGS: Nuclear Regulatory Commission. DATES: Weeks of October 6, 13, 20, 27, November 3, 10, 2008. PLACE: Commissioners’ Conference Room, 11555 Rockville Pike, Rockville, Maryland. STATUS: Public and Closed. Week of October 6, 2008 Monday, October 6, 2008 12:55 p.m. Affirmation Session (Public Meeting) (Tentative) a. Oyster Creek, Indian Point, Pilgrim, and Vermont Yankee License VerDate Aug<31>2005 18:23 Oct 06, 2008 Jkt 217001 Renewals, Docket Nos. 50–219–LR, 50–247–LR, 50–286–LR, 50–293– LR, 50–271–LR, Petition to Suspend Proceedings (Tentative). b. Pacific Gas and Electric Co. (Diablo Canyon ISFSI), Docket No. 72–26– ISFSI, Decision on the Merits of San Luis Obispo Mothers for Peace’s Contention 2 (Tentative). c. EnergySolutions (Radioactive Waste Import/Export)—EnergySolutions’ Applications for Low-Level Radioactive Waste Import and Export Licenses (Tentative). 1 p.m. Discussion of Security Issues (Closed—Ex. 1 and 3). Week of October 13, 2008—Tentative There are no meetings scheduled for the week of October 13, 2008. Week of October 20, 2008—Tentative Wednesday, October 22, 2008 9:30 a.m. Briefing on New Reactor Issues—Construction Readiness, Part 1 (Public Meeting) (Contact: Roger Rihm, 301 415–7807). 1:30 p.m. Briefing on New Reactor Issues—Construction Readiness, Part 2 (Public Meeting) (Contact: Roger Rihm, 301 415–7807). Both parts of this meeting will be Webcast live at the Web address— http://www.nrc.gov. Week of October 27, 2008—Tentative There are no meetings scheduled for the week of October 27, 2008. Week of November 3, 2008—Tentative Thursday, November 6, 2008 1:30 p.m. Briefing on NRC International Activities (Public Meeting) (Contact: Karen Henderson, 301 415–0202). This meeting will be webcast live at the Web address—http://www.nrc.gov. Friday, November 7, 2008 2 p.m. Meeting with Advisory Committee on Reactor Safeguards (Public Meeting) (Contact: Tanny Santos, 301 415–7270). This meeting will be Webcast live at the Web address—http://www.nrc.gov. Week of November 10, 2008—Tentative There are no meetings scheduled for the week of November 10, 2008. *The schedule for Commission meetings is subject to change on short notice. To verify the status of meetings, call (recording)—(301) 415–1292. Contact person for more information: Michelle Schroll, (301) 415–1662. The NRC Commission Meeting Schedule can be found on the Internet at: http://www.nrc.gov/about-nrc/policymaking/schedule.html. PO 00000 Frm 00156 Fmt 4703 Sfmt 4703 The NRC provides reasonable accommodation to individuals with disabilities where appropriate. If you need a reasonable accommodation to participate in these public meetings, or need this meeting notice or the transcript or other information from the public meetings in another format (e.g. braille, large print), please notify the NRC’s Disability Program Coordinator, Rohn Brown, at 301–492–2279, TDD: 301–415–2100, or by e-mail at rohn.brown@nrc.gov. Determinations on requests for reasonable accommodation will be made on a case-by-case basis. This notice is distributed by mail to several hundred subscribers; if you no longer wish to receive it, or would like to be added to the distribution, please contact the Office of the Secretary, Washington, DC 20555 (301–415–1969). In addition, distribution of this meeting notice over the Internet system is available. If you are interested in receiving this Commission meeting schedule electronically, please send an electronic message to darlene.wright@nrc.gov. Dated: October 2, 2008. R. Michelle Schroll, Office of the Secretary. [FR Doc. E8–23844 Filed 10–3–08; 4:15 pm] BILLING CODE 7590–01–P OFFICE OF PERSONNEL MANAGEMENT [OMB Control No. 3206–0206; Form RI 25– 37] Submission for OMB Review; Request for Comments on a Revised Information Collection Office of Personnel Management. ACTION: Notice. AGENCY: SUMMARY: In accordance with the Paperwork Reduction Act of 1995 (Pub. L. 104–13, May 22, 1995), this notice announces that the Office of Personnel Management (OPM) has submitted to the Office of Management and Budget (OMB) a request for review of a revised information collection. This information collection, ‘‘Evidence to Prove Dependency of a Child’’ (OMB Control No. 3206–0206; form RI 25–37), is designed to collect sufficient information for OPM to determine whether the surviving child of a deceased federal employee is eligible to receive benefits as a dependent child. Approximately 250 forms are completed annually. We estimate it takes approximately 60 minutes to assemble the needed documentation. E:\FR\FM\07OCN1.SGM 07OCN1
[Pages 58669-58684]
[FR Doc No: E8-23342]
proposed to be issued from September 11, 2008 to September 24, 2008.
The last biweekly notice was published on September 23, 2008 (73 FR
54862).
[[Page 58671]]
the Atomic Safety and Licensing Board that the petition and/or request
should be granted and/or the contentions should be admitted, based on a
balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii). To
be timely, filings must be submitted no later than 11:59 p.m. Eastern
Time on the due date.
Date of amendments request: August 28, 2008.
Description of amendments request: The amendment would relocate the
main steam isolation valve times in Technical Specification (TS)
section 3.7.2, ``Main Steam Isolation Valves (MSIVs)'' to the licensee
controlled document that is referenced in the Bases. In addition, the
valve isolation times in the TS are replaced with the phrase ``within
limits.'' The changes are consistent with the Nuclear Regulatory
Commission approved Technical Specification Task Force (TSTF)-491,
Revision 2, ``Removal of Main Steam and Main Feedwater Valve Isolation
Times From Technical Specifications.'' The availability of the TS
improvement was published in the Federal Register on December 29, 2006
(71 FR 250) as part of the consolidated item improvement process
with the phrase, ``within limits.''
the requirements of 10 CFR 50.59, ``Changes, tests and
experiments'', to ensure that such changes do not result in more
than minimal increase in the probability or consequences of an
of a New or Different Kind of Accident From Any Previously Evaluated
replaced in the TS with the phrase ``within limits''. The changes do
methods governing normal pant operation. The requirements in the TS
steam and main feedwater isolation valves. Changes to the Bases are
Attorney for licensee: Carey Fleming, Sr. Counsel-Nuclear
Date of amendments request: June 19, 2008.
Description of amendments request: The proposed change would: (1)
Revise Technical Specifications (TS) control rod notch surveillance
requirement (SR) frequency in TS 3.1.3, ``Control Rod Operability,''
and (2) revise Example 1.4-3 in Section 1.4, ``Frequency,'' to clarify
the applicability of the 1.25 surveillance test extension. The licensee
is proposing to adopt the approved Technical Specification Task Force
Testing Frequency.'' A notice of availability of TSTF-475, Revision 1,
was published in the Federal Register on November 13, 2007 (72 FR
63935).
In addition, the proposed amendment would remove Note 2 associated
with SR 3.1.3.3 for Unit 1, which is a cycle-specific note and has
expired. This change is administrative in nature and does not affect
the no significant hazards consideration (NSHC) determination.
50.91(a), the licensee, in its application dated June 19, 2008,
affirmed the applicability of the published model NSHC determination,
OPERABILITY'', (2) [not applicable to BSEP], and (3) revise Example
1.4-3 in Section 1.4 ``Frequency'' to clarify the applicability of
the 1.25 surveillance test interval extension. The consequences of
an accident after adopting TSTF-475, Revision 1 are no different
than the consequences of an accident prior to adoption. Therefore,
frequency in TS 3.1.3, ``Control Rod OPERABILITY'', (2) [not
applicable to BSEP], and (3) revise Example 1.4-3 in Section 1.4
test interval extension. The GE Nuclear Energy Report, ``CRD
Notching Surveillance Testing for Limerick Generating Station,''
dated November 2006, concludes that extending the control rod notch
test interval from weekly to monthly is not expected to impact the
proposed changes in TSTF-475, Revision 1 [. . .] do not involve a
Based on the review of the above analysis, the NRC staff finds that
the three standards in 10 CFR 50.92(c) are satisfied. Therefore, the
NRC staff proposes to determine that the amendment request involves
II-Legal Department, Progress Energy Service Company, LLC, Post Office
Limiting Condition for Operation (LCO) 3.10.1, and the associated
Bases, to expand its scope to include provisions for temperature
excursions greater than 212 degrees Fahrenheit ([deg]F) as a
consequence of inservice leak and hydrostatic testing, and as a
consequence of scram time testing initiated in conjunction with an
conditions to be in Mode 4.
Register on October 27, 2006 (71 FR 63050). The NRC also issued a
Federal Register notice on August 21, 2006 (71 FR 48561) that provided
a model safety evaluation and a model no significant hazards
consideration (NSHC) determination relating to modification of
requirements regarding LCO 3.10.1, ``Inservice Leak and Hydrostatic
Testing Operation.'' In its application dated June 19, 2008, the
licensee affirmed the applicability of the model NSHC determination.
Basis for proposed NSHC determination: As required by 10 CFR Part
Based on the above, the proposed change presents NSHCs under the
standards set forth in 10 CFR 50.92(c). Therefore, the NRC staff
proposes to determine that the amendment request involves NSHC.
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, NC 27602.
Entergy Nuclear Operations, Inc., Docket No. 50-3, Indian Point Energy
Center, Unit 1 Westchester County, New York
delete license conditions and Technical Specification (TS) requirements
which relate to the storage of spent nuclear fuel in the Indian Point
Unit 1 (IP1) Fuel Handling Building Spent Fuel Pool. The spent fuel is
to be transferred to, and stored at, the existing Indian Point
Independent Spent Fuel Storage Installation (ISFSI), Docket No. 72-51.
The removal of the stored spent fuel and drain down of the spent fuel
pools renders many of the license conditions and TS requirements
unnecessary and burdensome.
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
No. The proposed changes are all contingent on the prior removal
of the stored spent fuel from the IP1 Spent Fuel Pool (SFP) to the
Indian Point Energy Center (IPEC) ISFSI. The accidents previously
evaluated in the IP1 Final Safety Analysis Report (FSAR), which
consists of the IP1 Decommission Plan and Supplemental Environmental
Information, are stored fuel related accidents. The removal of the
stored fuel from the IP1 facility to the IPEC ISFSI precludes the
possibility of these accidents.
Consequently, the proposed changes to the license do not involve
a significant increase in the probability or the consequences of an
2. Does the proposed change create the probability of a new or
The proposed changes are all contingent on the prior removal of
the stored spent fuel from the IP1 SFP to the IPEC ISFSI. With the
removal of the stored spent fuel from the IP1 facility, and
considering the IP1 has been in a SAFESTOR mode for over thirty
years, no significant source term remains which could result in any
postulated radiological event that would impact the health and
Therefore, the proposed changes to the IP1 license consequently
of the stored spent fuel from the IP1 SFP to the IPEC ISFSI. Upon
the removal of spent fuel, the Technical Specifications being
deleted no longer are required to protect the health and safety of
the public or occupational workers from the potential adverse
conditions, hazards or accidents as discussed in the FSAR.
proposed amendments would not involve a significant reduction in the
Based upon the reasoning presented above it appears that the three
Attorney for licensee: Assistant General Counsel, Entergy Nuclear
Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.
NRC Branch Chief: Theodore Smith, Acting.
several surveillance requirements (SRs) and add SR 3.8.1.21 in
Sources--Operating,'' and TS 3.8.2, ``AC Sources--Shutdown.'' The
amendment would allow the slow-start testing sequence of the diesel
generators in order to reduce the stress and wear on the equipment.
The proposed change affects the surveillance requirements for
the Diesel Generators (DGs). The DGs are onsite standby power
sources intended to provide redundant and reliable power to ESF
[Engineered Safety Feature] systems credited as accident mitigating
features in design basis [accident] analyses. Per NRC Regulatory
Guide (RG) 1.9, Revision 3, which is referenced in Grand Gulf
Nuclear Station (GGNS) UFSAR [Updated Final Safety Analysis Report
Section] 8.3.1.2.1, the proposed change is intended to allow slower
starts of the DGs during testing in order to reduce DG aging effects
due to excessive testing conditions. As such, the proposed change
will result in improved DG reliability and availability, thereby
providing additional assurance that the DGs will be capable of
performing their safety function. The method of starting the
emergency diesel generators for testing purposes does not affect the
probability of any previously evaluated accident. Although the
change allows slower starts for the monthly tests, the more rapid
start function, assumed in the accident analysis, is unchanged and
will be verified on a 184 day frequency. Therefore the accident
analysis consequences are not affected [by the proposed change].
increase in the probability [or] consequences of any accident
the onsite ac sources, i.e. the Diesel Generators. Accordingly, the
proposed change does not involve any change to the configuration or
method of operation of any plant equipment that could cause an
accident. In addition, no new failure modes have been created nor
has any new limiting failure been introduced as a result of the
proposed surveillance changes.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
The proposed change is intended to bring the existing GGNS TS
requirements for the onsite AC sources in line with regulatory
guidance. Under the proposed change, the DGs will remain capable of
performing their safety function, and the effects of aging on the
DGs will be reduced by eliminating unnecessary testing. The DG start
times assumed in the current accident analyses are unchanged and
will be verified on a 184 day frequency.
Counsel--Nuclear Entergy Services, Inc., 1340
Echelon Parkway, Jackson, Mississippi 39213.
GE Hitachi Nuclear Energy (GEH), License No. DR-10, Docket No. 50-183,
ESADA Vallecitos Experimental Superheat Reactor (EVESR)
Date of amendment request: June 23, 2008.
would modify the Technical Specification (TS) requirements to revise
the scope of dismantling activities that GEH can perform under The
Vallecitos Nuclear Center Liabilities Reduction Project and specify
radiological control requirements of 10 CFR Part 20. Two TS changes are
proposed. The proposed changes to the TS:
Allow GEH to conduct dismantling activities below the 549-
ft elevation level within the containment building; and
Revise the physical security requirements for access to
areas below the 549-ft elevation level within the containment building.
The application for license amendment is available electronically
at the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/
the June 23, 2008, request is ML081780099.
electronically on the public computers located at the NRC's PDR, 01F21,
Proposed change one is an administrative change submitted to
clarify the area where dismantling activities will occur as
authorized by the facility license. The majority of the component
removal activities will occur in areas below the 549-ft. elevation.
Proposed change two removes the specific shielding and covering
requirements for the reactor vessel, shield plug storage pit and the
empty spent fuel storage pit and modifies the access control
requirements to be consistent with 10 CFR 20. The EVESR reactor was
shutdown in 1967 and has remained in a ``Possess Only'' status. All
fuel bundles were removed from the facility and the radiation and
contamination levels have been reduced by the removal of radioactive
material and natural decay. No aspect of the proposed changes will
Proposed change one is an administrative, therefore there it
cannot create a new or different kind of accident. Removal of the
specific shielding and covering requirements for the reactor vessel,
shield plug storage pit and the empty spent fuel storage pit and
modification of the access control requirements as described in
proposed change two will not impact the function or integrity of the
reactor pressure vessel, which is the primary safety system required
to be maintained by the license. The proposed changes do not create
accident evaluated.
Removal of the specific shielding, covering and access control
requirements will not result in a reduction of the margin of the
safety for the EVESR facility. These controls were implemented to
provide shielding and access controls to High Radiation Areas. Since
the reactor is no longer operating and the radiological conditions
have been significantly reduced, the specific controls specified in
the current technical specifications are not required. All areas in
the EVESR containment will be controlled in accordance with 10 CFR
20. High Radiation areas will be controlled in a manner consistent
with the requirements of 10 CFR 20.1601. The proposed changes do not
affect the margins of safety.
The NRC staff has reviewed the licensee's analysis and, based upon
the staff's review of the licensee's analysis, as well as the staff's
own evaluation, the staff concludes that the three standards of 10 CFR
GEH, Manager, Regulatory Compliance & EHS: LaTonya L. Mahlahla.
50-410, Nine Mile Point Nuclear Station Unit Nos. 1 and 2 (NMP 1 and
2), Oswego County, New York
Date of amendment request: June 24, 2008.
revise the Technical Specifications (TSs) by (1) replacing the
references to Section XI of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code with references to the
ASME Code for Operation and Maintenance of Nuclear Power Plants (OM
Code); and (2) revising the allowance to extend Inservice Testing (IST)
frequencies by 25 percent to clearly state that the allowance is
applicable to IST frequencies of 2 years or less. The proposed changes
are based on TS Task Force (TSTF) Standard Technical Specification
Change Traveler 479-A, Revision 0, ``Limit Inservice Testing Program SR
3.0.2 Application to Frequencies of 2 Years or Less.''
The proposed changes revise the IST Program sections of the NMP1
and NMP2 TS to maintain consistency with the requirements of 10 CFR
50.55a(f)(4) regarding the IST of pumps and valves that are
classified as ASME Code Class 1, Class 2, and Class 3. The proposed
changes incorporate revisions to the ASME Code that result in a net
improvement in the measures for testing pumps and valves. The
proposed changes also revise the allowance to extend IST frequencies
by 25 percent to clearly state that this allowance is applicable to
IST frequencies of 2 years or less.
The proposed TS changes are administrative in nature. They do
not impact any accident initiators, the ability to mitigate
previously evaluated accidents, or the assumptions used in
evaluating the radiological consequences of previously evaluated
accidents. The proposed changes do not involve the addition or
removal of any equipment, or any design changes to the facilities.
[[Page 58675]]
plants (i.e., no new equipment will be installed) or involve a
changes will not impose any new or different requirements or
introduce a new accident initiator, accident precursor, or failure
plants (i.e., no new equipment will be installed) or change the
methods governing normal plant operation. The proposed changes do
not modify the safety limits or setpoints at which protective
actions are initiated, and do not change the requirements governing
operation or availability of safety equipment assumed to operate to
preserve margins of safety. The incorporation of revisions to the
ASME Code results in a net improvement in the measures for testing
pumps and valves. The safety function of the affected pumps and
valves will be maintained.
revise the Technical Specification (TS) testing frequency for the
The proposed change revises the frequency of SR 3.1.4.2, control rod
scram time testing, from ``120 days cumulative operation in Mode 1'' to
``200 days cumulative operation in Mode 1.'' These changes are based on
TS Task Force (TSTF) change traveler TSTF-460 (Revision 0) that has
been approved generically for the Boiling-water reactor (BWR) Standard
TS, NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) by revising the frequency
of SR 3.1.4.2, control rod scram time testing, from ``120 days
cumulative operation in MODE 1'' to ``200 days cumulative operation in
MODE 1.'' The NRC staff issued a notice of availability of a model no
significant hazards consideration determination (NSHCD) for referencing
23, 2004 (69 FR 51864) using the consolidated line item improvement
process (CLIIP). The licensee affirmed the applicability of the model
NSHC determination and the model safety evaluation in its application
dated July 7, 2008.
of NSHC, based on the model NSHCD published in the Federal Register on
August 23, 2004 (69 FR 51864), is presented below:
probability or consequences of an accident, previously evaluated?
The proposed change continues to test the control rod scram time
to ensure the assumptions in the safety analysis are protected.
NRC Branch Chief: Mark Kowal.
Description of amendment request: PPL Susquehanna, LLC (the
licensee) requests adoption of the Nuclear Regulatory Commission (NRC)
approved Technical Specification Task Force (TSTF) change traveler
TSTF-475, (Revision 1), ``Control Rod Notch Testing Frequency and SRM
[Source Range Monitor] Insert Control Rod Action,'' to change the
Standard Technical Specifications (STS) for General Electric (GE)
Plants (NUREG-1433, BWR/4 to the plant specific TS, that allows: (1)
Revising the frequency of Surveillance Requirement (SR) 3.1.3.2, notch
testing of fully withdrawn control rod, from ``7 days after the control
rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM'' to
``31 days after the control rod is withdrawn and THERMAL POWER is
greater than the LPSP [Low Power Set Point] of the RWM [Rod With
Minimizer]'', and (2) revising Example 1.4-3 in Section 1.4
``Frequency'' to clarify that the 1.25 surveillance test interval
extension in SR 3.0.2 is applicable to time periods discussed in NOTES
in the ``SURVEILLANCE'' column in addition to the time periods in the
``FREQUENCY'' column.
The NRC staff issued a notice of availability in the Federal
Register on November 13, 2007, (72 FR 63935), which included a model
safety evaluation (SE) and model no significant hazards consideration
determination (NSHCD), using the consolidated line-item improvement
process (CLIIP), of possible amendments to revise the plant specific
TS, to allow: (1) Revising the frequency of SR 3.1.3.2, notch testing
of fully withdrawn control rod, from ``7 days after the control rod
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is withdrawn and THERMAL POWER is greater than the LPSP of RWM'' to
greater than the LPSP of the RWM'', (2) adding the word ``fully'' to
LCO 3.3.1.2 Required Action E.2 to clarify the requirement to fully
insert all insertable control rods in core cells containing one or more
fuel assemblies when the associated SRM instrument is inoperable, and
(3) revising Example 1.4-3 in Section 1.4 ``Frequency'' to clarify that
the 1.25 surveillance test interval extension in SR 3.0.2 is applicable
to time periods discussed in NOTES in the ``SURVEILLANCE'' column in
addition to the time periods in the ``FREQUENCY'' column. The licensee
affirmed the applicability of the model SE and model NSHC determination
in its application dated July 7, 2008.
November 13, 2007 (72 FR 63935), is presented below:
surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod
OPERABILITY'', (2) clarify the requirement to fully insert all
insertable control rods for the limiting condition for operation (LCO)
in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring
surveillance test interval extension. Implementing TSTF-475, Revision 1
does not change the control rod notch test method. Implementing TSTF-
475, Revision 1 decreases the performance frequency of the control rod
notch test. Therefore, the proposed change does not involve a
evaluated. The consequences of an accident after adopting TSTF-475,
Revision 1 are no different than the consequences of an accident prior
The proposed amendment will: (1) Revise the TS SR 3.1.3.2 frequency
in TS 3.1.3, ``Control Rod OPERABILITY'', and (2) revise Example 1.4-3
in Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. The GE Nuclear Energy Report,
``CRD Notching Surveillance Testing for Limerick Generating Station,''
decision to change the surveillance frequency. Therefore, the proposed
changes in TSTF-475, Revision 1 do not involve a significant reduction
delete the requirements related to plant staff working hours from
Section 6.0, ``Administrative Controls'' of the respective plants'
Technical Specifications (TSs). The current working hour requirements
were incorporated into the TSs as a result of the guidance in Nuclear
Regulatory Commission (NRC) Generic Letter (GL) 82-12, ``Nuclear Power
Plant Staff Working Hours.'' The guidance in GL 82-12 has been
superseded by the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), Part 26, ``Fitness for Duty Programs,'' Subpart
I, ``Managing Fatigue'' which was published in the Federal Register on
March 31, 2008, as part of the final rulemaking for Part 26. As
discussed in the Federal Register notice for the final rule (73 FR
16966), Subpart I must be implemented by licensees no later than
October 1, 2009. The licensee stated that the proposed amendments would
support implementation of the new requirements in Subpart I.
The removal of GL 82-12 administrative controls will not remove
the requirement to control work hours and manage fatigue. Removal of
TS controls required by GL 82-12 will be performed concurrently with
the implementation of the more conservative [10 CFR Part 26],
Subpart I, requirements. The proposed changes do not impact the
maintained, modified, tested, or inspected. The proposed changes do
not impact the initiators or assumptions of analyzed events, nor do
they impact the mitigation of accidents or transient events.
Because these new requirements are more conservative with
respect to work hour controls and fatigue management, this will not
significantly increase the probability or consequence of an accident
The proposed changes remove GL 82-12 administrative controls
from [the] TS to support the implementation of Subpart I to [10 CFR
Part 26]. The Subpart I regulations are more restrictive than the
current guidance in [the] TS and would add conservatism to work hour
controls and fatigue management. Work hours will continue to be
controlled in accordance with NRC requirements. The new rule
continues to allow for deviations from controls to mitigate or
restrict work hours at the
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expense of the health and safety of the public as well as plant
personnel. The proposed changes do not alter plant configuration,
require that new plant equipment be installed, alter assumptions
made about accidents previously evaluated, add any initiators, or
impact the function of plant SSCs or the manner in which SSCs are
Because the proposed changes do not remove the station's
requirement to control work hours and increases the conservatism of
work hour controls by changing administrative scheduling
requirements, the proposed changes do not create the possibility of
work hours in managing fatigue. Salem and Hope Creek Generating
Stations will continue their fitness-for-duty and behavioral
observation programs, both of which will be strengthened by
compliance with the new Part 26 regulation. The proposed changes add
conservatism to fatigue management and contribute to the margin of
safety. The proposed changes do not involve any physical changes to
plant SSCs or the manner in which SSCs are operated, maintained,
modified, tested, or inspected. The proposed changes do not involve
a change to any safety limits, limiting safety system settings,
limiting conditions of operation, or design parameters for any SSC.
The proposed changes do not impact any safety analysis assumptions
and do not involve a change in initial conditions, system response
this review, with changes in the areas noted above, it appears that the
relocate Technical Specification (TS) 3/4.7.5, ``Snubbers,'' to the
Hope Creek Generating Station (HCGS) Technical Requirements Manual
(TRM). TS 6.10.3.l, which specifies retention requirements for records
of snubber service life monitoring pursuant to TS 4.7.5, would also be
relocated to the TRM. In addition, the amendment would add new TS
Limiting Condition for Operation (LCO) 3.0.8, ``Inoperability of
Snubbers,'' and would modify LCO 3.0.1 to reference LCO 3.0.8.
the probability or consequences of an accident previously evaluate