Consolidated Guidance 

About Materials Licenses

Program-Specific Guidance About 

Licenses of Broad Scope

Final Report

Office of Nuclear Material Safety and Safeguards

NUREG-1556

Volume 11, Rev. 1 

Consolidated Guidance 

About Materials Licenses

Program-Specific Guidance About 

Licenses of Broad Scope

Final Report

Manuscript Completed: February 2017

Date Published: February 2017 

Prepared by: 

C. Frazier

P. Kern

G. Rivera-Capella

R. J. Torres

B. Ullrich

Office of Nuclear Material Safety and Safeguards

 

NUREG-1556

Volume 11, Rev. 1 

iii 

ABSTRACT 

This technical report contains information intended to provide program-specific guidance and 

assist applicants and licensees in preparing applications for materials licenses of broad scope. 

In particular, it describes the types of information needed to complete U.S. Nuclear Regulatory 

Commission (NRC) Form 313, “Application for Materials License.” This document describes 

both the methods acceptable to the NRC license reviewers in implementing the regulations and 

the techniques used by the reviewers in evaluating the application to determine if the proposed 

activities are acceptable for licensing purposes.

NUREG–1556, Volume 11, Revision 1, is not intended to be used alone. Because broad scope 

licensees may be involved in many different program areas (e.g., medicine, research and 

development, manufacturing and distribution), this document frequently refers the user to other 

relevant program-specific guidance documents in the NUREG–1556 series. 

Paperwork Reduction Act Statement 

This NUREG references information collection requirements that are subject to the Paperwork 

Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved 

by the Office of Management and Budget (OMB), approval numbers 3150-0044, 3150-0014, 

3150-0035, 3150-0017, 3150-0016, 3150-0001, 3150-0015, 3150-0007, 3150-0010, 3150-0158, 

3150-0214, 3150-0130, 3150-0020, 3150-0021, 3150-0135, 3150-0009, 3150-0008, 3150-0123 

and 3150-0120. 

Public Protection Notification 

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 

information or an information collection requirement unless the requesting document displays a 

currently valid OMB control number
.
v 

FOREWORD 

The U.S. Nuclear Regulatory Commission’s (NRC’s) NUREG–1556 technical report series 

provides a comprehensive source of reference information about various aspects of materials 

licensing and materials program implementation. These reports, where applicable, describe a 

risk-informed, performance-based approach to licensing consistent with the current regulations. 

The reports are intended for use by applicants, licensees, license reviewers, and other NRC 

personnel. The NUREG–1556 series currently includes the following volumes: 

Volume No. Volume Title 

1 Program-Specific Guidance About Portable Gauge Licenses 

2 Program-Specific Guidance About Industrial Radiography Licenses 

3 Applications for Sealed Source and Device Evaluation and Registration 

4 Program-Specific Guidance About Fixed Gauge Licenses 

5 Program-Specific Guidance About Self-Shielded Irradiator Licenses 

6 Program-Specific Guidance About 10 CFR Part 36 Irradiator Licenses 

7 Program-Specific Guidance About Academic, Research and Development, and 

Other Licenses of Limited Scope Including Electron Capture Devices and 

X-Ray Fluorescence Analyzers 

8 Program-Specific Guidance About Exempt Distribution Licenses 

9 Program-Specific Guidance About Medical Use Licenses 

10 Program-Specific Guidance About Master Materials Licenses 

11 Program-Specific Guidance About Licenses of Broad Scope 

12 Program-Specific Guidance About Possession Licenses for Manufacturing and 

Distribution 

13 Program-Specific Guidance About Commercial Radiopharmacy Licenses 

14 Program-Specific Guidance About Well Logging, Tracer, and Field Flood Study 

Licenses 

15 Guidance About Changes of Control and About Bankruptcy Involving 

Byproduct, Source, or Special Nuclear Materials Licenses 

16 Program-Specific Guidance About Licenses Authorizing Distribution to General 

Licensees 

17 Program-Specific Guidance About Special Nuclear Material of Less Than 

Critical Mass Licenses 

18 Program-Specific Guidance About Service Provider Licenses 

19 Guidance for Agreement State Licensees About NRC Form 241 “Report of 

Proposed Activities in Non-Agreement States, Areas of Exclusive Federal 

Jurisdiction, or Offshore Waters” and Guidance for NRC Licensees Proposing 

to Work in Agreement State Jurisdiction (Reciprocity) 

20 Guidance About Administrative Licensing Procedures 

21 Program-Specific Guidance About Possession Licenses for Production of 

Radioactive Material Using an Accelerator 

The current document, NUREG–1556, Volume 11, Revision 1, “Consolidated Guidance About 

Materials Licenses: Program-Specific Guidance About Licenses of Broad Scope,” is intended 

for use by applicants, licensees, and NRC staff. This revision provides a general update to the 

previous information contained in NUREG–1556, Volume 11, dated April 1999, to incorporate 

certain security requirements and other regulatory and policy changes that have been 

implemented since the last revision was published. 
vi 

This report takes a risk-informed, performance-based approach to licensing the use of 

radioactive material under a broad scope authorization. A team composed of staff from NRC 

Headquarters, NRC regional offices, and Agreement States prepared this document, drawing on 

their collective experience in radiation safety in general and as specifically applied to users of 

radioactive material under a broad scope license. 

NUREG–1556, Volume 11, Revision 1, is not a substitute for NRC or Agreement State 

regulations. The approaches and methods described in this report are provided for information 

only. Methods and solutions different from those described in this report may be acceptable if 

applicants include a basis for the staff to make the determinations needed to issue or renew a 

license. 

The comments received during the public comment period for NUREG–1556, Volume 11, 

Revision 1, were summarized and addressed in a document that can be located on the NRC’s 

Agencywide Documents and Management System (ADAMS) under ML15357A092. Access to 

ADAMS is available on the public Web site at: http://www.nrc.gov/reading-rm/adams.html. The 

comments received by NRC included general corrections, comments on training, and comments 

on safety culture. 

 

Daniel S. Collins, Director 

Division of Material Safety, State, Tribal, and Rulemaking Programs 

Office of Nuclear Material Safety and Safeguards 
vii 

TABLE OF CONTENTS 

ABSTRACT ................................................................................................................................. iii 

FOREWORD ................................................................................................................................. v 

FIGURES AND TABLES ............................................................................................................. xi 

ACKNOWLEDGMENTS ............................................................................................................ xiii 

ABBREVIATIONS ...................................................................................................................... xv 

1 PURPOSE OF REPORT .................................................................................................... 1-1

2 AGREEMENT STATES ...................................................................................................... 2-1

2.1 Jurisdiction Determination ......................................................................................... 2-1

2.2 Reciprocal Recognition of Specific Licenses ............................................................. 2-3

3 MANAGEMENT RESPONSIBILITY ................................................................................... 3-1

3.1 Commitments and Responsibilities ........................................................................... 3-1

3.2 Safety Culture ............................................................................................................ 3-2

4 APPLICABLE REGULATIONS .......................................................................................... 4-1

5 HOW TO FILE .................................................................................................................... 5-1

5.1 Application Preparation ............................................................................................. 5-1

5.2 Where to File ............................................................................................................. 5-1

5.3 Paper Applications ..................................................................................................... 5-2

5.4 Electronic Applications .............................................................................................. 5-2

6 IDENTIFYING AND PROTECTING SENSITIVE INFORMATION ...................................... 6-1

7 APPLICATION AND LICENSE FEES ................................................................................ 7-1

8 CONTENTS OF AN APPLICATION ................................................................................... 8-1

8.1 Item 1: License Action Type ..................................................................................... 8-2

8.2 Item 2: Name and Mailing Address of Applicant ....................................................... 8-2

8.2.1 Notification of Bankruptcy Proceedings ......................................................... 8-2

8.3 Item 3: Address(es) Where Licensed Material Will Be Used or Possessed ............. 8-3

8.4 Item 4: Person to be Contacted About This Application ........................................... 8-5

8.5 Item 5: Radioactive Material ..................................................................................... 8-5

8.5.1 Unsealed or Sealed Byproduct Material ........................................................ 8-5

8.5.2 Financial Assurance and Recordkeeping for Decommissioning .................... 8-9

8.5.3 Emergency Plan .......................................................................................... 8-12

8.6 Item 6: Purpose(s) for Which Licensed Material Will Be Used ............................... 8-12

8.7 Item 7: Individual(s) Responsible for Radiation Safety Program and Their 

Training and Experience .......................................................................................... 8-14

8.7.1 Executive Management ............................................................................... 8-14

8.7.2 Radiation Safety Committee ........................................................................ 8-15

8.7.3 Radiation Safety Officer ............................................................................... 8-18

8.7.4 Radiation Safety Office Staff ........................................................................ 8-22

8.8 Item 8: Training for Individuals Working in or Frequenting Restricted Areas ......... 8-22

8.9 Item 9: Facilities and Equipment ............................................................................ 8-24

8.10 Item 10: Radiation Safety Program ........................................................................ 8-27

8.10.1 Audit and Review of Program ...................................................................... 8-27
viii 

8.10.2 Radiation Monitoring Instruments ................................................................ 8-30

8.10.3 Material Receipt and Accountability ............................................................. 8-31

8.10.4 Occupational Dose ...................................................................................... 8-35

8.10.5 Public Dose .................................................................................................. 8-39

8.10.6 Safe Use of Radionuclides and Emergency Procedures ............................. 8-40

8.10.7 Surveys and Leak Tests .............................................................................. 8-43

8.10.8 Transportation .............................................................................................. 8-45

8.10.9 Security Program for Category 1 and Category 2 Radioactive 

Material ........................................................................................................ 8-47

8.11 Item 11: Waste Management ................................................................................. 8-48

8.12 Item 12: License Fees ............................................................................................. 8-54

8.13 Item 13: Certification ............................................................................................... 8-54

9 LICENSE AMENDMENTS AND RENEWALS ................................................................... 9-1

9.1 Timely Notification of Transfer of Control .................................................................. 9-1

10 APPLICATIONS FOR EXEMPTIONS .............................................................................. 10-1

11 TERMINATION OF ACTIVITIES ...................................................................................... 11-1

APPENDICES 

APPENDIX A U.S. NUCLEAR REGULATORY COMMISSION FORM 313 ...................... A–1 

APPENDIX B SUGGESTED FORMAT FOR PROVIDING INFORMATION 

 REQUESTED IN ITEMS 5 THROUGH 11 

 OF U.S. NUCLEAR REGULATORY COMMISSION FORM 313 ................ B–1 

APPENDIX C INFORMATION NEEDED FOR FIELD USE OF BYPRODUCT 

MATERIAL .................................................................................................. C–1 

APPENDIX D MODEL DELEGATION OF AUTHORITY FOR RADIATION 

SAFETY OFFICER ...................................................................................... D–1 

APPENDIX E FACILITIES AND EQUIPMENT CONSIDERATIONS .................................. E–1 

APPENDIX F SAMPLE AUDIT PROGRAM—NON-MEDICAL .......................................... F–1 

APPENDIX G TYPICAL U.S. NUCLEAR REGULATORY COMMISSION (NRC) 

NOTIFICATION AND REPORTING REQUIREMENTS FOR 

 INCIDENTS .................................................................................................. G–1 

APPENDIX H RADIATION MONITORING INSTRUMENT SPECIFICATIONS AND 

 MODEL RADIATION SURVEY INSTRUMENT AND AIR SAMPLER 

CALIBRATION PROGRAM......................................................................... H–1 

APPENDIX I MATERIAL RECEIPT AND ACCOUNTABILITY .......................................... I–1 

APPENDIX J PUBLIC DOSE ............................................................................................. J–1 

APPENDIX K GENERAL TOPICS FOR SAFE USE OF RADIONUCLIDES 

AND MODEL EMERGENCY PROCEDURES ............................................. K–1 

APPENDIX L RADIATION SAFETY SURVEY TOPICS ..................................................... L–1 

APPENDIX M MODEL LEAK TEST PROGRAM ............................................................... M–1 

APPENIDX N U.S. DEPARTMENT OF TRANSPORTATION REGULATIONS ................ N–1 

APPENDIX O MODEL WASTE MANAGEMENT PROCEDURES ..................................... O–1 

APPENDIX P COMMENCEMENT OF CONSTRUCTION AT EXISTING AND 

 PROPOSED BYPRODUCT (INCLUDING IRRADIATORS), 

 SOURCE, AND SPECIAL NUCLEAR MATERIAL FACILITIES 

 WITH SIGNIFICANT ENVIRONMENTAL IMPACTS .................................. P–1 
ix 

APPENDIX Q SAFETY CULTURE POLICY STATEMENT ............................................... Q–1 

APPENDIX R CHECKLIST FOR REQUESTS TO WITHHOLD PROPRIETARY 

INFORMATION FROM PUBLIC DISCLOSURE (UNDER 10 CFR 2.390) .. R–1 

xi 

FIGURES

Figure Page 

2-1 U.S. Map: Locations of NRC Offices and Agreement States ............................... 2-1 

8-1 Location of Use or Possession .............................................................................. 8-4 

8-2 Types of Records That Must be Maintained for Decommissioning ..................... 8-11 

8-3 Annual Dose Limits for Adult Radiation Workers ................................................ 8-37 

TABLES

Table Page 

2-1 Who Regulates the Activity? ................................................................................. 2-2 

3-1 Traits of a Positive Safety Culture ......................................................................... 3-3 

8-1 Sample Format for Providing Information About Requested Radioisotopes ......... 8-7 

8-2 Package Monitoring Requirements (10 CFR 20.1906) ....................................... 8-32 

8-3 Record Maintenance ........................................................................................... 8-34 

8-4 Guidance on Personnel Monitoring and Bioassay .............................................. 8-38 

H–1 Typical Survey Instruments (Instruments Used to Measure Radiological 

 Conditions at Licensed Facilities) ........................................................................ H–1 

J–1 Standard Occupancy Factors ............................................................................... J–3 

L–1 Suggested Contamination Survey Frequency ...................................................... L–3 

L–2 Acceptable Surface Contamination Levels ........................................................... L–3 

L–3 Screening Values for Building Surface Contamination ......................................... L–5 

xiii 

ACKNOWLEDGMENTS 

The working group thanks the individuals listed here for assisting in the review and update of 

the report. All participants provided valuable insights, observations, and recommendations. 

The working group would like to thank the staff in the regional offices of the U.S. Nuclear 

Regulatory Commission and all of the States who provided comments and technical information 

that assisted in the development of this report. 

The working group also thanks Tomas Herrera for the pivotal role that he played in the 

development of this report. 

The Participants for this Revision 

Frazier, Cassandra 

Kern, Phillip 

Rivera-Capella, Gretchen 

Torres, Roberto J. 

Ullrich, Betsy 

xv 

ABBREVIATIONS 

ADAMS Agencywide Documents Access and Management System 

AEA Atomic Energy Act 

ALARA as low as is reasonably achievable 

ALI annual limit on intake 

ANSI American National Standards Institute 

bkg background 

Bq becquerel 

CFR Code of Federal Regulations

Ci curie 

cm2 square centimeter 

cpm counts per minute 

DFP decommissioning funding plan 

DIS decay-in-storage 

DOT U.S. Department of Transportation 

dpm disintegrations per minute 

EA environmental assessment 

FA financial assurance 

FR Federal Register 

GBq gigabecquerel 

G-M Geiger-Mueller 

GPO Government Printing Office 

Gy Gray 

HAZMAT hazardous materials 

IAEA International Atomic Energy Agency 

IN information notice 

IP inspection procedure 

IROFS items relied on for safety 

ISG Interim Staff Guidance 

kBq kilobecquerel 

LLEA Local law enforcement agency 

LLW low-level radioactive waste 

LSA low specific activity 

LSC liquid scintillation counting 

MBq megabecquerel 

μCi microcurie 

μGy microgray 

μR microroentgen 

mCi millicurie 

MDA minimum detectable activity 

MDC minimum detectable concentration 

mGy milligray 

mrad millirad 

mrem millirem 

mSv millisievert 

ND not detectable 

NHPA National Historic Preservation Act of 1966, as amended 

NIST National Institute of Standards and Technology 

NMSS Office of Nuclear Material Safety and Safeguards 

N/A not applicable 
xvi 

NRC U.S. Nuclear Regulatory Commission 

NSTS National Source Tracking System 

NSTTR National Source Tracking Transaction Report 

NVLAP National Voluntary Laboratory Accreditation Program 

OMB Office of Management and Budget 

PII Personally Identifiable Information 

PG Policy and Guidance Directive 

Q quality factor 

RG regulatory guide 

RIS regulatory issue summary 

RSC radiation safety committee 

RSO radiation safety officer 

SCO surface contaminated objects 

SSD Sealed Source and Device [registration certificate] 

SNM special nuclear material 

std standard 

Sv sievert 

TBq terabecquerel 

TEDE total effective dose equivalent 

U.S.C. United States Code 
1-1 

1 PURPOSE OF REPORT 

This NUREG provides guidance to an applicant applying for a broad scope license and also 

provides the U.S. Nuclear Regulatory Commission (NRC) staff with the criteria for evaluating 

such applications. This document uses the terms “byproduct material,” “licensed material,” and 

“radioactive material” interchangeably. Although the applicant for a limited scope license 

generally must submit to the NRC for review and approval the specific training and experience 

of each proposed user and the facilities and equipment available to support each proposed use, 

the applicant for a broad scope license normally must submit to the NRC for review and 

approval a description of the internal review process and criteria that will be used to approve 

users and uses. As opposed to limited scope licenses, which typically identify specific isotopes 

that may be possessed, the broad scope license typically authorizes the possession and use of 

a wide range of byproduct radioactive materials. 

Because the NRC grants significant decision-making authority to broad scope licensees through 

the license, a broad scope license is not normally issued to a new licensee. An applicant for a 

broad scope license typically has several years of experience operating under a limited scope 

license and a good regulatory performance history. This document is intended to provide the 

additional guidance required by the experienced limited scope licensee to prepare an 

application for a broad scope license. Guidance related to specific program areas, which may 

not apply to all broad scope licensees, is not included in this document but can be found in other 

volumes of NUREG–1556, often referred to in this document as “the base NUREGs” or “the 

base documents,” or in guidance documents that have not yet undergone the consolidation 

process. 

Applicants should have established limited scope licensed programs in accordance with the 

guidance described in the appropriate base NUREG(s) before they apply for a broad scope 

license. For example, applicants for a broad scope license that use byproduct material for 

research and development should review NUREG–1556, Volume 7, “Consolidated Guidance 

About Materials Licenses: Program-Specific Guidance About Academic, Research and 

Development, and Other Licenses of Limited Scope Including Electron Capture Devices and XRay Fluorescence Analyzers,” for guidance. Similarly, applicants for broad scope licenses that 

use byproduct material for medical purposes should review NUREG–1556, Volume 9, 

“Consolidated Guidance About Materials Licenses: Program-Specific Guidance About Medical 

Use Licenses.” A list of the currently available base NUREGs is included in the “Foreword” to 

this document. 

Chapter 8, “Contents of an Application,” of this NUREG identifies the information needed to 

complete NRC Form 313, “Application for Materials License” (see Appendix A of this NUREG), 

for the use of byproduct material for licenses of broad scope. The Office of Management and 

Budget (OMB) has approved the information collection requirements in Title 10 of the Code of 

Federal Regulations (10 CFR) Part 30, “Rules of General Applicability to Domestic Licensing 

of Byproduct Material,” and NRC Form 313 under OMB Clearance Nos. 3150-0017 and 

3150-0120, respectively. 

The format within this document for each item of technical information is as follows: 

 Regulations—references the regulations applicable to the item 

 Criteria—outlines the criteria used to evaluate the applicant’s response 
1-2 

 Discussion—provides additional information about the topic 

 Response from Applicant—provides suggested response or responses, offers the option 

of an alternative reply, or indicates that no response is needed on that topic during the 

licensing process 

Notes and References are self-explanatory and may not be found for each item on 

NRC Form 313. 

NRC Form 313 does not have sufficient space for applicants to provide full responses to Items 5 

through 11, as indicated on the form. Applicants should address those items on separate 

sheets of paper and submit them along with the completed NRC Form 313. For the 

convenience and streamlined handling of applications, Appendix B of this NUREG, “Suggested 

Format for Providing Information Requested in Items 5 through 11 of U.S. Nuclear Regulatory 

Commission Form 313,” may be used to provide supporting information. Appendices C through 

R of this NUREG contain additional information on various radiation safety topics. 

In this document, “dose” or “radiation dose” means absorbed dose, dose equivalent, effective 

dose equivalent, committed dose equivalent, committed effective dose equivalent, or total 

effective dose equivalent (TEDE), as defined in 10 CFR Part 20, “Standards for Protection 

Against Radiation.” To describe units of radiation exposure or dose, rem and its International 

System of Units equivalent, sievert (Sv) (1 rem = 0.01 Sv), are used. This is done because 

10 CFR Part 20 sets dose limits in terms of rem (Sv), rather than rad or roentgen. When the 

radioactive material emits beta and gamma rays, 1 roentgen is assumed to equal 1 rad, which is 

assumed to equal 1 rem. For alpha and neutron-emitting radioactive material, 1 rad is not equal 

to 1 rem. Determination of dose equivalent (rem) from absorbed dose (rad) from alpha particles 

and neutrons requires the use of an appropriate quality factor (Q) value. These Q values are 

used to convert absorbed dose (rad) to dose equivalent (rem). Tables 1004(b).1 and .2 in 

10 CFR 20.1004, “Units of radiation dose,” address the Q values for alpha particles and 

neutrons. 

Overview of Broad Scope Programs 

Title 10 of the Code of Federal Regulations (10 CFR) Part 33, “Specific Domestic Licenses of 

Broad Scope for Byproduct Material,” provides for three distinct categories of broad scope 

license (i.e., Type A, Type B, and Type C), which are defined in 10 CFR 33.11, “Types of 

Specific Licenses of Broad Scope.” 

Type A licenses of broad scope are typically the largest licensed programs and encompass a 

broad range of uses. Type A broad scope licensees use a Radiation Safety Committee (RSC), 

radiation safety officer (RSO), and criteria developed and submitted by the licensee and 

approved by the NRC during the licensing process to review and approve all uses and users 

under the license. The requirements for issuance of a Type A broad scope license are 

described in 10 CFR 33.13, “Requirements for the Issuance of a Type A Specific License of 

Broad Scope.” 

An applicant for a Type A broad scope license must establish administrative controls and 

provisions related to organization and management, procedures, recordkeeping, material 

 
1-3 

control, and accounting and management review necessary to ensure safe operations, 

including: 

 establishment of an RSC 

 appointment of a qualified RSO 

 establishment of appropriate administrative procedures to ensure the following: 

— control of procurement and use of byproduct material 

— completion of safety evaluations of proposed uses that take into consideration 

adequacy of facilities and equipment, training and experience of the user, and 

operating and handling procedures 

— review, approval, and recording by the RSC of safety evaluations of proposed 

uses 

 use of byproduct material only by, or under the direct supervision of, individuals 

approved by the licensee’s RSC 

Type B broad scope licensed programs are normally smaller and less diverse than Type A 

broad scope programs. Type B broad scope licensees use an RSO and criteria developed and 

submitted by the licensee and approved by the NRC during the licensing process to review and 

approve all uses and users under the license. Because the RSO reviews and approves all uses 

and users under the license, rather than a full RSC, as established for Type A broad scope 

programs, the types and quantities of byproduct material authorized by the Type B broad scope 

license are limited to those described in 10 CFR 33.11(b) and 10 CFR 33.100, “Schedule A,” 

Column I. While the quantities of individual radionuclides described in Schedule A may be 

large, the “unity rule” further restricts total license possession limits (see Section 8.5.1, 

“Unsealed or Sealed Byproduct Material,” of this NUREG volume for additional information on 

license possession limits and the unity rule). Generally, the scope of authorization for Type B 

licenses is limited to the experience and knowledge of the RSO. The requirements for issuance 

of a Type B broad scope license are described in 10 CFR 33.14, “Requirements for the 

Issuance of a Type B Specific License of Broad Scope.” 

An applicant for a Type B broad scope license must also establish administrative controls and 

provisions related to organization and management, procedures, recordkeeping, material 

control, and accounting and management review that are necessary to ensure safe operations, 

including: 

 appointment of a qualified RSO 

 establishment of appropriate administrative procedures to ensure the following: 

— control of procurement and use of byproduct material 

— completion of safety evaluations of proposed uses that take into consideration 

adequacy of facilities and equipment, training and experience of the user, and 

operating and handling procedures 
1-4 

— review, approval, and recording by the RSO of safety evaluations of proposed 

uses 

 use of byproduct material only by, or under the direct supervision of, individuals 

approved by the licensee’s RSO 

Type C broad scope licensed programs typically are issued to institutions that do not require 

significant quantities of radioactive material but need the flexibility to possess a variety of 

different radioactive materials. Users of licensed material under these programs are approved 

by the licensee based on training and experience criteria described in 10 CFR 33.15(b). The 

types and quantities of byproduct material authorized by the Type C broad scope license are 

limited to those described in 10 CFR 33.11(c) and 10 CFR 33.100, Schedule A, Column II, 

again, considering the unity rule. The requirements for issuance of a Type C broad scope 

license are described in 10 CFR 33.15, “Requirements for the Issuance of a Type C Specific 

License of Broad Scope.” 

While 10 CFR 33.15 does not require Type C broad scope licensees to appoint an RSO, the 

licensee must establish administrative controls and provisions related to procurement of 

byproduct material, procedures, recordkeeping, material control and accounting, and 

management review to ensure safe operations. This should include the appointment of 

someone responsible for the day-to-day operation of the radiation safety program, such as an 

RSO. 

Except for activities specifically excluded from broad scope licenses by 10 CFR 33.17(a), a 

Type A broad scope license can include any licensed material the applicant needs and for which 

it qualifies. An application for a Type A broad scope license can include uses of source material 

and special nuclear material under the same program (e.g., laboratory-scale research and 

development or the use of uranium as shielding) as the byproduct material to be possessed 

under the provisions of 10 CFR 30.32(d). However, applicants should submit separate 

applications for the production of radioactive material in a cyclotron, or for the use of source and 

special nuclear materials for purposes not directly related to the use of byproduct material under 

the Type A broad scope license (e.g., use of source material in subcritical assemblies and 

special nuclear material in critical mass quantities ). 

In practice, 10 CFR Part 33 reduces the administrative burden for both licensees and the 

Commission without reducing the safety standards or lessening the licensing requirements for 

training, experience, facilities, and equipment. Both the NRC and the licensee benefit from the 

reduction in license amendments that might otherwise be needed to change authorized 

radionuclides, quantities, or names of individuals who may use, or supervise the use of, 

byproduct material. 

Part 33 does not specifically permit a broad scope licensee to make other types of changes to 

the radiation program as described in the application, such as changing the dosimetry provider, 

without amendment of the license. However, the NRC has permitted broad scope licensees, on 

a case-by-case basis, to build in limited program flexibility during the licensing process. 

For example, rather than requiring that the applicant identify the company that would provide 

personnel dosimetry, the broad scope licensee could specify that dosimetry would be provided 

by an organization holding current personnel dosimetry accreditation from the National 

Voluntary Laboratory Accreditation Program of the National Institute of Standards and 

Technology. The NRC will continue to allow licensees to build in this type of program flexibility. 
2-1 

2 AGREEMENT STATES 

2.1 Jurisdiction Determination 

Certain States, called Agreement States (see Figure 2-1), have entered into agreements with 

the U.S. Nuclear Regulatory Commission (NRC) that give them the authority to license and 

inspect byproduct, source, and special nuclear materials in quantities not sufficient to form a 

critical mass, which are used or possessed within their borders. Any applicant, other than a 

Federal entity, who wishes to possess or use licensed material in one of these Agreement 

States should contact the responsible officials in that State for guidance on preparing an 

application. These applications should be filed with State officials, not with the NRC. In areas 

under exclusive Federal jurisdiction within an Agreement State, NRC continues to be the 

regulatory authority. 

Figure 2-1. U.S. Map: Locations of NRC Offices and Agreement States 

In the special situation of work at federally controlled sites in Agreement States, it is necessary 

to ascertain the jurisdictional status of the area to determine whether the NRC or the Agreement 

State has regulatory authority. 

 
2-2 

These areas can also include Tribal lands of federally recognized Indian Tribes.2

 

The NRC has regulatory authority over land determined to be “exclusive Federal 

jurisdiction,” while the Agreement State may have jurisdiction over nonexclusive Federal 

jurisdiction land. 

Applicants are responsible for determining, in advance, the jurisdictional status of the specific 

areas where they plan to conduct licensed operations. Additional guidance on determining 

jurisdictional status is found in the Office of Nuclear Material Safety and Safeguards (NMSS) 

procedures in the State Agreement series, SA-500, “Jurisdiction Determination,” which is 

available at https://scp.nrc.gov. Once on the Web site, use the link for “NMSS Procedures” in 

the left-hand column under “Resources & Tools.” 

Table 2-1 provides a quick way to evaluate whether the NRC or an Agreement State has 

regulatory authority. 

Table 2-1. Who Regulates the Activity?

Applicant and Proposed Location of Work 

Regulatory 

Agency 

Federal agency regardless of location (except that the U.S. Department of 

Energy and, under most circumstances, its prime contractors are exempt 

from licensing, in accordance with Title 10 of the Code of Federal 

Regulations (10 CFR) 30.12, “Persons using byproduct material under 

certain Department of Energy and Nuclear Regulatory Commission 

contracts”; also, see 10 CFR 40.11, and/or 10 CFR 70.11, if applicable) 

NRC 

Non-Federal entity in non-Agreement State, District of Columbia, 

U.S. territory or possession, or in offshore Federal waters 

NRC 

Federally recognized Indian Tribe or Tribal member on Indian Tribal land NRC 

Non-Federal entity on federally recognized Indian Tribal land NRC3

Federally recognized Indian Tribe or Tribal member outside of Indian Tribal 

land in Agreement State. 

Agreement State 

Non-Federal entity in Agreement State Agreement State4

Non-Federal entity in Agreement State at federally controlled site not

subject to exclusive Federal jurisdiction 

Agreement State4

 

2

For the purposes of this guidance, an “Indian Tribe” is defined as an Indian or Alaska Native tribe, band, nation, 

pueblo, village, or community that the Secretary of the Interior acknowledges to exist as an Indian tribe pursuant to 

the Federally Recognized Indian Tribe List Act of 1994. A list of federally recognized tribes is available at 

www.bia.gov.

3

The NRC can exercise jurisdiction as the regulatory authority on Tribal land of a federally recognized Indian Tribe. 

Section 274b. agreements do not give States the authority to regulate nuclear material in these areas. However, 

there may be States that exercise regulatory authority over these areas based on treaties or agreements with specific 

tribes. Companies owned or operated by federally recognized Indian Tribe members or non-Indians that wish to 

possess or use licensed material on Tribal lands should contact the appropriate NRC regional office to determine the 

jurisdictional status of the Tribal lands and identify the appropriate regulatory agency for licensing and reciprocity. 

4

Section 274m. of the Atomic Energy Act (AEA) withholds to the NRC regulatory authority over radioactive materials 

covered under the Section 274b. agreements when the activity can affect the Commission’s authority to protect the 

common defense and security, to protect restricted data, or guard against the loss or diversion of special nuclear 

material. (This is an uncommon situation that NRC usually evaluates on a case-by-case basis.) Individuals or 

companies wishing to possess or use licensed material should contact the licensee to determine the jurisdictional 

status for specific AEA radioactive materials they intend to possess or use. 
2-3 

Table 2-1. Who Regulates the Activity? 

Applicant and Proposed Location of Work 

Regulatory 

Agency 

Non-Federal entity in Agreement State at federally controlled site subject to 

exclusive Federal jurisdiction 

NRC 

Non-Federal entity in Agreement State using radioactive materials (except 

industrial radiography) directly connected with 10 CFR Part 50 or 52 reactor 

operations or needed during the construction and preoperational phases of 

a reactor. 

NRC 

Non-Federal entity in Agreement State using radioactive materials not 

directly connected with 10 CFR Part 50 or 52 reactor operations or needed 

during the construction and preoperational phases of a reactor. 

Agreement State4

Reference: A current list of Agreement States (including names, addresses, and telephone 

numbers of responsible officials) is available at the NMSS public Web site, https://scp.nrc.gov. 

A request for the list can also be made to an NRC regional office. 

2.2 Reciprocal Recognition of Specific Licenses 

Performing licensed activities in other jurisdictions is possible through reciprocal recognition of 

specific licenses (i.e., reciprocity). Agreement States have reciprocity provisions that permit 

NRC licensees to perform licensed activities under circumstances when an Agreement State is 

the regulatory authority (See Section 2.1). NRC licensees and Agreement State licensees are 

subject to the regulations of the regulatory authority as indicated in Section 2.1. To ensure 

compliance with an Agreement State’s reciprocity requirements, licensees are advised to 

request authorization from the appropriate Agreement State radiation control program office well 

in advance of the scheduled use of licensed material. 

Agreement State licensees that wish to conduct licensed activities in areas under NRC 

jurisdiction must either obtain a specific NRC license or file for reciprocity with the appropriate 

NRC regional office for the Agreement State that issued their license. Failure to file for 

reciprocity or obtain a specific NRC license before working in areas under NRC jurisdiction can 

result in NRC enforcement action, which may include civil penalties. The reciprocity filing must 

be renewed annually. 

Specific guidance regarding NRC licensees filing for reciprocity in Agreement States and 

Agreement State licensees filing for reciprocity with the NRC or another Agreement State are 

provided in NUREG–1556, Volume 19, “Consolidated Guidance About Materials 

Licenses: Guidance for Agreement State Licensees About NRC Form 241 “Report of Proposed 

Activities in Non-Agreement States, Areas of Exclusive Federal Jurisdiction, or Offshore Waters” 

and Guidance for NRC Licensees Proposing to Work in Agreement State Jurisdiction 

(Reciprocity).”

3-1 

3 MANAGEMENT RESPONSIBILITY 

The U.S. Nuclear Regulatory Commission (NRC) recognizes that effective radiation safety 

program management is vital to achieving safe, secure, and compliant operations. Consistent 

compliance with NRC regulations provides reasonable assurance that licensed activities will be 

conducted safely and that effective management will result in increased safety, security, and 

compliance. 

“Management” as used in this volume refers to the processes for conduct and control of a 

radiation safety program and to the individuals who are responsible for those processes and 

who have authority to provide necessary resources to achieve regulatory compliance. 

3.1 Commitments and Responsibilities 

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 30.32(c), 40.31(b), and 

70.22(d), each application must be signed by the applicant or licensee or a person duly 

authorized to act for and on behalf of the applicant or licensee. If it is not clear whether the 

application was signed by someone duly authorized to act for and on the behalf of the applicant 

or licensee, NRC license reviewers may ask for additional assurances that the individual who 

signed the application is duly authorized to act for and on the behalf of the applicant or

licensee. The signature on an application acknowledges the licensee’s commitments and 

responsibilities for the following: 

 radiation safety, security and control of radioactive materials, and compliance with 

regulations 

 completeness and accuracy of the radiation safety records and all information provided 

to the NRC (10 CFR 30.9, “Completeness and accuracy of information”) 

 knowledge about the contents of the license and application 

 compliance with current NRC and Department of Transportation regulations, the 

licensee’s operating, emergency, and security procedures, and NRC license 

commitments 

 commitment to provide adequate resources (including space, equipment, personnel, 

time and, if needed, contractors) to the radiation protection program to ensure that the 

public and workers are protected from radiation hazards and compliance with regulations 

is maintained 

 commitment to report defects, noncompliances, or reportable events in accordance with 

regulations 

 selection and assignment of qualified individuals to serve on the radiation safety 

committee, if required, and to serve as radiation safety officer (RSO) for its licensed 

activities and confirmation that the RSO has independent authority to stop unsafe 

operations and will be given sufficient time to fulfill radiation safety duties and 

responsibilities 

 commitment to ensure that radiation workers have adequate training 
3-2 

 prevention of discrimination of employees engaged in protected activities and 

commitment to provide information to employees about employee protection provisions 

(10 CFR 30.7, 40.7, and 70.7, “Employee protection”) 

 commitment to provide information to employees about deliberate misconduct provisions 

(10 CFR 30.10, 40.10, and 70.10, “Deliberate misconduct”) 

 commitment to obtain NRC’s prior written consent before transferring control of the 

license (see Section 9.1, “Timely Notification of Transfer of Control,” of this NUREG) 

 notification of the appropriate NRC regional administrator, in writing, immediately 

following the filing of petition for voluntary or involuntary bankruptcy [10 CFR 30.34(h),

40.41(f), and 70.32(a)(9)], as discussed further in Section 8.2.1, “Notification of 

Bankruptcy Proceedings,” of this NUREG 

For information on NRC inspection, investigation, enforcement, and other compliance programs, 

see the current version of the NRC’s Enforcement Policy and Inspection Procedures available in 

the NRC’s online library, under “Document Collections,” at http://www.nrc.gov/reading-rm.html. 

3.2 Safety Culture 

Individuals and organizations performing regulated activities are expected to establish and 

maintain a positive safety culture commensurate with the safety and security significance of 

their activities and the nature and complexity of their organizations and functions. This applies 

to all licensees, certificate holders, permit holders, authorization holders, holders of quality 

assurance program approvals, vendors and suppliers of safety-related components, and 

applicants for a license, certificate, permit, authorization, or quality assurance program approval, 

subject to NRC authority. 

“Nuclear safety culture” is defined in the NRC’s safety culture policy statement (76 FR 34773; 

June 14, 2011) as the core values and behaviors resulting from a collective commitment by 

leaders and individuals to emphasize safety over competing goals to ensure protection of 

people and the environment. Individuals and organizations performing regulated activities bear 

the primary responsibility for safely handling and securing these materials. Experience has 

shown that certain personal and organizational traits are present in a positive safety culture. A 

trait, in this case, is a pattern of thinking, feeling, and behaving that emphasizes safety, 

particularly in goal conflict situations (e.g., production versus safety, schedule versus safety, 

and cost of the effort versus safety). Refer to Table 3-1 for the traits of a positive safety culture 

from NRC’s safety culture policy statement. 

Organizations should ensure that personnel in the safety and security sectors have an 

appreciation for the importance of each, emphasizing the need for integration and balance to 

achieve both safety and security in their activities. Safety and security activities are closely 

intertwined. While many safety and security activities complement each other, there may be 

instances in which safety and security interests create competing goals. It is important that 

consideration of these activities be integrated so as not to diminish or adversely affect either; 

thus, mechanisms should be established to identify and resolve these differences. A safety 

culture that accomplishes this would include all nuclear safety and security issues associated 

with NRC-regulated activities. 
3-3 

The NRC, as the regulatory agency with an independent oversight role, reviews the 

performance of individuals and organizations to determine compliance with requirements and 

commitments through its existing inspection and assessment processes. However, NRC’s 

safety culture policy statement and traits are not incorporated into the regulations. Safety 

culture traits may or may not be inherent to an organization’s existing radiation safety practices 

and programs. For instance, laboratory facilities that perform iodinations require that air 

monitoring be performed of the effluents released through the stack. Developing and 

implementing procedures for monitoring effluent releases may correspond with the safety 

culture trait specified in Table 3-1 as “Work Processes” (the process of planning and controlling 

work activities is implemented so that safety is maintained). Another example would be that 

broad scope licensees are required to develop and implement an audit program. Development, 

implementation, and revision of an audit program to effectively and promptly address issues 

potentially impacting safety may correspond with the safety culture trait identified in Table 3-1 as 

“Problem Identification and Resolution.” However, licensees should be aware that these are 

just examples, and should still consider reviewing their radiation safety programs in order to 

develop and implement a safety culture commensurate with the nature and complexity of their 

organizations and functions. 

Refer to Appendix Q of this NUREG for the NRC’s safety culture policy statement. More 

information on NRC activities relating to safety culture can be found at: 

http://www.nrc.gov/about-nrc/safety-culture.html. 

Table 3-1. Traits of a Positive Safety Culture 

Leadership Safety Values 

and Actions 

Problem Identification and 

Resolution Personal Accountability 

Leaders demonstrate a 

commitment to safety in their 

decisions and behaviors 

Issues with a potential impact 

on safety are promptly 

identified, fully evaluated, and 

promptly addressed and 

corrected commensurate with 

their significance 

All individuals take personal 

responsibility for safety 

Work Processes Continuous Learning 

Environment for Raising 

Concerns 

The process of planning and 

controlling work activities is 

implemented so that safety is 

maintained 

Opportunities to learn about 

ways to ensure safety are 

sought out and implemented 

A safety conscious work 

environment is maintained 

where personnel feel free to 

raise safety concerns without 

fear of retaliation, intimidation, 

harassment or discrimination 

Effective Safety 

Communications 

Respectful Work 

Environment Questioning Attitude 

Communications maintain a 

focus on safety 

Trust and respect permeate 

the organization 

Individuals avoid complacency 

and continuously challenge 

existing conditions and 

activities in order to identify 

discrepancies that might result 

in error or inappropriate action 

4-1 

4 APPLICABLE REGULATIONS 

It is the applicant’s or licensee’s responsibility to obtain and have available up-to-date copies of 

applicable regulations, to read and understand the requirements of each of these regulations, 

and to comply with each applicable regulation. The following parts of Title 10 of the Code of 

Federal Regulations (10 CFR) contain regulations applicable to the use of licensed material by 

broad scope licensees. Some of these parts are specific to one type of license, while others are 

general and will apply to many, if not all, licensees. 

The current versions of these parts can be found under the “Basic References” link at the 

U.S. Nuclear Regulatory Commission (NRC) online library at http://www.nrc.gov/readingrm.html; for viewing in a browser, the following list includes direct links to the rules: 

 10 CFR Part 2 “Agency Rules of Practice and Procedure” 

 10 CFR Part 19 “Notices, Instructions and Reports to Workers: Inspection 

 and Investigations” 

 10 CFR Part 20 “Standards for Protection Against Radiation” 

 10 CFR Part 21 “Reporting of Defects and Noncompliance” 

 10 CFR Part 30 “Rules of General Applicability to Domestic Licensing of 

 Byproduct Material” 

 10 CFR Part 33 “Specific Domestic Licenses of Broad Scope for 

 Byproduct Material” 

 10 CFR Part 37 “Physical Protection of Category 1 and Category 2 Quantities of 

 Radioactive Material” 

 10 CFR Part 51 “Environmental Protection Regulations for Domestic Licensing and 

 Related Regulatory Functions” 

 10 CFR Part 71 “Packaging and Transportation of Radioactive Material” 

 10 CFR Part 170 “Fees for Facilities, Materials, Import and Export Licenses, and 

 Other Regulatory Services Under the Atomic Energy Act of 1954, 

 as Amended” 

 10 CFR Part 171 “Annual Fees for Reactor Licenses and Fuel Cycle Licenses and 

 Materials Licenses, Including Holders of Certificates of 

 Compliance, Registrations, and Quality Assurance Program 

 Approvals and Government Agencies Licensed by the NRC” 

 
4-2 

The following Parts of 10 CFR Chapter 1 contain regulations that, depending on the type or 

types of activities authorized by the license, may be applicable to the use of licensed material by 

broad scope licensees: 

 10 CFR Part 31 “General Domestic Licenses for Byproduct Material” 

 10 CFR Part 32 “Specific Domestic Licenses to Manufacture or Transfer Certain 

 Items Containing Byproduct Material” 

 10 CFR Part 34 “Licenses for Industrial Radiography and Radiation Safety 

 Requirements for Industrial Radiographic Operations” 

 10 CFR Part 35 “Medical Use of Byproduct Material” 

 10 CFR Part 36 “Licenses and Radiation Safety Requirements for Irradiators” 

 10 CFR Part 39 “Licenses and Radiation Safety Requirements for Well Logging” 

 10 CFR Part 40 “Domestic Licensing of Source Material” 

 10 CFR Part 61 “Licensing Requirements for Land Disposal of Radioactive Waste” 

 10 CFR Part 70 “Domestic Licensing of Special Nuclear Material” 

 10 CFR Part 150 “Exemptions and Continued Regulatory Authority in Agreement 

 States and in Offshore Waters Under Section 274” 

Copies of these documents may be obtained by calling the Government Publishing Office 

Customer Contact Center toll-free at 866-512-1800, in Washington, DC; calling 202-512-1800; 

or ordering online at http://bookstore.gpo.gov. 

In addition, 10 CFR Parts 1 through 199 can be found on the NRC’s Web site at 

http://www.nrc.gov/reading-rm/doc-collections/ under “Regulations (10 CFR).” 

NRC regulations can also be accessed from the “NRC Library” link on the NRC’s public Web 

site at http://www.nrc.gov. Regulations are periodically amended, and the NRC (as well as all 

other Federal agencies) is required to publish notice of such amendments in the Federal 

Register. 
5-1 

5 HOW TO FILE 

5.1 Application Preparation 

Applicants for a materials license should do the following: 

 Use the most recent guidance in preparing an application. 

 Complete U.S. Nuclear Regulatory Commission (NRC) Form 313 (Appendix A of this 

NUREG), Items 1 through 4, 12, and 13 on the form itself. A link to the form is available 

at http://www.nrc.gov/reading-rm/doc-collections/forms/. 

 Complete NRC Form 313, Items 5 through 11, on supplementary pages or use 

Appendix B of this NUREG. 

 Provide sufficient detail for the NRC to determine that the equipment, facilities, training, 

experience, and radiation safety program are adequate to protect health and safety and 

minimize danger to life and property. 

 For each separate sheet other than NRC Form 313 and Appendix B pages, as 

applicable, identify and cross-reference submitted information to the item number on the 

application or the topic to which it refers. 

 Avoid submitting proprietary information and personally identifiable information. If 

submitted, proprietary, personal privacy, security-related, and other sensitive information 

should be clearly identified according to Title 10 of the Code of Federal Regulations 

(10 CFR) 2.390, “Public inspections, exemptions, requests for withholding” (see 

Chapter 6, “Identifying and Protecting Sensitive Information”). 

5.2 Where to File 

Applicants wishing to possess or use licensed material in any State, U.S. territory, or 

U.S. possession subject to NRC jurisdiction must file an application with the NRC regional office 

for the locale in which the material will be possessed or used. Figure 2-1 identifies the NRC’s 

four regional offices and their respective areas for licensing purposes and the Agreement 

States. Note that all materials applications are submitted to Regions I, III, or IV. All applicants 

for materials licenses located in the Region II geographical area should send their applications 

to Region I. 

In general, applicants wishing to possess or use licensed material in Agreement States must file 

an application with the Agreement State and not with the NRC. However, if work will be 

conducted at federally controlled sites, or federally recognized Indian Tribal lands, in Agreement 

States, applicants must first determine the jurisdictional status of the land in order to determine 

whether the NRC or the Agreement State has regulatory authority. See Chapter 2, “Agreement 

States,” for additional information. 

 
5-2 

5.3 Paper Applications 

Paper applications received by the NRC are scanned through an optical character reader and 

converted to an electronic format. To ensure a smooth transfer to an electronic format, 

applicants should do the following: 

 Submit all documents, typed, on 8 1⁄2 × 11-inch or legal sized paper that will feed easily 

into a document scanner. 

 Choose typeface designs that are sans serif, such as Arial, Helvetica, or Futura. 

 Use 11-point or larger font. 

 Avoid stylized characters, such as script or italics. 

 Ensure that the print is clear and sharp. 

 Ensure that there is high contrast between the ink and paper (black ink on white paper is 

best). 

Applications must be signed by the applicant, licensee, or a person duly authorized as required 

by 10 CFR 30.32(c) (see Section 8.13, “Certification”). 

5.4 Electronic Applications 

Applications may be submitted in electronic form via the NRC’s Electronic Information Exchange 

or CD-ROM. Detailed guidance on making electronic submissions can be obtained by visiting 

the NRC’s Web site at http://www.nrc.gov/site-help/e-submittals.html. The guidance discusses, 

among other topics, the formats the NRC can accept, the use of electronic signatures, and the 

treatment of non-public information. 
6-1 

6 IDENTIFYING AND PROTECTING SENSITIVE INFORMATION 

All licensing applications, except for portions containing sensitive information, will be made 

available for review in the U.S. Nuclear Regulatory Commission (NRC) Public Document Room 

and electronically at the NRC Library. For more information on the NRC Library, visit 

www.nrc.gov. 

The applicant or licensee should identify, mark, and protect sensitive information against 

unauthorized disclosure to the public. License applications that contain sensitive information 

should be marked as indicated in the list that follows in accordance with Title 10 of the Code of 

Federal Regulations (10 CFR 2.390) before the information is submitted to the NRC. Key 

examples are as follows: 

 Proprietary Information and Trade Secrets: If it is necessary to submit proprietary 

information or trade secrets, follow the procedure in 10 CFR 2.390(b). Failure to follow 

this procedure could result in disclosure of the proprietary information to the public or 

substantial delays in processing the application. Appendix R of this NUREG provides a 

checklist for requests for withholding proprietary information from public disclosure. 

 Personally Identifiable Information: Personally identifiable information (PII) about 

employees or other individuals should not be submitted unless specifically requested by 

the NRC. Examples of PII are social security number, home address, home telephone 

number, date of birth, and radiation dose information. If PII is submitted, a cover letter 

should clearly state that the attached documents contain PII and the top of every page of 

a document that contains PII should be clearly marked as follows: “Privacy Act 

Information—Withhold Under 10 CFR 2.390.” For further information, see Regulatory 

Issue Summary (RIS) 2007-04, “Personally Identifiable Information Submitted to the 

U.S. Nuclear Regulatory Commission,” dated March 9, 2007, and Information Notice 

2013-22, “Recent Licensing Submittals Containing Personally Identifiable Information,” 

dated November 15, 2013, which can be found on the NRC’s Generic Communications 

Web page under “Regulatory Issue Summaries” and “Information Notices,” respectively: 

http://www.nrc.gov/reading-rm/doc-collections/gen-comm/. 

 Security-Related Information: Following the events of September 11, 2001, the NRC 

changed its procedures to avoid the release of information that terrorists could use to 

plan or execute an attack against facilities or citizens in the U.S. As a result, certain 

types of information are no longer routinely released and are treated as sensitive 

unclassified information. For example, certain information about the quantities and 

locations of radioactive material at licensed facilities, and associated security measures, 

are no longer released to the public. Therefore, a cover letter should clearly state that 

the attached documents contain sensitive security-related information and the top of 

every page of a document that contains such information should be clearly marked: 

“Security-Related Information—Withhold under 10 CFR 2.390.” For the pages having 

security-related sensitive information, an additional marking should be included (e.g., an 

editorial note box) adjacent to that material. For further information, see RIS 2005-31, 

“Control of Security-Related Sensitive Unclassified Non-Safeguards Information Handled 

by Individuals, Firms, and Entities Subject to NRC Regulation of the Use of Source, 

Byproduct, and Special Nuclear Material,” dated December 22, 2005, which can be 

found on the NRC’s Generic Communications Web page under “Regulatory Issue 

Summaries”: http://www.nrc.gov/reading-rm/doc-collections/gen-comm/. Additional 
6-2 

information on procedures and any updates is available at http://www.nrc.gov/readingrm/sensitive-info.html. 

The regulations list various forms of information that can be protected from public disclosure. 

These include 

 trade secrets and commercial or financial information 

 interagency or intra-agency memoranda or letters that would not be available by law to a 

party other than an agency in litigation with NRC 

 certain records or information compiled for law enforcement purposes 

 geological and geophysical information and data, including maps, or information 

concerning wells 

 personnel, medical, and other information, the disclosure of which would constitute a 

clearly unwarranted invasion of personal privacy 

In 10 CFR 2.390, NRC specifies the procedures and requirements for persons to submit 

sensitive information to NRC so that it may be properly protected from disclosure. This 

regulation is available electronically on the NRC Web site: http://www.nrc.gov/reading-rm/doccollections/cfr. 

Except for personal privacy information, which is not subject to the affidavit requirement, if NRC 

determines that the application or affidavit is deficient (i.e., does not contain the required 

information as outlined in 10 CFR 2.390), the applicant will be notified that additional information 

is needed and that the review will continue when the required information is received. 

If the request is denied, in whole or in part, NRC will give the applicant the option of withdrawing 

the information or application, as permitted in 10 CFR 2.390. If the applicant decides not to 

withdraw the information or application, NRC will notify the applicant in writing that the request 

for withholding has been denied and that NRC will disregard any references concerning the 

proprietary status of the information. 

Any part of a license application or information provided by a licensee or applicant that the NRC 

determines should be withheld from public disclosure will be handled in accordance with 

Management Directive 12.6, “NRC Sensitive Unclassified Information Security Program,” and 

the licensee or applicant will be notified in writing that NRC plans to honor the request. 

Management Directive 12.6 is available electronically on the NRC Web site: 

http://www.nrc.gov/reading-rm/doc-collections/management-directives/. 
6-3 

Anyone submitting a request to withhold information from public disclosure should thoroughly 

review 10 CFR 2.390 and be familiar with its requirements and limitations. 

Withholding from public inspection will not affect the right, if any, of persons properly and 

directly concerned to inspect the documents. If the need arises, NRC may send copies of this 

information to NRC consultants working in that area. NRC will ensure that the consultants have 

signed the appropriate agreements for handling proprietary information. 

If the basis for withholding this information from public inspection should change in the future, 

such that the information could then be made available for public inspection, the licensee or 

applicant should promptly notify the NRC. The licensee or applicant also should understand 

that NRC may have cause to review this determination in the future; for example, if the scope of 

a Freedom of Information Act request includes the information in question. In all review 

situations, if NRC makes a determination adverse to the above, the licensee or applicant will be 

notified in advance of any public disclosure. Anyone submitting commercial or financial 

information they believe to be privileged, confidential, or a trade secret must remember that the 

NRC’s policy is to achieve an effective balance between legitimate concerns for the protection 

of competitive positions and the right of the public to be fully apprised of the basis for, and the 

effects of, licensing or rulemaking actions. It is within NRC’s discretion to withhold such 

information from public disclosure. 

7-1 

7 APPLICATION AND LICENSE FEES 

Each application for which a fee is specified must be accompanied by the appropriate fee. 

Refer to Title 10 of the Code of Federal Regulations (10 CFR) 170.31, “Schedule of fees for 

materials licenses and other regulatory services, including inspections, and import and export 

licenses,” to determine the amount of the fee. The U.S. Nuclear Regulatory Commission (NRC) 

will not issue a license until the fee is received. Consult 10 CFR 170.11, “Exemptions,” for 

information on exemptions from these fees. Once the technical review of an application has 

begun, no fees will be refunded. Application fees will be charged regardless of the NRC’s 

disposition of an application or the withdrawal of an application. 

Most NRC licensees are also subject to annual fees; refer to 10 CFR 171.16, “Annual fees: 

Materials licensees, holders of certificates of compliance, holders of sealed source and device 

registrations, holders of quality assurance program approvals, and government agencies 

licensed by the NRC.” Consult 10 CFR 171.11 for information on exemptions from annual fees 

and 10 CFR 171.16(c) on reduced annual fees for licensees that qualify as “small entities.” 

Note that in order to pay reduced fees, a licensee that qualifies as a “small entity” must provide 

proper certification of this status to the NRC each year along with its annual fee payment. 

Direct all questions about the NRC’s fees or completion of Item 12 of NRC Form 313 to the 

Office of the Chief Financial Officer at NRC Headquarters in Rockville, MD, 301-415-7554. 

Information about fees may also be obtained by calling NRC’s toll-free number, 

800-368-5642, extension 415-7554. The e-mail address is Fees.Resource@nrc.gov. 

 

8-1 

8 CONTENTS OF AN APPLICATION 

The following information applies to the indicated items on U.S. Nuclear Regulatory Commission 

(NRC) Form 313 (Appendix A of this NUREG). 

All items in the application should be completed in enough detail for the NRC to determine 

whether the proposed equipment, facilities, training and experience, and radiation safety 

program satisfy regulatory requirements and are adequate to protect public health and safety 

and minimize danger to life and property. Consideration should be given, when developing the 

application, to the concepts of keeping exposure as low as is reasonably achievable (ALARA), 

minimizing contamination, and maintaining control of radioactive materials. 

Title 10 of the Code of Federal Regulations (10 CFR) 20.1101(b) states: “The licensee shall 

use, to the extent practical, procedures and engineering controls based upon sound radiation 

protection principles to achieve occupational doses and doses to members of the public that are 

as low as is reasonably achievable (ALARA).” Regulatory Guide 8.10, Rev. 2, “Operating 

Philosophy for Maintaining Occupational Radiation Exposures as Low as Is Reasonably 

Achievable,” discusses the ALARA concept and philosophy. The application should document 

ALARA considerations, including establishing administrative action levels and monitoring 

programs. 

10 CFR 20.1406, “Minimization of contamination,” requires applicants for licenses to describe 

how facility design and procedures for operation will minimize, to the extent practicable, 

contamination of the facility and the environment; facilitate eventual decommissioning; and 

minimize, to the extent practicable, the generation of radioactive waste. As with ALARA 

considerations, applicants should address these concerns for all aspects of their programs. 

The application should include information on how the licensee will implement the security 

requirements in 10 CFR 20.1801, “Security of stored material,” and 10 CFR 20.1802, “Control of 

material not in storage.” 

After an application for broad scope authority has been reviewed by the NRC staff and found to 

be generally complete and responsive to NRC Form 313 (Appendix A of this NUREG) and this 

guidance, a pre-licensing visit may be scheduled by the NRC at the licensee’s facility. A visit or 

conference may also be scheduled as part of the license renewal process. A pre-licensing visit 

provides the NRC staff with an opportunity to better evaluate the proposed program and the 

necessity for a broad scope license. The pre-licensing visit also provides the NRC staff an 

opportunity to meet with licensee management and others responsible for the radiation 

protection program and stress the importance of their responsibilities under a broad license, as 

well as to discuss and agree on additional information and commitments that may be needed. If 

a broad license is not warranted, NRC staff may discuss the continuation of the program with an 

appropriate specific license. 

Refer to Appendix P of this NUREG for guidance regarding the definition of construction and the 

consideration of activities that can be performed by materials license applicants and potential 

applicants and licensees before the NRC has concluded its environmental review of the 

proposed licensing action. The majority of materials licensing actions will meet the criteria in 

10 CFR 51.22(c)(14) for a categorical exclusion. This means that the licensing action will not 

require an environmental assessment (EA) or environmental impact statement in accordance 

with 10 CFR 51.22(b), since the NRC has already determined that this type of licensing action 

does not have a significant impact on the environment. It is the applicant’s responsibility to 
8-2 

review the guidance in Appendix P to determine whether the categorical exclusion applies to the 

licensing action. 

All information submitted to the NRC during the licensing process may be incorporated as part 

of the license and will be subject to review during inspection. 

8.1 Item 1: License Action Type 

Item 1 of NRC Form 313 states the following: 

This is an application for (check appropriate item): 

Type of Action License No.

[ ] A. New License Not Applicable 

[ ] B. Amendment XX-XXXXX-XX 

[ ] C. Renewal XX-XXXXX-XX 

Check Box A for a new license request. Note that a pre-licensing visit may be conducted prior 

to issuance of the license. Also note that an initial security inspection may be conducted in 

accordance with NRC Inspection Manual Chapter 2800, “Materials Inspection Program,” before 

issuance of the license. 

Check box B for an amendment to an existing license, and provide the license number. 

Check box C for a renewal of an existing license, and provide the license number. 

See “License Amendments and Renewals” in Chapter 9 of this NUREG. 

8.2 Item 2: Name and Mailing Address of Applicant 

List the legal name of the applicant’s corporation or other legal entity with direct control over use 

of the radioactive material. A division or department within a legal entity may not be a licensee. 

An individual may be designated as the applicant only if the individual is acting in a private 

capacity and the use of the radioactive material is not connected with employment in a 

corporation or other legal entity. Provide the mailing address where correspondence should be 

sent. A post office box number is an acceptable mailing address. 

Notify the NRC of changes in the mailing address. These changes do not require a fee. 

Note: The NRC must be notified and the transfer approved before control of the license is 

transferred (see Section 9.1, “Timely Notification of Transfer of Control”). The NRC must also 

be notified when bankruptcy proceedings have been initiated (see Section 8.2.1, “Notification of 

Bankruptcy Proceedings”). 

8.2.1 Notification of Bankruptcy Proceedings 

Regulation: 10 CFR 30.34(h), 10 CFR 40.41(f)(1), 10 CFR 70.32(a)(9)(i) 

Criteria: Immediately following the filing of a voluntary or involuntary petition for bankruptcy for 

or against a licensee, the licensee must notify the appropriate NRC regional administrator in 

writing, identifying the bankruptcy court in which the petition was filed and the date of filing. 
8-3 

Discussion: Even though a licensee may have filed for bankruptcy, the licensee remains 

subject to all applicable NRC regulatory requirements. The NRC must be notified when 

licensees are in bankruptcy proceedings in order to determine whether all licensed material is 

accounted for and adequately controlled and whether there are any public health and safety 

concerns (e.g., contaminated facility). The NRC shares the results of its determinations with 

other involved entities (e.g., trustee), so that health and safety issues can be resolved before 

bankruptcy actions are completed and may request that the United States Department of 

Justice represent the NRC’s interests in the bankruptcy proceeding. 

Response from Applicant: None is required at the time of application for a new license. 

Licensees must immediately notify the NRC in writing following the filing of a voluntary or 

involuntary petition for bankruptcy by or against the licensee. 

Reference: See NUREG–1556, Volume 15, “Consolidated Guidance About Materials Licenses: 

Guidance About Changes of Control and About Bankruptcy Involving Byproduct, Source, or 

Special Nuclear Materials Licenses.” 

8.3 Item 3: Address(es) Where Licensed Material Will Be Used or 

Possessed 

“Permanent Location” 

With regard to areas where licensed materials are used or stored on a permanent basis, a 

“permanent location” may be: 

 an individual building or facility at one address 

 a contiguous, licensee-controlled geographic area, such as a campus or licenseeowned/operated/controlled business campus or park 

 a designated area at sea or on board a ship if it is a permanent jobsite 

Note: A temporary jobsite is not considered a “permanent location.” 

Specify the street address, city, and state or other descriptive address (e.g., Highway 10, 

5 miles east of the intersection of Highway 10 and State Route 234, Anytown, State) for each 

permanent location at which licensed material will be used or stored. Provide a list of activities 

to be conducted at each location. The descriptive address should be sufficient to allow an NRC 

inspector to find the permanent location. A post office box address is not acceptable. In 

addition, applicants are encouraged to provide global positioning system coordinates, as 

appropriate. 
8-4 

An acceptable location of use or possession specifies street address, city, State, and zip code and does 

not include a post office box number.

Figure 8-1. Location of Use or Possession 

If byproduct material is to be used at more than one permanent location, applicants should give 

the specific address of each location.1

 Applicants for a broad scope license need not identify 

each building or facility at a particular location. For example, applicants may specify that 

byproduct material will be used on the Main Campus of ABC University located in Anytown, 

State and at 123 First Street, Anothertown, State. 

Applicants should identify the location of all facilities designed or established for special uses, 

e.g., panoramic irradiators, interim or long-term waste storage facilities, high activity 

laboratories, iodination facilities, alpha laboratories, incinerators, field studies and animal 

facilities. 

If byproduct material (e.g., portable gauging devices, well logging tools and small quantities of 

sealed and unsealed material) will be used at temporary jobsites (i.e., areas where work is 

conducted for limited periods of time that may include uses at sea or on board a ship), the 

scope of these activities should be described and the address may be stated as “temporary 

jobsites anywhere in the U.S. where the NRC maintains jurisdiction.” 

If byproduct material is to be used in field studies, the activities must be specifically identified 

and authorized on the license. Appendix C of this NUREG contains information required of 

applicants prior to granting authorization for field use of licensed material. 

A license amendment is required before receiving, using, or storing licensed material at a 

permanent location not already listed on the license. 

 

1

Exceptionally, large Federal licensees with more than 20 locations may generically request to use or store licensed 

material at any of their facilities instead of giving specific addresses for each location. Please note that such 

licensees must track their locations and that a list of these locations must be available during NRC inspections. 

RPh17a.ppt

082898

32 Elm St.

City, ST 12345 P.O. Box XXXX

City, ST 12345

32 ELM St.
8-5 

An NRC license does not relieve a licensee from complying with other applicable Federal, State, 

or local regulations (e.g., local zoning requirements). 

If an applicant submits documents that give the exact location of use and storage for any 

amount of radioactive material, the applicant should mark these documents as “Security Related 

Information—Withhold under 10 CFR 2.390.” See Chapter 6, “Identifying and Protecting 

Sensitive Information,” for more details. 

Note: As discussed in Section 8.5.2, “Financial Assurance and Recordkeeping for 

Decommissioning,” licensees must maintain permanent records describing where licensed 

material was used or stored while the license was in effect. This is important for making future 

determinations about the release of these locations for unrestricted use (e.g., before the license 

is terminated). Acceptable records are sketches, written descriptions of the specific locations, 

or room numbers where licensed material is used or stored and any records of spills or other 

unusual occurrences involving the possible spread of contamination in or around the licensee’s 

facilities. 

8.4 Item 4: Person to be Contacted About This Application 

Identify the individual who can answer questions about the application, and include a telephone 

number where the individual may be contacted as well as business cell phone numbers and 

e-mail addresses. This individual, usually the radiation safety officer (RSO), will serve as the 

point of contact during the review of the application. If this individual is not a full-time employee 

of the licensed entity, his or her position and relationship to the licensee should be specified. 

The NRC should be notified if the person assigned to this function changes or if his or her 

telephone number, cell phone number, or e-mail address changes. Notification of a contact 

change is only provided for informational purposes and would not be considered an application 

for license amendment, unless the notification involves a change in the contact person who is 

also the RSO. 

As indicated on NRC Form 313 (see Appendix A of this NUREG), Items 5 through 11 should be 

submitted on separate sheets of paper. Applicants may use Appendix B of this NUREG for this 

purpose and should note that using the suggested wording of responses and committing to use 

the model procedures in this NUREG will facilitate the NRC’s review. 

8.5 Item 5: Radioactive Material 

8.5.1 Unsealed or Sealed Byproduct Material 

Regulations: 10 CFR 30.32(d), 10 CFR 30.32(g), 10 CFR 30.32(i), 10 CFR 30.33(a), 

10 CFR 30.72, 10 CFR 32.210, 10 CFR 33.11, 10 CFR 33.13, 10 CFR 33.14, 10 CFR 33.15, 

10 CFR 33.17, 10 CFR Part 37 

Criteria: An application for a specific license will be approved if the requirements of 10 CFR 

30.32; 10 CFR 30.33(a); 10 CFR 33.11; 10 CFR 33.12; 10 CFR 33.13; 10 CFR 33.14; 10 CFR 

33.15, as appropriate; and 10 CFR 33.17 are met. Licensees must also protect aggregated 

Category 1 and Category 2 quantities of radioactive material, as defined in 10 CFR 37.5, from 

theft, diversion, and sabotage. 

Discussion: Applicants for a Type A broad scope license typically request any form of 

byproduct material with atomic numbers from 1 through 83. The applicant should state the 
8-6 

maximum quantity of each radionuclide to be possessed at any one time and the total 

cumulative quantity for all radionuclides. When establishing individual radionuclide and total 

cumulative quantities, all materials possessed under the license should be included 

(i.e., materials received awaiting use, materials in use/process, and those categorized as waste 

awaiting disposal). The maximum quantity for each individual radionuclide and total cumulative 

possession should be commensurate with the applicant’s needs, facilities, procedures, and 

demonstrated experience/capability. If certain individual radionuclides will be needed in much 

larger quantities than described in the atomic number 1-83 request, they should be listed 

separately rather than increasing the possession limit for all radionuclides. Similarly, if 

applicants know that certain relatively more hazardous radionuclides (e.g., strontium-90) are 

needed only in smaller quantities, they should be listed separately. 

If needed, an applicant for a Type A broad scope license may request authorization to possess 

byproduct materials with atomic numbers greater than 83 (e.g., atomic numbers 84 to 96). For 

this request, the applicant should state the maximum quantity of each radionuclide to be 

possessed at any one time and the total cumulative quantity for all radionuclides. 

Note that authorization to possess byproduct materials with atomic numbers 84 through 96 does 

not authorize the possession of naturally occurring isotopes of uranium or thorium, or of 

plutonium because, even though these elements have atomic numbers within the range of 

84 through 96, these materials are either source material or special nuclear material and not 

byproduct material. Licensees may request source material and special nuclear material when 

use of these materials is directly related to the use of byproduct material under the broad scope 

license (e.g., laboratory-scale research and development or the use of depleted uranium as 

shielding). Applicants should submit separate applications for the production of radioactive 

material in a cyclotron, or for the use of source material; and special nuclear materials for 

purposes not directly related to the use of byproduct material under the broad scope license 

(e.g., subcritical assemblies and critical mass quantities of special nuclear material). 

A separate listing should also be submitted for sealed sources needed in quantities larger than 

that described in the atomic number 1-83 request (e.g., self-contained irradiators, instrument 

calibrators, sealed sources used for medical therapy, portable and non-portable gauging 

devices). Applicants must provide the manufacturer’s name and model number for each 

requested sealed source and device so that the NRC can verify that they have been evaluated 

in a Sealed Source and Device (SSD) registration certificate or provide the information 

requested by 10 CFR 30.32(g) to allow the NRC to conduct a case-by-case review. Sealed 

sources or devices containing sealed sources built to unique specifications of a given user 

(custom source) and that are intended for use solely under broad scope licenses, and are not 

transferred to another licensee, need not be evaluated by the NRC or Agreement State for 

registration if: (1) they contain less than 7.4 gigabecquerels (GBq) [200 millicurie (mCi)] of 

radioactive material or less than 740 GBq [20 curie (Ci)] of tritium, and (2) the licensee 

performs its own safety evaluation in accordance with the administrative procedures required by 

10 CFR 33.13(c)(3)(ii) and 33.14(b)(2)(ii), as appropriate. Custom sources and devices which 

contain an activity greater that these values must be submitted to the NRC or Agreement State 

for evaluation and registration in accordance with the guidance provided in NUREG–1556, 

Volume 3. Table 8-1 provides a sample of how byproduct material should be requested. 

 
8-7 

Table 8-1. Sample Format for Providing Information About Requested Radioisotopes 

Byproduct material Form Amount 

Any byproduct 

material with atomic 

numbers 3 through 83 

Any ___ gigabecquerels (GBq) 

[millicuries (mCi)] per 

radionuclide and 

___ GBq [curies (Ci)] total 

Any byproduct 

material with atomic 

numbers 3 through 83 and a 

half-life less than or equal to 

120 days 

Any ___ GBq [mCi] 

per radionuclide and 

___ GBq [Ci] total 

Hydrogen 3 Any ___ GBq [Ci] 

Carbon 14 Any ___ GBq [Ci] 

Phosphorus 32 Any ___ GBq [mCi] 

Sulfur 35 Any ___ GBq [mCi] 

Nickel 63 Sealed sources 

(insert manufacturer and 

model number) 

Not to exceed ___ GBq [mCi] 

per source and ___ GBq 

[mCi] total 

Cobalt 60 Sealed sources 

(insert manufacturer and 

model number) 

Not to exceed ___ GBq [mCi] 

per source and ___ GBq 

[mCi] total 

Applicant and licensee information on manufacturers, model numbers, and possession limits is 

sensitive and should be marked accordingly (see Chapter 6, “Identifying and Protecting 

Sensitive Information”). 

In most cases, the NRC or an Agreement State will perform a safety evaluation of sealed 

sources and devices before authorizing a manufacturer (or distributor) to distribute them to 

specific licensees. The safety evaluation is documented in an SSD registration certificate. 

Paragraph (g) of 10 CFR 32.210 provides some exceptions to this requirement. For additional 

guidance related to sealed sources and devices, see NUREG–1556, Volume 3, “Consolidated 

Guidance About Materials Licenses: Applications for Sealed Source and Device Evaluation and 

Registration.” 

Possession requests should be categorized into: (1) general areas of use (e.g., research and 

development activities, teaching, training, animal studies, and environmental studies); and (2) 

special areas of use, (e.g., gauging activities, irradiators, instrument calibrators, and medical 

applications). See Section 8.6 for more details. 

Applicants for Type A broad scope license should review the requirements for financial 

assurance (FA) and decommissioning before specifying possession limits for radionuclides with 

a half-life greater than 120 days. These requirements are discussed in Section 8.5.2 of this 

document. 

Applicants for a Type B or Type C broad scope license should request any chemical or physical 

form of byproduct material specified in 10 CFR 33.100, Schedule A. The possession limit for a 

Type B broad scope license, if only one radionuclide is possessed, is the quantity specified for 
8-8 

that radionuclide in 10 CFR 33.100, Schedule A, Column I. If two or more radionuclides are 

possessed, the possession limit is determined as follows. For each radionuclide, determine the 

ratio of the quantity to be possessed to the applicable quantity specified in 10 CFR 33.100, 

Schedule A, Column I, for that radionuclide. The sum of the ratios for all radionuclides 

possessed under the license may not exceed unity. The possession limit for a Type C broad 

scope license, if only one radionuclide is to be possessed, is the quantity specified for that 

radionuclide in 10 CFR 33.100, Schedule A, Column II. If two or more radionuclides are 

possessed, the sum of the ratios, determined in the same manner as discussed above, for all 

radionuclides possessed under the license may not exceed unity. This is sometimes referred to 

informally as the “unity rule” or the “rule of ratios.” Compliance with the unity rule and the 

Schedule A possession limits may be reviewed during inspection. 

Type B and Type C broad scope applicants/licensees that require materials not specified in 

Schedule A should either: (1) develop Type A broad scope programs, or (2) carry these 

additional materials under a separate specific license of limited scope. For the latter option, the 

applicant should review the base NUREG related to the planned specific use of this material 

and submit the information needed by the license reviewer as described in that document. For 

example, applicants that require materials not specified in Schedule A for purposes of research 

and development should review NUREG–1556, Volume 7, “Consolidated Guidance About 

Materials Licenses: Program-Specific Guidance About Academic, Research and Development, 

and Other Licenses of Limited Scope,” and submit the information described therein. 

Type B licensees that require quantities of material in excess of that permitted by 

10 CFR 33.11(b) should: (1) develop a Type A broad scope program and apply for a Type A 

license, or (2) apply for a separate specific license of limited scope for these additional 

quantities, as described in the previous paragraph. Type C licensees that require quantities of 

material in excess of that permitted by 10 CFR 33.11(c) should: (1) develop, as appropriate, a 

Type A or Type B broad scope program, and apply for such a license or (2) apply for a separate 

specific license of limited scope for these additional materials. 

Applicants for Type B or Type C broad scope license may consider limiting their possession of 

isotopes described in Schedule A with half-lives greater than 120 days below that amount 

permitted by 10 CFR 33.11(b) or 33.11(c), respectively, to avoid being required to submit 

certification of FA or a decommissioning funding plan. See Section 8.5.2 of this document for a 

discussion of Financial Assurance and Recordkeeping for Decommissioning. 

The regulations in 10 CFR Part 37 apply to licensees that possess an aggregated “Category 1 

quantity of radioactive material” or “Category 2 quantity of radioactive material.” These terms 

are defined in 10 CFR 37.5, and the radionuclides referenced in these 10 CFR 37.5 definitions 

are listed in Appendix A to 10 CFR Part 37. See Section 8.10.9, “Security Program for Category 

1 and Category 2 Radioactive Material,” of this NUREG for more information on the applicability 

and requirements of 10 CFR Part 37. 

Response from Applicant: Applicants for a Type A broad scope license should request any 

form of byproduct material with atomic numbers from 1 through 83. The applicant should state 

the maximum quantity of each radionuclide to be possessed at any one time and the total 

cumulative quantity for all radionuclides. The applicant should separately list individual 

radionuclides needed in much larger quantities or in much smaller quantities than that described 

in the atomic number 1 through 83 request. The maximum quantities of nuclides with atomic 

numbers above 83 also should be listed separately. A separate listing should also be submitted 

for sealed sources needed in larger quantities than that described in the atomic number 1 
8-9 

through 83 request. Applicants must provide the manufacturer’s name and model number for 

each requested sealed source and device so that the NRC can verify that they have been 

evaluated in a SSD registration certificate or specifically approved on a license. Applicants 

must also provide the maximum activity per source and the total possession limit. For sources 

and devices not registered, as allowed by 10 CFR 32.210(g)(2), the applicant must have 

adequate training and experience and facilities and equipment to handle comparable quantities 

of material in any form under 10 CFR 30.33(a)(2) and (3) and must provide information about 

the unregistered sealed sources and devices in accordance with 10 CFR 30.32(g)(4). 

Possession requests should be categorized into general areas of use (e.g., research and 

development activities, routine gauging activities, self-shielded irradiators, instrument 

calibrators, and medical applications). 

In accordance with 10 CFR 30.32(i), applications to possess radioactive materials in unsealed 

form, on foils or plated sources, or sealed in glass in excess of the quantities listed in 

10 CFR 30.72 must include either of the following: 

 an evaluation showing that the maximum offsite dose caused by a release of radioactive 

materials would not exceed 0.01 sievert (Sv) [1 rem] effective dose equivalent or 0.05 Sv 

[5 rem] to the thyroid; or 

 an emergency plan for responding to the release of radioactive material in accordance 

with the criteria listed in 10 CFR 30.32(i)(3) 

Applicants for a Type B or Type C broad scope license should request any chemical or physical 

form of byproduct material specified in 10 CFR 33.100, Schedule A. Type B applicants should 

request the quantity of material specified in 10 CFR 33.11(b). Type C applicants should request 

the quantity of material specified in 10 CFR 33.11(c). 

8.5.2 Financial Assurance and Recordkeeping for Decommissioning 

Regulations: 10 CFR 30.32(h), 10 CFR 30.34(b), 10 CFR 30.35, 10 CFR 30.36(e), 

10 CFR 30.36(g)(4)(v), 10 CFR 30.51, 10 CFR 40.31(i), 10 CFR 40.36, 10 CFR 40.42(e), 

10 CFR 40.42(g)(4)(v), 10 CFR 40.46, 10 CFR 40.61, 10 CFR 70.22(a)(9), 10 CFR 70.25, 

10 CFR 70.36, 10 CFR 70.38(e), 10 CFR 70.38(g)(4)(v), 10 CFR 70.51

Criteria: A licensee authorized to possess licensed material in excess of the limits specified in 

10 CFR 30.35, 10 CFR 40.36, and 10 CFR 70.25—all titled “Financial assurance and 

recordkeeping for decommissioning”— must submit a decommissioning funding plan (DFP) or 

provide a certification of financial assurance (FA) for decommissioning. 

All licensees are required to maintain records of information important to the decommissioning 

of the facility in an identified location until the site, or any area, is released for unrestricted use. 

Before licensed activities are transferred or assigned in accordance with 10 CFR 30.34(b), 

10 CFR 40.46, and/or 10 CFR 70.36, licensees must transfer records important to 

decommissioning to the new proposed licensee in accordance with 10 CFR 30.35(g), 

10 CFR 40.36(f) and/or 10 CFR 70.51(b)(3), respectively. Furthermore, before a license is 

terminated, the licensee must send records important to decommissioning that are required by 

10 CFR 30.35(g), 10 CFR 40.36(f) and/or 10 CFR 70.25(g) to the appropriate NRC regional 
8-10 

office in accordance with 10 CFR 30.51(f), 10 CFR 40.61(f), and/or 10 CFR 70.51(a)(3), 

respectively.

Discussion: The NRC seeks to ensure that decommissioning will be carried out with minimum 

impact on public and occupational health and safety and the environment. There are two parts 

to the rule: financial assurance, which applies to some licensees, and recordkeeping, which 

applies to all licensees. 

Financial Assurance 

NRC decommissioning FA regulations are designed to provide reasonable assurance that the 

decommissioning of licensed facilities will be accomplished in a safe and timely manner and that 

licensees will provide adequate funds to cover all costs associated with decommissioning. 

These requirements, if applicable, specify that a licensee either set aside funds for 

decommissioning activities or provide a guarantee that funds will be available. Applicants are 

required to provide FA when the possession of radioactive material of half-life greater than 

120 days exceeds certain limits. Criteria for determining if an applicant is required to submit a 

DFP or has an option of submitting either a DFP or a certification of FA are stated in 10 CFR 

30.35, 10 CFR 40.36, and 10 CFR 70.25. A DFP contains a site-specific cost estimate and a 

certification of FA. A certification of financial assurance includes a certification that the licensee 

has provided the required FA and an acceptable FA instrument. 

Acceptable FA includes prepayment option of a trust fund; surety, insurance, or other guarantee 

methods (letters of credit, surety bonds, parent company guarantees, insurance policies), and 

statements of intent from Government entities. Criteria for parent company guarantees and 

self-guarantees can be found in 10 CFR Part 30, Appendix A, Appendix C, Appendix D, and 

Appendix E. 

NUREG–1757, Volume 3, “Consolidated Decommissioning Guidance: Financial Assurance, 

Recordkeeping, and Timeliness” (NUREG–1757, Vol. 3), provides guidance acceptable to the 

NRC staff on the information to be provided for establishing FA for decommissioning and a 

standard format for presenting the information. Note that FA is required for four types of 

licensed materials: unsealed byproduct material, sealed byproduct material, dispersible source 

material, and unsealed special nuclear material. The total amount of FA required to be provided 

is the sum of the FA required for each of these types of materials. 

Recordkeeping for Decommissioning 

The following regulations state the requirements for maintaining records important to 

decommissioning: 10 CFR 30.35(g), 10 CFR 40.36(f), and 10 CFR 70.25(g). All licensees are 

required to maintain these records in an identified location until the site is released for 

unrestricted use. Careful recordkeeping of radionuclides used, including form, amount, and 

area used, will facilitate area release and license termination. In the event that the licensed 

activities are transferred to another person or entity, these records must be transferred to the 

new licensee before the transfer of the licensed activities takes place. 

In accordance with 10 CFR 30.35(g) and corresponding requirements in Parts 40 and 70, 

licensees must transfer records important to decommissioning to the new licensee before 

licensed activities are transferred or assigned according to 10 CFR 30.34(b) and corresponding 

requirements in Parts 40 and 70. 
8-11 

Furthermore, pursuant to 10 CFR 30.51(f), 10 CFR 40.61(f), and/or 10 CFR 70.51(a)(3), prior to 

license termination, each licensee must forward the records required by 10 CFR 30.35(g), 

10 CFR 40.36(f) and/or 10 CFR 70.25(g), respectively, to the appropriate NRC regional office. 

Figure 8-2. Types of Records That Must be Maintained for Decommissioning

Response from Applicant: 

 State the following: “Pursuant to 10 CFR 30.35(g), 10 CFR 40.36(f), and/or 

10 CFR 70.25(g) and 10 CFR 70.51(b)(3), we will maintain records important to 

decommissioning and transfer these records to an NRC or Agreement State 

licensee before licensed activities are transferred or assigned in accordance with 

10 CFR 30.34(b), 10 CFR 40.46, and/or 10 CFR 70.36. Furthermore, pursuant to 

10 CFR 30.51(f), 10 CFR 40.61(f), and /or 10 CFR 70.51(a)(3), prior to license 

termination, we will forward the records required by 10 CFR 30.35(g), 10 CFR 40.36(f), 

and/or 10 CFR 70.25(g) to the appropriate NRC Regional Office.”

AND 

 If financial assurance is required, submit evidence of financial assurance following the 

guidance of NUREG–1757, Volume 3, “Consolidated Decommissioning Guidance: 

Financial Assurance, Recordkeeping, and Timeliness.” 

BS04a.ppt

041299

Structural

Modifications As-built

Drawings

Surveys Spills

Deco

R

mmissioning

ecords

References:

Blueprints stored at

Location XYZ
8-12 

8.5.3 Emergency Plan 

Regulations: 10 CFR 30.32(i); 10 CFR 30.72 

Criteria: Applicants who will be authorized to possess radioactive material in unsealed form, on 

foils or plated sources, or sealed in glass in excess of the quantities listed in 10 CFR 30.72 must 

be prepared for the potential release of radioactive material. 

Discussion: When requesting authorization for possession limits in excess of the quantities 

listed in Schedule C of 10 CFR 30.72, the applicant must provide in conjunction with the license 

application either: 

 an evaluation showing that the maximum offsite dose due to a release of radioactive 

materials would not exceed 0.01 Sv (1 rem) effective dose equivalent or 0.05 Sv [5 rem] 

to the thyroid; or 

 an emergency plan for responding to the release of radioactive material in accordance 

with the criteria listed in 10 CFR 30.32(i)(3). 

For NRC to grant authorization to possess quantities equal to the activities specified in 

Schedule C of 10 CFR 30.72, it is necessary to provide the information outlined in 

10 CFR 30.32(i) sufficient to evaluate the need for an emergency plan. 

Response from Applicant: If an emergency plan is required, provide either: 

 an evaluation showing that the maximum offsite dose due to a release of radioactive 

materials would not exceed 0.01 Sv [1 rem] effective dose equivalent or 0.05 Sv [5 rem] 

to the thyroid; or 

 an emergency response plan for responding to the release of radioactive material in 

accordance with the criteria listed in 10 CFR 30.32(i)(3). 

References:

 Regulatory Guide 3.67, Rev. 1, “Standard Format and Content for Emergency Plans for 

Fuel Cycle and Materials Facilities.” 

8.6 Item 6: Purpose(s) for Which Licensed Material Will Be Used 

Regulations, 10 CFR 30.32(d), 10 CFR 30.33(a)(1), 10 CFR 33.13, 10 CFR 33.14, or 

10 CFR 33.15, as appropriate, 10 CFR 33.17(a) 

Criteria: An application for a license will be approved if the proposed activity is authorized by 

the Atomic Energy Act of 1954, as amended. Sealed sources and devices containing licensed 

material must be used only for the purpose for which they are designed and according to the 

manufacturer’s and distributor’s instructions and recommendations for use, as specified in the 

SSD registration certificate, unless otherwise authorized in the license. Certain types of sealed 

sources and devices containing licensed material that are intended for use solely under broad 

scope licenses, and that will not be transferred to another licensee, need not be evaluated by 

the NRC prior to use if the licensee is authorized to possess the requested quantity of 

radioactive material in unsealed form and has administrative controls and provisions established 

for ensuring safety in accordance with 10 CFR 33.13, 10 CFR 33.14, or 10 CFR 33.15, as 
8-13 

appropriate. Additional licensing guidance for the possession and use of certain types of sealed 

sources and devices that are not required to have an SSD registration certificate is provided in 

Chapter 5 of NUREG–1556, Volume 3, “Consolidated Guidance About Materials Licenses: 

Applications for Sealed Source and Device Evaluation and Registration.” 

Note: Applicants desiring certain radionuclide activities and type of uses disallowed by 

10 CFR 33.17(a) should apply for specific authorization. Refer to the “Discussion” section for 

further explanation. 

Discussion: The applicant should describe in general terms the purposes for which the 

licensed material will be used. New applicants should describe why a broad scope license is 

needed rather than amendments to an existing limited scope license. The uses should be 

consistent with prior licensed activities. Sufficient information should be provided to enable the 

reviewer to have a clear understanding of each use and to determine the potential for exposure 

of workers and members of the public to radiation and radioactive materials. The information 

provided regarding “Purpose of Use” is understood by the NRC staff as a self-imposed limitation 

contained within the application. If a broad scope licensee desires to initiate a use other than 

those described in its application and committed to in its license, the licensee must submit an 

amendment to the license to modify or expand the “purpose.” 

If the material will be used in tracer/field studies in which licensed material is deliberately 

released into the environment, or in animal studies that may result in the release of licensed 

material into the environment, an EA may be needed in accordance with 10 CFR 51.21. 

NUREG–1748, “Environmental Review Guidance for Licensing Actions Associated with NMSS 

Programs,” addresses procedures that staff should use in conducting environmental reviews. A 

memorandum dated October 20, 2009 (ADAMS Accession No. ML092321078), provides further 

guidance. 

The exclusions stated in 10 CFR 33.17(a) provide that, unless specifically authorized by other 

parts of the regulations, persons licensed under broad scope licenses are not permitted to do 

any of the following: 

 conduct tracer studies in the environment involving the direct release of radioactive 

material (applies to field users) 

 receive, acquire, own, possess, use, transfer, or import devices containing 

3.7 × 1015 becquerels (Bq) [100,000 Ci] or more of byproduct material in sealed sources 

for irradiation of materials 

 conduct activities for which a specific license issued by the Commission under 

10 CFR Part 32, “Specific Domestic Licenses to Manufacture or Transfer Certain Items 

Containing Byproduct Material”; 10 CFR Part 34, “Licenses for Industrial Radiography 

and Radiation Safety Requirements for Industrial Radiographic Operations”; or 

10 CFR Part 35, “Medical Use of Byproduct Material,” is required 

 add or cause the addition of byproduct material to any food, beverage, cosmetic, drug, 

or other product designed for ingestion or inhalation by, or application to, a human being 

Applicants desiring these activities should review the appropriate base NUREG (i.e., that 

volume of NUREG–1556 that most closely applies) and request specific authorization in 

accordance with the guidance contained therein. For example, broad scope licensees that wish 
8-14 

to perform industrial radiography should review NUREG–1556, Volume 2, “Consolidated 

Guidance About Materials Licenses: Program-Specific Guidance About Industrial Radiography 

Licenses,” and provide necessary information, as specified. 

Response from Applicant: Describe in general terms the purposes for which the licensed 

material will be used. 

8.7 Item 7: Individual(s) Responsible for Radiation Safety Program and 

Their Training and Experience 

Executive management, the radiation safety committee (RSC), if required, and the RSO and his 

or her staff, as necessary, work as a team to oversee the broad scope program. Each plays a 

critical role within a given area of responsibility. The roles and responsibilities of executive 

management, the RSC, the RSO, and the radiation safety office staff are discussed in the 

sections that follow. 

Note: NUREG–1516, “Management of Radioactive Material Safety Programs at Medical 

Facilities,” published May 1997, describes the role of executive management, the RSC, and the 

RSO at medical facilities but contains information pertinent to all broad scope programs. 

8.7.1 Executive Management 

Regulations: 10 CFR 20.1101(c), 10 CFR 33.13(c), 10 CFR 33.14(b), 10 CFR 33.15(c)

Criteria: The applicant must have administrative controls and provisions relating to 

organization and management and management review necessary to ensure safe operations. 

Discussion: Executive management is the individual at the senior management level who is 

responsible for oversight of the facility’s radiation safety program and has the ultimate 

responsibility for the license and the activities associated with the license. Executive 

management has an important role in implementing and managing the radiation safety program. 

Because of the various structures of different organizations, the NRC recognizes that executive 

management may need to delegate certain responsibilities to other managers for the day-to-day 

oversight of the program. For example, a large company may have several licenses issued to it 

for different facilities or for different activities. In this case, a company may choose to establish 

a senior level manager with responsibility for all of the licenses issued to that company, with the 

day-to-day responsibility for each license designated to the facility senior manager or program 

area senior manager. There are numerous ways in which an applicant may wish to structure its 

management oversight to meet the needs of the organization. However, there must still be one 

level of management, as the licensee’s representative, with ultimate responsibility for the 

radiation safety program. 

In a Type A broad scope program, executive management or her or his delegate is a vital 

member of the RSC and should attend committee meetings. In all licensed programs, executive 

management should be knowledgeable of the results of periodic audits and the annual review of 

the licensed program to ensure all activities comply with regulatory requirements and the 

conditions of the license and that activities are being conducted in a safe manner. Annual 

reviews and audits are discussed in more detail in Section 8.10.1 of this guidance document. 
8-15 

The licensee should consider several factors when selecting executive management for the 

radiation safety program. This individual represents the highest level of facility management 

and has the authority to delegate resources for the program and appropriate funds in a timely 

manner. This individual must be available to facilitate effective and immediate action on behalf 

of management, the RSC (for Type A broad scope programs), and the RSO, particularly in the 

event of an emergency. Executive management must have the authority to make prompt 

decisions without having to consult with higher management officials, including the authority to 

take whatever action is necessary to ensure that all radiation safety practices are in accordance 

with the regulations and conditions of the license. 

Executive management is involved in selecting the chairperson and members of the RSC 

(for Type A broad scope) and the RSO (for Type A and Type B broad scope), and defines the 

role, duties, and responsibilities of each. Executive management should support the RSC and 

the RSO, creating an atmosphere of cooperation and professionalism such that individuals feel 

comfortable raising radiation safety concerns. Authority will be enhanced if authorized users 

clearly understand that there is strong management support for, and participation in, the 

licensed program. Many problems can be avoided if management takes proactive steps before 

radiation safety problems escalate. Individuals should understand management’s expectations 

regarding internal enforcement of program requirements and the consequences for 

noncompliance. 

NUREG–1516, “Management of Radioactive Material Safety Programs at Medical Facilities,” 

Chapter 1, describes the role of executive management at medical facilities, but it also contains 

information pertinent to all broad scope programs. 

Response from Applicant: The applicant must describe administrative controls and provisions 

relating to organization and management and management review necessary to ensure safe 

operations. The applicant should submit an organizational chart that describes the 

management structure, reporting paths, and the flow of authority between executive 

management, the RSC (for Type A broad scope), and the RSO (For Type A and Type B broad 

scope). 

8.7.2 Radiation Safety Committee 

Regulations: 10 CFR 33.13(c)(1), 33.13(c)(3)(iii)

Criteria: Type A broad scope licensees must establish an RSC that works with executive 

management and the RSO in implementing the radiation safety program. Type B and Type C 

broad scope licensees are not required to establish an RSC. 

Discussion: An applicant for a Type A broad scope license must establish an RSC pursuant to 

10 CFR 33.13(c)(1). The RSC works with executive management and the RSO to implement 

the radiation safety program, and will be involved in establishing policies and procedures for 

managing the radiation safety program. The RSC, through the executive management, must 

have the authority and flexibility necessary to effectively fulfill its role in managing the radiation 

safety program. 

The RSC for a Type A broad scope program is composed of such persons as the RSO, 

executive management, and persons trained and experienced in the safe use of radioactive 

material. Each area of use under the license should be represented on the RSC. 
8-16 

The licensee should select a chairperson for the committee. There are several factors to 

consider when making this selection. An individual with a knowledge of radiation safety issues, 

good leadership abilities, the authority and credibility by virtue of his or her position within the 

facility, and a desire to serve as chairperson will facilitate the effectiveness of the RSC. 

Additionally, the individual chosen as the chairperson must have the time to devote to the 

position in addition to other responsibilities he or she might have within the facility. Executive 

management should delegate a level of authority to the position so that the chairperson is 

effective. In general, the RSO should not be appointed as the chairperson of the committee, 

since the RSO is responsible for the day-to-day operation of the radiation safety program and 

may be too closely involved with the licensed activities to be objective. 

The RSC should establish a quorum for RSC meetings. The NRC considers as acceptable the 

following: a quorum consisting of the chairperson of the committee (or his or her designee), the 

RSO, the executive management (or his or her alternate), a representative from each area of 

use from which specific issues will be discussed, and any other member whose field of 

expertise is necessary for the discussion. 

Regulations in 10 CFR Part 33 do not specify the meeting frequency for RSC meetings for 

broad scope programs. The RSC should meet as often as needed to ensure the radiation 

safety program is operating in compliance with the license, established procedures, and the 

regulations. For most programs, quarterly RSC meetings are needed to adequately oversee the 

program. 

The RSC should maintain minutes of its meetings. The minutes should include the date of the 

meeting, the members present and absent to demonstrate a quorum was present, a summary of 

the discussions, recommendations, and the results of votes. The RSC should also document its 

review of new users, uses, and program changes. The minutes should also include information 

related to the ALARA program reviews and the annual radiation safety program review. 

Duties and Responsibilities 

The radiation safety committee is required, pursuant to 10 CFR 33.13(c)(3)(iii), to review, 

approve, and record safety evaluations of proposed uses of byproduct materials. Pursuant to 

10 CFR 33.17(b), the material possessed under the broad scope program may only be used by, 

or under the direct supervision of, individuals approved by the RSC. Therefore, one of the 

primary responsibilities of the RSC for a broad scope program is to evaluate new users and new 

uses of byproduct material. The RSC should consider all available information in making 

decisions. This includes evaluating the training and experience of applicants that request 

authorization to use radioactive material at the facility, using criteria developed by the RSC. 

The RSC members should be aware of the regulatory training and experience criteria that apply 

to each type of use at their institution. The criteria developed by the committee should include 

such things as the requester’s training and experience, the proposed facilities, the protocol for 

using radioactive material to ensure that all procedures are in accordance with good radiation 

safety practices, and waste disposal. 

Broad scope programs that also include activities that are under other NRC regulations must 

meet all applicable requirements of those regulations. Most common are medical licensees of 

broad scope, which must meet the requirements of 10 CFR Part 35, as well as those of Part 33. 

Broad scope licensees should review other base NUREGs that may apply to their licensed 

program, such as NUREG–1556, Volume 9, “Consolidated Guidance About Materials 

Licensees: Program Specific Guidance About Medical Use Licenses,” which provides guidance 
8-17 

for licensees that possess radioactive material for medical use. Similarly, guidance can be 

found for other uses of radioactive materials commonly found at broad scope programs, such as 

self-shielded irradiators (NUREG–1556, Vol. 5), use in animals including veterinary treatment 

(NUREG-1556, Vol. 7), and uses of small sealed sources in devices such as x-ray fluorescence 

devices or gas chromatographs (NUREG–1556, Vol. 7) or portable gauges (NUREG–1556, 

Vol. 1) and fixed gauges (NUREG–1556, Vol. 4). 

In addition, the committee is responsible for reviewing personnel dosimetry results, and 

discussing the results of required radiation surveys and any significant incidents, including spills, 

contamination, and medical events. Since the licensee is required under 10 CFR 20.1101 to 

maintain a radiation program based upon sound radiation protection principles to achieve doses 

that are ALARA, the RSC should review the program for maintaining doses ALARA and provide 

any necessary recommendations to ensure this. The overall compliance status for authorized 

users should be thoroughly reviewed. The RSC, working with the executive management, 

shares responsibility with the RSO for conducting periodic audits of the radiation safety 

program. Additionally, the RSC reviews any consultant’s audit findings and documents the 

acceptance or rejection of the consultant’s findings in the RSC Committee minutes. The RSC 

also reviews the results of the annual review of the radiation safety program. Licensees should 

analyze possible trends and implement timely corrective actions as needed. 

Problems should be clearly defined and reviewed in the future as open items. An assessment 

of the effectiveness of corrective actions is also helpful in deterring or eliminating future 

problems and violations. 

For Type A broad scope licensees or applicants for a Type A broad scope license that desire 

the flexibility to make certain program changes and changes to certain procedures as discussed 

in Section 1 of this document, the RSC, along with executive management and the RSO, will 

review and approve program and procedural changes in accordance with criteria developed and 

approved by the RSC. The criteria for reviewing and approving such changes should include 

provisions for training staff before implementing new procedures and ensure that the proposed 

changes will not degrade the effectiveness of the currently approved program. Additionally, the 

audit program should include an evaluation process that will ensure that changes have been 

properly implemented by the staff and will determine the effectiveness of changes made in 

achieving program goals. 

Response from Applicant: Applicants for a Type A broad scope license should submit 

the following: 

 description of the duties and responsibilities of the RSC 

 criteria used for selecting members of the RSC, including what members and the 

number of members constituting a quorum. Members should be indicated by position 

title, rather than by name 

 criteria and procedure describing the approval process used by the RSC and RSO for 

authorizing new users and new uses 
8-18 

In addition, applicants for a Type A broad scope license that request the flexibility to make some 

program changes and revise some procedures previously approved by the NRC without 

amendment of the license should submit the following: 

 a description of the duties and responsibilities of the RSC, including: 

— review and approval of permitted program and procedural changes prior to 

implementation 

— implementation of program and procedural changes 

— audit of licensed operations to determine compliance 

— appropriate actions taken when noncompliance is identified, including analysis of 

the cause, corrective actions, and actions to prevent recurrence 

 a description of the process for procedure and program review and approval, including 

documentation of the specific change (At a minimum, documentation should state the 

reason for the change and summarize the radiation safety matters that were considered 

prior to approval of the change.) 

8.7.3 Radiation Safety Officer 

Regulations: 10 CFR 30.33(a)(3), 10 CFR 33.13(c)(2), 10 CFR 33.14(b)(1)

Criteria: 10 CFR Part 33, “Specific Domestic Licenses of Broad Scope for Byproduct Material,” 

requires that Type A and Type B broad scope licensees must have a radiological safety officer, 

more commonly known as an RSO, who is qualified by training and experience in radiation 

protection and who is available for advice and assistance on radiological safety matters. The 

RSO’s training and experience should be applicable to and generally consistent with the types 

and quantities of licensed material listed on the license for which the individual’s authorization 

as an RSO is requested. While regulation does not require Type C broad scope licensees to 

have an RSO, 10 CFR 33.15, “Requirements for the issuance of a Type C specific license of 

broad scope,” requires that the licensee establish administrative controls and provisions related 

to procurement of byproduct material, procedures, recordkeeping, material control and 

accounting, and management review necessary to assure safe operations. Type C broad scope 

licensee management should appoint someone responsible for the day-to-day operation of the 

radiation safety program, who will be named as the RSO on the license. 

Discussion: Each Type A and Type B program that uses byproduct materials must appoint an 

RSO who is responsible for radiation safety and compliance with the regulations for the use of 

byproduct material. Each Type C broad scope program should appoint an individual who is 

responsible for the day-to-day operation of the radiation safety program and who will be listed 

on the license as the RSO. In a Type A broad scope license, the RSO is a member of the RSC 

and works closely with the RSC and executive management to implement the radiation safety 

program. In a Type B or Type C broad scope license, the RSO is the individual responsible for 

implementing the radiation safety program. 

Duties and Responsibilities 

For all broad scope licenses, the RSO must ensure that the licensee’s radioactive material is 

used and stored safety and securely according to approved policies and procedures, and that all 

regulatory requirements are met. The RSO must have full access to all activities that involve 

the use of byproduct material and the authority to terminate any activity in which health and 
8-19 

safety appear to be compromised without consulting with executive management or the RSC. 

The applicant should submit a “Radiation Safety Officer Delegation of Authority” signed by 

executive management. Appendix D of this NUREG contains a model “delegation of authority” 

that the NRC considers acceptable. 

 In a Type A broad scope licensed program, the RSO typically performs a preliminary 

review of proposed new uses and users, prior to formally discussing the proposal with 

the RSC. The RSC grants the formal approval of new users and uses in a Type A broad 

scope license. 

 In a Type B broad scope licensed program, the task of reviewing and approving 

proposed uses and users is the responsibility of the RSO. The Type B broad scope 

RSO should be qualified by training and experience to provide advice and assistance to 

others working with materials under the Type B broad scope license. The Type B broad 

scope RSO also is required to perform the review, approval, and recording of safety 

evaluations of proposed uses prepared in accordance with 10 CFR 33.14(b)(2)(ii) before 

use of the byproduct material. 

 In a Type C broad scope program, individuals are qualified as users if they meet the 

training and experience criteria described in 10 CFR 33.15(b). While no licensee, 

committee, or individual is required by regulation to make the determination that an 

individual is qualified to use the material possessed under the Type C broad scope 

license, or that a particular use of byproduct material is safe, licensee management 

ultimately is responsible for assuring safe operations. The licensee should designate 

one individual to be named as the RSO on the license. This individual will be the NRC’s 

primary technical contact for matters related to the Type C broad scope license. 

The RSO performs audits of all areas of use and individuals who are authorized to use 

byproduct material to ensure work is done in accordance with the license, regulations, and user 

permit conditions. Specific duties and responsibilities of the RSO include but are not limited to: 

 monitoring and surveys of all areas in which radioactive material is used 

 overseeing ordering, receipt, surveys, and delivery of byproduct material 

 packaging, labeling, surveys, etc., of all shipments of byproduct material leaving the 

institution 

 monitoring programs, including determining the need for and evaluating bioassays, 

monitoring personnel exposure records, and developing corrective actions for those 

exposures approaching maximum permissible limits 

 developing and implementing an ALARA program 

 training all personnel 

 overseeing the waste disposal program 

 monitoring inventory and leak tests of sealed sources 

 overseeing decontamination 
8-20 

 investigating incidents, responding to emergencies and notifying the appropriate 

agencies 

 for licensees possessing an aggregated Category 1 or Category 2 quantity of radioactive 

material, participating in the development and implementation of a security program for 

radioactive material in accordance with 10 CFR Part 37 (a “Category 1 quantity of 

radioactive material” and a “Category 2 quantity of radioactive material” are defined 

terms in 10 CFR 37.5, and the radionuclides referenced in these 10 CFR 37.5 definitions 

are listed in Appendix A to 10 CFR Part 37) 

 maintaining all required records 

The applicant or licensee may not transfer the responsibilities of the RSO to other individuals. 

Many tasks and duties associated with managing the program may be assigned or delegated to 

other qualified individuals. The responsibility for these tasks and duties, however, lies with the 

RSO. The NRC does recognize that a qualified individual will have to fill in for the RSO when 

the RSO will be away for short periods of time for professional conferences, vacation, or illness. 

However, this should not occur for extended or indefinite periods of time. The applicant or 

licensee should have a plan in place for contacting the RSO in the event of an emergency. 

When selecting an RSO, the applicant should keep in mind the duties and responsibilities of the 

position and select an individual who is qualified to serve. The RSO will need sufficient 

technical knowledge to understand, in general, the majority of the work being done with 

byproduct materials under his or her responsibility. The NRC recognizes that an RSO may not 

be an expert in all areas that might be involved in a broad scope program. The RSO should be 

qualified by training and experience to perform the duties required for the position. Executive 

management should ensure that enough time is allocated to the individual selected as the RSO 

to carry out the responsibilities of the position. 

RSO Qualifications and Training 

In order to demonstrate adequate training and experience, the RSO should have: (1) at a 

minimum, a college degree at the bachelor level, or equivalent training and experience in the 

physical or biological sciences or in engineering; and (2) training and experience commensurate 

with the scope of proposed activities. Training should include the following subjects: 

 radiation protection principles 

 characteristics of ionizing radiation 

 units of radiation dose and quantities 

 radiation detection and measurement instrumentation 

 biological hazards of exposure to radiation (appropriate to types and forms of licensed 

material to be possessed and used) 

 NRC regulatory requirements and standards 

 hands-on use of radioactive materials commensurate with the uses proposed by the 

applicant 
8-21 

The amount of training and experience will depend on the type, form, quantity, and proposed 

use of the licensed material requested. For instance, in addition to a college degree, RSOs at 

most broad scope licensees should be specialists in the field of radiation protection and may 

need at least 40 hours of radiation safety training specific to their job duties, as well as a year of 

experience with similar types, forms, quantities, and uses of radioactive material before the 

individual is qualified to be an RSO. 

The RSO designee should have obtained the above training in formal course(s) designed for 

RSOs, presented by an academic institution, commercial radiation safety consulting company, 

or a professional organization of radiation protection experts. In addition, the proposed RSO’s 

experience should be sufficient to identify and control the anticipated radiation hazards. For 

example, the RSO should have experience planning and conducting evaluations, surveys, and 

measurements similar to those required by the licensee’s radiation safety program. 

The applicant should also be aware of specific regulatory requirements for the RSO that may 

apply to its licensed program. For example, 10 CFR Part 35 contains specific requirements for 

an RSO in a medical program. However, an individual who qualifies as a medical RSO may not 

necessarily be qualified to be an RSO in a broad scope program. 

Chapters 3 and 4 of NUREG–1516, “Management of Radioactive Material Safety Programs at 

Medical Facilities,” describe the role of the RSO and selection of the RSO at medical facilities, 

but it also contains information pertinent to all broad scope programs.

Response from Applicant:

For Type A, Type B, and Type C applicants: 

 Submit the name of the proposed RSO. 

 Describe the training for the proposed RSO that demonstrates the individual is qualified 

to perform the duties required under the license. 

 Address the RSO’s experience in performing each of the duties listed in the “Duties and 

Responsibilities” section, when and where the experience was gained, and the type, 

form, and quantity of radionuclides involved. 

 Submit a statement delineating the RSO’s duties and responsibilities. 

 Submit a radiation safety officer delegation of authority memorandum signed by the 

licensee’s executive management. 

In addition, for Type B applicants, submit the criteria used by the RSO to approve new users 

and uses of byproduct material. Also, submit the criteria that the RSO will use to evaluate the 

radiation safety aspects of proposed uses, prior to approval. 
8-22 

Applicants should provide specific information about the proposed RSO’s training and 

experience that is relevant to the requested licensed material types, forms, and quantities 

requested in the application. 

Applicants should not submit extraneous information, such as unrelated lists of publications, 

research grants, committee and society memberships, etc. This only serves to slow down the 

review process. 

Applicants should not submit personal information, such as birth dates, social security numbers, 

home addresses, personal telephone or cell phone numbers, for the proposed RSO. Such 

information is not required to be submitted and under the Privacy Act, documents containing 

such information must be handled separately, which may delay completion of the review 

process. 

Note: Notify the NRC and obtain a license amendment before making changes in the 

designation of the RSO listed on the license. Applicants should review the regulations for 

specific program areas, such as medical uses, that have specific requirements regarding 

changes in the RSO.

8.7.4 Radiation Safety Office Staff 

Criteria: Licensees should provide sufficient staff to assist the RSO in implementing the 

radiation safety program.

Discussion: The licensee should provide the RSO with a sufficient staff of professional and 

administrative support personnel. The number of staff and their qualifications will vary 

depending on the scope of the program. For small programs, the RSO may not require any 

assistance. Licensees should evaluate the licensed program and ensure that the RSO has 

adequate resources to effectively manage the program. 

Chapters 6 and 7 of NUREG–1516, “Management of Radioactive Material Safety Programs at 

Medical Facilities,” discuss the subjects of radiation safety program resources and the use of 

consultants and service companies at medical facilities. However, these chapters also contain 

information pertinent to all broad scope programs. 

Response from Applicant: No response is required. 

8.8 Item 8: Training for Individuals Working in or Frequenting Restricted 

Areas 

Regulations: 10 CFR 19.12, 10 CFR 19.13, 10 CFR 30.33(a)(3), 10 CFR 30.34(e), 

10 CFR 37.5, 10 CFR 37.43 

Criteria: Before beginning work with or in the vicinity of licensed material, all individuals who 

are likely to receive an occupational dose in excess of 1 millisievert (mSv) [100 millirem (mrem)] 

in a year must receive radiation safety training commensurate with their assigned duties and 

specific to the licensee’s radiation safety program. Each individual should also receive periodic 

refresher training. Any licensee that possesses an aggregated Category 1 or Category 2 

quantity of radioactive material (as defined in 10 CFR 37.5) must implement a training program 

for those individuals implementing the security program. 
8-23 

Discussion: Requirements in 10 CFR 19.12(a) establish the training that licensees are 

required to provide to individuals who, in the course of their employment, are likely to receive in 

a year an occupational dose in excess of 1 mSv [100 mrem]. Section 19.12(b) of 10 CFR 

requires that the licensee, in determining which individuals are subject to the training 

requirements of 19.12(a), consider assigned activities during both normal and abnormal 

situations involving exposure to radiation and/or radioactive material that can reasonably be 

expected to occur during the life of a licensed facility. Many licensees can demonstrate that it is 

not likely during a normal situation for a laboratory worker, manufacturing technician, or 

environmental services worker at the facility to receive in a year an occupational dose in excess 

of 1 mSv [100 mrem]. However, such individuals could reasonably receive this level of 

exposure during abnormal situations (e.g., radioactive material left unsecured, a contamination 

event, or improper disposal of radioactive material in the regular trash) or, by their actions, 

cause others to receive this level of exposure. Untrained workers represent a potential hazard 

to themselves, other individuals, and property. 

Licensees should not assume that safety instruction has been adequately covered by prior 

employment or academic training. Practical, site-specific training should be provided for all 

individuals prior to beginning work with or in the vicinity of licensed material. Periodic refresher 

training should also be provided. Topics covered must, at a minimum, include those described 

in 10 CFR 19.12(a). The training may take any form. Many licensees use interactive media or 

offline computer programs to provide training. The licensee should determine if the training 

succeeded in conveying the desired information and adjust the training program as necessary. 

The person conducting the training should be a qualified individual who is familiar with the 

licensee’s program. 

Retraining should be performed whenever there is a change in duties or the work environment 

and at a frequency sufficient to ensure that all staff members are adequately trained. 

Applicants should review the model training program described in the appropriate base NUREG 

corresponding to the particular type of licensed program. For example, NUREG–1556, 

Volume 7, “Consolidated Guidance About Materials Licenses: Program-Specific Guidance 

About Academic, Research and Development, and Other Licenses of Limited Scope,” describes 

a training program that is acceptable to the NRC for licensees that are involved in research and 

development, and Volume 9, “Consolidated Guidance About Materials Licensees: Program 

Specific Guidance About Medical Use Licenses,” describes a training program that is 

acceptable to the NRC for licensees that possess radioactive material for medical use. 

The applicant should also be aware of additional specific training requirements that may apply 

to their licensed program. For example, 10 CFR Part 35 contains specific requirements for the 

training of individuals who will work under the supervision of medical authorized users. 
8-24 

In accordance with 10 CFR Part 37, any licensee that possesses an aggregated Category 1 or 

Category 2 quantity of radioactive material must implement a training program in accordance 

with 10 CFR 37.43, “General security program requirements,” and specifically, must comply with 

10 CFR 37.43(c), “Training,” to ensure that those individuals who may have a responsibility to 

implement portions of the security program possess and maintain the knowledge, skills, and 

abilities to carry out their assigned duties and responsibilities effectively. Additionally, in 

accordance with 10 CFR 37.43(c)(3), refresher training must be provided at a frequency not to 

exceed 12 months and when significant changes have been made to the security program. For 

additional guidance on implementing 10 CFR Part 37 requirements, see NUREG–2155, 

“Implementation Guidance for 10 CFR Part 37, “Physical Protection of Category 1 and Category 

2 Quantities of Radioactive Material.”” Additional information regarding best practices for 

protection of risk-significant radioactive material is available in NUREG–2166, “Physical Security 

Best Practices for the Protection of Risk-Significant Radioactive Material.” 

Please note, under 10 CFR Part 37, security plans are not submitted to the NRC, but may 

be subject to review and inspection. 

Response from Applicant:

Applicants should do one of the following: 

 Submit a description of the radiation safety training program developed for each group of 

workers, including: topics covered, qualifications of the instructors, method of training, 

method for assessing the success of the training, and the frequency of training and 

refresher training. 

 Identify the model training program described in the appropriate base NUREG 

corresponding to the particular type of licensed program and submit a statement that this 

training program will be implemented. 

In addition, Type A broad scope licensees or applicants that want the flexibility to revise their 

radiation safety training program without amendment of the license (as discussed in Chapter 1, 

“Purpose of Report,” and Section 8.7.2, “Radiation Safety Committee,” of this NUREG) should 

describe the process that will be used to revise and implement their submitted training 

programs. 

8.9 Item 9: Facilities and Equipment 

Regulations: 10 CFR 20.1101, 10 CFR 20.1406, 10 CFR 30.33(a)(2), 10 CFR 30.34(e), 

10 CFR 30.35(g), 10 CFR 33.13, 10 CFR 33.14, 10 CFR 33.15, 10 CFR Part 37, 10 CFR 37.5, 

10 CFR 37.49, 10 CFR 37.53

Criteria: Facilities and equipment must be adequate to protect health and minimize danger to 

life or property and provide enhanced physical protection of aggregated Category 1 and 

Category 2 quantities of radioactive material, as defined in 10 CFR 37.5. They must minimize 

the possibility of contamination and keep exposures to workers and the public ALARA.

Discussion: Applicants for all broad scope licenses need to demonstrate that their facilities 

and equipment provide sufficient engineering controls and barriers to protect the health and 

safety of the public and their employees, keep exposures to radiation and radioactive materials 
8-25 

ALARA, and minimize the danger to life and property from the types and quantities of 

radioactive materials to be used. Facilities and equipment designed to control exposure can 

range from a vial that contains licensed material to buildings, fences, or exclusion areas that are 

between the source and the maximally exposed member of the public. These facilities not only 

reduce the exposure from the source but may also limit access to the source. The licensee 

should list and describe these facilities for the following purposes: 

 to show compliance with a regulation 

 to demonstrate the use of the material will be within the ALARA concept 

 to meet emergency response requirements 

Licensees should consider controlling exposures through available engineering options, as well 

as through limiting the releases of effluents to the environment. The licensee should describe 

all facilities and equipment essential to achieving these goals. The licensee will also need to 

describe the criteria that will be used by the RSC and/or RSO, as appropriate, to review and 

approve proposed facilities. Facilities and equipment used for special applications in which the 

effect on workers or the public could be significant if radioactive material were to be released 

accidentally, should be specifically described. These would include, for example, self-shielded 

irradiators, specialized radio-labeling (iodination or tritiation) facilities, alpha laboratories, 

radioactive waste processing facilities (including incinerators, compactors, liquid reclamation 

processors, etc.), radioactive waste storage facilities (including decay-in-storage locations), 

individual laboratories processing 3.7 GBq [100 mCi] or more of radioactive materials per 

experiment or process, nuclear pharmacies, specially designed therapy rooms, and sealed 

source storage areas. Significant modifications that affect facilities and equipment should have 

prior RSO review and RSC approval before being adopted. 

Also note that if radioactive materials will be used in or on animals, licensees should provide a 

description of the animal handling and housing facilities. NUREG–1556, Volume 7, 

“Consolidated Guidance About Materials Licenses: Program-Specific Guidance About 

Academic, Research and Development, and Other Licenses of Limited Scope,” provides 

guidance on the information that should be addressed concerning the use of radioactive 

materials in animals. 

In discussing the criteria used to evaluate their facilities and equipment, applicants should 

include information on how a laboratory or facility classification scheme relates to toxicity and 

quantity of radioactive material and the facility and equipment requirements. 

Appendix E of this NUREG provides additional guidance regarding facilities and equipment 

used to handle radioactive materials in a laboratory setting. 
8-26 

In accordance with 10 CFR Part 37, any licensee that possesses an aggregated Category 1 or 

Category 2 quantity of radioactive material must, among other things, 

 implement the physical protection requirements in 10 CFR Part 37 for material in use 

and storage, at both permanent and temporary jobsites; and 

 in accordance with 10 CFR 37.49, be able to monitor, detect without delay, assess, and 

respond to any unauthorized entries into security zones, including those surrounding 

mobile devices, and immediately detect any unauthorized removal of Category 1 

quantities of radioactive material from the security zone. (Monitoring and detection 

systems may include, among other methods, monitored video surveillance systems and 

electronic devices for intrusion detection alarms.) 

 for mobile devices containing Category 1 or Category 2 quantities of radioactive 

material, have two independent physical controls to secure the material from 

unauthorized removal when the device is not under direct control and constant 

surveillance in accordance with 10 CFR 37.53. “Mobile device” is defined in 

10 CFR 37.5 

For additional guidance on implementing 10 CFR Part 37 requirements, see NUREG–2155, 

“Implementation Guidance for 10 CFR Part 37, “Physical Protection of Category 1 and 

Category 2 Quantities of Radioactive Material.”” Additional information regarding best 

practices for protection of risk-significant radioactive material is available in NUREG–2166, 

“Physical Security Best Practices for the Protection of Risk-Significant Radioactive Material.” 

Please note, under 10 CFR Part 37, security plans are not submitted to the NRC, but may 

be subject to review and inspection.

Response from Applicant: Describe the criteria the RSC or RSO, as appropriate, will use to 

review and approve facilities and equipment (research laboratories, iodination facilities, waste 

storage facilities, survey and counting equipment, etc.). The description will need to include the 

method of classifying laboratories based on type, toxicity, and quantity of byproduct material 

being requested. Sample diagrams should be provided for each classification scheme that take 

into consideration shielding, the proximity of radiation sources to unrestricted areas, and other 

items related to radiation safety. When reviewing facilities where radioactive materials may 

become airborne, sample diagrams should take into consideration descriptions of the ventilation 

systems—including pertinent airflow rates, pressures, filtration equipment, and monitoring 

systems. For facilities and equipment used in special applications, such as those described 

above, the application will include their locations (i.e., buildings and room numbers) and special 

considerations that the RSC or RSO (or both) will use in authorizing byproduct material use. 

Also describe the procedures for control, review, and approval of significant facilities or 

equipment modifications. 

Reference: For further information on facility design, see Chapter 4 of NCRP Report No. 127, 

“Operational Radiation Safety Program.” 

 
8-27 

8.10 Item 10: Radiation Safety Program 

8.10.1 Audit and Review of Program 

Regulations: 10 CFR 33.13(c), 10 CFR 33.14(b), 10 CFR 33.15(c), 10 CFR 20.1101, 

10 CFR 20.2102, 10 CFR 37.33, 10 CFR 37.55 

Criteria: Applicants for Type A, Type B, and Type C broad scope licenses are required by 

10 CFR 33.13(c), 33.14(b), and 33.15(c), respectively, to establish administrative controls and 

provisions related to management review necessary to ensure safe operations. 

10 CFR 20.1101(c) requires the licensee to review the radiation program content and 

implementation, periodically (at least annually). 10 CFR 20.2102, “Records of radiation 

protection programs,” requires licensees to maintain records of the radiation protection program, 

including: (1) the provisions of the program; and (2) audits and other reviews of the program 

contents and implementation. Licensees that are subject to the requirements in 10 CFR Part 37 

must annually review their access authorization program and security program. 

Discussion: 

Management and Radiation Safety Committee Audits 

The application for a Type A, B, or C broad scope license should discuss executive 

management oversight of the licensed program and the mechanisms they will use to ensure 

awareness of NRC regulations, the provisions of the license, and the compliance status of the 

institution’s licensed program. This oversight may include independent audits of the program, 

frequent meetings with the RSC or RSO (or both), as appropriate, and periodic tours of selected 

facility areas. 

In a Type A broad scope program, the RSC assists executive management in performing this 

oversight function. Detailed written procedures should be developed and implemented for the 

operation of the RSC to ensure that appropriate oversight is provided. The RSC should be fully 

aware of the operations and activities of the RSO. The RSC should conduct periodic interactive 

management audits and evaluations of the radiation safety program’s performance, including: 

nonconformance reports, corrective action, status reports and audits, incident investigation 

reports, ALARA program development and implementation, effluent releases, qualification and 

radiological safety training, and performance of the RSO. Licensees should report results of the 

RSC’s audit and program reviews to executive management to allow for timely remedial actions 

sufficient in scope to ensure safe operations and compliance with NRC regulations and license 

conditions. 

Appendix F of this NUREG contains an annual audit program that the NRC finds acceptable for 

use in the review of most non-medical broad scope programs. Because all areas indicated in 

Appendix F may not be applicable to every licensee and all items may not need to be addressed 

during each audit, licensees may wish to develop a program-specific audit checklist. 

10 CFR 20.1101(c) requires the licensee to review the radiation program content and 

implementation periodically (at least annually). Reviews should be conducted at least once 

every 12 months. 
8-28 

Internal Audits 

The application should describe the audit mechanism implemented by the RSO and her or his 

staff, or other individual who is responsible for the day-to-day operation of the licensed program, 

to determine user compliance with NRC regulations, the terms and conditions of the NRC 

license, the requirements of the RSC- or RSO-approved permits (as appropriate), good health 

physics practices, and ALARA principles. The audit program should include performance based 

routine unannounced inspections of each user’s facility and practices to supplement and audit 

the routine monitoring performed by the user. Facility inspections should include the following: 

 review of user inventory and survey records 

 evaluation of user and technician training through discussion and observation of work 

practices 

 performance of independent surveys of user work areas 

 evaluation of compliance with NRC regulations, the conditions of the license, and the 

RSC and RSO permit and safety manual requirements 

 evaluation of performance based instruction for users and technical-level staff 

The types and frequencies of monitoring performed by the RSO should be indicated. The 

intervals of surveys and audits should be frequent enough to ensure close communications and 

proper surveillance of individual radioactive material users. Applicants should consider 

developing survey and audit schedules based on activity and use (e.g., high use facilities and 

users of volatile radioactive materials may be audited weekly or biweekly; intermediate use 

facilities may be audited monthly; and low-level facilities may be audited quarterly). 

The NRC encourages licensee management to conduct performance-based reviews by 

observing work in progress, interviewing staff, and spot-checking required records. As part of 

the audit program, licensees should consider including unannounced audits of licensed material 

users to observe whether radiation safety procedures are being followed. 

It is essential that once problems are identified, comprehensive corrective actions are taken in a 

timely manner. Information Notice (IN) 96-28, “Suggested Guidance Relating to Development 

and Implementation of Corrective Action,” dated May 1, 1996, provides guidance on this 

subject. The NRC routinely reviews licensee’s records to verify whether appropriate corrective 

actions were implemented in a timely manner to prevent recurrence. It is in the best interest of 

the licensee to identify potential violations of regulatory requirements and take necessary steps 

to correct them. The NRC can opt to exercise discretion and may elect not to cite the licensee 

for these violations if prompt and effective corrective actions are implemented. 

The NRC’s Enforcement Policy may be found online at http://www.nrc.gov/aboutnrc/regulatory/enforcement/enforce-pol.html. The Enforcement Manual may be found online at 

http://www.nrc.gov/about-nrc/regulatory/enforcement/guidance.html. For examples of the 

NRC’s use of discretion in issuing a notice of violation, refer to the most recent version of NRC’s 

enforcement documents at http://www.nrc.gov/reading-rm/doc-collections/enforcement/. 

 
8-29 

Recordkeeping 

With regard to audit records, 10 CFR 20.2102 requires, in part, that licensees maintain records 

of “audits and other reviews of program content and implementation” for 3 years after the record 

is made. The NRC has found audit records that contain the following information to be 

acceptable: date of audit, name of person(s) who conducted audit, persons contacted by the 

auditor(s), areas audited, audit findings, corrective actions, and followup. 

In accordance with 10 CFR Part 37, any licensee that possesses an aggregated Category 1 or 

Category 2 quantity of radioactive material must, among other things, 

 in accordance with 10 CFR 37.33, review its access authorization programs at least 

annually to confirm compliance with the requirements of Subpart B of 10 CFR Part 37 

and ensure that comprehensive actions are taken to correct any noncompliance that is 

identified; and 

 in accordance with 10 CFR 37.55, review its security program at least annually to 

confirm compliance with the requirements of Subpart C of 10 CFR Part 37 and ensure 

that comprehensive actions are taken to correct any noncompliance that is identified. 

For additional guidance on implementing 10 CFR Part 37 requirements, see NUREG–2155, 

“Implementation Guidance for 10 CFR Part 37, “Physical Protection of Category 1 and 

Category 2 Quantities of Radioactive Material.”” Additional information regarding best practices 

for protection of risk-significant radioactive material is available in NUREG–2166, “Physical 

Security Best Practices for the Protection of Risk-Significant Radioactive Material.” 

Please note, under 10 CFR Part 37, security plans are not submitted to the NRC, but may 

be subject to review and inspection.

Response from Applicant:

 Describe the mechanisms executive management uses to ensure adequate oversight of 

the program. In addition, if a licensee is upgrading its limited scope license to a Type A 

broad scope license or is renewing its Type A broad scope license, describe the RSC’s 

involvement in these oversight mechanisms. 

 Describe the audit mechanism implemented by the RSO or other responsible individual 

to determine user compliance with NRC regulations, the terms and conditions of the 

NRC license, the requirements of the RSC- or RSO-approved permits (as appropriate), 

and good health physics practices. 

 The applicant is not required to, and should not, submit its program for conducting the 

annual audit required by 10 CFR 20.1101, “Radiation protection programs,” to the NRC 

for review as part of a license application. The NRC will review this audit program during 

inspection. 

In addition, Type A broad scope licensees or applicants that want the flexibility to revise the 

audit mechanism implemented by the RSO without amendment of the license, as discussed in 

Chapter 1,”Purpose of Report” and Section 8.7.2, “Radiation Safety Committee,” of this 

NUREG, should describe the process they will use to revise and implement their audit program. 
8-30 

References: The NRC’s Enforcement Policy, which is located on the NRC’s Web site at 

http://www.nrc.gov/about-nrc/regulatory/enforcement.html, and IN 96-28, “Suggested Guidance 

Relating to Development and Implementation of Corrective Action,” which is available in the 

“NRC Library” under Document Collections/Generic Communications on the NRC’s Web site, 

http://www.nrc.gov. 

8.10.2 Radiation Monitoring Instruments 

Regulations: 10 CFR 20.1501, 10 CFR 20.2103(a), 10 CFR 30.33(a)(2), 10 CFR 33.13, 

10 CFR 33.14, 10 CFR 33.15 

Criteria: Pursuant to 10 CFR 20.1501, “General,” licensees must possess and periodically 

calibrate radiation monitoring instruments that are necessary to protect health and minimize 

danger to life or property.

Discussion: Licensees must possess an adequate number of radiation detection and 

measurement instruments, as necessary, to maintain radiation safety and to comply with 

regulations. Licensees must ensure instruments are calibrated periodically for the radiation 

being measured. For purposes of this document, survey instruments are defined as any device 

used to measure the radiological conditions at a licensed facility. The choice of instrument 

needs to be appropriate for the type of radiation to be measured and for the type of 

measurement to be taken (count rate, dose rate, etc.). 

The applicant should submit the criteria used in determining what radiation detection and 

monitoring equipment will be required for each type of use by authorized users and the 

availability of a sufficient quantity of these instruments to both the radiation safety office and 

authorized users (e.g., ionization chambers, Geiger-Mueller (G-M) detectors, air samplers, liquid 

scintillation counters). 

The NRC requires that survey instruments used for quantitative measurements be calibrated 

periodically. Calibrations requiring the use of radioactive sources should be performed by the 

instrument manufacturer or persons specifically authorized by the NRC or an Agreement State. 

Survey instruments should be calibrated at least annually (every 12 months), unless another 

frequency is specified by regulation or license condition. Licensees seeking authorization to 

perform their own survey instrument calibrations will need to submit calibration procedures for 

review. The licensee may wish to review available industry standards for calibration of 

instruments, such as American National Standards Institute (ANSI) N323AB-2013, “Radiation 

Protection Instrumentation Test and Calibration, Portable Survey Instruments.” Regardless of 

whether an applicant is authorized to calibrate radiation survey meters or contracts an 

authorized firm to perform calibrations, the licensee must retain records of the calibration of 

instruments and equipment used for quantitative radiation measurements for 3 years after the 

record is made, in accordance with 10 CFR 20.2103(a). Appendix H of this NUREG provides 

useful information about instrument specifications and model calibration procedures that the 

NRC considers to be acceptable. 

Some instruments may only need to be checked periodically for operability and response to 

radiation rather than receive full calibration. For example, G-M detection instruments used to 

identify contamination in laboratories may only need to be checked for ability to detect low-level 

contamination. 
8-31 

Applicants will need to submit their method for assuring that instruments are checked and 

calibrated at proper frequencies. 

Response from Applicant:

 Provide the criteria used by the RSC or RSO (or both), as appropriate, to review and 

approve radiation monitoring instrumentation to ensure that appropriate radiation 

monitoring equipment will be used during licensed activities. 

 Discuss how the RSC or RSO, as appropriate, will ensure that instruments are properly 

calibrated at prescribed frequencies. 

 Submit procedures for instrument calibration or state that instruments will be calibrated 

by a vendor licensed by the NRC or an Agreement State to perform instrument 

calibrations. Licensees that want authorization to calibrate their own survey instruments 

may commit to implementing the model procedures published in Appendix H of this 

NUREG. 

 State the frequency at which instruments will be calibrated. 

In addition, Type A broad scope licensees or applicants that want the flexibility to revise their 

instrument specifications and procedure for calibration of instruments without amendment of the 

license, as discussed in Chapter 1, “Purpose of Report,” and Section 8.7.2, “Radiation Safety 

Committee,” of this NUREG, should describe the process that they will use to revise and 

implement these submitted procedures. 

Note: If an applicant or licensee wishes to perform instrument calibration as a commercial 

service, they will need to either amend their existing broad scope license or apply for a new 

NRC license authorizing commercial calibration service. 

8.10.3 Material Receipt and Accountability 

Regulations: 10 CFR 20.1501(a), 10 CFR 20.1801, 10 CFR 20.1802, 10 CFR 20.1906, 

10 CFR 20.2001, 10 CFR 20.2108(b), 10 CFR 20.2201, 10 CFR 20.2207, 10 CFR 30.34(e), 

10 CFR 30.35(g), 10 CFR 30.41, 10 CFR 30.51, 10 CFR 33.13, 10 CFR 33.14, 10 CFR 33.15, 

10 CFR 37.49, 10 CFR 37.71, 10 CFR 37.75, 10 CFR 37.77, 10 CFR 40.36(f), 10 CFR 70.25(g) 

Criteria: Licensees must—pursuant to 10 CFR Parts 20, 30, and 33—develop, implement, and 

maintain written procedures for all of the following: 

 purchasing and receipt of radioactive material 

 safely receiving and opening packages 

 ensuring control and accountability of licensed material 

Licensees must also do the following: 

 Maintain records of receipt, transfer, and disposal of licensed material. 

 Update transactions in the National Source Tracking System (NSTS), including 

performing annual inventory reconciliation, if applicable. 
8-32 

 Before transferring aggregated Category 1 or Category 2 quantities of radioactive 

material listed in Appendix A to 10 CFR Part 37, use NRC’s license verification system 

to verify that the recipient licensee is authorized to possess the radioactive material. 

 Preplan, coordinate and provide advance notification of shipment of Category 1 

quantities of radioactive material and coordinate shipment of Category 2 quantities of 

radioactive material listed in Appendix A to 10 CFR Part 37. 

Discussion: Licensed materials must be tracked from receipt to transfer to ensure 

accountability at all times; identify when licensed material may be lost, stolen, or misplaced; and 

ensure that the possession limit stated on the license is not exceeded. The NRC requires 

applicants for a broad scope license to establish appropriate administrative controls and 

provisions that are necessary to ensure safe operations, including procedures to regulate the 

control of procurement and use of byproduct material. Administrative procedures must ensure 

that only authorized individuals receive radioactive materials and then only the types and 

quantities of radioactive material that they are authorized to receive. 

Applicants for a broad scope license are strongly encouraged to develop an administrative 

procedure to control procurement and use of radioactive material that emphasizes centralized 

purchasing and receipt. The NRC has found centralized purchasing and receipt, involving the 

radiation safety officer as an active part of the process, to be effective in controlling licensed 

materials entering the licensed institution through normal commercial channels, particularly for 

larger institutions. Procedures must also be established to control licensed materials obtained 

outside of the normal channels (e.g., through the loan or transfer of materials without purchase 

or through surplus). Appendix I of this NUREG describes a model procedure for controlling 

procurement and use of specifically licensed and generally licensed radioactive material that is 

acceptable to the NRC. 

Licensees are required to develop, implement, and maintain written procedures for safely 

receiving and opening packages in accordance with 10 CFR 20.1906, “Procedures for receiving 

and opening packages.” Appendix I of this NUREG describes a model procedure for safely 

receiving and opening packages containing licensed materials that the NRC finds to be 

acceptable. Table 8-2 provides a list of package monitoring requirements. 

Table 8-2. Package Monitoring Requirements (10 CFR 20.1906) 

Package Contents Survey Type Survey Time*

Damaged Licensed Material Radiation Level 

and Radioactive 

Contamination 

[§20.1906(b)(3)] 

As soon as practicable, 

but not later than 

3 hours after receipt of 

package 

Labeled (White I, 

Yellow II, 

Yellow III) 

Not Gas nor Special Form 

Greater Than Type A† 

Radiation Level 

[§20.1906(b)(2)] 

and Radioactive 

Contamination 

[§20.1906(b)(1)] 

As soon as practicable, 

but not later than 

3 hours after receipt of 

package 

Labeled (White I, 

Yellow II, 

Yellow III) 

Gas or Special Form 

Greater Than Type A† 

Radiation Level 

[§20.1906(b)(2)] 

As soon as practicable, 

but not later than 

3 hours after receipt of 

package 
8-33 

Table 8-2. Package Monitoring Requirements (10 CFR 20.1906) 

Package Contents Survey Type Survey Time*

Labeled 

(White I, Yellow II, 

Yellow III) 

Not Gas nor Special Form 

Less Than or Equal to 

Type A†

Radioactive 

Contamination 

[§20.1906(b)(1)] 

As soon as practicable, 

but not later than 

3 hours after receipt of 

package 

Labeled 

(White I, Yellow II, 

Yellow III) 

Gas or Special Form 

Less Than or Equal to 

Type A†

None 

[§20.1906(b)(1)] 

None 

Not Labeled Licensed Material None 

[§20.1906(b)] 

None‡

*Assumes packages are received during normal working hours. If packages are received outside of normal working 

hours, the licensee has 3 hours after the beginning of the next work day to perform the required surveys 

[§20.1906(c)]. 

 

†

Type A Quantity means a quantity of radioactive material, the aggregate radioactivity of which does not exceed 

A1 for special form radioactive material, or A2, for normal form radioactive material, where A1 and A2 are given in 

Table A–1 of 10 CFR Part 71, or may be determined by procedures described in Appendix A of 10 CFR Part 71. 

‡

Excepted Packages and limited quantity packages received by many laboratories are required to have the 

appropriate identification number from the Hazardous Materials Table in 49 CFR 172.101 (i.e., the “UN number”) on 

the outside of the box, identifying it as containing radioactive materials. It is good health physics practices to perform 

an incoming survey on these packages, even though transportation regulations do not require it. 

Applicants for a broad scope license are required to establish appropriate administrative 

controls and provisions related to material control and accounting that are necessary to ensure 

safe operations. Licensees use various methods to account for receipt, use, transfer, disposal, 

and radioactive decay of unsealed licensed material (e.g., computer programs, manual ledgers, 

log books). These methods help to ensure that licensee and individual authorized user 

possession limits are not exceeded. Licensees that possess sealed sources are required to 

perform periodic inventories. The frequency of these inventories is normally established by 

license condition at intervals not to exceed 6 months; however, regulations may specify a 

different inventory frequency. For aggregated Category 1 and Category 2 quantities of 

radioactive material, licensees must, according to 10 CFR 37.49(a)(1), continuously monitor and 

detect, without delay, all unauthorized entries into security zones. Additionally, for Category 1 

quantities of radioactive material, 10 CFR 37.49(a)(3)(i) requires immediate detection of any 

attempted unauthorized removal of the radioactive material from the security zone. For 

Category 2 quantities of radioactive material, 10 CFR 37.49(a)(3)(ii) requires weekly verification 

through physical checks, tamper indicating devices, use, or other means to ensure that the 

radioactive material is present. 

The NRC considers licensed material to become part of the licensee’s inventory at the time that 

it is received by the licensee, be it during normal working hours or after hours when delivered by 

the carrier in accordance with procedures established by the licensee. If through some error, 

the licensee receives material it is unauthorized to possess or receives quantities of material 

that would result in the total inventory being in excess of license possession limits, the licensee 

should place the package in secure storage and arrange for the return of these materials in a 

timely manner. If return of the materials is not possible, the licensee should contact the NRC 

regional office and request issuance of an expedited license amendment. The materials must 

not be used until the amendment is granted. 

Licensees are required under 10 CFR 20.1801 and 20.1802 to secure radioactive materials 

from unauthorized removal or access while in storage in controlled or unrestricted areas and to 
8-34 

control and maintain constant surveillance over licensed material that is in a controlled or 

unrestricted area and is not in storage. Applicants should establish policies and procedures for 

ensuring accountability of licensed materials. Licensed materials should be tracked from receipt 

to transfer to ensure accountability at all times; to identify when licensed material may be lost, 

stolen, or misplaced; and to ensure that the possession limit stated on the license is not 

exceeded. 

Licensees must maintain records of receipt, use, transfer, and disposal of all licensed material. 

Table 8-3 below lists each type of record and how long the record must be maintained. 

Table 8-3. Record Maintenance 

Type of Record How Long Record Must be Maintained

Receipt For as long as the material is possessed and for 3 years 

following the transfer or disposal of the material 

Transfer For 3 years after each transfer unless a specific requirement 

dictates otherwise 

Disposal Until the NRC terminates the license 

Important to decommissioning* Until the site is released for unrestricted use 

*Information about locations where licensed material is used or stored are among the records important to 

decommissioning and required by 10 CFR 30.35(g), 10 CFR 40.36(f), and/or 10 CFR 70.25(g). See also the section 

on “Financial Assurance and Recordkeeping for Decommissioning.” 

Category 1 and Category 2 sealed sources listed in Appendix E to 10 CFR Part 20 

(i.e., nationally tracked sources) must be tracked in the National Source Tracking System 

(NSTS) in accordance with 10 CFR 20.2207. The regulations in 10 CFR 20.2207 require that 

each licensee that manufactures, transfers, receives, disassembles, or disposes of a nationally 

tracked source shall complete and submit a National Source Tracking Transaction Report 

(NSTTR) to the NRC. The NSTTRs are maintained in the NSTS, a secure computer system 

that tracks Category 1 and Category 2 nationally tracked sources from the time they are 

manufactured or imported through the time of their disposal or export, or until the source activity 

decays to below Category 2.
8-35 

There are additional security requirements for shipment and transfer of a Category 1 and 

Category 2 quantity of radioactive material listed in Appendix A to 10 CFR Part 37. Prior to 

transferring Category 1 or Category 2 quantities of radioactive material, licensees must use 

NRC’s license verification system (or contact the licensing authority) to verify that the recipient 

licensee is authorized to possess the radioactive material. Licensees that ship Category 1 or 

Category 2 quantities of radioactive material must preplan and coordinate such shipments in 

accordance with 10 CFR 37.75. Shipments of Category 1 quantities are also subject to the 

10 CFR 37.77 advance notification requirements. For additional guidance on 

implementing 10 CFR Part 37 requirements, see NUREG–2155, “Implementation Guidance for 

10 CFR Part 37, “Physical Protection of Category 1 and Category 2 Quantities of Radioactive 

Material.”” Additional information regarding best practices for protection of risk-significant 

radioactive material is available in NUREG–2166, “Physical Security Best Practices for the 

Protection of Risk-Significant Radioactive Material.” 

Please note, under 10 CFR Part 37, security plans are not submitted to the NRC, but may 

be subject to review and inspection. 

Response from Applicant:

 Describe the administrative procedures used to ensure control of procurement and use 

of byproduct material. 

 While the applicant is required to develop and implement safe opening procedures for 

packages containing radioactive material, the applicant need not submit the procedures 

during the licensing process. The NRC will review activities subject to these procedures 

during inspection. 

 Provide the following statement: “We will develop, implement and maintain procedures 

for ensuring accountability of licensed materials at all times.” 

 Describe the administrative controls and provisions related to materials control, 

accounting and security. Describe the method for maintaining accountability of licensed 

material at all times. 

 If applicable, provide the following statement: “We will comply with the National Source 

Tracking System (NSTS) reporting requirement as described in 10 CFR 20.2207.” 

In addition, Type A broad scope licensees or applicants that want the flexibility to revise their 

administrative procedures concerning control of procurement and use of byproduct material 

without amendment of their licenses (as discussed in Chapter 1, “Purpose of Report,” and 

Section 8.7.2, “Radiation Safety Committee,” of this NUREG) should describe the process their 

RSC will use to revise these administrative procedures, controls, and provisions. Licensees and 

applicants should also do this when making revisions to their administrative controls and 

provisions related to material control, accounting, and security. 

8.10.4 Occupational Dose 

Regulations: 10 CFR 19.13, 10 CFR 20.1201, 10 CFR 20.1202, 10 CFR 20.1203, 

10 CFR 20.1204, 10 CFR 20.1207, 10 CFR 20.1208, 10 CFR 20.1501, 10 CFR 20.1502, 

10 CFR 20.1703, 10 CFR 20.2104, 10 CFR 20.2106, 10 CFR 20 Appendix B
8-36 

Criteria: Licensees must evaluate the potential occupational exposure of all workers and 

monitor occupational exposure. 

The use of individual monitoring devices for external dose is required, pursuant to 

10 CFR 20.1502(a), for: 

 adults who are likely to receive an annual dose from sources external to the body in 

excess of any of the following (each evaluated separately): 

— 5 mSv [0.5 rem] deep-dose equivalent 

— 15 mSv [1.5 rems] lens (of the eye) dose equivalent 

— 50 mSv [5 rems] shallow-dose equivalent to the skin 

— 50 mSv [5 rems] shallow-dose equivalent to any extremity 

 minors who are likely to receive an annual dose from sources external to the body in 

excess of any of the following (each evaluated separately): 

— 1.0 mSv [0.1 rem] deep-dose equivalent 

— 1.5 mSv [0.15 rem] lens (of the eye) dose equivalent 

— 5 mSv [0.5 rem] shallow-dose equivalent to the skin 

— 5 mSv [0.5 rem] shallow-dose equivalent to any extremity 

 declared pregnant women who are likely to receive a dose from radiation sources 

external to the body during the entire pregnancy in excess of 1.0 mSv [0.1 rem] 

deep-dose equivalent 

 individuals entering a high or very high radiation area 

Internal exposure monitoring is required, pursuant to 10 CFR 20.1502(b), for the following: 

 adults likely to receive, in 1 year, an intake in excess of 10 percent of the applicable 

annual limit on intake for ingestion and inhalation 

 minors likely to receive, in 1 year, a committed effective dose equivalent in excess of 

1.0 mSv [0.1 rem] and declared pregnant women likely to receive, during the entire 

pregnancy, a committed effective dose equivalent in excess of 1 mSv [0.1 rem] 
8-37 

Figure 8-3. Annual Dose Limits for Adult Radiation Workers

Discussion:

Total effective dose equivalent (TEDE) equals the effective dose equivalent (for external 

exposures) plus the committed effective dose equivalent (for internal exposures). 

If an adult radiation worker is likely to receive in 1 year a dose greater than 10 percent of any 

applicable limit (See Figure 8-2 for annual dose limits), monitoring for occupational exposure is 

required. Monitoring is required for minors and declared pregnant females as shown in the 

criteria section. The licensee should perform an evaluation of the dose the individual is likely to 

receive prior to allowing the individual to receive the dose. This evaluation need not be made 

for every individual; evaluations can be made for employees with similar job functions or work 

areas. An example of such an evaluation for a licensee possessing only sealed sources in 

portable gauges is provided in Appendix H of NUREG–1556, Volume 1, Revision 2, 

“Consolidated Guidance About Materials Licenses: Program-Specific Guidance About Portable 

Gauge Licenses.” Also, an example of such an evaluation for a licensee possessing only 

sealed sources in fixed gauges is provided in Appendix G of NUREG–1556, Volume 4, Revision 

1, “Consolidated Guidance About Materials Licenses: Program-Specific Guidance About Fixed 

Gauge Licenses.” 

If this prospective evaluation shows that an adult individual’s dose is not likely to exceed 

10 percent of any applicable limit, there are no recordkeeping or reporting requirements in 

regard to the individual’s exposure. For individuals who have received doses at other facilities 

in the current year, the previous dose need not be considered in this prospective evaluation. 

Only dose that could be received at the facility performing the evaluation need be considered 

when determining the need for monitoring, and therefore, recordkeeping and reporting 

requirements. If it was determined that monitoring was not required and a subsequent 

evaluation shows that the 10 percent threshold has or will be exceeded, the dose received when 

monitoring was not provided should be estimated, recorded, and reported. These estimates can 

be based on any combination of work location radiation monitoring or survey results, monitoring 

results of individuals in similar work situations, or other estimates to produce a “best estimate” of 

the actual dose received. 

Dose Limits for 

Adult Radiation Worker

(10 CFR 20.1201) 

Dose Limits for

Adult Radiation Worker

(10 CFR 20.1201) 
8-38 

Licensees should use NRC Form 4, “Cumulative Occupational Dose History,” and NRC Form 5, 

“Occupational Dose Record for a Monitoring Period,” to record individual dose. If monitoring is 

not required to demonstrate compliance with all limits but is required relative to one or more 

specific limits, the licensee should enter “N/A” for “not applicable” in the blocks on NRC Form 4, 

“Cumulative Occupational Dose History,” and NRC Form 5, “Occupational Dose Record for a 

Monitoring Period,” to indicate the areas for which monitoring was not required (e.g., extremity 

or skin doses). Where monitoring was provided but not measurable, the licensee should enter 

“ND” for “not detectable.” 

If the prospective evaluation shows that the individual adult is likely to exceed 10 percent of an 

applicable limit, then monitoring and reporting of the results of monitoring performed—

regardless of the actual dose received—is required. If air sampling or bioassay is required, 

discussion of air sampling or bioassay should provide enough detail so that the NRC staff is 

assured that appropriate steps will be taken to manage and monitor such exposure. Licensees 

must provide individual radiation exposure data to each worker as required by 10 CFR 19.13. 

Licensees should also perform prospective evaluations of the doses that may be received by 

occupationally exposed minors and declared pregnant women. As with individual adult workers, 

licensees must supply and require the use of individual monitoring devices to monitor external 

exposures and monitor the occupational intake of radioactive material when the results of 

prospective dose evaluations exceed the doses specified in 10 CFR 20.1502. 

When personnel dosimeters that require processing to determine the radiation dose are used to 

comply with the individual monitoring requirement for external doses in 10 CFR 20.1502(a), 

licensees must use dosimeters supplied by a NVLAP-approved processor. The exchange 

frequency for dosimeters is typically monthly or quarterly. Applicants should consult with their 

NVLAP-approved processor for its recommendations for exchange frequency and proper use of 

the dosimeter. 

For guidance about methodologies for determination of internal occupational dose and 

summation of occupational dose, refer to Table 8-4. 

Table 8-4. Guidance on Personnel Monitoring and Bioassay 

Regulatory Guide 8.7, Revision 2 Instructions for Recording and Reporting Occupational 

Radiation Exposure Data 

Regulatory Guide 8.9, Revision 1 Acceptable Concepts, Models, Equations, and 

Assumptions for a Bioassay Program 

Regulatory Guide 8.20, Revision 2 Applications of Bioassay for Radioiodine 

Regulatory Guide 8.21, Revision 1 Health Physics Surveys for Byproduct Material at 

NRC-Licensed Processing and Manufacturing Plants 

Regulatory Guide 8.23, Revision 1 Radiation Safety Surveys at Medical Institutions 

Regulatory Guide 8.25, Revision 1 Air Sampling in the Workplace 

Regulatory Guide 8.32 Criteria for Establishing a Tritium Bioassay Program 

Regulatory Guide 8.34 Monitoring Criteria and Methods to Calculate Occupational 

Radiation Doses 
8-39 

Table 8-4. Guidance on Personnel Monitoring and Bioassay 

Regulatory Guide 8.35, Revision 1 Planned Special Exposures 

Regulatory Guide 8.36 Radiation Dose to the Embryo/Fetus 

Regulatory Guide 8.37 ALARA Levels for Effluents from Materials Facilities 

ANSI N13.30-2011 Performance Criteria for Radiobioassay 

Information Notice 2000-10 Recent Events Resulting in Extremity Exposures 

Exceeding Regulatory Limits 

Additional Reference for Further Reading:

U.S. Department of Energy G 441.1-1C, “Radiation Protection Programs Guide,” May 19, 2008 

Response from Applicant: Provide one of the following statements:

“We will maintain, for inspection by the NRC, documentation demonstrating that unmonitored 

individuals are not likely to receive a radiation dose in excess of the limits in 10 CFR 20.1502.” 

OR 

“We will monitor individuals in accordance with the criteria in the section titled, ‘Radiation Safety 

Program–Occupational Dose’ in NUREG–1556, Vol. 11, Rev. 1, ‘Consolidated Guidance About 

Materials Licenses: Program-Specific Guidance About Licenses of Broad Scope.’”

OR, IN LIEU OF THESE STATEMENTS, 

Provide a description of an alternative method for demonstrating compliance with the referenced 

regulations. 

In addition, Type A broad scope licensees or applicants that want the flexibility to revise their 

personnel monitoring program without amendment of the license, as discussed in Chapter 1, 

“Purpose of Report,” and Section 8.7.2, “Radiation Safety Committee,” of this NUREG, should 

describe the process they will use to revise and implement their submitted personnel monitoring 

program. 

Reference: The NIST maintains a directory of laboratories that are NVLAP-approved at 

http://ts.nist.gov/standards/scopes/dosim.htm. 

8.10.5 Public Dose 

Regulations: 10 CFR 20.1003, 10 CFR 20.1301, 10 CFR 20.1302, 10 CFR 20.1801, 

10 CFR 20.1802, 10 CFR 20.2107

Criteria: Licensees must ensure that licensed material will be used, transported, stored, and 

disposed in such a way that the total effective dose equivalent (TEDE) to members of the public 

will not exceed more than 1 mSv [100 mrem] in a year, and the dose in any unrestricted area 

will not exceed 0.02 mSv [2 mrem] in any one hour.
8-40 

Discussion: Public dose is defined in 10 CFR 20.1003 as “the dose received by a member of 

the public from exposure to radiation or to radioactive material released by a licensee, or to any 

other source of radiation under the control of a licensee.” Public dose excludes occupational 

doses or doses received from background radiation, from any medical administration the 

individual has received, from exposure to individuals administered radioactive material and 

released in accordance with 10 CFR 35.75, “Release of individuals containing unsealed 

byproduct material or implants containing byproduct material,” from voluntary participation in 

medical research programs, and from the licensee’s disposal of radioactive material into 

sanitary sewerage in accordance with 10 CFR 20.2003, “Disposal by release into sanitary 

sewerage.” Whether the dose to an individual is an occupational dose or a public dose depends 

on the individual’s assigned duties. It does not depend on the area (restricted, controlled, or 

unrestricted) where the individual is in when he or she receives the dose. 

There are both external exposure components and internal exposure components of public 

dose. The licensee should review all possible internal and external exposure pathways and 

decide which are applicable to its operations. 

Licensees should design a monitoring program to ensure compliance with 10 CFR 20.1301. 

Specific requirements on demonstrating compliance with dose limits for individual members of 

the public are provided in 10 CFR 20.1302. The extent and frequency of monitoring will depend 

on the nature of the licensee’s operations, potential release pathways, and potential exposures 

that can contribute to public dose or environmental releases. For additional guidance on 

monitoring of effluents, refer to the section titled “Radiation Safety Program – Surveys.” 

Regulations in 10 CFR 20.2107, “Records of dose to individual members of the public,” require 

that licensees maintain survey and monitoring records that demonstrate compliance with the 

dose limits for members of the public until the Commission terminates the license. 

Response from Applicant: No response is required from the applicant, but records and written 

materials documenting compliance will be examined during inspection. During NRC 

inspections, licensees must be able to demonstrate, by measurement or calculation, that the 

total effective dose equivalent to the individual likely to receive the highest dose from the 

licensed operation does not exceed the annual limit for members of the public. 

For guidance about accepted methodologies for determining doses to members of the public, 

see Appendix J of this NUREG, and, for examples of such methodologies as applied only to 

sealed sources in fixed gauges, see Appendix G of NUREG–1556, Volume 4, Revision 1, 

“Consolidated Guidance About Materials Licenses: Program-Specific Guidance About Fixed 

Gauge Licenses.” 

8.10.6 Safe Use of Radionuclides and Emergency Procedures 

Regulations: 10 CFR 19.11(a)(3), 10 CFR 20.1101, 10 CFR 20.1406; 10 CFR 20.1801, 

10 CFR 20.1802, 10 CFR 20.2201-2203, 10 CFR 21.21, 10 CFR 30.32(i), 10 CFR 30.34(e), 

10 CFR 30.50, 10 CFR 30.72, 10 CFR 33.13, 10 CFR 33.14, 10 CFR 33.15, 10 CFR 37 

(Subpart B), 10 CFR 37.21(a), 10 CFR 37.45, 10 CFR 37.49 

Criteria: Licensees must, pursuant to the regulations stated above, keep radiation doses to 

workers and members of the public ALARA; ensure security of licensed material; and make 

required notifications to the NRC of events. Licensees that possess an aggregated Category 1 

or Category 2 quantity of radioactive material, listed in Appendix A to 10 CFR Part 37, must also 
8-41 

establish, implement, and maintain its access authorization program; coordinate, to the extent 

practicable, with local law enforcement authorities for responding to threats to the licensee’s 

facility; and be able to monitor, detect without delay, assess, and respond to any unauthorized 

entries into security zones. 

Discussion: Licensees are responsible for developing and implementing procedures to ensure 

the security and safe use of all licensed material from the time it arrives at their facilities until it is 

used, transferred, and/or disposed. The written procedures should provide reasonable 

assurance that only appropriately trained personnel will handle and use licensed material 

without undue hazard to themselves, other workers, or members of the public. 

All licensed materials stored in controlled or unrestricted areas must be secured from 

unauthorized access or removal so that individuals who may not be knowledgeable about 

radioactive materials cannot be exposed to or contaminated by the material, and so that 

individuals cannot take the material. When any licensed materials are in use in controlled or 

unrestricted areas, they must be under constant surveillance so that the radiation worker can 

prevent others from becoming contaminated by or exposed to the material and prevent 

unauthorized persons from removing the material from the area. 

Acceptable methods for securing material will vary from one facility to another. Some 

alternatives used by licensees include: 

 storing and using licensed materials only in restricted areas 

 limiting access to an entire facility or building or portion of the building to radiation 

workers 

 providing storage areas that can be locked to prevent access to the material 

 implementing procedures that require a radiation worker to be within “line of sight” of the 

materials whenever licensed materials are in use 

Licensees should develop procedures that clearly state acceptable methods to secure licensed 

material at a facility. Particular attention may be required at facilities that have unusual needs 

because of the activities performed, such as hot cells, animal care facilities, and waste 

processing facilities. Security procedures may be in a separate document or included in the 

“General Safety Procedures.” 

Applicants should develop radionuclide-specific procedures based on the respective hazards 

associated with the radionuclides. Appendix K of this NUREG describes general safety 

guidelines. Licensees are encouraged to use these guidelines in developing procedures for the 

safe use of radionuclides. 

Licensees must identify all areas that require posting in accordance with 10 CFR 20.1902, 

“Posting requirements,” unless they meet the exemptions listed in 10 CFR 20.1903, 

“Exemptions to posting requirements.” Also, containers of licensed material (including 

radioactive waste) must be labeled in accordance with 10 CFR 20.1904, “Labeling containers,” 

unless they meet the exemptions in 10 CFR 20.1905, “Exemptions to labeling requirements.” 

Applicants need to establish written procedures to handle emergencies ranging from a minor 

spill to a major accident that may require intervention by outside emergency response 
8-42 

personnel. These procedures should include provisions for immediate response, after-hours 

notification, handling of each type of emergency, equipment, and the appropriate roles of users 

and the radiation safety staff. Except for minor spills or releases of radioactivity that can be 

controlled and cleaned up by the user, individual users should have a clear understanding of 

their limitations in an emergency, with step-by-step instructions and clear direction for whom to 

contact. Appendix K of this NUREG describes model emergency procedures that the NRC 

considers acceptable. 

Emergency spill kits should be strategically placed in well-marked locations for use by all users 

and the radiation safety staff. All equipment should be periodically inspected for proper 

operation and replenished, as necessary. The licensee should also consider establishing an 

emergency response team composed of individuals experienced in various emergency 

response functions (e.g., radiological, medical, emergency management, security, and fire 

protection). 

Regulations in 10 CFR 20.2201–20.2203, 10 CFR 21.21, “Notification of failure to comply or 

existence of a defect and its evaluation,” and 10 CFR 30.50, “Reporting requirements,” require 

certain incidents and emergencies be reported to the NRC. Appendix G of this NUREG 

provides examples of some events that require notification and/or reports. Note that 

Appendix G is not all inclusive, as there are other notification and/or reporting requirements that 

may apply to specific programs (e.g., 10 CFR Parts 32, 34, 35, 36, and 39). 

If licensees submit a license application requesting to possess quantities of material in excess 

of the applicable amounts listed in 10 CFR 30.72, “Schedule C,” then they may also be required 

to submit an “emergency plan for responding to a release of radioactive material.” See Section 

8.5.1, “Unsealed or Sealed Byproduct Material,” of this document for specific information related 

to this requirement. 

In accordance with 10 CFR Part 37, any licensee that possesses an aggregated Category 1 or 

Category 2 quantity of radioactive material must, among other things, 

 in accordance with 10 CFR 37.21(a), establish, implement, and maintain its access 

authorization program in accordance with the requirements of 10 CFR Part 37, 

Subpart B; 

 in accordance with 10 CFR 37.45, coordinate with their local law enforcement agency 

(LLEA) for responding to threats to a licensee’s facility; and 

 in accordance with 10 CFR 37.49, be able to monitor, detect without delay, assess, and 

respond to any unauthorized entries into security zones, including those surrounding 

mobile devices. 

For additional guidance on implementing 10 CFR Part 37 requirements, see NUREG–2155, 

“Implementation Guidance for 10 CFR Part 37, “Physical Protection of Category 1 and 

Category 2 Quantities of Radioactive Material.”” Additional information regarding best 

practices for protection of risk-significant radioactive material is available in NUREG–2166, 

“Physical Security Best Practices for the Protection of Risk-Significant Radioactive Material.” 

Please note, under 10 CFR Part 37, security plans are not submitted to the NRC, but may 

be subject to review and inspection.
8-43 

Response from Applicant: Submit procedures for safe use of radionuclides and emergencies. 

Submissions should include procedures for maintaining security of licensed radioactive 

materials. As an alternative, state, “We will adopt the procedures for the safe use of 

radionuclides and emergencies as published in Appendix K of NUREG–1556, Volume 11, 

Revision 1, ‘Consolidated Guidance About Materials Licenses: Program-Specific Guidance 

About Licenses of Broad Scope.’” 

In addition, Type A broad scope licensees or those applying for a Type A broad scope license 

that want the flexibility to revise their safe use and emergency procedures without amendment 

of the license, as described in Chapter 1, “Purpose of Report,” and Section 8.7.2, “Radiation 

Safety Committee,” of this NUREG, should discuss the process that will be used to revise and 

implement their submitted safe use and emergency procedures. 

8.10.7 Surveys and Leak Tests 

Regulations: 10 CFR 20.1501, 10 CFR 20.1906; 10 CFR 20.2103, 10 CFR 30.53, 

10 CFR 33.13, 10 CFR 33.14, 10 CFR 33.15, 10 CFR 35.67(d), 10 CFR 35.2067(a), 

10 CFR 36.81(h), 10 CFR 39.35(a)

Criteria: Licensees are required, pursuant to the regulations listed above, to make surveys of 

potential radiological hazards in their workplace. The NRC requires testing to determine 

whether there is any radioactive leakage from sealed sources. Licensees must retain records of 

surveys and leak test results in accordance with license conditions and NRC regulations.

Discussion: Survey is defined as an evaluation of the radiological conditions and potential 

hazards incident to the production, use, transfer, release, disposal, or presence of radioactive 

material or other sources of radiation. These evaluations may be measurements (e.g., radiation 

levels measured with a survey instrument or results of wipe tests for contamination), 

calculations, or a combination of measurements and calculations. The selection and proper use 

of appropriate instruments is one of the most important factors in ensuring that surveys 

accurately assess the radiological conditions. Licensees should also use surveys to plan work 

in areas where licensed material or radiation exists and to evaluate doses to workers and 

individual members of the public. In certain cases, environmental monitoring may be required to 

demonstrate compliance with 10 CFR Part 20. 

Surveys are required when it is necessary for the licensee to comply with the regulations or to 

evaluate a radiological hazard. Many different types of surveys may need to be performed 

because of the particular use of licensed materials. Typical surveys may include: 

 measurements of radioactive contamination that could be present on surfaces of floors, 

walls, laboratory furniture, and equipment 

 measurements of radioactive material concentrations in air for areas where radioactive 

materials are handled or processed in unsealed form and where operations could 

expose workers to the inhalation of radioactive material or where licensed material is or 

could be released to unrestricted areas 

 measurements of radioactive material concentrations in water that are released to the 

environment or to the sanitary sewer 
8-44 

 bioassays to determine the kinds, quantities, or concentration—and in some cases, the 

location of—radioactive material in the human body (A bioassay can be made by direct 

measurement, in vivo counting, or by analysis and evaluation of material excreted or 

removed from the human body.) 

 measurements of external radiation exposure levels in both restricted and unrestricted 

areas 

The frequency of routine surveys depends on the nature, quantity, and use of radioactive 

materials, as well as the specific protective facilities, equipment, and procedures that are 

designed to protect the worker from external and internal exposure. Also, the frequency of the 

survey depends on the type of survey, such as those listed above. 

Appendix L of this NUREG describes survey procedures that are acceptable to the NRC. 

The regulations in 10 CFR Part 20 do not specify limits for surface contamination. Each 

applicant should propose and justify which removable surface contamination limits will be 

allowable before decontamination will be performed in each work area. However, Subpart E of 

10 CFR Part 20 specifies dose limits for residual contamination at the time of license 

termination. These criteria should also be used at the time of decommissioning of facilities that 

are to be released for unrestricted use, even as the license continues in effect at other locations. 

Guidance for survey criteria to meet the license termination criteria may be found in 

NUREG–1757, “Consolidated Decommissioning Guidance.” 

Also, NUREG–1575, “Multi-Agency Radiation Survey and Site Investigation Manual 

(MARSSIM),” dated August 2000, contains additional guidance concerning surveys during the 

decommissioning of facilities. Licensees that have facilities to decommission should review this 

document. 

Leak Test 

When issued, a license will require performance of leak tests of sealed or plated foil sources at 

intervals as approved by the NRC or an Agreement State and specified by the SSD registration 

certificate. The measurement of the leak-test sample is a quantitative analysis requiring that 

instrumentation used to analyze the sample be capable of detecting 185 Bq [0.005 μCi 

(microcurie)] of radioactivity. 

Because the types, forms, and quantities of licensed materials in sealed sources can vary 

significantly for applicants, leak test requirements usually are specified in a license condition. 

Typically, leak tests are not required if: 

 sources contain only hydrogen-3 (tritium) 

 sources contain only byproduct material with a half-life of less than 30 days 

 sources contain only a radioactive gas 

 sources contain 3.7 megabecquerels [100 μCi] or less of beta-emitting or gammaemitting material or 370 kilobecquerels (kBq) [10 μCi] or less of alpha-emitting material 
8-45 

 sources are stored and are not being used (but must be leak tested before use or 

transfer, or if stored more than 10 years) 

For more information regarding leak tests, see Appendix M of this NUREG. 

Response from Applicant:

 Surveys 

 Submit procedures to evaluate radiological hazards, both external and internal. As an 

alternative, the applicant may state, “We will survey our facility and maintain 

contamination levels and perform bioassays of occupationally exposed workers in 

accordance with the survey frequencies and contamination levels published in Appendix 

L of NUREG–1556, Volume 11, Revision 1, ” Consolidated Guidance About Materials 

Licenses: Program-Specific Guidance About Licenses of Broad Scope.”” 

 Leak Testing 

 Submit leak test procedures. As an alternative, the applicant may state, “We will 

implement the model leak test program published in Appendix M of NUREG–1556, 

Volume 11, Revision 1, ” Consolidated Guidance About Materials Licenses: ProgramSpecific Guidance About Licenses of Broad Scope.”” 

In addition, Type A broad scope licensees or those applying for a Type A broad scope license 

that want the flexibility to revise their survey or leak test program without amendment of the 

license, as described in Chapter 1, “Purpose of Report,” and Section 8.7.2, “Radiation Safety 

Committee,” of this NUREG, should discuss the process that will be used to revise and 

implement their submitted survey and leak test program. 

8.10.8 Transportation 

Regulations: 10 CFR 20.1101, 10 CFR Part 20, Appendix G, 10 CFR 30.41, 10 CFR 30.51, 

10 CFR 33.13, 10 CFR 33.14, 10 CFR 33.15, 10 CFR 34.35, 10 CFR Part 37 (Subpart D), 

10 CFR 71.5, 10 CFR 71.12, 10 CFR 71.13, 10 CFR 71.14, 10 CFR 71.47, 10 CFR 71.87, 

49 CFR Parts 171-185 

Criteria: Broad scope licensees that will transport or ship licensed material, including 

radioactive waste, must develop, implement, and maintain safety programs for the transport of 

radioactive material to ensure compliance with NRC and U.S. Department of Transportation 

(DOT) regulations. In accordance with 10 CFR Part 37 (Subpart D), licensees must also 

preplan, coordinate and provide advance notification of the shipment of Category 1 quantities of 

radioactive material and coordinate the shipment of Category 2 quantities of radioactive 

material. 

Discussion: DOT regulations [Title 49 of the Code of Federal Regulations (49 CFR), 

“Transportation”] were written to help ensure that hazardous materials in commerce were 

transported uniformly and safely. NRC licensees that transport byproduct material (hazardous 

material) in commerce are, therefore, required to comply with all applicable DOT regulations. 

However, many NRC licensees routinely transport byproduct material that is not in commerce. 

In 1979, 10 CFR 71.5, “Transportation of Licensed Material,” was codified to ensure that all 

NRC licensees that transport byproduct material, in commerce or not, transport it in a uniformly 
8-46 

safe manner. Some broad scope licensees have applied for certain exemptions to 

10 CFR 71.5; however, the NRC did not grant these exemption requests. The NRC’s position 

regarding the transportation regulations in 49 CFR is that these regulations already allow for 

flexibility and are thus considered to be performance-based requirements rather than 

prescriptive requirements. For example, packages shipped by broad scope licensees frequently 

meet the “limited quantity” criteria as described in 49 CFR 173.421, “Excepted Packages for 

Limited Quantities of Class 7 (Radioactive) Materials,” and are therefore excepted from certain 

DOT requirements, provided that certain other less-restrictive requirements are met. 

Appendix N of this NUREG provides an overview of the transportation requirements that 

commonly affect NRC licensees. Broad scope licensees’ radiation safety staff should be 

thoroughly familiar with 10 CFR 71.5 and 49 CFR and how they interrelate in order to be able to 

comply with and to take full advantage of the flexibility inherent in DOT requirements. 

Licensed material, including radioactive waste, must be packaged and transported in 

accordance with NRC and DOT requirements if the transportation involves the use of public 

highways. In addition, broad scope licensees should develop and maintain their own radiation 

safety procedures for transporting licensed material within their facilities if such transportation 

does not involve the use of public highways. 

Licensees also should consider the safety of all individuals who may handle or may come in 

contact with the packages containing licensed material. Therefore, the licensee’s primary 

concern in transporting licensed material should be to ensure that the package is properly 

prepared prior to transport, that package integrity is not compromised during transport and that 

the radiation levels (including removable contamination levels) not only meet the regulatory 

requirements of 10 CFR 71.47, “External radiation standards for all packages,” but also are 

ALARA. 

Licensees shipping radioactive waste for disposal must prepare appropriate documentation as 

specified in 10 CFR Part 20, Appendix G, “Requirements for transfers of low-level radioactive 

waste intended for disposal at licensed land disposal facilities and manifests.” 

Licensees shipping or transferring a Category 1 or Category 2 quantity of radioactive material 

are subject to the requirements in 10 CFR Part 37, Subpart D (“Physical Protection in Transit”). 

For additional guidance on implementing 10 CFR Part 37 requirements, see NUREG–2155, 

“Implementation Guidance for 10 CFR Part 37, “Physical Protection of Category 1 and Category 

2 Quantities of Radioactive Material.”” Additional information regarding best practices for 

protection of risk-significant radioactive material is available in NUREG–2166, “Physical Security 

Best Practices for the Protection of Risk-Significant Radioactive Material.” 

Please note, under 10 CFR Part 37, security plans are not submitted to the NRC, but may 

be subject to review and inspection. 

Response from Applicant: No response is needed from applicants during the licensing phase. 

Compliance with transportation requirements will be reviewed during NRC inspections. 

Reference: “Radioactive Material Regulations Review (December 2008)” is available from the 

DOT at 

www.phmsa.dot.gov/staticfiles/PHMSA/DownloadableFiles/Files/RAM_Regulations_Review_12-

2008.pdf. The Memorandum of Understanding with DOT on the Transportation of Radioactive 

Material (signed June 8, 1979) is available from the NRC on its public Web site under 

“Enforcement Guidance” at http://www.nrc.gov/about-nrc/regulatory/enforcement/moudot.pdf.
8-47 

8.10.9 Security Program for Category 1 and Category 2 Radioactive Material 

Regulations: 10 CFR Part 37 

Criteria: Licensees must ensure the security of Category 1 and Category 2 radioactive 

material. 

Note: The regulations in 10 CFR Part 37 apply to licensees that possess an aggregated 

Category 1 or Category 2 quantity of radioactive material. The specific radionuclides subject to 

10 CFR Part 37 requirements are listed in Table 1 of Appendix A to 10 CFR Part 37. 

Discussion: 

Requirements in 10 CFR Part 37, “Physical Protection of Category 1 and Category 2 Quantities 

of Radioactive Material” 

In accordance with 10 CFR Part 37, licensees that possess aggregated Category 1 or 

Category 2 quantities of radioactive material must establish, implement, and maintain an access 

authorization program (Subpart B) and a security program (Subpart C) to ensure physical 

protection of the radioactive material. 

Table 1 of Appendix A, “Category 1 and Category 2 Radioactive Materials,” to 10 CFR Part 37, 

lists Category 1 and Category 2 threshold quantities of radioactive material. The applicant 

should refer to this table to determine whether its proposed activities would be subject to the 

10 CFR Part 37 requirements. 

Before giving individuals unescorted access to Category 1 or Category 2 quantities of 

radioactive material (as defined in 10 CFR 37.5), licensees must conduct background 

investigations of these individuals, to determine that they are trustworthy and reliable, in 

accordance with 10 CFR 37.25. 

In accordance with 10 CFR 37.41(b), licensees must establish a security program designed to 

monitor and, without delay, detect, assess, and respond to any actual or attempted 

unauthorized access to Category 1 or Category 2 quantities of radioactive material. 

Per 10 CFR Part 37, Subpart D, licensees must provide for physical protection of Category 1 or 

Category 2 quantities of radioactive materials in transit. These requirements apply to licensees 

delivering such material to a carrier for transport, as well as cases in which licensees are 

transporting such material. Please note that the Subpart D requirements applicable to the 

transport of Category 1 quantities of radioactive material are more stringent than those 

applicable to Category 2 quantities. 
8-48 

Applicants and licensees are required to implement the 10 CFR Part 37 security requirements 

before they take possession of an aggregated Category 1 or Category 2 quantity of radioactive 

material. 

Any licensee that has not previously been made subject to the provisions of 10 CFR Part 37, 

Subpart C must notify the NRC regional office specified in 10 CFR 30.6 in writing at least 90 

days before aggregating radioactive material to a quantity that equals or exceeds the Category 

2 threshold. Pursuant to 10 CFR 37.43(b), as part of the security program, the licensee must 

develop and maintain written procedures that document how the requirements of Subpart C will 

be met. These written procedures may be subject to NRC review and inspection. 

For additional guidance on implementing 10 CFR Part 37 requirements, see NUREG–2155, 

“Implementation Guidance for 10 CFR Part 37, “Physical Protection of Category 1 and Category 

2 Quantities of Radioactive Material.”” Additional information regarding best practices for 

protection of risk-significant radioactive material is available in NUREG–2166, “Physical Security 

Best Practices for the Protection of Risk-Significant Radioactive Material.” 

Response from Applicant:

No response is required from an applicant or licensee. Compliance with access authorization 

and security program requirements may be reviewed during NRC inspections.

8.11 Item 11: Waste Management 

Regulations: 10 CFR 20.1501, 10 CFR 20.2001, 10 CFR 20.2002, 10 CFR 20.2003, 

10 CFR 20.2004, 10 CFR 20.2005, 10 CFR 20.2006, 10 CFR 20.2007, 10 CFR 20.2108, 

10 CFR Part 20, Appendix B, 10 CFR 30.51, 10 CFR 37.11(c)

Criteria: Radioactive waste must be disposed of in accordance with regulatory requirements 

and license conditions. Appropriate records of waste disposal must be maintained. 

Discussion: The applicant should discuss the methods for management and disposal of 

radioactive waste. The program should include procedures for the handling of waste, safe and 

secure storage, waste characterization, waste minimization, and disposal of radioactive waste. 

Appropriate training should be provided to waste handlers. Licensees may not receive 

radioactive waste from other licensees for processing, storage, or disposal unless specifically 

authorized by the NRC. 

The U.S. Environmental Protection Agency issued guidance for developing a comprehensive 

program to reduce hazardous waste. The NRC transmitted this guidance to licensees in 

IN-94-23, “Guidance to Hazardous, Radioactive, and Mixed Waste Generators on the Elements 

of a Waste Minimization Program,” dated March 1994. NRC Regulatory Issue Summary 2011-

09, “Available Resources Associated with Extended Storage of Low-Level Radioactive Waste,” 

also contains useful information. Applications should include, where appropriate for the types of 

waste involved, provisions for monitoring and segregating waste materials (radioactive from 

non-radioactive, short from long half-life, liquid from solid waste, etc.). 

 
8-49 

Waste Volume Reduction 

Licensees may implement procedures to reduce the volume of radioactive waste for final 

disposal in an authorized low-level radioactive waste (LLW) disposal facility. These procedures 

include volume reduction by segregating, consolidating, compacting, or allowing certain waste 

to decay-in-storage. Waste compaction or other treatments can reduce the volume of 

radioactive waste, but such processes may pose additional radiological hazards (e.g., airborne 

radioactivity or increased radiation levels) to workers, members of the public, and the 

environment. Safety procedures to address these concerns should be implemented. Applicants 

must describe in detail waste volume reduction operations that could create a radiological 

hazard to licensee employees or the general public. Appendix O of this NUREG describes a 

model procedure for waste compaction. 

The following methods of waste disposal may be considered and should be addressed in the 

application, as appropriate. 

Transfer to an Authorized Recipient 

Waste may be transferred to a recipient (usually a waste disposal service company or the 

original supplier) that is properly licensed to receive such waste in accordance with 

10 CFR 20.2001(a). Each shipment must comply with all applicable NRC and DOT 

requirements. 

Decay-in-Storage 

The NRC has concluded that materials with half-lives of less than or equal to 120 days are 

appropriate for decay-in-storage (DIS). The holding time of the waste should be based on the 

radionuclide(s), half-life, and the activity present when the waste was placed into storage. Such 

waste may be disposed of as ordinary trash if radiation surveys of the waste indicate that 

radiation levels are indistinguishable from background. The surveys should be performed with 

an appropriate radiation detection meter set on its most sensitive scale in a low background 

area and without any interposed shielding. In accordance with 10 CFR 20.1904(b), all radiation 

labels must be defaced or removed from containers and packages prior to disposal as ordinary 

trash, except for radiation labels on materials that are within containers and that will be 

managed as biomedical waste after they have been released. If the decayed waste is 

compacted, all labels that are visible in the compacted mass must also be defaced or removed. 

Applicants must maintain accurate records of such disposals. 

Appendix O of this NUREG provides a model procedure for DIS. 

Extended Interim Storage 

The NRC does not consider storage as a substitute for final disposal of radioactive wastes. 

Other than storage for radioactive decay, LLW should be stored only when disposal capacity is 

unavailable, and for no longer than is necessary. NRC IN 90-09, “Extended Interim Storage of 

Low-Level Radioactive Waste for Fuel Cycle and Material Licensees,” dated February 5, 1990, 

provides guidance to licensees for requesting an amendment to authorize extended interim 

storage of LLW. Regulatory Issue Summary 2008-12, “Considerations for Extended Interim 

Storage of Low-Level Radioactive Waste by Fuel Cycle and Materials Licensees,” dated May 9, 

2008, updates information provided in IN 90-09. In addition, the NRC issued Regulatory Issue 
8-50 

Summary 2011-09, “Available Resources Associated with Extended Storage of Low-Level 

Radioactive Waste,” dated August 16, 2011, which refers to other helpful guidance documents. 

Release into Air and Water 

Release of radioactive material into air and water must conform to the requirements described 

in 10 CFR 20.1302(b)(2). The applicant should discuss the monitoring and control mechanisms 

in place to ensure compliance with the requirements. Applicants are reminded of the 

“constraint” on air emissions of radioactive material required by 10 CFR 20.1101(d), which 

effectively reduces the limits specified in 10 CFR 20.1302(b)(2) for release of gaseous effluents. 

Applicants that are considering release of radioactive material into air and water should review 

Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials Facilities,” dated July 1993. 

Regulatory Guide 8.37 deals with the application of ALARA in controlling gaseous and liquid 

effluents and references documents containing acceptable methods of effluent monitoring. 

Regulatory Guide 4.20, Rev. 1, “Constraint on Releases of Airborne Radioactive Materials to the 

Environment for Licensees Other than Power Reactors,” dated April 2012, also contains useful 

information. 

Licensees considering disposal by release to the sanitary sewerage system must comply with 

the requirements of 10 CFR 20.2003. The regulations in 10 CFR 20.2003 authorize disposal of 

licensed material by release into a public sanitary sewerage system if certain conditions are 

met. Licensees are responsible for demonstrating that licensed materials discharged into the 

public sewerage system are readily soluble in water or are biological materials that are readily 

dispersible in water. NRC IN 94-07, “Solubility Criteria for Liquid Effluent Releases to Sanitary 

Sewerage under the Revised 10 CFR Part 20,” dated January 1994, provides the criteria for 

evaluating solubility of waste. Licensees should carefully consider the possibility of 

reconcentration of radionuclides that are released into the sewer. The NRC alerted licensees to 

the potentially significant problem of reconcentration of radionuclides released to sanitary 

sewage systems in IN 84-94, “Reconcentration of Radionuclides Involving Discharges into 

Sanitary Sewage Systems Permitted under 10 CFR 20.303 [now 10 CFR 20.2003],” dated 

December 1984. 

Applicants should provide procedures that will ensure that all releases of radioactive waste into 

the sanitary sewerage meet the criteria stated in 10 CFR 20.2003 and do not exceed the 

monthly and annual limits specified in regulations. Licensees are required to maintain accurate 

records of all releases of licensed material into the sanitary sewerage. Appendix O of this 

NUREG provides a model procedure for disposal of radioactive waste via sanitary sewer and 

maintenance of records. 

If the applicant’s or licensee’s facility maintains a private sewerage treatment system, a septic 

system, or leach fields, the regulations of 10 CFR 20.2003 are not applicable for releases to 

these systems (see 10 CFR 20.1003, “Definitions,” for the definition of “sanitary sewerage”). 

Applicants may make releases of liquids to private sewerage systems, septic systems, or leach 

fields as effluents released to unrestricted areas pursuant to 10 CFR 20.1302(b)(2)(i). 

If liquid releases are made to a private sewerage treatment system, septic system, or leach 

field, the sludges or other solids from these systems may become contaminated with radioactive 

material. Applicants should describe the monitoring planned for these systems in Item 8.10.7, 

“Surveys and Leak Tests,” of the application. Contaminated sludges will be required to be 

disposed of as radioactive waste using one of the methods described in this section. Applicants 
8-51 

may obtain approval of alternative disposal methods through application to the Commission, as 

described in 10 CFR 20.2002, “Method for obtaining approval of proposed disposal procedures.” 

Incineration 

Applicants that wish to treat or dispose of licensed material by incineration must comply with the 

requirements of 10 CFR 20.2004, “Treatment or disposal by incineration.” Applicants proposing 

incineration should be aware that a notice in the Federal Register may be required before 

disposal of ash as ordinary waste can be approved. However, approval of incineration pursuant 

to 10 CFR 20.2004 does not require notice in the Federal Register if the ash is disposed as 

radioactive waste or transferred to a specific licensee. Policy and Guidance Directive PG 8-10, 

“Disposal of Incineration Ash as Ordinary Waste,” dated January 1997, provides guidance on 

the disposal of ash. Appendix O of this NUREG offers a model procedure for incineration of 

waste. 

Applicants that are considering disposal of radioactive material by incineration should review 

Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials Facilities,” dated July 1993. 

Regulatory Guide 8.37 deals with the application of ALARA in controlling gaseous and liquid 

effluents and references documents containing acceptable methods of effluent monitoring. 

Disposal of Specific Waste as if it Were Not Radioactive 

The following radioactive wastes may be disposed of as non-radioactive waste pursuant to 

10 CFR 20.2005, “Disposal of specific wastes”: 

 liquid scintillation medium containing no more than 1.85 kBq [0.05 μCi] of H-3 or C-14 

per gram of the medium 

 animal carcasses or animal tissue containing no more than 1.85 kBq [0.05 μCi] of H-3 or 

C-14 per gram averaged over the weight of the entire animal 

Applicants should have procedures to ensure that the above limits are not exceeded and that 

the disposal of animal tissue or carcasses containing licensed material is in a manner that will 

not permit their use either as food for humans or as animal feed. Applicants must maintain 

accurate records of these disposals. 

Burial 

Licensees previously authorized to bury radioactive materials pursuant to 10 CFR 20.304 before 

January 28, 1981, should describe the locations, condition, and current status of these former 

sites (i.e., controlled or uncontrolled, active monitoring of the site, and current condition of the 

burial site). If the licensee plans to decommission a burial site prior to termination of the license, 

the licensee may be required to notify the NRC and develop a decommissioning plan in 

accordance with 10 CFR 30.36. Additional guidance is provided in NUREG–1757, 

“Consolidated Decommissioning Guidance.” 

Other Methods Specifically Approved by the NRC Pursuant to 10 CFR 20.2002 

Applicants may also request alternate methods for the disposal of radioactive waste generated 

at their facilities. Such requests must describe the waste containing licensed material, including 

the physical and chemical properties that may be important to assess risks associated with the 
8-52 

waste, and the proposed manner and conditions of waste disposal. Additionally, the applicant 

must submit its analysis and evaluation of pertinent information on the nature of the 

environment, nature and location of other affected facilities, and procedures to ensure that 

radiation doses are maintained ALARA and within regulatory limits. 

The application should describe the ALARA considerations taken before disposal of radioactive 

materials. The request should discuss the potential for unmonitored or unanticipated release of 

radioactive materials to work areas and from release points (i.e., hoods and incinerator stacks). 

To be in compliance with the ALARA philosophy stated in 10 CFR 20.1101, “Radiation 

protection programs,” radioactive material waste stream concentrations should be a fraction 

(generally 10 to 20 percent) of the limits specified in Appendix B, “Annual Limits on Intake (ALIs) 

and Derived Air Concentrations of Radionuclides for Occupational Exposure; Effluent 

Concentrations; Concentrations for Release to Sewerage,” Table II, 10 CFR Part 20. 

Furthermore, because of the variability of inventory control programs for monitoring disposal 

and releases of byproduct material in use, a program for physically measuring releases should 

be in place whenever releases exceed the specified ALARA action point. 

Note: Guidance concerning alternate waste disposal procedures that involve multiple 

jurisdictions has been provided in in the State and Tribal Communication Letter FSME 12-025, 

dated March 13, 2012, “Clarification of the Authorization for Alternative Disposal of Material 

Issued Under 10 CFR 20.2002 and Exemption Provisions in 10 CFR.” (ADAMS Accession No. 

ML12065A038) 

Note: Before licensed activities are transferred or assigned in accordance with 

10 CFR 30.34(b), 10 CFR 40.46, and/or 10 CFR 70.36, if licensees are authorized to possess 

byproduct material with a half-life greater than 120 days in an unsealed form, source material in 

an unsealed form, and/or special nuclear material, the licensees must, in accordance with 

10 CFR 30.51(e), 10 CFR 40.61(e), and/or 10 CFR 70.51(b)(1)&(2), respectively, transfer the 

following records to the new licensee: 

 records of disposal of licensed material made under: 

— 10 CFR 20.2002, “Method for obtaining approval of proposed disposal 

procedures“ 

— 10 CFR 20.2003, “Disposal by release into sanitary sewerage” 

— 10 CFR 20.2004, “Treatment or disposal by incineration 

— 10 CFR 20.2005, “Disposal of specific wastes” 

 records required by 20.2103(b)(4) of the results of measurements and calculations used 

to evaluate the release of radioactive effluents to the environment 

Because of the difficulties and costs associated with disposal of sealed sources, applicants 

should preplan the disposal. Applicants may want to consider contractual arrangements with 

the source supplier as part of a purchase agreement. 
8-53 

In accordance with 10 CFR 37.11(c), a licensee that possesses radioactive waste that contains 

Category 1 or Category 2 quantities of radioactive material as defined in 10 CFR 37.5 is 

exempt from the requirements of 10 CFR Part 37, Subparts B, C, and D. However, any 

radioactive waste that contains discrete sources, ion-exchange resins, or activated material 

that weighs less than 2,000 kg [4,409 lbs] is not exempt from the requirements of 

10 CFR Part 37. For additional guidance on implementing 10 CFR Part 37 requirements, see 

NUREG–2155, “Implementation Guidance for 10 CFR Part 37, “Physical Protection of 

Category 1 and Category 2 Quantities of Radioactive Material.”” Additional information 

regarding best practices for protection of risk-significant radioactive material is available in 

NUREG–2166, “Physical Security Best Practices for the Protection of Risk-Significant 

Radioactive Material.” 

A licensee possessing radioactive waste that is exempt under 10 CFR 37.11(c) from the 

requirements of 10 CFR Part 37, Subparts B, C, and D must implement the following 

requirements to secure the radioactive waste: 

 use continuous physical barriers that allow access to the radioactive waste only through 

established access control points; 

 use a locked door or gate with monitored alarm at the access control point; 

 assess and respond to each actual or attempted unauthorized access to determine 

whether an actual or attempted theft, sabotage, or diversion occurred; and 

 immediately notify the LLEA and request an armed response from the LLEA upon 

determination that there was an actual or attempted theft, sabotage, or diversion of the 

radioactive waste that contains Category 1 or Category 2 quantities of radioactive 

material. 

Please note, under 10 CFR Part 37, security plans are not submitted to the NRC, but may 

be subject to review and inspection.

Response from Applicant: 

Provide procedures for waste collection, storage, and the disposal by any of the authorized 

methods described in this section. Applicants should contact the appropriate regional office of 

the NRC for guidance to obtain approval of any method(s) of waste disposal other than those 

discussed in this section. 

Note: Applicants do not need to provide information to the NRC if they plan to dispose of LLW 

through transfer to an authorized recipient or to dispose of liquid scintillation media or animals 

containing low levels of H-3 or C-14, as authorized by 10 CFR 20.2005. 

References: 

 Information Notice 94-23, “Guidance to Hazardous, Radioactive, and Mixed Waste 

Generators on the Elements of a Waste Minimization Program,” dated March 1994 

 Information Notice 94-07, “Solubility Criteria for Liquid Effluent Releases to Sanitary 

Sewerage Under the Revised 10 CFR Part 20,” dated January 1994 
8-54 

 Information Notice 84-94, “Reconcentration of Radionuclides Involving Discharges into 

Sanitary Sewage Systems Permitted Under 10 CFR 20.203 (now 10 CFR 20.2003),” 

dated December 1984 

 Information Notice 90-09, “Extended Interim Storage of Low-Level Radioactive Waste by 

Fuel Cycle and Materials Licensees,” dated February 1990 

 Regulatory Issue Summary 2008-12, “Considerations For Extended Interim Storage Of 

Low-Level Radioactive Waste By Fuel Cycle And Materials Licensees” 

 Regulatory Issue Summary 2011-09, “Available Resources Associated With Extended 

Storage Of Low-Level Radioactive Waste” 

 Policy and Guidance Directive PG 8-10, “Disposal of Incineration Ash as Ordinary 

Waste,” January 1997, ADAMS Accession Nos. ML003744979 and ML003752866 and 

Addendum, ADAMS Accession Nos. ML003744984 and ML003744988. 

 NRC Regulatory Issue Summary 2004-17, Revision 1, “Revised Decay-In-Storage 

Provisions For The Storage Of Radioactive Waste Containing Byproduct Material,” 

September 2005 

 State and Tribal Communication Letter FSME 12-025 dated March 13, 2012 

“Clarification of the Authorization for Alternative Disposal of Material Issued Under 

10 CFR 20.2002 and Exemption Provisions in 10 CFR” (ADAMS Accession No. 

ML12065A038) 

 Division of Waste Management and Environmental Protection, Environmental and 

Performance Assessment Directorate, Operating Procedures, EPPAD 3.5 (Draft for 

Interim Use), “Review, Approval, and Documentation of Low-Activity Waste Disposals in 

Accordance with 10 CFR 20.2002 and 10 CFR 40.13(a),” dated August 2009 

 Regulatory Issue Summary 2016-11, “Requests to Dispose of Very Low-Level 

Radioactive Waste Pursuant to 10 CFR 20.2002,” November 13, 2016 

8.12 Item 12: License Fees 

On NRC Form 313, enter the appropriate fee category from 10 CFR 170.31 and the amount of 

the fee enclosed with the application. 

Direct all questions about the NRC’s fees or the completion of Item 12 of NRC Form 313 to 

the Office of the Chief Financial Officer at NRC Headquarters in Rockville, MD, 301-415-7554. 

Information about fees may also be obtained by calling the NRC’s toll-free number, 

800-368-5642, extension 415-7554. The e-mail address for fees questions is 

Fees.Resource@nrc.gov. 

8.13 Item 13: Certification 

A representative of the corporation or legal entity filing the application should sign and date 

NRC Form 313. The representative signing the application must be authorized to make 

binding commitments and to sign official documents on behalf of the applicant. As discussed 
8-55 

previously in Chapter 3, “Management Responsibility,” signing the application acknowledges 

management’s commitment to and responsibility for the radiation protection program. The 

NRC will return all unsigned applications for proper signature. 

Notes:

 It is a criminal offense to knowingly and willfully make a false statement or 

representation on applications or correspondence (18 U.S.C. 1001). 

 When an application references commitments, those items will be incorporated into the 

license and therefore become binding regulatory requirements. 

9-1 

9 LICENSE AMENDMENTS AND RENEWALS 

It is the licensee’s obligation to keep the license current. If any of the information provided in 

the original application is to be modified or changed, the licensee must submit an application for 

a license amendment before the change takes place. The change is not in effect until the 

amendment has been issued. Also, to continue the license after its expiration date, the licensee 

must submit an application for a license renewal at least 30 days before the expiration date 

[10 CFR 2.109(a), 10 CFR 30.36(a)]. 

Applicants for license amendment or renewal should do the following: 

 Use the most recent guidance in preparing an amendment or renewal request. 

 Submit either an U.S. Nuclear Regulatory Commission (NRC) Form 313 or a letter 

requesting amendment or renewal. 

 Provide the license number and docket number. 

 For renewals, provide a complete and up-to-date application including all required 

program elements outlined in Appendix B of this NUREG. Training documentation for 

personnel currently listed on the license does not need to be submitted as part of the 

renewal application. 

9.1 Timely Notification of Transfer of Control 

Regulations: 10 CFR 30.34(b), 10 CFR 40.46, 10 CFR 70.36. 

Criteria: Licensees must provide all supporting information and obtain the NRC’s prior, written 

consent before transferring control of the license, also referred to as a “change of ownership” 

and/or “transferring the license.” 

Discussion: Transferring control may be the result of mergers, buyouts, or majority stock 

transfers. Although it is not the NRC’s intent to interfere with the business decisions of 

licensees, it is necessary for licensees to obtain prior NRC written consent to ensure the 

following: 

 radioactive materials are possessed, used, or controlled only by persons who have valid 

NRC licenses or Agreement State licenses 

 materials are properly handled and secured 

 persons using these materials are capable, competent and committed to implementing 

appropriate radiological controls 

 a clear chain of custody is established to identify who is responsible for disposition of 

records and licensed material 

 the transferee has the financial resources to decommission the license, if necessary
9-2 

 public health and safety are not compromised by the use of such materials 

 adequate financial assurance is provided for compliance with the applicable NRC 

requirements, if required 

Response from Applicant: No response is required from an applicant for a new license. 

However, current licensees should refer to NUREG–1556, Volume 15, “Consolidated Guidance 

About Materials Licenses: Guidance About Changes of Control and About Bankruptcy Involving 

Byproduct, Source, or Special Nuclear Materials Licenses,“ for more information about transfer 

of control (i.e., ownership). 

Reference: For further information, see Regulatory Issue Summary (RIS) 2014-08, Rev. 1, 

“Regulatory Requirements for Transfer of Control (Change of Ownership) of Specific Materials 

Licensees,” dated May 5, 2016, which can be found on the NRC’s Generic Communications 

Web page under “Regulatory Issue Summaries”: http://www.nrc.gov/reading-rm/doccollections/gen-comm/. 
10-1 

10 APPLICATIONS FOR EXEMPTIONS 

Regulations: 10 CFR 19.31, 10 CFR 20.2301, 10 CFR 30.11 

Criteria: Licensees may request exemptions from U.S. Nuclear Regulatory Commission (NRC) 

regulations. The licensee must demonstrate that the exemption is authorized by law; will not 

endanger life, property, or the common defense and security; and is otherwise in the public 

interest. Licensees may also use existing specific exemptions outlined in the Title 10 of the 

Code of Federal Regulations (10 CFR) regulations if they meet the established criteria. 

Discussion: Various sections of the NRC’s regulations address requests for exemptions 

(e.g., 10 CFR 19.31, “Application for exemptions”; 10 CFR 20.2301, “Applications for 

exemptions”; 10 CFR 30.11, “Specific exemptions”). These regulations state that the NRC may 

grant an exemption, acting on its own initiative or on an application from an interested person. 

Exemptions are not intended to revise regulations or apply to large classes of licensees and are 

generally limited to unique situations. Requests for exemptions submitted to the NRC must 

identify the regulation for which the exemption is being requested and include a justification for 

the requested exemption. 

Unless the NRC has granted an exemption in writing, licensees must comply with all applicable 

regulations. 

 11-1 

11 TERMINATION OF ACTIVITIES 

Regulations: 10 CFR 20.2002, 10 CFR 20.2003, 10 CFR 20.2004, 10 CFR 20.2005, 

10 CFR 30.34(b), 10 CFR 30.35(g), 10 CFR 30.36, 10 CFR 30.51, 10 CFR 40.36(f), 

10 CFR 40.42, 10 CFR 40.46, 10 CFR 40.61, 10 CFR 70.25(g), 10 CFR 70.36, 10 CFR 70.38, 

10 CFR 70.51 

Criteria: The licensee must do the following: 

 Notify the U.S. Nuclear Regulatory Commission (NRC), in writing, within 60 days of the 

occurrence of any of the following: 

— expiration of its license 

— a decision to permanently cease principal activities1

 at the entire site 

— for licensees subject to Title 10 of the Code of Federal Regulations (10 CFR) 

30.36, a decision to permanently cease principal activities in any separate 

building or outdoor area that contains residual radioactivity such that the building 

or area is unsuitable for release according to NRC requirements 

— for licensees subject to 10 CFR 40.42 or 10 CFR 70.38, a decision to 

permanently cease principal activities in any separate building or outdoor area 

— no principal activities under the license have been conducted for a period of 

24 months 

— no principal activities have been conducted for a period of 24 months in any 

separate building or outdoor area that contains residual radioactivity such that 

the building or outdoor area is unsuitable for release according to NRC 

requirements 

 Submit a decommissioning plan, if required by 10 CFR 30.36(g), 10 CFR 40.42(g), 

and/or 10 CFR 70.38(g). 

 Conduct decommissioning, as required by 10 CFR 30.36(h) and (j), 10 CFR 40.42(h) 

and (j), and/or 10 CFR 70.38(h) and (j). 

 Submit to the appropriate NRC regional office a completed NRC Form 314, “Certificate 

of Disposition of Materials” (or equivalent information), and information demonstrating 

that the premises are suitable for release for unrestricted use (e.g., results of final 

surveys and results of leak tests of sealed sources). 

 Before a license is terminated, send records important to decommissioning that are 

required by 10 CFR 30.35(g), 10 CFR 40.36(f) and/or 10 CFR 70.25(g) to the 

 

1

‘Principal activities’ are activities which are essential to achieving the purpose(s) for which the license was issued or 

amended. Storage during which no licensed material is accessed for use or disposal and activities incidental to 

decontamination or decommissioning are not principal activities. 
 11-2 

appropriate NRC regional office in accordance with 10 CFR 30.51(f), 10 CFR 40.61(f), 

and/or 10 CFR 70.51(a)(3), respectively. 

 Before a license is terminated, send records of disposal of licensed material made under 

10 CFR 20.2002, 10 CFR 20.2003, 20.2004, 20.2005, and the results of measurements 

and calculations used to evaluate the release of radioactive effluents to the environment 

to the appropriate NRC regional office in accordance with 10 CFR 30.51(d), 

10 CFR 40.61(d), and/or 10 CFR 70.51(a)(1)&(2), if authorized to possess byproduct 

material with a half-life greater than 120 days in an unsealed form, source material in an 

unsealed form, and/or special nuclear material, respectively. 

Discussion: To comply with the above criteria, before a licensee can decide whether it must 

notify the NRC under 10 CFR 30.36(d), 10 CFR 40.42(d), and/or 10 CFR 70.38(d), the licensee 

must determine whether residual radioactivity is present and, if so, whether the levels make the 

building or outdoor area unsuitable for release, according to NRC requirements. A licensee’s 

determination that a facility is not contaminated is subject to verification by NRC inspection. 

The permanent cessation of principal activities in an individual room or laboratory may require 

the licensee to notify the NRC if no other licensed activities are being performed in the building. 

This requirement also applies to buildings that were approved by the broad scope licensee as 

locations of use but not specifically named on the broad scope license. 

For information about requirements that apply to the timeliness of decommissioning, see 

Regulatory Issue Summary (RIS) 2015-19, Rev. 1, “Decommissioning Timeliness Rule 

Implementation and Associated Regulatory Relief,” dated September 27, 2016, which can be 

found on the NRC’s Generic Communications Web page under “Regulatory Issue Summaries”: 

http://www.nrc.gov/reading-rm/doc-collections/gen-comm/reg-issues/2015/. 

For guidance on the disposition of licensed material, see Section 8.11 “Waste Management.” 

For guidance on decommissioning records, see Section 8.5.2, “Financial Assurance and 

Recordkeeping for Decommissioning.” 

NUREG–1757, “Consolidated Decommissioning Guidance,” contains the current regulatory 

guidance concerning decommissioning of facilities and termination of licenses. Licensees that 

have large facilities to decommission should review NUREG–1575, “Multi-Agency Radiation 

Survey and Site Investigation Manual (MARSSIM).” The computer code “DandD” offers an 

acceptable method for calculating screening values to demonstrate compliance with the 

unrestricted dose limits. Supplemental information on the implementation of the final rule on 

radiological criteria for license termination was published in the Federal Register (63 FR 64132) 

on November 18, 1998. 

Supplemental information on the implementation of the final rule on radiological criteria for 

license termination also was published in the Federal Register on December 7, 1999, 

(64 FR 68395) which addresses screening values in soils for the most common radionuclides, 

and in the Federal Register on June 13, 2000, (65 FR 37186) for screening values for building 

surfaces and soils contaminated with radionuclides not addressed in the prior Federal Register

notices. 

Response from Applicant: The applicant is not required to submit a response to the NRC 

during the initial application. The licensee’s obligations in this matter begin when the license 
 11-3 

expires or at the time the licensee ceases operations, whichever is earlier. These obligations 

are to undertake the necessary decommissioning activities, to submit NRC Form 314 or 

equivalent information, and to perform any other actions summarized in “Criteria” above.

Reference: NRC Form 314 is available at http://www.nrc.gov/reading-rm/doc-collections/forms. 

APPENDIX A 

U.S. NUCLEAR REGULATORY COMMISSION FORM 313 

A–1 

U.S. Nuclear Regulatory Commission Form 313 

Please use the most current version of this form, which may be found at: 

http://www.nrc.gov/reading-rm/doc-collections/forms/

APPENDIX B 

SUGGESTED FORMAT FOR PROVIDING INFORMATION REQUESTED IN 

ITEMS 5 THROUGH 11 OF 

U.S. NUCLEAR REGULATORY COMMISSION FORM 313 

B–1 

Suggested Format for Providing Information Requested in 

Items 5 Through 11 of 

U.S. Nuclear Regulatory Commission Form 313

Instructions: 

If an applicant checks a box in the column marked “Description Attached,” then they must 

provide that information on separate sheets. 

Item 

No.

Suggested Response Description 

Attached

5. RADIOACTIVE MATERIAL

Unsealed and Sealed Byproduct Material

Applicants for a Type A broad scope license should request any form of 

byproduct material with atomic numbers from 1 through 83. The 

applicant should state the maximum quantity of each radionuclide to be 

possessed at any one time and the total cumulative quantity for all 

radionuclides. The applicant should separately list individual 

radionuclides needed in much larger quantities or in much smaller 

quantities than that described in the atomic number 1 through 83 

request. The maximum quantities of nuclides with atomic numbers 

above 83 also should be listed separately. 

[ ] 

 A separate listing should be submitted for sealed sources needed in 

larger quantities than described in the atomic number 1 through 83 

request. Applicants must provide the manufacturer’s name and model 

number for each requested sealed source and device so that the NRC 

can verify that they have been evaluated in a SSD registration 

certificate or specifically approved on a license. Applicants must also 

provide the maximum activity per source and the total possession 

limit. For sources and devices not registered, as allowed by 

10 CFR 32.210(g)(2), the applicant must have adequate training and 

experience and facilities and equipment to handle comparable 

quantities of material in any form under 10 CFR 30.33(a)(2) and (3) and 

must provide information about the unregistered sealed sources and 

devices in accordance with 10 CFR 30.32(g)(4). 

[ ] 
B–2 

Item 

No.

Suggested Response Description 

Attached

5. RADIOACTIVE MATERIAL (Cont’d)

Unsealed and Sealed Sources (Cont’d)

Possession requests should be categorized into general areas of use 

(e.g., research and development activities, routine gauging activities, 

self-shielded irradiators, instrument calibrators, and medical 

applications). 

Applicants for licenses to possess radioactive materials in unsealed 

form, on foils or plated sources, or sealed in glass in excess of the 

quantities listed in 10 CFR 30.72 must include either of the following: 

(1) an evaluation showing that the maximum offsite dose caused by a 

release of radioactive materials would not exceed 0.01 Sv (1 rem) 

effective dose equivalent or 0.05 Sv (5 rem) to the thyroid, or (2) an 

emergency response plan for responding to the release in accordance 

with the criteria listed in 10 CFR 30.32(i)(3). 

Applicants for a Type B or Type C broad scope license should request 

any chemical or physical form of byproduct material specified in 

10 CFR 33.100, Schedule A. Type B applicants should request the 

quantity of material specified in 10 CFR 33.11(b). Type C applicants 

should request the quantity of material specified in 10 CFR 33.11(c). 
B–3 

Item 

No.

Suggested Response Description 

Attached

Financial Assurance and Recordkeeping for Decommissioning

State the following: “Pursuant to 10 CFR 30.35(g), 10 CFR 40.36(f), 

and/or 10 CFR 70.25(g) and 10 CFR 70.51(b)(3), we will maintain 

records important to decommissioning and transfer these records to an 

NRC or Agreement State licensee before licensed activities are 

transferred or assigned in accordance with 10 CFR 30.34(b), 

10 CFR 40.46, and/or 10 CFR 70.36. Furthermore, pursuant to 

10 CFR 30.51(f), 10 CFR 40.61(f), and /or 10 CFR 70.51(a)(3), prior to 

license termination, we will forward the records required by 

10 CFR 30.35(g), 10 CFR 40.36(f), and/or 10 CFR 70.25(g) to the 

appropriate NRC Regional Office.” 

If financial assurance is required, submit evidence of financial 

assurance following the guidance of NUREG–1757, Volume 3, 

“Consolidated Decommissioning Guidance: Financial Assurance, 

Recordkeeping, and Timeliness.” 

Emergency Plan 

If an emergency plan is required, provide either: 

an evaluation showing that the maximum offsite dose due to a release 

of radioactive materials would not exceed 0.01 Sv [1 rem] effective 

dose equivalent or 0.05 Sv [5 rem] to the thyroid; or 

an emergency plan for responding to the release of radioactive material 

in accordance with the criteria listed in 10 CFR 30.32(i)(3). 

[ ] 

[ ] 

6. PURPOSE(S) FOR WHICH LICENSED MATERIAL WILL BE USED

Describe in general terms the purposes for which the licensed material 

will be used. 

[ ] 

7. INDIVIDUAL(S) RESPONSIBLE FOR RADIATION SAFETY 

PROGRAM AND THEIR TRAINING AND EXPERIENCE

Executive Management

The applicant must describe administrative controls and provisions 

relating to organization and management and management review 

necessary to ensure safe operations. The applicant should submit an 

organizational chart that describes the management structure, reporting 

paths, and the flow of authority between executive management, the 

RSC (for Type A broad scope), and the RSO (For Type A and Type B 

broad scope). 

[ ] 
B–4 

Item 

No.

Suggested Response Description 

Attached

7. INDIVIDUAL(S) RESPONSIBLE FOR RADIATION SAFETY 

PROGRAM AND THEIR TRAINING AND EXPERIENCE (Cont’d)

Radiation Safety Committee

Applicants for a Type A broad scope license should submit the 

following: 

 description of the duties and responsibilities of the RSC 

 criteria used for selecting members of the RSC, including what 

members and the number of members constituting a quorum. 

Members should be indicated by position title rather than by name 

 criteria and procedure describing the approval process used by the 

RSC and RSO for authorizing new users and new uses 

In addition, applicants for a Type A broad scope license that request 

the flexibility to make some program changes and revise some 

procedures previously approved by the NRC without amendment of the 

license should submit the following: 

 a description of the duties and responsibilities of the RSC, 

including: 

– review and approval of permitted program and procedural 

changes prior to implementation 

– implementation of program and procedural changes 

– audit of licensed operations to determine compliance 

– the appropriate actions taken when noncompliance is identified, 

including analysis of the cause, corrective actions, and actions to 

prevent recurrence 

 a description of the process for procedure and program review and 

approval, including documentation of the specific change. (At a 

minimum, documentation should state the reason for the change 

and summarize the radiation safety matters that were considered 

prior to approval of the change.) 

[ ] 

[ ] 

[ ] 
B–5 

Item 

No.

Suggested Response Description 

Attached

7. INDIVIDUAL(S) RESPONSIBLE FOR RADIATION SAFETY 

PROGRAM AND THEIR TRAINING AND EXPERIENCE (Cont’d)

Radiation Safety Officer

For Type A, Type B, and Type C applicants: 

 Submit the name of the proposed RSO. 

 Describe the training for the proposed RSO that demonstrates the 

individual is qualified to perform the duties required under the 

license. 

 Address the RSO’s experience in performing each of the duties 

listed in the “Duties and Responsibilities” section in section 8.7.3, 

“Radiation Safety Officer,” when and where the experience was 

gained, and the type, form, and quantity of radionuclides involved. 

 Submit a statement delineating the RSO’s duties and 

responsibilities. 

 Submit a radiation safety officer delegation of authority 

memorandum signed by the licensee’s executive management. 

In addition, for Type B applicants: 

 Submit the criteria used by the RSO to approve new users and 

uses of byproduct material. 

 Submit the criteria that the RSO will use to evaluate the radiation 

safety aspects of proposed uses, prior to approval. 

[ ] 

[ ] 
B–6 

Item 

No.

Suggested Response Description 

Attached

8. TRAINING FOR INDIVIDUALS WORKING IN OR FREQUENTING 

RESTRICTED AREAS

Submit a description of the radiation safety training program developed 

for each group of workers, including: topics covered, qualifications of 

the instructors, method of training, method for assessing the success of 

the training, and the frequency of training and refresher training. 

Alternately, identify the model training program described in the 

appropriate base NUREG corresponding to the particular type of 

licensed program and submit a statement that this training program will 

be implemented. 

In addition, Type A broad scope licensees or applicants that want the 

flexibility to revise their radiation safety training program without 

amendment of the license (as discussed in Chapter 1, “Purpose of 

Report,” and Section 8.7.2, “Radiation Safety Committee,” of NUREG1556, Volume 11, Revision 1) should describe the process that will be 

used to revise and implement the submitted training programs. 

[ ] 

[ ] 

9. FACILITIES AND EQUIPMENT

Describe the criteria the RSC or RSO, as appropriate, will use to review 

and approve facilities and equipment (research laboratories, iodination 

facilities, waste storage facilities, survey and counting equipment, etc.). 

The description will need to include the method of classifying 

laboratories based on type, toxicity, and quantity of byproduct material 

being requested. Sample diagrams should be provided for each 

classification scheme that take into consideration shielding, the 

proximity of radiation sources to unrestricted areas, and other items 

related to radiation safety. When reviewing facilities where radioactive 

materials may become airborne, sample diagrams should take into 

consideration descriptions of the ventilation systems—including 

pertinent airflow rates, pressures, filtration equipment, and monitoring 

systems. For facilities and equipment used in special applications such 

as those described in Section 8.9 of NUREG-1556, Volume 11, 

Revision 1, the application will include their locations, (i.e., buildings 

and room numbers) and special considerations that the RSC or RSO 

(or both) will use in authorizing byproduct material use. Also, describe 

the procedures for control, review, and approval of significant facilities 

or equipment modifications. 

[ ] 
B–7 

Item 

No.

Suggested Response Description 

Attached

10. RADIATION SAFETY PROGRAM

Audit and Review of Program

Describe the mechanisms executive management uses to ensure 

adequate oversight of the program. In addition, if a licensee is 

upgrading its limited scope license to a Type A broad scope license or 

renewing its Type A broad scope license, describe the RSC’s 

involvement in these oversight mechanisms. 

[ ] 

 Describe the audit mechanism implemented by the RSO or other 

responsible individual to determine user compliance with NRC 

regulations, the terms and conditions of the NRC license, the 

requirements of the RSC- or RSO-approved permits (as appropriate), 

and good health physics practices. 

 The applicant is not required to, and should not, submit its program for 

conducting the annual audit required by 10 CFR 20.1101, “Radiation 

protection programs,” to the NRC for review as part of a license 

application. The NRC will review this audit program during inspection. 

 In addition, Type A broad scope licensees or applicants that want the 

flexibility to revise the audit mechanism implemented by the RSO 

without amendment of the license, as discussed in Chapter 1, “Purpose 

of Report,” and Section 8.7.2, “Radiation Safety Committee,” of 

NUREG-1556, Volume 11, Revision 1, should describe the process 

they will use to revise and implement their audit program. 
B–8 

Item 

No.

Suggested Response Description 

Attached

10. RADIATION SAFETY PROGRAM (Cont’d)

Radiation Monitoring Instruments

Provide the criteria used by the RSC or RSO (or both), as appropriate, 

to review and approve radiation monitoring instrumentation to ensure 

that appropriate radiation monitoring equipment will be used during 

licensed activities. 

Discuss how the RSC or RSO, as appropriate, will ensure that 

instruments are properly calibrated at prescribed frequencies. 

 

Submit procedures for instrument calibration or state that instruments 

will be calibrated by a vendor licensed by the NRC or an Agreement 

State to perform instrument calibrations. Licensees that want 

authorization to calibrate their own survey instruments may commit to 

implementing the model procedures published in Appendix H of 

NUREG-1556, Volume 11, Revision 1. 

State the frequency at which instruments will be calibrated. 

In addition, Type A broad scope licensees or applicants that want the 

flexibility to revise their instrument specifications and procedure for 

calibration of instruments without amendment of the license, as 

discussed in Chapter 1, “Purpose of the Report,” and Section 8.7.2, 

“Radiation Safety Committee,” of NUREG-1556, Volume 11, Revision 

1, should describe the process that will be used to revise and 

implement these submitted procedures. 

[ ] 

[ ] 
B–9 

Item 

No.

Suggested Response Description 

Attached

10. RADIATION SAFETY PROGRAM (Cont’d)

Material Receipt and Accountability

Describe the administrative procedures to ensure control of 

procurement and use of byproduct material. 

While the applicant is required to develop and implement safe opening 

procedures for packages containing radioactive material, the applicant 

need not submit the procedures during the licensing process. The NRC 

will review activities subject to these procedures during inspection. 

Provide the following statement: “We will develop, implement and 

maintain procedures for ensuring accountability of licensed materials at 

all times.” 

Describe the administrative controls and provisions related to materials 

control, accounting, and security. Describe the method for maintaining 

accountability of licensed material at all times. 

If applicable, provide the following statement: “We will comply with the 

National Source Tracking System (NSTS) reporting requirement as 

described in 10 CFR 20.2207.” 

In addition, Type A broad scope licensees or applicants that want the 

flexibility to revise their administrative procedures concerning control of 

procurement and use of byproduct material without amendment of their 

license should describe the process that their radiation safety 

committee will use to revise these administrative procedures, controls, 

and provisions. Licensees and applicants should also do this when 

making revisions to their administrative controls and provisions related 

to material control, accounting, and security. The flexibility to revise 

these administrative procedures is discussed in Chapter 1, “Purpose of 

Report,” and Section 8.7.2, “Radiation Safety Committee,” of NUREG1556, Volume 11, Revision 1. 

[ ] 
B–10 

Item 

No.

Suggested Response Description 

Attached

10. RADIATION SAFETY PROGRAM (Cont’d)

Occupational Dose

Provide one of the following statements:

“We will maintain, for inspection by the NRC, documentation 

demonstrating that unmonitored individuals are not likely to receive a 

radiation dose in excess of the limits in 10 CFR 20.1502.” 

OR 

“We will monitor individuals in accordance with the criteria in the section 

titled, ‘Radiation Safety Program–Occupational Dose’ in NUREG–1556, 

Vol. 11, Revision 1, ‘Consolidated Guidance About Materials Licenses: 

Program-Specific Guidance About Licenses of Broad Scope.’”

OR, IN LIEU OF THESE STATEMENTS, 

provide a description of an alternative method for demonstrating 

compliance with the referenced regulations. 

In addition, Type A broad scope licensees or applicants that want the 

flexibility to revise their personnel dosimetry program without 

amendment of the license, as discussed in Chapter 1, “Purpose of 

Report,” and Section 8.7.2, “Radiation Safety Committee,” of NUREG1566, Volume 11, Revision 1, should describe the process they will use 

to revise and implement their submitted personnel dosimetry program. 

Public Dose

No response is required from the applicant, but records and written 

materials documenting compliance will be examined during inspection. 

During NRC inspections, licensees must be able to demonstrate, by 

measurement or calculation, that the total effective dose equivalent to 

the individual likely to receive the highest dose from the licensed 

operation does not exceed the annual limit for members of the public. 

[ ] 

[ ] 
B–11 

Item 

No.

Suggested Response Description 

Attached

10. RADIATION SAFETY PROGRAM (Cont’d)

Safe Use of Radionuclides and Emergency Procedures

Submit the procedures for safe use of radionuclides and emergencies. 

Submissions should include procedures for maintaining security of 

licensed radioactive materials. As an alternative, the applicant or 

licensee may state that, “We will adopt the procedures for the safe use 

of radionuclides and emergencies as published in Appendix K of 

NUREG–1556, Volume 11, Revision 1, “Consolidated Guidance About 

Materials Licenses: Program-Specific Guidance About Licenses of 

Broad Scope.”” 

In addition, Type A broad scope licensees or those applying for a Type 

A broad scope license that want the flexibility to revise their safe use 

and emergency procedures without amendment of the license, as 

described in Chapter 1, “Purpose of Report,” and Section 8.7.2, 

“Radiation Safety Committee,” of NUREG-1556, Volume 11, Revision 

1, should discuss the process that will be used to revise and implement 

their submitted safe use and emergency procedures. 

Surveys

Submit procedures to evaluate radiological hazards, both external and 

internal. As an alternative, the applicant may state, “We will survey our 

facility and maintain contamination levels and perform bioassays of 

occupationally exposed workers in accordance with the survey 

frequencies and contamination levels published in Appendix L of 

NUREG–1556, Volume 11, Revision 1, “Consolidated Guidance About 

Materials Licenses: Program-Specific Guidance About Licenses of 

Broad Scope.”” 

Leak Testing 

Submit leak test procedures. As an alternative, the applicant may 

state, “We will implement the model leak test program published in 

Appendix M of NUREG–1556, Volume 11, Revision 1, “Consolidated 

Guidance About Materials Licenses: Program-Specific Guidance About 

Licenses of Broad Scope.”” 

In addition, Type A broad scope licensees or those applying for a Type 

A broad scope license that want the flexibility to revise their survey or 

leak test program without amendment of the license, as described in 

Chapter 1, “Purpose of Report,” and Section 8.7.2, “Radiation Safety 

Committee,” of NUREG-1556, Volume 11, Revision 1, should discuss 

the process that will be used to revise and implement their submitted 

survey and leak test program. 

[ ] 

[ ] 

[ ] 

[ ] 

[ ] 
B–12 

Item 

No.

Suggested Response Description 

Attached

10. RADIATION SAFETY PROGRAM (Cont’d)

Transportation

No response is needed from applicants during the licensing phase. 

Compliance with transportation requirements will be reviewed during 

NRC inspections. 

Security Program for Category 1 and Category 2 Material

No response is required from an applicant or licensee. Compliance 

with access authorization and security program requirements may be 

reviewed during NRC inspections.

11. WASTE MANAGEMENT

Provide procedures for waste collection, storage, and the disposal by 

any of the authorized methods described in this section. Applicants 

should contact the appropriate regional office of the NRC for guidance 

to obtain approval of any method(s) of waste disposal other than those 

discussed in this section. 

[ ] 
APPENDIX C 

INFORMATION NEEDED FOR FIELD USE OF BYPRODUCT MATERIAL 

C–1 

Information Needed for Field Use of Byproduct Material 

Title 10 of the Code of Federal Regulations (CFR) 51.22(c)(14)(v) identifies as a categorical 

exclusion (from the requirement to prepare an environmental assessment (EA) or impact 

statement) the use of radioactive material for research and development and for educational 

purposes. However, this categorical exclusion does not encompass, among other things, 

performance of field studies in which licensed material is deliberately released directly into the 

environment for purposes of the study (e.g., tagging of animals or insects that remain in the 

wild, and use of tagged pesticides on crops grown outdoors in fields). 

These types of requests may require an environmental report filed by the applicant and an EA 

by the U.S. Nuclear Regulatory Commission (NRC), pursuant to 10 CFR Part 51, 

“Environmental Protection Regulations for Domestic Licensing and Related Regulatory 

Functions.” Field studies that do not deliberately release radioactive material into the 

environment, such as tagging of animals and penning them to prevent escape or plant studies in 

greenhouse conditions, may be eligible for a categorical exclusion, pursuant to 

10 CFR 51.22(c)(14)(xvi). 

Applicants and licensees that desire to perform field studies in which licensed material is 

deliberately released to the environment for the purposes of studies should provide the following 

information: 

1. a complete application describing the type and amount of material to be used, the 

location of use, and training and experience of the individual using the material 

2. a complete experimental protocol 

3. a description of the amount of radioactive material to be released in the field, 

decontamination procedures at the conclusion of the experiment, if appropriate, and 

procedures for minimizing releases 

4. a description of the expected radiation dose to humans 

5. written permission from the property owner to use radioactive materials at the proposed 

site 

6. a letter from the appropriate state health authorities indicating that they have reviewed 

the applicant’s or licensee’s application and concur with the request 

APPENDIX D 

MODEL DELEGATION OF AUTHORITY FOR RADIATION 

SAFETY OFFICER

D–1 

Model Delegation of Authority for Radiation Safety Officer 

Model Delegation of Authority 

Memo To: Radiation Safety Officer 

From: Chief Executive Officer 

Subject: Delegation of Authority 

You, _______________________________, have been appointed radiation safety officer and 

are responsible for ensuring the safe and secure use of radiation. You are responsible for 

managing the Radiation Protection Program, identifying radiation protection problems, initiating, 

recommending, or providing corrective actions, verifying implementation of corrective actions, 

stopping unsafe activities, and ensuring compliance with regulations. You are hereby delegated 

the authority necessary to meet those responsibilities, including prohibiting the use of byproduct 

material by employees who do not meet the necessary requirements and shutting down 

operations, when justified, to maintain radiation safety. You are required to notify management 

if staff does not cooperate and does not address radiation safety issues. In addition, you are 

free to raise issues with the U.S. Nuclear Regulatory Commission at any time. It is estimated 

that you will spend _____ hours per week conducting radiation protection activities. 

____________________________________ ____________________ 

Signature of Management Representative Date 

I accept the above responsibilities, 

_____________________________________ ____________________ 

Signature of Radiation Safety Officer Date 

cc: Affected department heads 

APPENDIX E 

FACILITIES AND EQUIPMENT CONSIDERATIONS

E–1 

Facilities and Equipment Considerations 

Below is a list of topics that should be considered when developing a description of the facilities 

and equipment that a broad scope applicant will use or otherwise have available. Not every 

applicant will need to address each topic in its application. 

 Restricted areas are defined as areas where the licensee limits access to protect 

individuals against undue risks from exposure to radiation and radioactive materials. 

The application should contain detailed descriptions and diagrams of the facilities, 

including information about the shielding properties of the construction materials used. 

Scaled drawings and sketches should be submitted showing the relationship between 

restricted areas and unrestricted areas and the location of all pertinent safety-related 

equipment. 

 Security zones are defined as any temporary or permanent area determined and 

established by the licensee for the physical protection of aggregated Category 1 or 

Category 2 quantities of radioactive material listed in Appendix A to 10 CFR Part 37. 

The security zone should be designed so that the licensee can monitor, detect without 

delay, assess, and respond to any unauthorized entries into security zones and any 

unauthorized removal of radioactive material from the security zone. Monitoring and 

detection systems may include video surveillance systems and electronic devices for 

intrusion detection alarms. 

 Bench top or open work areas may be used for sealed sources, small quantities of solid 

materials in a form not likely to become airborne or dispersed, and small quantities of 

liquids of such low volatility as not to cause airborne contamination or toxicity problems. 

Trays and absorbent surface covers to catch and retain spilled liquids should be used on 

these open work surfaces and inside closed systems. Surfaces should be smooth and 

non-porous to facilitate decontamination. 

 Radioactive materials that are handled or used in unsealed forms should be confined to 

control the release of material and to prevent the spread of contamination. Gaseous, 

volatile, and fine particulate solid materials should be handled in closed or isolated 

systems such as fume hoods or glove boxes with controlled, and possibly filtered, 

exhaust systems. Ventilation systems for these facilities should be designed so that 

airborne radioactive material work areas are at negative pressure compared to nonradioactive work areas. 

 Chemical-type fume hoods provide a working area with controlled inward airflow from 

the room to the hood exhaust system. Hoods are used for gases, unsealed volatile 

licensed materials, and processes such as evaporation that may release gases and 

vapors. Fume hoods provide emergency ventilation and exhaust for unplanned 

releases, such as accidental spills and ruptures, as well as routine exhaust of effluents. 

Filters may be required in the exhaust stream unless monitoring or calculations 

demonstrate that any planned or likely effluent will be in accordance with the limits found 

in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Appendix B. 

 Glove boxes are sealed boxes with transparent viewing windows, sealable ports or 

doors for transferring materials and equipment, and gloves sealed to the box through 

which licensed materials are handled. Glove boxes are used for the containment during 

storage and use of liquids and solids that can become airborne particulates or aerosols. 
E–2 

Glove boxes can be closed or exhausted, with filtration systems if appropriate, to prevent 

contamination. 

 For the most efficient operation of hoods and glove boxes, minimize storage of materials 

and equipment inside the work areas. 

 Where appropriate, ventilation systems should be designed such that, in the event of an 

accident, they can be shut down to prevent the spread of radioactivity. If appropriate, 

supply and exhaust fans can be interlocked such that if exhaust fans shutdown, the 

shutdown of supply fans is also triggered, this interlock system to prevent laboratory and 

work areas from becoming positively pressurized with respect to the surrounding parts of 

the facility. 

 Sink faucets should be designed, where possible, for operation by foot, knee, or elbow, 

rather than by hand. 

 Plumbing and ductwork should be designed to avoid radioactive contamination build-up. 

This build-up of contamination can create external radiation exposure hazards and 

problems for decommissioning. 

 To reduce radiation exposure from gamma-emitting radioactive materials, shielding 

consisting of lead or other high-density material in the form of bricks, panels, L-shields, 

storage containers, or other shapes may be used on bench tops, in fume hoods, or in 

glove boxes. 

 To reduce the exposure from high-energy beta-emitting materials, shielding of low 

atomic number material, such as high-density plastic, may be used. 

 Shielded shipping containers frequently are used for continued storage after receipt of 

materials. 

 Remote handling tools, such as forceps or extension handles, should be used to provide 

distance in the handling of radioactive materials (as low as is reasonably achievable). In 

addition, shielded handling devices, such as shielded syringes, can be used to protect 

workers from materials that cannot be handled remotely. Pipetting should be done using 

appropriate devices. Pipetting by mouth should be strictly forbidden. 

 Designated areas should be provided for coats and personal belongings, to avoid 

contamination. 

 Areas with low background radiation levels should be designated for personnel 

dosimetry storage when not in use. 

 Areas of use should be well-lighted to avoid spills and other accidents that could result in 

contamination buildup. 

 Observation of activities conducted behind shielding with remote tools (or with extended 

arms and hands, within limits consistent with permissible occupational exposures) can 

be accomplished by mirrors, through shielded (e.g., leaded glass) windows, through 

transparent plastic beta shields, or by remote video monitoring. 
E–3 

 The combination of containment, shielding, and handling devices proposed for any use 

of radioactive materials should be appropriate to the type and quantity of materials to be 

used and to the type and duration of operations to be conducted. 

 If respiratory protective equipment will be used to limit inhalation of airborne licensed 

material, follow the provisions of 10 CFR Part 20, Subpart H, “Respiratory protection and 

controls to restrict internal exposure in restricted areas.” 

 A particular sink should be designated for disposal of liquid radioactive waste to the 

sanitary sewerage system. In some cases, depending on the number of users and 

distance between areas of use, more than one sink may need to be designated. 

 Labeled waste containers should be used. These containers may be shielded as 

necessary, placed near the waste-generating areas and away from areas that personnel 

frequently occupy. Additionally, these containers should be effectively enclosed to 

prevent airborne contamination from radioactive materials deposited. If radioactive 

waste materials are volatile, the containers should be stored in ventilated areas. 

 If compaction of waste is performed, ensure that facilities are adequate for the ventilation 

of the area where the waste is compacted. In addition, also ensure that air sampling for 

internal exposures is available, if needed per 10 CFR 20.1204, “Determination of internal 

exposure.” 

 Adequate air and/or water effluent monitoring equipment should be used to demonstrate 

compliance with the limits found in 10 CFR Part 20, Appendix B, if applicable, and tested 

for operability at the frequency established by the manufacturer. 

APPENDIX F 

SAMPLE AUDIT PROGRAM—NON-MEDICAL 

F–1 

Sample Audit Program–Non-medical 

Licensees may use the following audit form to self-assess the adequacy of the licensed 

program, identify program weaknesses, and allow licensees to take early corrective actions 

[before an inspection by the U.S. Nuclear Regulatory Commission (NRC)]. This form is not 

intended to be all inclusive. During an audit, the auditor needs to keep in mind not only the 

requirements of the NRC’s regulations, but also the licensee’s commitments in its applications 

and other correspondence with the NRC. Licensees are encouraged to modify the audit form as 

needed to include items specific to their licensed program (e.g., implementation of a security 

program). The auditor should also evaluate whether the licensee is maintaining exposures to 

workers and the general public as low as is reasonably achievable (ALARA) and, if not, make 

suggestions for improvement. This audit form includes references at the end. 

1. MANAGEMENT OVERSIGHT: 

[Management support to radiation safety, radiation safety committee, radiation safety 

officer (RSO), program audits (including annual reviews of program and ALARA 

reviews), control by authorized users, appropriate follow-up on events, implementation of 

corrective actions from previous audit/inspection findings, and active communication and 

interaction with other committees if conducting human and/or animal research] 

2. AMENDMENTS AND PROGRAM CHANGES: 

(Amendments to the license properly implemented; if applicable, program and 

procedural changes were approved and implemented in accordance with license 

condition) 

3. FACILITIES: 

(Facilities as described in license, uses, control of access, engineering controls, 

calibration facilities, shielding, and air flow) 

4. EQUIPMENT AND INSTRUMENTATION: 

[Operable and calibrated survey and effluent monitoring equipment, and Title 10 of the 

Code of Federal Regulations (10 CFR) Part 21.21 requirements] 

5. MATERIAL USE, CONTROL, AND TRANSFER: 

(Materials and uses authorized, security and control of licensed materials, and 

procedures for receipt and transfer of licensed material) 

6. AREA RADIATION SURVEYS AND CONTAMINATION CONTROL: 

(Radiological surveys, air sampling, leak tests, inventories, handling of radioactive 

materials, contamination controls, records, and public doses) 

7. TRAINING AND INSTRUCTIONS TO WORKERS: 

(Training and retraining requirements and documentation, interviews and observations of 

routine work, staff knowledge of all routine activities, requirements in 10 CFR Part 19 

and 10 CFR Part 20, emergency situations, and supervision by authorized users) 

8. RADIATION PROTECTION: 

(Radiation protection program with ALARA provisions, external and internal dosimetry, 

exposure evaluations, dose and survey records and reports, annual notifications to 

workers, and bulletins and other generic communications) 
F–2 

9. RADIOACTIVE WASTE MANAGEMENT: 

(Disposal; effluent pathways and control; storage areas; transfer; packaging, control, and 

tracking procedures; equipment; incinerators, hoods, vents, and compactors; and license 

conditions for special disposal method) 

10. DECOMMISSIONING: 

(Records relevant to decommissioning; decommissioning plan and schedule, notification 

requirements, cost estimates, funding methods, financial assurance; “timeliness rule” 

requirements, and changes in radiological conditions since decommissioning plan was 

submitted–NUREG–1757, Volume 3, provides guidance on decommissioning) 

11. TRANSPORTATION: 

[Quantities and types of licensed material shipped, special form and packaging design 

requirements and documentation, shipping papers, hazardous materials (HAZMAT) 

communication procedures, return of sources, procedures for monitoring radiation and 

contamination levels of packages, HAZMAT training, and records and reports] 

12. NOTIFICATIONS AND REPORTS: 

(Reporting and follow-up of theft, loss, incidents, and overexposures. Notification of 

change in RSO or authorized user. Radiation exposure reports provided to individuals.) 

13. POSTING AND LABELING: 

(Notices; license documents; regulations; bulletins, and generic information; posting of 

radiation areas; and labeling of containers of licensed material) 

14. INDEPENDENT AND CONFIRMATORY MEASUREMENTS: 

(Areas surveyed, both restricted and unrestricted; measurements made; comparison of 

data with staff’s results and regulations) 

15. AUDIT FINDINGS: 
F–3 

REFERENCES 

A. MANAGEMENT OVERSIGHT 

1. Radiation Safety Committee 

 Applicable license conditions 

2. Radiation Safety Officer 

 Applicable license conditions 

3. Audits, Reviews, or Inspections 

 10 CFR 20.1101, “Radiation protection programs” 

10 CFR 20.2102, “Records of radiation protection programs” 

Applicable license conditions 

4. ALARA 

10 CFR 20.1101, “Radiation protection programs” 

 

5. Authorized Users 

 Applicable license conditions 

B. AMENDMENTS AND PROGRAM CHANGES 

 Applicable license conditions 

C. FACILITIES 

1. Access Control 

 10 CFR 20.1601, “Control of access to high radiation areas” 

 10 CFR 20.1602, “Control of access to very high radiation areas” 

10 CFR 20.1801, “Security of stored material” 

10 CFR 20.1802, “Control of material not in storage” 

10 CFR Part 37, Subpart B, “Background Investigations and Access 

Authorization Program” 

Applicable license conditions 

2. Engineering Controls 

 10 CFR 20.1101, “Radiation protection programs” 

10 CFR 20.1701, “Use of process or other engineering controls” 

10 CFR Part 37, Subpart C, “Physical Protection Requirements During Use” 

Applicable license conditions 

D. EQUIPMENT AND INSTRUMENTATION 

1. Survey Instruments 

 10 CFR 20.1501, “General” 

10 CFR 20.1701, “Use of process or other engineering controls” 

10 CFR 20.2103, “Records of surveys” 

Applicable license conditions 
F–4 

2. Safety Component Defects 

 10 CFR 21.21, “Notification of failure to comply or existence of a defect and its 

 evaluation” 

E. MATERIAL USE, CONTROL, AND TRANSFER 

1. License and Applicable License Conditions 

2. Security and Control 

10 CFR 20.1003, “Definitions” 

10 CFR 20.1801, “Security of stored material” 

10 CFR 20.1802, “Control of material not in storage” 

10 CFR Part 37, Subpart C, “Physical Protection Requirements During Use” 

 

3. Receipt and Transfer of Licensed Material 

 10 CFR 20.1302, “Compliance with dose limits for individual members of the 

 public” 

10 CFR 20.1906, “Procedures for receiving and opening packages” 

10 CFR 20.1501, “General” 

 10 CFR 20.2103, “Records of surveys” 

 10 CFR 20.2207, “Reports of transactions involving nationally tracked sources” 

 10 CFR 30.41, “Transfer of byproduct material” 

 10 CFR 30.51, “Records” 

 10 CFR Part 37, Subpart D, “Physical Protection in Transit” 

 10 CFR 40.51, “Transfer of source or byproduct material” 

 10 CFR 70.42, “Transfer of special nuclear material” 

 10 CFR 74.13, “Material status reports” 

 10 CFR 74.15, “Nuclear material transaction reports” 

F. AREA RADIATION SURVEYS AND CONTAMINATION CONTROL 

1. Area Surveys 

 10 CFR 20.1302, “Compliance with dose limits for individual members of the 

public” 

10 CFR 20.1501, “General” 

10 CFR 20.2103, “Records of surveys” 

10 CFR 20.2107, “Records of dose to individual members of the public” 

Applicable license conditions 

2. Leak Tests and Inventories 

Applicable license conditions 

G. TRAINING AND INSTRUCTIONS TO WORKERS 

1. General 

 10 CFR 19.12, “Instruction to workers” 

Knowledge of 10 CFR Part 20, “Standards for Protection Against Radiation” 

Applicable license conditions 

H. RADIATION PROTECTION 

1. Radiation Protection Program 

a. Exposure evaluation 

10 CFR 20.1501, “General” 
F–5 

b. Programs 

10 CFR 20.1101, “Radiation protection programs” 

2. Dosimetry 

a. Dose Limits 

10 CFR 20.1201, “Occupational dose limits for adults” 

10 CFR 20.1202, “Compliance with requirements for summation of 

 external and internal doses” 

10 CFR 20.1207, “Occupational dose limits for minors” 

10 CFR 20.1208, “Dose equivalent to an embryo/fetus” 

b. External 

10 CFR 20.1203, “Determination of external dose from airborne 

 radioactive material” 

10 CFR 20.1501, “General” 

10 CFR 20.1502, “Conditions requiring individual monitoring of external 

 and internal occupational dose” 

Applicable license conditions 

c. Internal 

10 CFR 20.1204, “Determination of internal exposure” 

10 CFR 20.1502, “Conditions requiring individual monitoring of external 

and internal occupational dose” 

10 CFR Part 20, Subpart H, “Respiratory protection and controls to 

restrict internal exposure in restricted areas” 

3. Source Security Protection 

 10 CFR 37.43, “General security program requirements” 

4. Records 

 10 CFR 20.2102, “Records of radiation protection programs” 

10 CFR 20.2103, “Records of surveys” 

10 CFR 20.2104, “Determination of prior occupational dose” 

10 CFR 20.2106, “Records of individual monitoring results” 

I. RADIOACTIVE WASTE MANAGEMENT 

1. Disposal 

10 CFR 20.1904, “Labeling containers” 

10 CFR 20.2001, “General requirements” 

10 CFR 20.2103, “Records of surveys” 

10 CFR 20.2108, “Records of waste disposal” 

10 CFR 20.2003, “Disposal by release into sanitary sewerage” 

2. Effluents 

a. General 

 Applicable license conditions 

b. Release to septic tanks 

 10 CFR 20.1003, “Definitions” (sanitary sewerage) 

 10 CFR Part 20, Appendix B, Table 2, “Effluent Concentrations” 
F–6 

c. Incineration of waste 

 10 CFR 20.2004, “Treatment or disposal by incineration” 

d. Control of air effluents and ashes 

 10 CFR 20.1201, “Occupational dose limits for adults” 

 10 CFR 20.1301, “Dose limits for individual members of the public” 

 10 CFR 20.1501, “General” 

 10 CFR 20.1701, “Use of process or other engineering controls” 

 Applicable license conditions 

3. Waste Management 

a. General 

10 CFR 20.2001, “General Requirements” 

Information Notice (IN) 90-09, “Extended Interim Storage of Low-Level 

 Radioactive Waste by Fuel Cycle and Materials Licensees” 

b. Waste compacted 

 Applicable license conditions 

c. Waste storage areas 

10 CFR 20.1801, “Security of stored material” 

10 CFR 20.1902, “Posting requirements” 

10 CFR 20.1904, “Labeling containers” 

Applicable license conditions 

d. Packaging, control, and tracking 

10 CFR Part 20, Appendix G, “Requirements for Transfers of Low-Level 

Radioactive Waste Intended for Disposal at Licensed Land Disposal 

Facilities and Manifests” 

10 CFR 20.2006, “Transfer for disposal and manifests” 

 10 CFR 61.55, “Waste classification” 

10 CFR 61.56, “Waste characteristics” 

e. Transfer 

10 CFR Part 20, Appendix G, “Requirements for Transfers of Low-Level 

Radioactive Waste Intended for Disposal at Licensed Land Disposal 

Facilities and Manifests” 

10 CFR 20.2001, “General requirements” 

10 CFR 20.2006, “Transfer for disposal and manifests” 

f. Records 

10 CFR 20.2103, “Records of surveys” 

10 CFR 20.2108, “Records of waste disposal” 

J. DECOMMISSIONING 

 10 CFR Part 20, Subpart E, “Radiological Criteria for License Termination” 

 10 CFR 30.35, “Financial assurance and recordkeeping for decommissioning” 

10 CFR 30.36, “Expiration and termination of licenses and decommissioning of sites 

and separate buildings or outdoor areas” 
F–7 

K. TRANSPORTATION 

1. General 

10 CFR 71.5, “Transportation of licensed material” 

2. Shippers—Requirements for Shipments and Packaging 

a. General requirements 

49 CFR Part 173, “Shippers—general requirements for shipments and 

 packagings,” Subpart I, “Class 7 (Radioactive) Materials” 

49 CFR 173.24, “General requirements for packaging and packages” 

49 CFR 173.448, “General transportation requirements” 

49 CFR 173.435, “Table of A1 and A2 values for radionuclides” 

b. Transport quantities 

 10 CFR 71.4, “Definitions” 

 i. All quantities 

 10 CFR 71.4, “Definitions” 

 49 CFR 173.410, “General design requirements” 

49 CFR 173.431, “Activity limits Type A and Type B 

 packages” 

49 CFR 173.441, “Radiation level limitations and exclusive use 

 provisions” 

 49 CFR 173.443, “Contamination control” 

49 CFR 173.475, “Quality control requirements prior to each 

 shipment of Class 7 (radioactive) materials)” 

49 CFR 173.476, “Approval of special form Class 7 (radioactive) 

 materials” 

 ii. Limited quantities 

49 CFR 173.421, “Excepted packages for limited quantities of 

 Class 7 (radioactive) materials” 

49 CFR 173.422, “Additional requirements for excepted packages 

 containing Class 7 (radioactive) materials” 

 iii. Type A quantities 

49 CFR 173.412, “Additional design requirements for Type A 

 packages” 

49 CFR 173.415, “Authorized Type A packages” 

49 CFR 178.350, “Specification 7A; general packaging, Type A” 

 iv. Type B quantities 

49 CFR 173.416, “Authorized Type B packages” 

49 CFR 173.467, “Tests for demonstrating the ability of Type B 

 and fissile materials packagings to withstand accident 

 conditions in transportation” 

 v. Low specific activity (LSA) material and surface contaminated 

 objects (SCOs) 

49 CFR 173.403, “Definitions” 
F–8 

49 CFR 173.427, “Transport requirements for low specific activity 

 (LSA) Class 7 (radioactive) materials and surface 

 contaminated objects (SCO)” 

vi. International Atomic Energy Agency (IAEA) Code of Conduct 

 Category 1 and 2 materials including Highway Route 

 Controlled quantities as defined in 49 CFR 173.403 or known 

 radionuclides in forms listed as RAM-QC by the Nuclear 

 Regulatory Commission 

 49 CFR 172.802, “Components of a security plan” 

 

c. HAZMAT Communication Requirements 

49 CFR Part 172, “Hazardous Materials Table, Special Provisions, 

 Hazardous Materials Communications, Emergency Response 

 Information, Training Requirements, and Security Plans” 

49 CFR Part 172, Subpart C, “Shipping Papers” 

49 CFR Part 172, Subpart D, “Marking” 

49 CFR Part 172, Subpart E, “Labeling” 

49 CFR Part 172, Subpart F, “Placarding” 

49 CFR Part 172, Subpart G, “Emergency Response Information” 

3. HAZMAT Training 

49 CFR 172.702, “Applicability and responsibility for training and testing” 

49 CFR 172.704, “Training requirements” 

4. Transportation by Public Highway 

49 CFR 171.15, “Immediate notice of certain hazardous materials incidents” 

49 CFR 171.16, “Detailed hazardous materials incident reports” 

49 CFR 177.800, “Purpose and scope of this part and responsibility for 

 compliance and training” 

49 CFR 177.816, “Driver training” 

49 CFR 177.842, “Class 7 (radioactive) material” 

L. NOTIFICATIONS AND REPORTS 

10 CFR 19.13, “Notifications and reports to individuals” 

10 CFR 20.2201, “Reports of theft or loss of licensed material” 

10 CFR 20.2202, “Notification of incidents” 

10 CFR 20.2203, “Reports of exposures, radiation levels, and concentrations of 

 radioactive material exceeding the constraints or limits” 

10 CFR 30.50, “Reporting requirements” 

M. POSTING AND LABELING 

10 CFR 19.11, “Posting of notices to workers” 

10 CFR 21.6, “Posting requirements” 

10 CFR 20.1902, “Posting requirements” 

10 CFR 20.1903, “Exemptions to posting requirements” 

10 CFR 20.1904, “Labeling containers” 

10 CFR 20.1905, “Exemptions to labeling requirements” 
APPENDIX G 

TYPICAL U.S. NUCLEAR REGULATORY COMMISSION (NRC) 

NOTIFICATION AND REPORTING REQUIREMENTS FOR INCIDENTS 

G–1 

Typical U.S. Nuclear Regulatory Commission (NRC) 

Notification and Reporting Requirements for Incidents

Note: The following list of notification and reporting requirements is provided to inform 

licensees about typical notification and reporting requirements that apply to their licensed 

activities. Licensees should note that the list is incomplete in that not all potentially applicable 

requirements have been included. Also, notification and reporting requirements change; 

therefore, licensees should consult the regulations for definitive information about current 

requirements. 

Event

Telephone 

Notification

Written 

Report

Regulatory 

Requirement

Package received with removable 

radioactive surface contamination 

exceeding the limits of 10 CFR 71.87(i) 

or external radiation levels exceeding 

the limits of 10 CFR 71.47 

immediate 

(NRC and 

final delivery 

carrier must 

be notified) 

none 10 CFR 20.1906(d) 

Theft or loss of licensed material immediate 30 days 10 CFR 20.2201(a)(1)(i) 

10 CFR 20.2201(b)(1) 

Whole body dose greater than 0.25 Sv 

[25 rems] 

immediate 30 days 10 CFR 20.2202(a)(1)(i) 

10 CFR 20.2203(a)(1) 

Extremity dose greater than 2.5 Gy 

[250 rads] 

immediate 30 days 10 CFR 20.2202(a)(1)(iii) 

10 CFR 20.2203(a)(1) 

Whole body dose greater than 0.05 Sv 

[5 rems] in 24 hours 

24 hours 30 days 10 CFR 20.2202(b)(1)(i) 

10 CFR 20.2203(a)(1) 

Extremity dose greater than 0.5 Sv 

[50 rems] in 24 hours 

24 hours 30 days 10 CFR 20.2202(b)(1)(iii) 

10 CFR 20.2203(a)(1) 

Whole body dose greater than 0.05 Sv 

[5 rems] 

none 30 days 10 CFR 20.2203(a)(2)(i) 

Dose to individual member of public 

greater than 1 mSv [0.1 rem] 

none 30 days 10 CFR 20.2203(a)(2)(iv) 

Defect in equipment that could create a 

substantial safety hazard 

2 days 30 days 10 CFR 21.21(d)(3)(i) & 

(ii) 

Event that prevents immediate 

protective actions necessary to avoid 

exposures to radiation or radioactive 

materials that could exceed regulatory 

limits 

immediate 30 days 10 CFR 30.50(a) & (c)(2); 

40.60(a) & (c)(2); and 

70.50(a) & (c)(2) 
G–2 

Event

Telephone 

Notification

Written 

Report

Regulatory 

Requirement

Unplanned contamination event that 

requires restricted access for more than 

24 hours and involves a quantity of 

material greater than five times the 

lowest annual limit on intake for the 

material as specified in Appendix B of 

10 CFR Part 20 and requires the area to 

be restricted for a reason other than to 

allow radionuclides with half-lives less 

than 24 hours to decay 

24 hours 30 days 10 CFR 30.50(b)(1) & 

(c)(2); 40.60(b)(1) & 

(c)(2); and 70.50(b)(1) & 

(c)(2) 

Equipment is disabled or fails to function 

as designed when required to prevent 

radiation exposure in excess of 

regulatory limits 

24 hours 30 days 10 CFR 30.50(b)(2) & 

(c)(2); 40.60(b)(2) & 

(c)(2); and 70.50(b)(2) & 

(c)(2) 

Unplanned fire or explosion that affects 

the integrity of any licensed material or 

device, container, or equipment with 

licensed material 

24 hours 30 days 10 CFR 30.50(b)(4) & 

(c)(2); 40.60(b)(4) & 

(c)(2); and 70.50(b)(4) & 

(c)(2) 

Determination that any licensee that has 

not previously implemented the Security 

Orders (i.e., orders issued by the NRC 

to require licensees to implement interim 

security measures) or been subject to 

the provisions of 10 CFR Part 37, 

Subpart C will aggregate radioactive 

material to a quantity that equals or 

exceeds the Category 2 threshold 

none 90 days before 

aggregating 

radioactive 

material to a 

quantity that 

equals or 

exceeds the 

Category 2 

threshold 

10 CFR 37.41(a)(3) 

Coordination with local law enforcement 

agency (LLEA) has failed, either 

because the LLEA has not responded or 

because the LLEA does not plan to 

participate 

3 business 

days 

Submittal of a 

written report 

concerning 

failures of 

coordination 

with LLEA as 

described in 

10 CFR 

37.45(b) is not 

required; 

however, 

licensees must 

document their 

efforts to 

coordinate 

with the LLEA 

and keep this 

documentation 

for 3 years 

10 CFR 37.45(b) 
G–3 

Event

Telephone 

Notification

Written 

Report

Regulatory 

Requirement

Determination that an unauthorized 

entry resulted in an actual or attempted 

theft, sabotage, or diversion of Category 

1 or Category 2 quantity of radioactive 

material 

As soon as 

possible (but 

not at the 

expense of 

causing delay 

or interfering 

with the LLEA 

response), 

but no later 

than 4 hours 

after 

discovery 

30 days 10 CFR 37.57(a)&(c) 

Assessment of any suspicious activity 

related to possible theft, sabotage, or 

diversion of Category 1 or Category 2 

quantities of radioactive material 

As soon as 

possible, but 

no later than 

4 hours after 

notifying the 

LLEA 

none 10 CFR 37.57(b) 

Determination that a shipment 

containing a Category 1 quantity of 

material is lost or missing in transport 

Within 1 hour 

of the 

determination

. Also, notify 

LLEA within 1 

hour of 

determination

30 days and 

periodic 

updates 

10 CFR 37.81(a)&(g) 

Determination that a shipment 

containing a Category 2 quantity of 

material is lost or missing in transport 

Within 4 

hours of the 

determination 

and again 

within 24 

hours if the 

material has 

not yet been 

located and 

secured 

30 days 10 CFR 37.81(b)&(g) 

Discovery along the route of any actual 

or attempted theft or diversion, or 

suspicious activity, related to a Category 

1 quantity of material in transport 

Upon 

discovery, as 

soon as 

possible. 

Also, notify 

LLEA as soon 

as possible 

upon 

discovery 

30 days 

(except no 

report for 

suspicious 

activity) 

10 CFR 37.81(c)&(g) 
G–4 

Event

Telephone 

Notification

Written 

Report

Regulatory 

Requirement

Discovery of any actual or attempted 

theft or diversion, or suspicious activity, 

related to a Category 2 quantity of 

material in transport 

As soon as 

possible 

30 days 

(except no 

report for 

suspicious 

activity) 

10 CFR 37.81(d)&(g) 

Upon recovery of any lost or missing 

Category 1 quantity of material 

As soon as 

possible. 

Also, notify 

LLEA as soon 

as possible 

To be included 

in the 30-day 

report of an 

event 

described in 

10 CFR 

37.81(g), if 

recovered 

during that 

time 

10 CFR 37.81(e)&(h) 

Upon recovery of any lost or missing 

Category 2 quantity of material 

As soon as 

possible 

To be included 

in the 30-day 

report of an 

event 

described in 

10 CFR 

37.81(g), if 

recovered 

during that 

time 

10 CFR 37.81(f)&(h) 

Note: Telephone notifications must be made to the NRC Operations Center at 301-816-5100, 

except as noted. The Center is staffed 24 hours a day and accepts collect calls. 
APPENDIX H 

RADIATION MONITORING INSTRUMENT SPECIFICATIONS 

AND MODEL RADIATION SURVEY INSTRUMENT AND 

AIR SAMPLER CALIBRATION PROGRAM 

H–1 

Radiation Monitoring Instrument Specifications and Model Radiation 

Survey Instrument and Air Sampler Calibration Program 

Radiation Monitoring Instrument Specifications 

The specifications in Table H–1 will help applicants and licensees choose the proper radiation 

detection equipment for monitoring the radiological conditions at their facility(ies). Additional 

information about instruments and their uses also can be found in NUREG–1575, “Multi Agency 

Radiation Survey and Sited Investigation Manual (MARSSIM),” Chapter 6 and Appendix H. 

Table H–1. Typical Survey Instruments* (Instruments Used to Measure Radiological 

Conditions at Licensed Facilities) 

Portable Instruments Used for Contamination and Ambient Radiation Surveys

Detectors Radiation Energy Range Efficiency 

Exposure Rate 

Meters 

Gamma, 

X-ray 

μR-R N/A 

Count Rate Meters 

GM Alpha All energies (dependent on window thickness) Moderate 

 Beta All energies (dependent on window thickness) Moderate 

Gamma All energies < 1% 

NaI Scintillator Gamma All energies (dependent on crystal thickness) Moderate 

Plastic Scintillator Beta C-14 or higher (dependent on window thickness) Moderate 

BF3 Proportional 

Tube* 

Neutron Thermal neutron High 

Stationary Instruments Used to Measure Wipe, Bioassay, and Effluent Samples

Detectors Radiation Energy Range Efficiency 

LSC* Alpha All energies High 

 Beta All energies High 

Gamma Moderate 

Gamma Counter 

(NaI)* 

Gamma All energies High 

Gas Proportional Alpha All energies High 

 Beta All energies Moderate 

Gamma All energies < 1% 

*Table from “The Health Physics and Radiological Health Handbook, Revised Edition,” edited by Bernard Shleien, 

1992 (except for * items). 
H–2 

In addition to selecting an instrument that is appropriate for the radiation(s) of interest, it is 

important to know if the instrument is sufficiently sensitive so as to make measurements at the 

required level. This is particularly important for measurements such as for leak test samples, 

bioassay measurement, and decommissioning of facilities or equipment. The “minimum 

detectable activity” (MDA) for the instrument should be a fraction (10 to 50 percent) of the 

criteria that must be met. 

Example 1: A sealed source is considered to be leaking if a removable contamination exceeds 

185 becquerels (Bq) [0.005 microcurie (μCi), or 11,100 disintegrations per minute (dpm)]. The 

instrument used to measure wipe test samples should have an MDA of 10 percent of that limit, 

or 1,100 dpm for the radionuclide being tested; this is usually easy to detect for cobalt-60 or 

cesium-137, but more difficult to detect for nickel-63, depending on the instrument used to 

analyze the sample. 

Example 2: The licensee is closing a laboratory where uranyl acetate (generally licensed 

pursuant to 10 CFR 40.22, “Small quantities of source material”) was used. The total residual 

contamination screening value for uranium-238 is 101 dpm/100 square centimeters (cm2

). The 

MDA for direct measurements of uranium-238 should be made at 10 to 50 percent of the 

screening value for uranium-238, or 10 to 50 dpm/100 cm2

. 

When the sample count time and the background count time are the same, a simplified 

calculation can be used to determine the MDA for a static measurement. This simplified 

calculation assumes that the Type I error (false positive) and Type II error (false negative) are 

both selected to be equal in probability and at the 95 percent confidence error. 

Note 1: This calculation can be modified for more complex situations as described in 

NUREG–1575, Chapter 6, “Field Measurements Methods and Instrumentation.” 

Note 2: This equation applies only to instruments used in scalar mode, accumulating 

counts of radiation detected over a defined period of time. It is NOT applicable to survey 

instruments used in rate meter mode. 

This simplified equation is: 

MDA =

2.71 + 4.65 ඥbkg × t

t × E

where: MDA = minimum detectable activity in disintegrations per minute (dpm) 

 bkg = background count rate in counts per minute (cpm) 

 t = background counting time and sample counting time in minutes (min) 

 E = detector efficiency in counts per disintegration (c/d) 

 
H–3 

Example: 

A gas-flow proportional counter is used in scalar mode to make 1-minute counts of samples. 

background count rate (bkg) = 300 cpm 

sample counting time (t) = 1 min 

background counting time (t) = 1 min 

efficiency (E) = 0.15 c/d 

MDA = 

2.71 + 4.65 ඥbkg × t

t × E = 

2.71 + 4.65 ඥ300 cpm × 1 min

1 min × 0.15 (c/d) = 555 dpm

According to this calculation, the licensee would be confident that 95 percent of the time, the 

instrument can reliably detect measurements as low as 555 dpm. This is the minimum activity 

that the instrument can detect; results below this number are not reliable at the 95 percent 

confidence interval. However, all numerical results should be recorded. From the basic MDA, 

the licensee can determine the minimum detectable concentration (MDC) for the actual 

measurement conditions. 

For example, suppose the above measurement was made with a radiation survey meter probe 

with a surface area of 15 square centimeters (cm2

); then the MDC would be calculated as 

follows: 

 MDC = 555 dpm/15 cm2

 = 37 dpm/cm2 or 3700 dpm/100 cm2

Determining the MDA or MDC for instruments used in rate meter mode and for scanning 

surveys is more complicated. If the licensee will be performing surveys for decommissioning, 

which require direct measurement surveys, scanning measurement surveys, and surveys for 

removable contamination, they should review NUREG–1757, “Consolidated Decommissioning 

Guidance.” Additional information related to determining the MDA and MDC for direct 

measurements and scanning measurements may be found in Chapter 6 and Appendix H of 

NUREG–1575. 

 
H–4 

Model Radiation Survey Instrument Calibration Program

Training 

Before independently calibrating radiation survey instruments, an individual should complete 

both classroom and on-the-job training as follows: 

 Classroom training may be in the form of lecture, video, computer-based, or self-study 

and will cover the following subject areas: 

— principles and practices of radiation protection 

— radioactivity measurements, monitoring techniques, and the use of radiation 

detection instruments 

— mathematics related to the use and measurement of radioactivity 

— biological effects of radiation 

 On-the-job training will consist of the following: 

— observing authorized personnel performing radiation survey 

instrument calibration 

— conducting radiation survey meter calibrations under the supervision and in the 

physical presence of an individual already authorized to perform calibrations 

Facilities and Equipment for Calibration of Dose and Dose Rate Measuring Instruments 

To reduce doses received by individuals not calibrating radiation survey instruments, 

calibrations will be conducted in an isolated area of the facility or at times when no one else is 

present. 

The calibration source should be well-collimated, and the calibration area should be designed to 

minimize scatter of radiation, which could affect the calibration process. 

The calibration area should be appropriately controlled so that persons entering the area will be 

aware if a radiation source is in use. 

Evaluate posting of the calibration area with appropriate radiation warning signs, as required by 

Subpart J of 10 CFR 20. 

Individuals conducting calibrations of radiation survey instruments will wear assigned dosimetry. 

Individuals conducting calibrations will use a calibrated and operable radiation survey 

instrument to ensure that unexpected changes in exposure rates are identified and corrected. 
H–5 

Frequency of Calibration of Radiation Measurement Instruments and Equipment 

A licensee committed to a routine or emergency radiation survey program should perform an 

acceptable calibration of all radiation measurement instruments and equipment at the frequency 

specified in U.S. Nuclear Regulatory Commission (NRC) regulations, annually, or at the 

frequency recommended by the manufacturer, whichever period is shorter. 

Special calibrations should be performed at any time there is reason to believe that the 

operating characteristics of a radiation measurement instrument have changed, by repair or 

alteration, or whenever system performance is observed to change significantly. 

Routine maintenance of radiation measurement instruments should be performed as 

recommended by the manufacturer. 

Primary or secondary standard instruments used to calibrate radiation measurement 

instruments should be inspected frequently for consistency of performance. 

Calibration Sources for Dose and Dose Rate Measuring Instruments 

A radioactive sealed source(s) will be used for calibrating dose and dose rate measuring 

radiation survey instruments, and this source will have the following characteristics: 

 The source should approximate a point source. 

 Calibration fields from gamma sources should be known with an accuracy when 

compared to secondary or primary national standards of 5 percent for dose rates greater 

than or equal to 1.0 microgray/hour (μGy/h) [0.1 millirad/hour (mrad/h)] and 10 percent 

for dose rates less than 1.0 μGy/h [0.1 mrad/h]. 

 The source should contain a radionuclide that emits radiation of identical or similar type 

and energy as the environment in which the calibrated device will be used. 

 The source should be strong enough to give an exposure rate of at least 

7.7 microcoulomb per kilogram per hour [30 milliroentgen per hour] at 100 centimeters 

{e.g., 3.1 gigabecquerels [(85 mCi (millicuries)] of cesium-137 or 780 megabecquerels 

[21 mCi] of cobalt-60}. 

Note: Inverse square and radioactive decay laws should be used to correct changes in 

exposure rate due to changes in distance or source decay. 

Calibration of Dose or Dose Rate Measuring Instruments 

There are three kinds of scales frequently used on dose and dose-rate survey meters. These 

are calibrated as follows: 

 Linear readout instruments with a single calibration control for all scales should be 

adjusted at the point recommended by the manufacturer or at a point within the normal 

range of use. Instruments with calibration controls for each scale should be adjusted on 

each scale. After adjustment, check the response of the instrument at approximately 20 

percent and 80 percent of full scale. Instrument readings should be within ±x of the 

conventionally true value for the following ranges: 
H–6 

— Background to 10 μGy/h [1.0 mrad/h]; ±x = ±30% 

— 10 μGy/h [1.0 mrad/h] to 1.0 mGy/h [100 mrad/h]; ±x = ±20% 

— 1.0 mGy/h [100 mrad/h] to 10 Gy/h [1,000 Rad/h]; ±x = ±10% 

 Logarithmic readout instruments, which commonly have a single readout scale 

spanning several decades, normally have two or more adjustments. Adjust the 

instrument for each scale according to site specifications or the manufacturer’s 

specifications. After adjustment, check the calibration at a minimum of one point on 

each decade. Instrument readings should have a maximum deviation from the 

conventionally true value as described for linear readout instruments. 

 Digital readout instruments should be calibrated the same as linear 

readout instruments. 

Note: Readings above 2.58 × 10−4

 coulomb/kilogram/hour [1 roentgen/h] need not be 

calibrated, unless the licensee expects to make measurements at higher dose rates; regardless, 

such scales should be checked for operation and response to radiation. 

Calibration of Surface Contamination Measurement Instruments 

Instruments used to detect surface contamination usually consist of a count-rate meter and a 

detector that is appropriate for the type of radiation(s) being measured. 

The efficiency of radiation survey meters must be determined by using radiation sources with 

similar energies and types of radiation that users of the radiation survey instrument intend to 

measure. 

If each scale has a calibration potentiometer, the reading should be adjusted to respond to the 

calibration source at approximately 80 percent of full scale, and the response at approximately 

20 percent of full scale should be observed. If only one calibration potentiometer is available, 

the response should be adjusted at mid-scale on one of the scales, and response on the other 

scales should be observed. The instrument efficiency factor (e.g., cpm/dpm) thus obtained 

should have a signal-to-noise ratio, including the compilation of source and instrument 

uncertainties, of ±x for the following ranges: 

 alpha measurement

0.01 Bq/cm2

 to 2.0 Bq/cm2 [60 to 12,000 dpm/100 cm2]; ±x = ±20% 

 2.0 Bq/cm2

 to 200 Bq/cm2 [12,000 to 1,200,000 dpm/100 cm2]; ±x = ±10% 

 beta measurement

0.05 Bq/cm2

 to 2.0 Bq/cm2 [300 to 12,000 dpm/100 cm2]; ±x = ±20% 

2.0 Bq/cm2

 to 200 Bq/cm2 [12,000 to 1,200,000 dpm/100 cm2]; ±x = ±10% 

Calibration of Analytical Instruments Such As Liquid Scintillation Counters, Gamma 

Counters, Gas Flow Proportional Counters, and Multichannel Analyzers 

Analytical instruments used to determine radioactivity in a sample may be specialized 

equipment according to the type of samples to be analyzed and the types and quantities of 

radioactivity to be measured. Typically, the sample sizes and activities are very small, and can 
H–7 

be difficult to measure. Sample collection and preparation may differ for the various analytical 

instruments, so manufacturer procedures and industry standard practices should be followed. 

Such analytical instruments should be calibrated in accordance with the manufacturer’s 

instructions. Analytical instruments typically require routine maintenance and verification 

procedures to ensure that they are operating properly when used. 

As with calibration of other radiation measurement instruments, calibration of analytical 

instruments use a radioactive sealed source(s). These should be suitable for the geometry of 

the sample(s) to be analyzed. The calibration source(s) should have a known activity(ies) and 

be of similar type and energy as the radioactive materials to be analyzed. The analysis should 

be sensitive enough to detect the lowest levels of radioactivity desired. Correction of results for 

quenching, self-absorption, and other factors may be required, depending on the analytical 

instrument, the samples type, and other environmental conditions. 

Calibration Records 

Calibration records for all radiation survey instruments should indicate the procedure used and 

the results of the calibration. The records should include the following: 

 the owner or user of the radiation survey instrument 

 a description of the radiation survey instrument that includes the manufacturer’s name, 

model number, serial number, and type of detector 

 a description of the calibration source, including the exposure rate at a specified 

distance or activity on a specified date 

 for each calibration point, the calculated exposure rate or count rate, the indicated 

exposure rate or count rate, the deduced correction factor (the calculated exposure rate 

or count rate divided by the indicated exposure rate or count rate), and the scale 

selected on the radiation survey instrument 

 the exposure reading indicated with the radiation survey instrument in the “battery 

check” mode (if available on the instrument) 

 for radiation survey instruments with external detectors, the angle between the radiation 

flux field and the detector (i.e., parallel or perpendicular) 

 for radiation survey instruments with internal detectors, the angle between the radiation 

flux field and a specified surface of the instrument 

 for radiation detectors with removable shielding, an indication of whether the shielding 

was in place or removed during the calibration procedure 

 the exposure rate or count rate from a check source, if used 

 the name and signature of the individual who performed the calibration and the date on 

which the calibration was performed 
H–8 

The following information will be attached to the radiation survey instrument as a calibration 

sticker or tag: 

 for dose and dose rate measuring instruments, the source radionuclide used to calibrate 

the radiation survey instrument (with correction factors) for each scale 

 for surface contamination measurement instruments, the efficiency of the radiation 

survey instrument, for each radionuclide the instrument will be used to measure 

(if efficiency is not calculated before each use) 

 for each scale or decade not calibrated, an indication that the scale or decade was 

checked only for function but not calibrated 

 the date of calibration and the next calibration due date 

 the apparent exposure rate or count rate from the check source, if used 

Air Sampler Calibration 

To assess accurately the air concentration of radioactive materials in a given location, the 

volume of air sampled and the quantity of contaminant in the sample must be determined. 

Accurate determination of the volume of air sampled requires standard, reproducible, and 

periodic calibration of the air metering devices that are used with air sampling instruments. 

Licensees can find guidance on total air sample volume calibration methods acceptable to NRC 

staff in the publication titled “Air Sampling Instruments,” which can be found in the 9th Edition, 

American Conference of Governmental Industrial Hygienists, 2001. This information is 

supplemented below. 

Frequency of Calibration of Air Sampling Equipment 

 A licensee committed to a routine or emergency air sampling program should perform an 

acceptable calibration of all airflow or volume metering devices at least annually 

(See Regulatory Guide 8.25, Rev. 1, “Air Sampling in the Workplace”). 

 Special calibrations should be performed at any time there is reason to believe that the 

operating characteristics of a metering device have been changed, by repair or 

alteration, or whenever system performance is observed to have changed significantly. 

 Routine instrument maintenance should be performed as recommended by the 

manufacturer. 

 Primary or secondary standard instruments used to calibrate air sampling instruments 

should be inspected frequently for consistency of performance. 

Error Limit for Measurement of Air Sample Volume 

Most methods of calibrating airflow or air volume metering devices require direct comparison to 

a primary or secondary standard instrument to determine a calibration curve or a correction 

factor. An example of a primary standard is a spirometer that measures total air volume directly 
H–9 

with high precision by liquid displacement. An example of a secondary standard is a wet-test 

meter that has been calibrated against a primary standard. 

The following are significant errors associated with determining the total air volume sampled: 

EC: The error in determining the calibration factor. (An acceptable estimate is the 

percentage error associated with the standard instrument used in the calibration.)1

ES: Intrinsic error in reading the meter scale. (An acceptable estimate is the percentage 

equivalent of one-half of the smallest scale division, compared to the scale reading.) 

Et: The percentage error in measurement of sampling time that should be kept within 

1 percent. 

EV: The most probable value of the cumulative percentage error in the determination of the 

total air volume sampled. EV can be calculated from the following equation, provided 

there are no additional significant sources of errors: 

 EV = [ES

2

 + EC

2 + Et2

]

1/2

The most probable value of the cumulative error EV, in the determination of total volume, should 

be less than 20%. 

A sample calculation of the most probable value of the cumulative error in total volume 

measured is as follows: If accuracies of the scale reading, the calibration factor, and sample 

time are ± 4, 2, and 1 percent, respectively, and there are no other significant sources of error, 

the cumulative error would be: 

EV = [42

 + 22 + 12]

1/2 = 4.58% or approx. 5% 

If there are significant differences in pressure and temperature between the calibration site and 

the sampling site, appropriate corrections should be made using the ideal gas laws provided 

below: 

Vs = V1 * (P1/760) * (273/T1) 

where Vs = volume at standard pressure and temperature (760 mm Hg and 273̊K 

 V1 = volume measured at conditions P1 and T1

 T1 = temperature of V1 in K 

 P1 = pressure of V1 in mm Hg 

Documentation of Calibration of Air Metering Devices 

 

1

The calibration factor should be based on two kinds of determinations. First, correction factors should be 

determined at several flow rates distributed over the full-scale range. Each flow rate correction factor should be 

determined while adjusting flow rates upscale and again while adjusting flow rates downscale, and the two sets of 

data should be compared. Second, subsequent calibrations should compare the new correction factors to those 

determined during the previous calibration. If observed differences are significant compared to the overall volume 

error limit of 20 percent, licensees should include an additional error term in the calculation above. 
H–10 

The licensee should maintain records of all routine and special calibrations of airflow or volume 

metering devices, including the primary or secondary standard used, method employed, and 

estimates of accuracy of the calibrated metering devices. All instruments should be clearly 

labeled as to the date and results of the most recent calibration and should include the 

appropriate correction factors to be used. 

References: 

 Regulatory Guide 8.25, Revision 1, “Air Sampling in the Workplace,” June 1992 

 NUREG–1400, “Air Sampling in the Workplace,” September 1993 (available at the 

ADAMS Accession No. ML13051A671) 

 Health Physics and Radiological Health, 4th Edition. Edited by Thomas E. Johnson and 

Brian Kent Birky, 2012 

 American National Standards Institute (ANSI) N323AB-2013, “American National 

Standard for Radiation Protection Instrumentation Test and Calibration, Portable Survey 

Instruments” 

 “Air Sampling Instruments,” American Conference of Governmental Industrial Hygienists, 

9th Edition, 2001
APPENDIX I 

MATERIAL RECEIPT AND ACCOUNTABILITY 

I–1 

Material Receipt and Accountability 

The broad scope licensee is authorized to possess only the radionuclides in the types and forms 

listed on the license, and the total quantity possessed under the license must not exceed the 

maximum possession limit listed on the license. Therefore, the radiation safety officer (RSO) 

must know how much material is possessed under the license, in all locations, at any time. The 

licensed inventory includes all radioactive materials in use, in storage, and in waste. The 

regulations in Title 10 of the Code of Federal Regulations (10 CFR) 30.51, 40.61, 70.51 and 

10 CFR Part 74 require the licensee to maintain records of receipt, transfer, and disposal of all 

licensed materials. 

Sample Procedure for Ordering and Receiving Radioactive Material 

 The RSO should approve or place all orders for radioactive material and should ensure 

that the requested material, quantities, manufacturer, and model are authorized by the 

license and do not exceed the possession limits. 

 During normal working hours, carriers should be instructed to deliver radioactive 

packages directly to the RSO (or designated receiving area). 

 During off-duty hours, security or other designated trained personnel should accept 

delivery of radioactive packages in accordance with the procedure outlined in the sample 

memorandum below: 

Sample Memorandum 

Memorandum for Security Personnel 

From: RSO, President, Vice President, etc. 

Subject: Procedures for Receipt of Packages Containing Radioactive Material 

If the package appears to be damaged, immediately contact the RSO. Ask the carrier to 

remain at the facility until it can be determined that neither the carrier nor the vehicle is 

contaminated. 

Any packages containing radioactive material that arrive between [state times 

(e.g., 4:30 p.m. and 7:00 a.m. or on Saturdays or Sundays) must be signed for by the 

security guard (or other designated trained individual] on duty and taken immediately to the 

designated receiving area. Security personnel (or other designated trained individual) 

should unlock the door, place the package in the designated secured storage area, and 

re-lock the door. 

Radiation Safety Officer: 

Office Phone: 

Home Phone: 
I–2 

Sample Instructions to Personnel Involved in Material Receipt 

Shipping and Receiving Personnel 

During normal working hours, immediately upon receipt of any package of licensed material, 

each package must be visually inspected for any signs of shipping damage, such as crushed 

or punctured containers or signs of dampness. Any obvious damage must be reported to the 

RSO immediately. Do not touch any package suspected of leaking. Request the person 

delivering the package to remain until monitored by the RSO. 

Outside of normal working hours (e.g., nights, weekends, and holidays), deliveries usually will 

be handled by security personnel (or other trained individuals), as described in the above 

procedures. Because certain packages of licensed material will have detectable external 

radiation, they should be sent immediately to a designated storage area, where they will be 

checked for contamination and external radiation level as soon as practical. They should not 

be allowed to remain in the receiving area any longer than necessary, as they may be a 

source of exposure for receiving personnel. 

If the instructions are not clear, or if there are questions regarding receiving packages 

containing radioactive material, please contact: 

Name 

Phone 

For additional information on worker training, see the section titled, “Training for Individuals 

Working In or Frequenting Restricted Areas.” 

Materials Possessed Under a General License or Received from a General Licensee 

Individuals at a licensee’s facility may receive and use material pursuant to a general license as 

authorized in 10 CFR Parts 31, 40, or 70. Generally licensed materials are distributed by 

manufacturers authorized by the U.S. Nuclear Regulatory Commission (NRC) to distribute 

materials directly to the persons who will use them under a general license. Some common 

items include nickel-63 sources in electron capture detectors in certain gas chromatographs, 

tritium gas contained in self-luminous exit signs, calibration sources in liquid scintillation 

counters, and uranyl acetate used for staining electron microscope samples. The licensee 

should develop a policy for how their institution will require responsible use and tracking of this 

material. 

If the licensee possesses generally licensed materials and wishes to transfer them to a specific 

license, they should review the regulations in 10 CFR Parts 30, 31, 40, or 70, as applicable, to 

determine how this may be done. 
I–3 

SAMPLE PROCEDURE FOR SAFELY OPENING PACKAGES CONTAINING 

LICENSED MATERIALS 

For packages received under the specific license, authorized individuals must implement 

procedures for opening each package, as follows: 

 Wear gloves to prevent hand contamination. 

 Approach the package with a radiation survey meter to ensure that no unusual or 

unexpected radiation levels are present. 

 Visually inspect the package for any sign of damage (e.g., crushed, punctured). If 

damage is noted, stop and notify the RSO. 

 Check U.S. Department of Transportation (DOT) White I, Yellow II, or Yellow III label or 

packing slip for activity of contents so shipment does not exceed license possession 

limits. 

 Monitor the external surfaces of a labeled package in accordance with 10 CFR 20.1906. 

 Open the outer package (following supplier’s directions, if provided) and remove packing 

slip. Open inner package to verify contents (compare requisition, packing slip, and label 

on the bottle or other container). Check integrity of the final source container 

(e.g., inspect for breakage of seals or vials, loss of liquid, discoloration of packaging 

material, high count rate on smear). Check again that the shipment does not exceed 

license possession limits. If an authorized individual finds anything other than expected, 

they should stop and notify the RSO. 

 Survey the packing material and packages for contamination before discarding. If 

contamination is found, treat as radioactive waste. If no contamination is found, 

obliterate the radiation labels prior to discarding in the regular trash. 

 Maintain records of receipt, package survey, and wipe test results. 

 Notify the final delivery carrier and the NRC Operations Center, 301-816-5100, by 

telephone when removable radioactive surface contamination exceeds the limits of 

10 CFR 71.87(i) or the external radiation levels exceed the limits of 10 CFR 71.47. 

Sample Transfer Policy Statements 

Internal Transfers 

Licensed materials that may be transferred from one department or laboratory or authorized 

user’s control to another should have prior approval from the RSO. A written transfer procedure 

should be developed by the RSO to ensure that transfers are done in accordance with the 

conditions of the license. All transfers must be done in a way that minimizes the probability of 

spillage or breakage. Double containers should be used, including suitable shielding, for such 

transfers. 

 
I–4 

External Transfers 

Licensed material should not be transferred or shipped from one institution to another without 

the approval of the RSO. Such transfers/shipments must be packaged and labeled in 

accordance with DOT, NRC, or U.S. Postal Service Regulations, whichever is applicable. Prior 

to any transfer from the license, the licensee must verify that the recipient is authorized to 

receive the licensed material, as required by 10 CFR 30.41, 40.51 and 70.42. 

Gifts 

On occasion, licensees may be offered or have donated licensed materials by other individuals 

as gifts (e.g., a retiring medical practitioner donating his cesium needles to the university 

medical center). All such gifts of radioactive materials must be transferred to the licensee and 

handled in accordance with NRC requirements and the conditions of the license. In any case, 

the RSO should approve the gift prior to the transfer and package receipt and accountability 

procedures should be followed and documented as for all licensed material. 
APPENDIX J 

PUBLIC DOSE 

J–1

Public Dose 

This appendix describes methods for determining radiation doses to members of the public. 

Licensees must ensure that: 

 The radiation dose received by individual members of the public does not exceed 

1 millisievert (mSv) [100 millirem (mrem)] in 1 calendar year resulting from the licensee’s 

possession and/or use of licensed materials. 

 The radiation dose in unrestricted areas does not exceed 0.02 mSv (2 mrem) in any 

one hour. 

 Air emissions of radioactive materials do not result in doses greater than 0.1 mSv 

(10 mrem) per year total effective dose equivalent (TEDE). As required by Title 10 of the 

Code of Federal Regulations (10 CFR) 20.1101(d), if the licensee exceeds the 10 mrem 

per year air emission dose constraint, the licensee must report the exceedance as 

provided in 10 CFR 20.2203, and promptly take appropriate corrective action to ensure 

against recurrence. 

Members of the public include persons who live, work, study, or may be near locations where 

licensed material is used or stored and employees whose assigned duties do not include the 

use of byproduct material but may work in the vicinity where such materials are used or stored. 

Doses to Members of the Public

INCLUDE doses from: 

 radiation and/or radioactive material 

released by a licensee 

 sources of radiation under the control of a 

licensee 

 effluents from sources of licensed 

radioactive materials 

 licensed material in transportation or 

storage at the licensee’s facility 

DO NOT INCLUDE doses from: 

 sanitary sewerage discharges from licensee 

action taken in accordance with 

10 CFR 20.2003 

 natural background radiation 

 medical administration of radioactive material 

including patients released under 10 CFR 

35.75 

 voluntary participation in medical research 
J–2

As defined in 10 CFR 20.1003, the term unrestricted area means “an area, access to which is 

neither limited nor controlled by the licensee.” For purposes of this definition in 20.1003, an 

“unrestricted area” is an area where access is neither limited nor controlled by the licensees for 

purposes of limiting exposures to radiation and radioactive materials. An “unrestricted area” for 

purposes of 20.1003 may be controlled for other purposes, such as for security purposes (see, 

for ex., 10 CFR 20.1801 and 20.1802), and still be considered an “unrestricted area” as long as 

it is not required to be controlled for limiting exposure to radiation and radioactive materials. 

Typical unrestricted areas may include offices, shops, areas outside buildings, property, and 

storage areas for non-radioactive materials, and other facilities and laboratories where licensed 

materials are not normally used or stored. 

The licensee may show compliance with the annual dose limit for individual members of the 

public by: 

 demonstrating by measurement or calculation that the TEDE to the individual likely to 

receive the highest dose at the boundary of the unrestricted area does not exceed 

1 mSv [100 mrem] in a year 

 demonstrating that the annual average concentration of radioactive material released in 

gaseous and liquid effluents at the boundary of the unrestricted area does not exceed 

the values specified in Table 2, “Effluent Concentrations,” of Appendix B, to 

10 CFR Part 20 {The licensee also should show that if an individual were continuously 

present in an unrestricted area, the dose from external sources would not exceed 

0.02 mSv [2 mrem] in any one hour and 0.5 mSv [0.05 rem] in a year.} 

To perform a dose assessment, the licensee should identify all potential sources of external and 

internal radiation exposure to members of the public and all locations of use, transport, and 

storage of radioactive material at the facility. The licensee must then take radiation 

measurements or perform calculations to demonstrate compliance. 

Measurements 

The licensee may use measurements to demonstrate that the average annual releases are 

within regulatory limits, as well as demonstrate that the TEDE to the individual likely to receive 

the highest dose at the boundary of the unrestricted area does not exceed 1 mSv [100 mrem] in 

a year. These measurements may include: 

 dose rate surveys for radiation exposures from external radiation sources 

 measurements of radionuclides in air and water effluent 

The method used to measure dose will depend on the nature of the radiation source. If the 

source of radiation is constant, it may be adequate to measure the dose rate and integrate it 

over time. If the source of radiation differs or changes over time, it may be necessary to 

perform continuous measurements. 

Radioactivity releases may be determined by effluent monitoring or by effluent sampling and 

analysis. Airborne effluents may be discharged when volatile materials are used, such as 

during iodinations, but the discharge itself usually is not continuous since volatile materials often 

are used periodically rather than continuously. Liquid effluents may be discharged continuously 
J–3

or may be stored and subsequently discharged on a batch basis. For each type of source and 

for each route of potential exposure, consider the location of measurement points, whether 

continuous or periodic monitoring is required, the frequency of sampling and measurement, and 

any additional information. For discharges of airborne radionuclides, for example, it may be 

necessary to obtain information on the efficiency of filters and the air flow rate of the discharge 

system, as well as meteorological data and the distance to the nearest individual member of the 

public. 

Calculation Method 

Using a calculation method, the licensee must determine the highest dose an individual is likely 

to receive at the boundary of the unrestricted area. The licensee must take into account the 

individual’s exposure from external sources and the concentration of radionuclides in gaseous 

and liquid releases. In practice, the licensee may wish to make conservative assumptions to 

simplify the dose calculation. 

The public dose limit applies to the individual who is likely to receive the highest dose from 

licensed operations. Therefore, the dose calculations must consider the location with the 

potential for the highest internal and external exposures. An extremely conservative calculation 

would assume that the individual was continuously present 24 hours a day, 365 days a year, or 

an occupancy factor of 1 (see Table J–1). If the result of the calculation using an occupancy 

factor of 1 demonstrates that the public dose limit is not exceeded, then there is no need for 

further evaluation. 

If, however, the licensee would rather choose a more realistic assumption of the individual’s 

occupancy at the points of highest internal and external exposures, then the licensee may use 

the occupancy factors in Table J–1 or may calculate a specific occupancy factor by determining 

the likely fraction of time that the individual is present. 

Table J–1. Standard Occupancy Factors 

Occupancy Factor Description

1 Work areas such as offices, laboratories, shops, and occupied space in 

nearby buildings or outdoor areas 

1/4 Corridors, lounges, elevators using operators, unattended parking lots 

1/16 Waiting rooms, rest rooms, stairways, unattended elevators, janitor’s 

closets, outside areas used only for pedestrians or vehicular traffic 

Records 

In accordance with 10 CFR 20.2107, the licensee must maintain records to demonstrate 

compliance with the dose limit for individual members of the public, until the Commission 

terminates the license. In general, survey and monitoring records of ambient radiation and 

effluent radioactivity should be adequate. 

Records demonstrating the dose to an individual member of the public should identify the 

instruments used in the survey, the name of the surveyor, the date of the survey, the location of 

the survey(s) including a description or drawing of the area surveyed, survey results, and if 

applicable, the occupancy factors used and justification for their use. In addition, records 
J–4

demonstrating the dose to an individual member of the public that involve effluent sampling 

analysis should include information on concentrations of specific radionuclides, minimum 

detectable activity of the system, and the estimated uncertainty of measurements.
APPENDIX K 

GENERAL TOPICS FOR SAFE USE OF RADIONUCLIDES 

AND MODEL EMERGENCY PROCEDURES 

K–1

General Topics for Safe Use of Radionuclides 

and Model Emergency Procedures

General Topics for Safe Use of Radionuclides 

Each laboratory or area where radioactive material is used or stored should have general rules 

so that workers know what is required. Typical instructions should include: 

 Wear a laboratory coat or other protective clothing at all times in areas where licensed 

materials are used. 

 Wear disposable gloves at all times when handling licensed materials. 

 After each procedure or before leaving the area, monitor hands, shoes, and clothing for 

contamination in a low-background area. 

 Do not eat, drink, smoke, or apply cosmetics in any area where licensed material is 

stored or used. 

 Do not store food, drink, or personal effects in areas where licensed material is stored or 

used. 

 Wear personnel monitoring devices, if required, at all times while in areas where 

licensed materials are used or stored. 

 Dispose of radioactive waste only in designated, labeled, and properly shielded 

receptacles. 

 Never pipette by mouth. 

 Store radioactive solutions in clearly labeled containers. 

 Safely handle sealed sources. 

 Secure all licensed material when it is not under the constant surveillance and 

immediate control of the user(s). 

Security of Radioactive Materials 

 Licensed materials in use in controlled or unrestricted areas must be under constant 

surveillance 

 Licensed materials will be secured by one or more of the following methods: 

— storing and using licensed materials only in restricted areas 

— limiting access to an entire facility or building or portion of the building to radiation 

workers 

— providing storage areas that can be locked to prevent access to the licensed 

material 
K–2

— implementing procedures that require a radiation worker to be within “line of 

sight” of the materials whenever licensed materials are in use 

Radionuclide-specific Procedures 

Licensees should develop written procedures for use of different radionuclides so that users 

know the appropriate types of shielding, protective clothing, survey instruments, surveys, and 

decontamination activities that are required. Safety procedures may vary, depending on the 

chemical form of the radionuclide. See examples of such procedures included below. 

Example 1: 

If requesting more than 37 megabecquerels (MBq) [(1 millicurie (mCi)] of unbound (“free”) 

iodine-125 or iodine-131, special safety instructions should be provided to users, including the 

following: 

 a mandatory radiation survey and wipe test for radioactive contamination after each use 

 bioassay procedures for individuals working with mCi quantities of unbound radioiodine or 

volatile compounds labeled with radioiodine 

 the use of vented hoods for iodination procedures and for the storage of millicurie quantities 

of potentially volatile forms radioiodine 

 a dry run prior to the performance of unfamiliar procedures to preclude unexpected 

complications. In addition, the U.S. Nuclear Regulatory Commission (NRC) recommends 

that the radiation safety officer (RSO) be present during new procedures. 

 procedures for measuring the concentration of radioiodine effluents from the hoods 

Example 2: 

If requesting more than 37 MBq [1 mCi] of phosphorus-32, special safety instructions should be 

provided to users, including the following: 

 the use of shielding of low atomic number material, such as high-density plastic, to keep 

bremsstrahlung radiation to a minimum 

 a mandatory radiation survey and wipe test for radioactive contamination after each use 

 the use of extremity monitors for procedures that involve 37 MBq [1 mCi] or more 

 a dry run prior to the performance of unfamiliar procedures to preclude unexpected 

complications. In addition, the NRC recommends that the RSO be present during new 

procedures. 

 the use of eye protection for procedures that involve 370 MBq [10 mCi] or more 
K–3

Sample Procedures for Handling Emergencies 

Licensees should not neglect, delay, or ignore appropriate first aid and other immediate medical 

needs of injured individuals suspected of having been contaminated. 

General Safety Procedures to Handle Spills 

 Name and telephone number of RSO or an alternate person(s) should be posted 

conspicuously in areas of use so that they are readily available to workers in case of 

emergencies. Licensee should have emergency equipment readily available for 

handling spills. Spill kits should include the following: 

— disposable gloves 

— housekeeping gloves 

— disposable lab coats 

— disposable head coverings 

— disposable shoe covers 

— roll of absorbent paper with plastic backing 

— masking tape 

— plastic trash bags with twist ties 

— “radioactive material” labeling tape 

— marking pen 

— pre-strung “radioactive material” labeling tags 

— box of wipes 

— instructions for “emergency procedures” 

— clipboard with copy of the radioactive spill report form for the facility 

— pencil 

— appropriate survey instruments, including batteries (for radiation survey 

meters) 

The decision to implement a major spill procedure instead of a minor spill procedure depends 

on many incident-specific variables, such as the number of individuals affected, other hazards 

present, the likelihood of spread of contamination, the types of surfaces contaminated, and the 

radiotoxicity of the spilled material. For some spills of short-lived radionuclides, the best spill 

procedure may be restricted access pending complete decay. The applicant should establish 

criteria for determining when to use the major spill procedure versus the minor spill procedure 

Minor Spills of Liquids and Solids 

 Instructions to Workers 

— Notify persons in the area that a spill has occurred. 

— Prevent the spread of contamination by covering the spill with absorbent paper. 

(Paper should be dampened if solids are spilled.) 

— Clean up the spill, wearing disposable gloves and using absorbent paper. 

— Carefully fold the absorbent paper with the clean side out and place in a plastic 

bag for transfer to a radioactive waste container. Put contaminated gloves and 

any other contaminated disposable material in the bag. 
K–4

— Survey the area with an appropriate low-range radiation survey meter or other 

appropriate technique. Check the area around the spill for contamination. Also 

check hands, clothing, and shoes for contamination. 

— Promptly report the incident to the RSO. 

— Allow no one to return to work in the area unless approved by the RSO. 

— Cooperate with the RSO and the RSO’s staff (e.g., investigation of root cause, 

provision of requested bioassay samples). 

— Follow the instructions of the RSO and the RSO’s staff (e.g., decontamination 

techniques, surveys, provision of bioassay samples, requested documentation). 

 Reminders to RSO 

— Follow up on the decontamination activities and document the results. 

— As appropriate, determine cause and corrective actions needed; consider 

bioassays if licensed material may have been ingested, inhaled, or absorbed 

through the skin. 

— If necessary, notify the NRC. 

Major Spills of Liquids and Solids 

 Instructions to Workers 

— Clear the area. If appropriate, survey all persons not involved in the spill and 

vacate the room. 

— Prevent the spread of contamination by covering the spill with absorbent paper 

(paper should be dampened, if solids are spilled), but do not attempt to clean it 

up. To prevent the spread of contamination, limit the movement of all personnel 

who may be contaminated. 

— Shield the source only if it can be done without further contamination or 

significant increase in radiation exposure. 

— Close the room and lock or otherwise secure the area to prevent entry. Post the 

room with a sign to warn anyone trying to enter that a spill of radioactive material 

has occurred. 

— Notify the RSO immediately. 

— Survey all personnel who could possibly have been contaminated. 

Decontaminate personnel by removing contaminated clothing and flushing 

contaminated skin with lukewarm water and then washing with a mild soap. 

— Allow no one to return to work in the area unless approved by the RSO. 
K–5

— Cooperate with the RSO and/or the RSO’s staff (e.g., investigation of root cause, 

provision of requested bioassay samples). 

— Follow the instructions of the RSO and/or the RSO’s staff (e.g., decontamination 

techniques, surveys, provision of bioassay samples, requested documentation). 

 Reminders to RSO 

— Confirm decontamination of personnel. If decontamination of personnel was not 

fully successful, consider inducing perspiration by covering the area with plastic. 

Then wash the affected area again to remove any contamination that was 

released by perspiration. 

— Supervise decontamination activities and document the results. Documentation 

should include location of surveys and decontamination results. 

— Determine cause and needed corrective actions; consider need for bioassays if 

licensed material is suspected to have been ingested, inhaled, or absorbed 

through or injected under the skin. 

– If necessary, notify the NRC. 

Incidents Involving Radioactive Dusts, Mists, Fumes, Organic Vapors, and Gases 

 Instructions to Workers 

— Notify all personnel to vacate the room immediately. 

— Shut down ventilation system, if possible, unless it is determined that the room 

ventilation system needs to be used to clear the air for access purposes. 

— Vacate the room. Seal the area, if possible. 

— Notify the RSO immediately. 

— Ensure that all access doors to the area are closed and posted with appropriate 

warning signs, or post guards (trained) at all access doors to prevent accidental 

opening of the doors or entry to the area. 

— Survey all persons who could possibly have been contaminated. Decontaminate 

as directed by the RSO. 

— Promptly report suspected inhalations and ingestions of licensed material to the 

RSO. 

— Decontaminate the area only when advised and/or supervised by the RSO. 

— Allow no one to return to work in the area unless approved by the RSO. 

— Cooperate with the RSO and/or the RSO’s staff (e.g., investigation of root cause, 

provision of requested bioassay samples). 
K–6

— Follow the instructions of the RSO and the RSO’s staff (e.g., decontamination 

techniques, surveys, provision and collection of bioassay samples, requested 

documentation). 

 Reminders to RSO 

— Supervise decontamination activities. 

— Perform air sample surveys in the area before permitting resumption of work with 

licensed materials. 

— Provide written directions to potentially contaminated individuals about providing 

and collecting urine, breath, blood, or fecal samples, etc. 

— Consider need for medical exam and/or whole body count before permitting 

involved individuals to return to work with licensed material. 

— Determine cause and corrective actions needed; consider need for bioassays if 

licensed material is suspected to have been ingested, inhaled, or absorbed 

through or injected under the skin. Document incident. 

— If necessary, notify the NRC. 

Minor Fires 

 Instructions to Workers (Licensees should develop procedures and instructions in 

accordance with local fire safety requirements, Occupational Safety and Health 

Administration regulations, etc. and provide training as needed.) 

— Immediately attempt to put out the fire by approved methods (e.g., fire 

extinguisher) if other fire hazards or radiation hazards are not present. 

— Notify all persons present to vacate the area and have one individual immediately 

call the RSO and fire department (as instructed by RSO). 

— Once the fire is out, isolate the area to prevent the spread of possible 

contamination. 

— Survey all persons involved in combating the fire for possible contamination. 

— Decontaminate personnel by removing contaminated clothing and flushing 

contaminated skin with lukewarm water, then washing with a mild soap. 

— In consultation with the RSO, determine a plan of decontamination and the types 

of protective devices and survey equipment necessary to decontaminate the 

area. 

— Allow no one to return to work in the area unless approved by the RSO. 

— Cooperate with the RSO and/or the RSO’s staff (e.g., investigation of root cause, 

provision of requested bioassay samples). 
K–7

— Follow the instructions of the RSO and/or the RSO’s staff (e.g., decontamination 

techniques, surveys, provision of bioassay samples, requested documentation). 

 Reminders to RSO 

— Supervise decontamination activities. 

— If decontamination of personnel was not fully successful, consider inducing 

perspiration by covering the area with plastic. Then wash affected area again to 

remove any contamination that was released by the perspiration. 

— Consult with fire safety officials to ensure that there are no other possibilities of 

another fire starting. 

— Determine cause and needed corrective actions; consider need for bioassays if 

licensed material is suspected to have been ingested, inhaled, or absorbed 

through or injected under the skin. Document incident. 

— If necessary, notify the NRC. 

Fires, Explosions, or Major Emergencies 

 Instructions to Workers 

— Notify all persons in the area to leave immediately. 

— Notify the fire department. 

— Notify the RSO and other facility safety personnel. 

— Upon arrival of firefighters, inform them where radioactive materials are stored or 

where radionuclides were being used; inform them of the present location of the 

licensed material and the best possible entrance route to the radiation area, as 

well as any precautions to avoid exposure or risk of creating radioactive 

contamination by use of high pressure water, etc. 

— Cooperate with the RSO and/or the RSO’s staff (e.g., investigation of root cause, 

provision of requested bioassay samples). 

— Allow no one to return to work in the area unless approved by the RSO. 

— Follow the instructions of the RSO and/or the RSO’s staff (e.g., decontamination 

techniques, surveys, provision of bioassay samples, requested documentation). 

 Reminders to RSO 

— Coordinate activities with facility’s industrial hygienist or environmental health 

and safety office and with local fire department. 
K–8

— Consult with the firefighting personnel and set up a controlled area where the 

firefighters can be surveyed for contamination of their protective clothing and 

equipment after the fire is extinguished. 

— Once the fire is extinguished, advise firefighters not to enter potentially 

contaminated areas where radioactive sources may be present or radiation areas 

until a thorough evaluation and survey are performed to determine the extent of 

the damage to the licensed material use and storage areas. 

— Perform thorough contamination surveys of the firefighters and their equipment 

before they leave the controlled area and decontaminate, if necessary. 

— Supervise decontamination activities. 

— Consider bioassays if licensed material is suspected to have been ingested, 

inhaled, or absorbed through or injected under the skin. Document incident. 

— If necessary, notify the NRC. 

Incidents Involving Sealed Sources 

For an emergency situation that may occur concerning a sealed source that has been exposed 

unintentionally, is unshielded or compromised, the following safety instructions should be 

considered: 

 Immediately secure and post the restricted area; maintain continuous surveillance and 

restrict access to the restricted area. 

 Notify the RSO, RSO designee, and management personnel immediately. 

 Retrieval operations should be supervised by the RSO. 

 No source or suspected source should be handled directly with bare hands. 

 Determine if additional dosimetry will be required during source retrieval. 

 Appropriate survey instruments should be used for the response activity. 

 Expedient methods of reducing unintended exposure to staff and the public, such as 

lead shot bags, sandbags, steel plates, and remote handling devices. 

 The RSO should make required notifications to the NRC. 

Copies of emergency procedures should be provided to all users. Post a current copy in each 

laboratory or other area where radioactive material is used. 

Procedures for Collecting Bioassay Samples 

In the event of an emergency in which an individual may become contaminated and radioactive 

material was taken into the body through skin absorption or other means, or is suspected of 
K–9

having ingested or inhaled radioactive material, an estimate of the amount of material taken into 

the body may be required. The following items should be considered in developing procedures 

for collecting bioassay samples: 

 the type of bioassay that must be performed (direct or indirect) 

 the number of samples or data points to be collected 

 the frequency of sampling (e.g., hourly, daily, weekly, once) 

 the size of the sample to be collected (e.g., 24-hour urine collection) 

 the ease or difficulty of sample collection 

 the need for written instructions to be provided to the sample collector, who may be the 

contaminated individual 

APPENDIX L 

RADIATION SAFETY SURVEY TOPICS

L–1 

Radiation Safety Survey Topics 

This appendix provides applicants and licensees with additional information on surveys, 

including training requirements, survey frequency, contamination limits, and bioassays. 

Training 

Before independently performing radiation surveys, an individual should complete both 

classroom and on-the-job training as follows: 

 Classroom training may be in the form of lecture, video, computer-based, or self-study 

and will cover the following subject areas: 

— principles and practices of radiation protection 

— radioactivity measurements, monitoring techniques, and using instruments 

— usage and basic mathematics and calculations for measuring radioactivity 

— biological effects of radiation 

 Appropriate on-the-job-training consists of the following: 

— observing authorized personnel using survey equipment, collecting samples, and 

analyzing samples 

— using survey equipment, collecting samples, and analyzing samples under the 

supervision and in the physical presence of an individual authorized to perform 

surveys 

Facilities and Equipment 

 To ensure achieving the required sensitivity of measurements, survey samples will be 

analyzed in a low-background area. 

 A gamma counter system with a single or multichannel analyzer can be used to count 

samples containing gamma-emitters (e.g., iodine-125, cesium-137, cobalt-60). 

 A liquid scintillation counting (LSC) system can be used to count samples containing 

most beta-emitters and gamma-emitters (if efficiency is great enough to achieve the 

required sensitivity for measurements). The LSC is the most common instrument used 

for measurement of tritium (H-3) contamination and other low-energy beta-emitters 

commonly used in laboratory research and development, such as carbon-14, sulfur-35, 

and phosphorus-32. 

 Licensees may use a gas-flow proportional counting system to count samples containing 

alpha-emitters, beta-emitters, and gamma-emitters (if efficiency is great enough to 

achieve the required sensitivity for measurements). 

 
L–2 

Ambient Radiation Level Surveys 

 Dose-rate surveys, at a minimum, should be performed in locations where workers are 

exposed to radiation levels that might result in radiation doses in excess of 10 percent of 

the occupational dose limits or where an individual is working in a dose rate of 

0.025 millisievert (mSv) [2.5 millirem/hour (mrem/h)] or more (50 mSv/year divided by 

2,000 h/year). 

 10 CFR 20.1301 requires that the total effective dose equivalent to an individual member 

of the public from the licensed operation does not exceed 1 mSv [0.1 rem] in a year and 

the dose in any unrestricted area from external sources does not exceed 0.02 mSv 

[2 mrem] in any one hour. 

The frequency of ambient surveys depends on the quantity and use of radioactive materials, as 

well as the specific protective facilities, equipment, and procedures that are designed to protect 

the worker and members of the public from external exposure to radiation. While the 

regulations do not specify a specific survey frequency, the licensee must conduct such surveys 

as will ensure that the dose rate limits in Title 10 of the Code of Federal Regulations (10 CFR) 

Part 20 Subparts C and D are not exceeded. 

Contamination Surveys 

Licensees’ contamination surveys should be sufficient to identify areas of contamination that 

might result in doses to workers or to the public. Combined removable and fixed contamination 

should be surveyed using appropriate radiation detection equipment. Removable contamination 

can be detected and measured through a wipe test of the surface, which is counted in an 

appropriate counting instrument, such as a liquid scintillation counter, a sodium iodide or 

germanium gamma counter, or a proportional alpha/beta counter. 

Contamination surveys will be performed: 

 to evaluate radioactive contamination that could be present on surfaces of floors, walls, 

laboratory furniture, and equipment 

 after any spill or contamination event 

 when procedures or processes have changed 

 to evaluate contamination of users and the immediate work area, at the end of the day or 

before leaving the area of use, when licensed material is used 

 in unrestricted areas at frequencies consistent with the types and quantities of materials 

in use but generally not less frequently than quarterly 

 in areas adjacent to restricted areas and in all areas through which licensed materials 

are transferred and temporarily stored before shipment 

 
L–3 

Contamination Survey Frequency 

Personnel should survey for contamination in locations where individuals are working with an 

unsealed form of radioactive material. These surveys should be done at a frequency 

appropriate to the types and quantities of radioactive materials in use. If the activity used is 

greater than or equal to the smallest annual limit on intake (ALI) (for either inhalation or 

ingestion) as identified in 10 CFR Part 20, Appendix B, then documented surveys should be 

performed at least daily and retain records accordance with 10 CFR 20.2103. 

Table L–1 contains suggested contamination survey frequencies based on ALIs. The 

suggested frequency of surveys is based upon the amount of licensed material “in use” at any 

one time at any particular location. If licensed material has not been used for a period of time 

greater than the required survey frequency, then it is considered to be “not in use.” 

Table L–1. Suggested Contamination Survey Frequency 

< 0.1 ALI ≥ 0.1 ALI < 1.0 ≥ 1.0 ALI 

In Use Monthly Weekly Daily 

Not in Use Every 6 Months 

Contamination in Unrestricted Areas 

Contamination found in unrestricted areas should be immediately decontaminated to 

background levels. When it is not possible to get to background levels, the licensee must 

ensure that the amounts do not exceed the contamination levels listed in Table L–2, taken from 

the “Guidelines for Decontamination of Facilities and Equipment Prior to Release for 

Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material” 

(August 1987) (ADAMS Accession No. ML030590504). Note that, for the purposes of release 

of facilities for unrestricted use or termination of the license, these values have been 

superseded by 10 CFR 20, Subpart E, “Radiological Criteria for License Termination,” and 

cannot be used for that purpose. In particular, the acceptable contamination levels listed below 

for most alpha emitters exceed the levels which will meet the 10 CFR 20, Subpart E criteria. 

Table L–2 levels can continue to be used for release of equipment and material from licensed 

material facilities during operational activities prior to license termination. (63 FR 64132; 

November 18, 1998) 

Table L–2. Acceptable Surface Contamination Levels 

Nuclide1 Average2, 3 Maximum2, 4 Removable2, 5

U-nat, U-235, U-238, 

and associated decay 

products

83.3 Bq/100 cm2 

(5,000 dpm/100 cm2)

250 Bq/100 cm2 

(15,000 dpm /100 cm2)

16.7 Bq/100 cm2

(1,000 dpm/100 cm2)

Transuranics, Ra-226, 

Ra-228, Th-230, 

Th-228, Pa-231, 

Ac-227, I-125, I-129 

1.7 Bq/100 cm2

(100 dpm/100 cm2) 

5.0 Bq/100 cm2

(300 dpm/100 cm2) 

0.3 Bq/100 cm2

(20 dpm/100 cm2) 

Th-nat, Th-232, Sr-90, 

Ra-223, Ra-224, 

U-232, I-126, I-131, 

I-133 

16.7 Bq/100 cm2

(1,000 dpm/100 cm2) 

50.0 Bq/100 cm2

(3,000 dpm/100 cm2) 

3.3 Bq/100 cm2

(200 dpm/100 cm2) 
L–4 

Table L–2. Acceptable Surface Contamination Levels 

Nuclide1 Average2, 3 Maximum2, 4 Removable2, 5

Beta-gamma emitters 

(nuclides with decay 

modes other than 

alpha emission or 

spontaneous 

fission) except Sr-90 

and others noted 

above. 

83.3 Bq/100 cm2 

(5,000 dpm/100 cm2) 

250 Bq/100 cm2 

(15,000 dpm /100 cm2)

16.7 Bq/100 cm2

(1,000 dpm/100 cm2

) 

1

Where surface contamination by both alpha- and beta-gamma-emitting nuclides exists, the limits established for 

alpha- and beta-gamma-emitting nuclides should apply independently. 

2

As used in this table, disintegrations per minute (dpm) means the rate of emission by radioactive material as 

determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and 

geometric factors associated with the instrumentation. 3

Measurements of average contaminant should not be averaged over more than 1 square meter. For objects of less 

surface area, the average should be derived for each such object. 

4

The maximum contamination level applies to an area of not more than 100 square centimeters (cm2). 5

The amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that area 

with filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on 

the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less 

surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be 

wiped. 

For equipment that is potentially contaminated and is to be released for unrestricted use, 

Table L–2 provides the maximum acceptable residual levels for equipment. Additional guidance 

for release of equipment can be found in NUREG–1575, Supplement 1, “Multi-Agency Radiation 

Survey and Assessment of Materials and Equipment Manual (MARSAME).” Table L–2 values 

also may be acceptable criteria for contamination in facilities during facilities in operation. 

A standardized method for smear testing of a relatively uniform area should be used to aid in 

comparing contamination at different times and places. A smear taken from an area of about 

100 cm2

 is acceptable to indicate levels of removable contamination. 

Decommissioning Surveys for Release for Unrestricted Use 

When a facility will be closed and released for unrestricted use, the values in Table L–3 provide 

acceptable residual contamination levels, known as “screening values” for building surfaces. To 

the extent practicable facilities should be decontaminated to below these levels [as low as is 

reasonably achievable (ALARA)]. Surveys should be conducted for both removable 

contamination (not to exceed 10 percent of the values in Table L–3) and for total residual 

contamination before the facilities or equipment are released from restricted to unrestricted use, 

to ensure that they meet the applicable limits. 

 
L–5 

Table L–3. Screening Values for Building Surface Contamination* 

Radionuclide Symbol Screening levels for unrestricted release (dpm/100 cm2)

Hydrogen-3 (Tritium) H-3 1.2 × 108

Carbon-14 C-14 3.7 × 106

Sodium-22 Na-22 9.5 × 103

Sulfur-35 S-35 1.3 × 107

Chlorine-36 Cl-36 5.0 × 105

Manganese-54 Mn-54 3.2 × 104

Iron-55 Fe-55 4.5 × 106

Cobalt-57 Co-57 2.1 × 105

Cobalt-60 Co-60 7.1 × 103

Nickel-63 Ni-63 1.8 × 106

Zinc-65 Zn-65 4.8 × 104

Strontium-90 Sr-90 8.7 × 103

Technetium-99 Tc-99 1.3 × 106

Iodine-129 I-129 3.5 × 104

Cesium-137 Cs-137 2.8 × 104

Europium-152 Eu-152 1.3 × 104

Tungsten-181 W-181 1.1 × 106

Iridium-192 Ir-192 7.4 × 104

*

Screening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. 

For cases when the fraction of removable contamination is undetermined or higher than 0.1, users may assume, for 

screening purposes, that 100 percent of surface contamination is removable; and therefore the screening levels 

should be decreased by a factor of 10. Alternatively, users having site-specific data on the fraction of removable 

contamination (e.g., within 10 percent to 100 percent range) may calculate site-specific screening levels using 

DandD, Version 1 computer code. 

Units are disintegrations per minute per 100 cm2

 (dpm/100 cm2). One dpm is equivalent to 

0.0167 Bq. The screening values represent surface concentrations of individual radionuclides 

that would be deemed in compliance with the 0.25 mSv [25 mrem] in a year unrestricted release 

dose limit in 10 CFR 20.1402, “Radiological criteria for unrestricted use.” For radionuclides in a 

mixture, the “sum of fractions” rule applies; see 10 CFR Part 20, Appendix B, Note 4 for an 

example of the “sum of fractions” calculation. Refer to NUREG–1757, “Consolidated 

Decommissioning Guidance,” for further information on application of the values in this table. 

Table L–3 was derived using the DandD screening code, Version 1, (DandD, v1.0) and its 

default input parameters. Table L–3 provides criteria that permit licensees to demonstrate 

compliance with the unrestricted release dose criterion in the License Termination Rule in 

Subpart E of 10 CFR Part 20. Sites with building surface contamination levels below those 

listed in Table L–3 would be deemed acceptable for release for unrestricted use in accordance 

with the dose criteria in 10 CFR 20.1402, provided that residual radioactivity has been reduced 

to ALARA levels. The table is intended for use as criteria to facilitate license termination for 

many simple routine decommissioning cases without a site-specific dose assessment. For 

facilities with contamination levels above those in Table L–3, additional site-specific dose 

assessments may be necessary, and licensees should refer to NUREG–1757 regarding 

acceptable methods for conducting the appropriate dose assessment, such as using the current 

version of DandD to develop site-specific screening criteria. The most recent version of the 

DandD code can be installed by downloading the self-extracting program file, setup.exe, 

accessed through the Web site: http://www.marssim.com/Dose_Modeling.htm. Links to other 

useful software and guidance documents are also found at that Web site. 
L–6 

Table L–3 does not include screening values for radionuclides that emit alpha particles, or for 

soil contamination. Screening values for radionuclides not listed above may be found in 

“Supplemental Information on the Implementation of the Final Rule on Radiological Criteria for 

License Termination” (63 FR 64132; November 18, 1998) for building surfaces; “Supplemental 

Information on the Implementation of the Final Rule on Radiological Criteria for License 

Termination” (64 FR 68395; December 7, 1999) for soils; and “Use of Screening Values to 

Demonstrate Compliance With the Final Rule on Radiological Criteria for License Termination” 

(65 FR 37186; June 13, 2000), which references Tables 5.19 (surface contamination) and 6.91 

(surface soil) from NUREG/CR–5512, Vol. 3, “Residual Radioactive Contamination from 

Decommissioning, Parameter Analysis, Draft Report for Comment, October 1999.” Tables 5.19 

(surface contamination) and 6.91 (surface soil) are for use in determining acceptable screening 

values are for radionuclides not listed in the first two Federal Register notices. 

The type of surveys to be performed for decommissioning of facilities, and the number and 

locations of survey points or samples to be collected, should be determined using guidance 

found in NUREG–1757. Many broad scope licensees will be able to use the “Simple 

Approaches for Conducting Final Radiological Surveys” found in Appendix B of NUREG–1757, 

Volume 2. If the decommissioning of a facility is too complex to allow use one of the “simple 

approaches,” then a licensee may have to develop a more formal decommissioning plan. 

Survey Record Requirements 

Each survey record will include the following: 

 a diagram of the area surveyed 

 a list of items and equipment surveyed 

 specific locations on the survey diagram where wipe test was taken 

 ambient radiation levels with appropriate units 

 contamination levels with appropriate units 

 make and model number of instruments used 

 background levels 

 name of the person making the evaluation and recording the results and date 

Licensees should record contamination levels observed and procedures followed for incidents 

involving contamination of individuals. The record should include names of individuals involved, 

description of work activities, calculated dose, probable causes (including root causes), steps 

taken to reduce future incidents of contamination, times and dates, and the surveyor’s 

signature. In addition, 10 CFR 30.35(g), 40.36(f) and 70.25(g) state, in part, that records of 

information important to the decommissioning of a facility, including records of spills or other 

unusual occurrences involving the spread of contamination in and around the facility, 

equipment, or site, must be maintained. 

Air Monitoring in the Workplace 

Air monitoring can be used to do the following: 

 determine whether the confinement of radioactive materials is effective 

 measure airborne radioactive material concentrations in the workplace 

 estimate worker intakes of radioactive material 

 determine posting requirements 
L–7 

 determine what protective equipment and measures are appropriate 

 warn of significantly elevated levels of airborne radioactive materials 

If bioassay measurements are used to determine worker doses of record, air sampling may be 

used to determine time of intake and to determine which workers should have bioassay 

measurements. The use of engineering controls and a good air sampling program can 

eliminate the need for bioassays. 

Refer to Regulatory Guide 8.25, Revision 1, “Air Sampling in the Workplace,” dated June 1992, 

and NUREG–1400, “Air Sampling in the Workplace,” dated September 1993, which is available 

in the NRC Agencywide Documents Access and Management System at Accession No. 

ML13051A671, for further guidance on air sampling. 

Airborne Effluent Release Monitoring 

When practicable, airborne radioactive effluents should be released from monitored release 

points (e.g., monitored stacks, discharges, vents) to provide accurate measurements to 

estimate public exposure. Licensees should verify the performance of effluent monitoring 

systems by regular calibration (at least annually) to ensure their reliability. 

Regulatory Guide 4.20, Revision 1,“Constraint on Releases of Airborne Radioactive Materials to 

the Environment for Licensees Other Than Power Reactors,” dated April 2012, provides 

guidance on methods acceptable (calculation or COMPLY code) to NRC for compliance with the 

constraint on air emissions to the environment. 

Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials Facilities,” dated July 1993, 

provides guidance on designing an acceptable program for establishing and maintaining ALARA 

levels for gaseous and liquid effluents at materials facilities. 

For release points for which monitoring is not practicable, the licensee should estimate the 

magnitude of the unmonitored effluents. These unmonitored releases will occur anytime 

unsealed material is handled outside a fume hood or other device that will control the releases. 

The licensee should include these estimates when demonstrating compliance with dose limits 

and ALARA goals. Unmonitored releases may be estimated based on the quantity of material 

used in these areas, the number of procedures performed, or other appropriate methods. The 

unmonitored effluents should not exceed 30 percent of the total estimated effluent releases or 

10 percent of the permissible air effluent concentrations found on column 1 of Table 2 in 

10 CFR Part 20, Appendix B, whichever is greater. 

Effluent monitoring systems should be designed in accordance with ANSI N13.1 (2011), 

“Sampling And Monitoring Releases Of Airborne Radioactive Substances From The Stacks And 

Ducts Of Nuclear Facilities,” and ANSI N42.18 (2004), “Specification and Performance of Onsite Instrumentation for Continuously Monitoring Radioactivity in Effluents.” 

Liquid Effluent Release Monitoring 

The licensee should evaluate the concentrations of radioactive material in water that is released 

to the sanitary sewer and to the environment. The licensee must show that these releases meet 

the limits in 10 CFR 20.2003 and 10 CFR 20.1302, respectively. 

The topic of sanitary sewerage releases is more fully discussed in 10 CFR Part 20, Appendix B. 
L–8 

Bioassay Monitoring 

Frequency of Required Bioassay Measurements 

Determining the appropriate frequency of routine bioassay measurements depends upon the 

exposure potential and the physical and chemical characteristics of the radioactive material and 

the route of entry to the body. Consider the following elements: 

 potential exposure of the individual 

 retention and excretion characteristics of the radionuclides 

 sensitivity of the measurement technique 

 acceptable uncertainty in the estimate of intake and committed dose equivalent 

Bioassay measurements used for demonstrating compliance with the occupational dose limits 

should be conducted often enough to identify and quantify potential exposures and resultant 

intakes that, during any year, are likely to collectively exceed 0.1 times the ALI. The 10 percent 

ALI criterion is consistent with 10 CFR 20.1502(b), which requires licensees to monitor intakes 

and assess occupational doses for exposed individuals who are likely to exceed 10 percent of 

the applicable limit (i.e., intakes likely to exceed 0.1 ALI for adults). 

Separate categories of bioassay measurements, routine measurements and special 

measurements further determine the frequency and scope of measurements. 

Routine Measurements 

Routine measurements include baseline measurements, periodic measurements, and 

termination measurements. These measurements should be conducted to confirm that 

appropriate controls exist and to assess dose. The method of bioassay selected (for example, 

whole body counting, urinalysis, etc.) and the samples collected will vary according to the 

radionuclides and the compound to which it is attached. Sample collection procedures should 

be developed to ensure that appropriate types, sizes, and numbers of samples are collected 

that will provide appropriate physiological information for the dose assessment. 

An individual’s baseline measurement of radioactive material within the body should be 

conducted before beginning work that involves exposure to radiation or radioactive materials for 

which monitoring is required. 

In addition to the baseline measurements, periodic bioassay measurements should be 

performed. The frequency of periodic measurements should be based on the likelihood of 

significant exposure of the individual. In determining the worker’s likely exposure, consider such 

information as the worker’s access, work practices, measured levels of airborne radioactive 

material, and exposure time. Periodic measurements should be made when the cumulative 

exposure to airborne radioactivity since the most recent bioassay measurement is > 0.02 ALI 

(40 derived air concentration hours). Noble gases and airborne particulates with a radioactive 

half-life of less than 2 hours should be excluded from the evaluation, since external exposure 

generally controls these radionuclides. 

At a minimum, periodic measurements should be conducted annually. Periodic measurements 

provide additional information on any long-term accumulation and retention of radioactive 

material in the body, especially for exposures to concentrations of airborne radioactive material 

below monitoring thresholds. 
L–9 

When an individual is no longer subject to the bioassay program because of change in 

employment status, termination bioassay measurement should be made, when practicable, to 

ensure that any unknown intakes are quantified. 

Special Monitoring 

Because of uncertainty in the time of intakes and the absence of other data related to the 

exposure (e.g., physical and chemical forms, exposure duration), correlating positive results to 

actual intakes for routine measurements can sometimes be difficult. Abnormal and inadvertent 

intakes from situations such as a failed respiratory protective device, inadequate engineering 

controls, inadvertent ingestion, contamination of a wound, or skin absorption, should be 

evaluated on a case-by-case basis. When determining whether potential intakes should be 

evaluated, consider the following circumstances: 

 the presence of unusually high levels of facial and/or nasal contamination 

 entry into airborne radioactivity areas without appropriate exposure controls 

 operational events with a reasonable likelihood that a worker was exposed to unknown 

quantities of airborne radioactive material (e.g., loss of system or container integrity) 

 known or suspected incidents of a worker ingesting radioactive material 

 incidents that result in contamination of wounds or other skin absorption 

 evidence of damage to or failure of a respiratory protective device 

References:

 Regulatory Guide 4.20, Revision 1, “Constraint on Releases of Airborne Radioactive 

Materials to the Environment for Licensees Other than Power Reactors,” April 2012 

 Regulatory Guide 8.9, Revision 1, “Acceptable Concepts, Models, Equations, and 

Assumptions for a Bioassay Program,” July 1993 

 Regulatory Guide 8.20, Revision 2, “Applications of Bioassay for Radioiodine,” 

September 2014, ADAMS Accession No. ML14064A060 

 Regulatory Guide 8.23, Revision 1, “Radiation Safety Surveys at Medical Institutions,” 

January 1981 

 Regulatory Guide 8.25, Revision 1, “Air Sampling in the Workplace,” June 1992 

 Regulatory Guide 8.32, “Criteria for Establishing a Tritium Bioassay Program,” July 1988 

 Regulatory Guide 8.34, “Monitoring Criteria and Methods to Calculate Occupational 

Radiation Doses,” July 1992 

 Regulatory Guide 8.37, “ALARA Levels for Effluents from Materials Facilities,” July 1993 

 NUREG–1400, “Air Sampling in the Workplace,” September 1993 
L–10 

 NUREG–1549, Draft Report for Comment, “Decision Methods for Dose Assessment to 

Comply with Radiological Criteria for License Termination,” July 1998, ADAMS 

Accession No. ML993250291 

 NUREG–1575, “Multi-Agency Radiation Survey and Site Investigation Manual 

(MARSSIM),” Revision 1, August 2000 

 NUREG–1575, Supplement 1, “Multi-Agency Radiation Survey and Assessment of 

Materials and Equipment Manual (MARSAME),” January 2009 

 NUREG–1757, “Consolidated Decommissioning Guidance,” 

— Volume 1, Decommissioning Process for Materials Licensees (Revision 2), 

September 2006 

— Volume 2, Characterization, Survey, and Determination of Radiological Criteria 

(Revision 1), September 2006 

 NUREG/CR–4884, “Interpretation of Bioassay Measurements,” July 1987 

 NUREG/CR–5512, Volume 2, “Residual Radioactive Contamination from 

Decommissioning: User’s Manual DandD Version 2.1,” April 2001, 

ADAMS Accession No. ML010940257 

 NUREG/CR–5512, Volume 3, “Residual Radioactive Contamination from 

Decommissioning, Parameter Analysis, Draft Report for Comment,” October 1999 

[containing Tables 5.19 (surface contamination) and 6.91 (surface soil)], 

ADAMS Accession No. ML082460902 

 Federal Register: “Supplemental Information on the Implementation of the Final Rule on 

Radiological Criteria for License Termination,” 63 FR 67132-34, November 18, 1998 

 Federal Register: “Supplemental Information on the Implementation of the Final Rule on 

Radiological Criteria for License Termination,” 64 FR 68395–96, December 7, 1999 

 Federal Register: “Use of Screening Values to Demonstrate Compliance With the Final 

Rule on Radiological Criteria for License Termination,” 65 FR 37186, June 13, 2000 

 ANSI N13.1-2011, “Sampling and Monitoring Releases of Airborne Radioactive 

Substances from the Stacks and Ducts of Nuclear Facilities,” 2011 

 ANSI N13.30-2011, “Performance Criteria for Radiobioassay,” 2011 

 ANSI N42.18-2004, “Specification and Performance of On-site Instrumentation for 

Continuously Monitoring Radioactivity in Effluents,” 2004 

 NCRP Commentary No. 3, “Screening Techniques for Determining Compliance with 

Environmental Standards—Releases of Radionuclides to the Atmosphere,” published in 

January 1989, and the addendum published October 1989 
L–11 

 U.S. Department of Energy, DOE G 441.1-1C, Admin Chg 1, “Radiation Protection 

Programs Guide for Use with Title 10, Code of Federal Regulations, Part 835, 

Occupational Radiation Protection,” July 8, 2011

APPENDIX M 

MODEL LEAK TEST PROGRAM 

M–1 

Model Leak Test Program 

Training 

Before allowing an individual to perform leak testing, the licensee must ensure that he or she 

has sufficient classroom and on-the-job training to show competency in performing leak testing 

and sample analysis independently. 

Classroom training may be in the form of lecture, online, video, hands-on, or self-study and 

should cover the following subject areas: 

 principles and practices of radiation protection 

 radioactivity measurements, monitoring techniques, and instrument use 

 mathematics and calculations used for measuring radioactivity 

 biological effects of radiation 

Appropriate on-the-job-training consists of the following: 

 observing authorized personnel collecting and analyzing leak-test samples 

 collecting and analyzing leak-test samples under the supervision and in the physical 

presence of an individual authorized to perform leak testing and sample analysis 

Facilities and Equipment 

 To ensure achieving the required sensitivity of measurements, analyze leak tests in a 

low-background area. 

 Use a calibrated and operable radiation survey instrument to check leak-test samples 

for gross contamination before they are analyzed. 

 Analyze the leak-test sample using an instrument that is appropriate for the type of 

radiation to be measured [e.g., NaI (Tl) well-counter system for gamma emitters, liquid 

scintillation for beta emitters, and gas-flow proportional counter for alpha emitters]. 

 If the sensitivity of the counting system is unknown, determine the minimum detectable 

activity (MDA). The MDA may be determined using the following formula: 

MDA =

2.71 + 4.65 ඥbkg × t

t × E

where: MDA = minimum detectable activity in disintegrations per minute (dpm) 

 bkg = background count rate in counts per minute (cpm) 

 t = background counting time in minutes 

 E = detector efficiency in counts per disintegration 

 
M–2 

For example: 

where: bkg = 200 cpm 

 E = 0.1 counts per disintegration (10 percent efficient) 

 t = 2 minutes 

MDA = 

2.71 + 4.65 ඥ200 cpm × 2 minutes

2 × 0.1 = 

2.71 + 4.65 √400

0.2

=		

2.71 + 4.65ሺ20ሻ

0.2 		=		

2.71 + 93

0.2 		=		

95.71

0.2

=		

478.55 disintegrations

minute

becquerels ሺBqሻ		=		

1 disintegration

second

MDA =		

478.55 disintegration

minutes

×

minute

60 seconds = 7.976 Bq

Note: The MDA equation shown assumes that counting times for the background 

measurement and for the sample will be equal. MDA equations for non-equal counting times, 

as well as derivations of equations and discussions of limitations, can be found in 

“Decommissioning Health Physics—A Handbook for MARSSIM Users,” Eric W. Abelquist, 

published by Taylor & Francis Group, 2001. 

Frequency for Conducting Leak Tests of Sealed Sources 

Leak tests will be conducted at the frequency specified in the respective Sealed Source and 

Device Registration certificate. If a sealed source is not registered, leak tests should be 

conducted at 6 month intervals, unless a different interval is established during the licensing 

process. Leak testing of sealed sources may be required by license condition. 

Procedure for Performing Leak Testing and Analysis 

 For each sealed source to be tested, list identifying information such as the sealed 

source serial number, manufacturer, model number, radionuclide, and activity. 

 Use a radiation survey meter to monitor exposure. 

 Prepare a separate wipe sample (e.g., cotton swab or filter paper) for each source. 

 Number each wipe to correlate with identifying information for each source. 

 Wipe the most accessible area where contamination would accumulate if the sealed 

source were leaking, but do not wipe the surface of a plated or foil source (see 

manufacturer’s instructions). 

 Select instrumentation that is sensitive enough to detect 185 becquerels 

[0.005 microcurie] of the radionuclide contained in the sealed source. 
M–3 

 Using the selected instrument, count and record background count rate. 

 Check the instrument’s counting efficiency using a standard source of the same 

radionuclide as the source being tested or one with similar energy characteristics. The 

calibration source should be in the same configuration as the sample. Accuracy of 

standards should be within plus or minus 5 percent of the stated value and traceable to 

primary radiation standards such as those maintained by the National Institute of 

Standards and Technology. 

 Calculate the counting efficiency of the detector. 

Efficiency in cpm/Bq =

[(cpm from std) – (cpm from bkg)]

activity of std in Bq

where: cpm = counts per minute 

 std = standard 

 bkg = background 

 Bq = becquerel 

 Count each wipe sample; determine net count rate. 

 For each sample, calculate and record estimated activity in Bq (or millicuries). The activity 

of the sample in becquerels may be calculated using the following formula: 

Activity of sample [Bq] = [(cpm from wipe sample) - (cpm from bkg)]

efficiency in cpm/Bq

 Sign and date the list of sources, data, and calculations. Retain records for 3 years 

[under Title 10 of the Code of Federal Regulations (10 CFR) 20.2103(a)]. 

 If the wipe test activity is 185 Bq [0.005 microcurie] or greater, notify the radiation safety 

officer so that the source can be withdrawn from use and disposed of properly. Also 

notify the U.S. Nuclear Regulatory Commission. 

Reference: See NUREG–1556, Volume 18, “Consolidated Guidance About Materials Licenses: 

Program-Specific Guidance About Service Provider Licenses,” if the applicant wants to provide 

leak testing and sample analysis as a commercial service provider.

APPENDIX N 

U.S. DEPARTMENT OF TRANSPORTATION REGULATIONS 

N–1 

U.S. Department of Transportation Regulations 

Note: The following list of U.S. Department of Transportation (DOT) regulations is provided to 

inform licensees about typical requirements that apply to the transportation of licensed material 

including the preparation of shipments of licensed material. Licensees should note that the list 

is incomplete in that not all potentially applicable requirements have been included. Also, 

transportation requirements change; therefore, licensees should consult the regulations for 

definitive information about current requirements. Additional information on transportation 

requirements may be found at the DOT Web site: http://www.dot.gov/. 

 Table of Hazardous Materials and Special Provisions—49 CFR 172, Subpart B 

— 49 CFR 172.101—Hazardous Materials Table [proper shipping name, hazard 

class, identification number] 

— 49 CFR 172.101—List of Hazardous Substances and Reportable Quantities, 

Table 2 to Appendix A—Radionuclides 

 Shipping Papers—49 CFR 172, Subpart C 

— 49 CFR 172.201—Preparation and retention of shipping papers 

— 49 CFR 172.202—Description of hazardous material on shipping papers 

— 49 CFR 172.203—Additional description requirements 

— 49 CFR 172.204—Shipper’s certification 

 Marking—49 CFR 172, Subpart D 

— 49 CFR 172.300—Applicability 

— 49 CFR 172.301—General marking requirements for non-bulk packagings 

— 49 CFR 172.304—Marking requirements 

— 49 CFR 172.310—Class 7 (radioactive) materials 

— 49 CFR 172.324—Hazardous substances in non-bulk packagings [designation of 

“reportable quantities” with the letters “RQ”] 

 Labeling—49 CFR 172, Subpart E 

— 49 CFR 172.400—General labeling requirements 

— 49 CFR 172.400a—Exceptions from labeling 

— 49 CFR 172.401—Prohibited labeling 

— 49 CFR 172.403—Class 7 (radioactive) material 

— 49 CFR 172.406—Placement of labels 

— 49 CFR 172.436—RADIOACTIVE WHITE-I label 

— 49 CFR 172.438—RADIOACTIVE YELLOW-II label 

— 49 CFR 172.440—RADIOACTIVE YELLOW-III label 

 
N–2 

 Placarding—49 CFR 172, Subpart F 

— 49 CFR 172.500—Applicability of placarding requirements 

— 49 CFR 172.504—General placarding requirements 

— 49 CFR 172.516—Visibility and display of placards 

— 49 CFR 172.556—RADIOACTIVE placard 

 Emergency Response Information—49 CFR 172, Subpart G 

— 49 CFR 172.600—Applicability and general requirements 

— 49 CFR 172.602—Emergency response information 

— 49 CFR 172.604—Emergency response telephone number 

 Training—49 CFR 172, Subpart H 

— 49 CFR 172.702—Applicability and responsibility for training and testing 

— 49 CFR 172.704—Training requirements 

 Safety and Security Plans—49 CFR 172, Subpart I 

— 49 CFR 172.800—Purpose and applicability 

— 49 CFR 172.802—Components of a security plan 

 Shippers—General Requirements for Shipments and Packagings—49 CFR Part 173 

— 49 CFR 173.25—Authorized packagings and overpacks 

— 49 CFR 173.403—Definitions 

— 49 CFR 173.411—Industrial packages 

— 49 CFR 173.412—Additional design requirements for Type A packages 

— 49 CFR 173.413—Requirements for Type B packages 

— 49 CFR 173.415—Authorized Type A packages 

— 49 CFR 173.416—Authorized Type B packages 

— 49 CFR 173.433—Requirements for determining basic radionuclide values, and 

for the listing of radionuclides on shipping papers and labels 

— 49 CFR 173.435—Table of A1 and A2 values for radionuclides 

— 49 CFR 173.441—Radiation level limitations and exclusive use provisions 

— 49 CFR 173.471—Requirements for U.S. Nuclear Regulatory Commission 

approved packages 

— 49 CFR 173.475—Quality control requirements prior to each shipment of Class 7 

(radioactive) materials 

— 49 CFR 173.476—Approval of special form Class 7 (radioactive) materials 

 
N–3 

 Carriage by Public Highway—49 CFR Part 177 

— 49 CFR 177.817—Shipping papers 

— 49 CFR 177.842—Class 7 (radioactive) material [includes requirement for 

blocking and bracing during transport] 

Note: The following reference charts are for reference only and are not a substitute for DOT

and U.S. Nuclear Regulatory Commission transportation regulations. 

 
N–4 

1. Minimum Required Packaging for Class 7 (Radioactive) Material:[1] (49 CFR 173 and 10 CFR 71)[2]

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements. 

Minimum Packaging Required for Radioactive Materials other than Low Specific Activity (LSA) Material and Surface 

Contaminated Objects (SCO) based on Activity of Package Contents 

Radioactive Material Quantity[3] Limited Quantities and 

Articles Type A[4] [9] Type B 

Activity Restrictions ≤ the limits specified in

Table 4 of § 173.425

≤ A1 for special form

≤ A2 for normal form 

> A1 for special form 

> A2 for normal form 

Contents of 

Package 

Non-fissile and 

Fissile Excepted Excepted Package Type A Package Type B(U) or Type B(M) package 

Fissile N/A Type AF[10] package Type B(U)F or Type B(M)F package

Minimum Packaging Required for LSA Material and SCO[5,6]

Type(s) of LSA 

and/or SCO LSA-I LSA-II LSA-III SCO-I SCO-II

Category of 

Package for 

Domestic or 

International 

Transport[7,8]

Unpackaged[8] - - Unpackaged[8] -

IP-1: solids or liquids/exclusive use - - IP-1 -

IP-2: liquids/non-exclusive use IP-2: exclusive use[9] IP-2: exclusive use - IP-2

Specification tank cars or cargo tank 

motor vehicles: liquids/exclusive use

IP-3: liquids or 

gases/non-exclusive 

use[9]

IP-3: non-exclusive 

use - - 

Alternative 

Provisions for 

Domestic only 

Transport[8]

Packaging shall meet the requirements of §§ 173.24, 24a, and 173.410. 

Transportation shall be an exclusive use shipment. 

Activity per shipment must be less than an A2 quantity (see § 173.427(b)(4)). 

[1] Additional provisions may apply for radioactive materials that are pyrophoric, oxidizing, fissile excepted, or uranium hexafluoride. 

[2] Each NRC licensee shall comply with the applicable requirements of the DOT regulations in 49 CFR parts 107, 171 through 180, 

and 390 through 397 (see § 71.5). 

[3] Materials that contain radionuclides, where both the activity concentration and the total activity in the consignment exceed either the 

values specified in the table in § 173.436 or the values derived according to the instructions in § 173.433, must be regulated in 

transport as Class 7 (radioactive) material. 

[4] Except for LSA material and SCO, a Type A package may not contain a quantity of Class 7 (radioactive) material greater than A1 or 

A2 (see § 173.431(a)). See A1 and A2 definitions in § 173.403. 

[5] The external dose rate from LSA material or SCO in a single package may not exceed 10 mSv/h (1 rem/h) at 3 meters from the 

unshielded material or objects (see §§ 173.427(a)(1) and (d)). 

[6] LSA material and SCOs that are or contain fissile material in quantities that are not fissile excepted must be packaged in appropriate 

Type AF or Type BF packages, and not classified as LSA material or SCO. For alternate domestic transport provisions, see 

§ 173.427(b)(4). For comprehensive guidance on packaging and transportation of LSA material and SCO, see NUREG-1608. 

[7] For the quantity of LSA material and SCO transported in a single conveyance, see the limits specified in § 173.427(a)(2). 

[8] LSA material or SCO shall be appropriately packaged in accordance with § 173.427(b) or (d). Certain LSA-I material and SCO-I may 

be transported unpackaged under the conditions in § 173.427(c). 

[9] See §§ 173.411(c) and 173.415(a) for requirements related to package record retention (2 years) and associated documentation of 

physical tests. 

[10] See §§ 71.22(a), 71.23(a) and 173.417(a) for regulations regarding the use of non-AF packages for fissile materials. 

 
N–5 

2. Radiation Level, TI and CSI Limits for Transportation by Mode:[1] (49 CFR 173 - 177, and 10 CFR 71)[10]

Type of Transport Non-exclusive use Exclusive use 

Mode of Transport Road, Rail, Vessel and Air[9] Road and Rail Vessel Air (cargo 

only) 

Radiation Level Limits[2]

Package Surface 2 mSv/h (200 mrem/h)

2 mSv/h (200 mrem/h): other than 

closed vehicles 

10 mSv/h (1000 mrem/h): closed 

vehicles 

2 mSv/h[11] 

(200 mrem/h) 

2 mSv/h 

(200 mrem/h)[3]

Conveyance[4] N/A 

2 mSv/h (200 mrem/h): outer 

surfaces (sides, top and underside) 

of vehicle[5]

N/A N/A 

0.1 mSv/h (10 mrem/h): at any 

point two (2) m (6.6 ft) from sides 

of the vehicle[5]

N/A N/A 

Occupied position N/A 0.02 mSv/h (2 mrem/h): in any 

normally occupied area[6]

Requirements 

of 

§ 176.708 apply 

N/A 

Transport Index (TI) Limits[2]

Package[7]

3: passenger aircraft 

10: road, rail, vessels and cargo 

aircraft 

No limit 10 

Conveyance[4]

50: road, rail and passenger aircraft 

50 to No limit: vessels[8] 

200: cargo aircraft 

No limit 200 

Overpack 

N/A: for road, rail 

50 to 200: vessel[8]

3: passenger aircraft; 

10: cargo aircraft 

N/A No limit[8] N/A 

Criticality Safety Index (CSI) Limit for fissile material[2]

Package[7] 50 100 100 100 

Conveyance[4]

50: road, rail and air 

50: for holds, compartments or 

defined deck areas of vessels[8] 

200 to No limit: for a total vessel[8]

100 

200 to No limit: 

for a total 

vessel[8]

100 

Overpack 50: road, rail, vessels[8] and air N/A

[1] Radiation level, TI, and CSI are defined in § 173.403. 

[2] In addition to any applicable radiation level, TI and CSI limits, separation distance requirements apply to packages, 

conveyances, freight containers and overpacks; to occupied positions; and to materials stored in transit. Separation distances

are based on the sum of the TIs and, for fissile materials, the sum of the CSIs. [see applicable 49 CFR references for: Rail -

§ 174.700; Air – §§ 175.700 through 175.703; Vessel - §§ 176.700 through 176.708; and Highway - § 177.842]. 

[3] Higher package surface radiation levels may be allowed through an approved special arrangement. 

[4] Conveyance is, for transport by public highway or rail, any transport vehicle or large freight container; and for transport by air, 

any aircraft. See definitions in § 173.403. 

[5] The outer surfaces (sides, top and underside) of vehicles are specified for road and rail vehicles in § 173.441. 

[6] For rail, normally occupied areas include the transport vehicle and adjacent rail cars. The 0.02 mSv/h (2 mrem/h) limit does not 

apply to carriers operating under a State or federally regulated radiation protection program where personnel wear radiation 

dosimetry devices. 

[7] Additional TI and CSI limits apply for individual packages when non-fissile radioactive material packages are mixed with fissile 

material packages (see § 173.459). 

[8] For details on TI and CSI limits for transport by vessel, see § 176.708.

[9] Only excepted packages and packages intended for use in research, medical diagnosis, and treatment are permitted on 

passenger aircraft (see §§ 173.448(f) and 175.700). 

[10] The limits in this table do not apply to excepted packages. See the following references for the radiation level limits for: limited 

quantities, § 173.421; instruments and articles, § 173.424; articles containing natural uranium or thorium, § 173.426; or empty 

packaging, § 173.428.

[11] 2 mSv/h (200 mrem/h) other than intermodal transport of closed transport vehicles or exclusive use vessel. 
N–6 

 3. Contamination Limits and Quality Control for Class 7 (Radioactive) Materials: 

(49 CFR 173.443 and 173.475, and 10 CFR 71)

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements. 

Maximum Permissible Limits for Non-fixed Radioactive Contamination on Packages When Offered for Transport

The level of non-fixed (removable) radioactive contamination on the external surface of each package, conveyance, freight container, and overpack offered 

for transport must be kept as low as reasonably achievable, and shall not exceed the values shown in the following table: 

Contaminant 

Maximum permissible limits (§ 173.443(a), Table 9) 

Bq/cm2 μCi/cm2 dpm/cm2

Beta and gamma emitters and low toxicity alpha emitters 4 10-4 240 

All other alpha emitting radionuclides 0.4 10-5 24 

The non-fixed contamination shall be determined by: 

(a) wiping, with an absorbent material using moderate pressure, sufficient areas on the package to obtain a representative sampling of the non-fixed 

contamination; 

(b) ensuring each wipe area is 300 cm2

 in size; 

(c) measuring the activity on each single wiping material and dividing that value by the surface area wiped and the efficiency of the wipe procedure, 

where an actual wipe efficiency may be used, or it may be assumed to be 0.10. 

Alternatively, the contamination level may be determined using alternative methods of equal or greater efficiency. 

A conveyance used for non-exclusive use shipments is not required to be surveyed unless there is reason to suspect that it exhibits contamination (see 

§ 173.443(a)(2)).

Provisions for Control of Contamination on Radioactive Material Packages Offered for Transport and at the Time of Receipt

 When offered for transport, the non-fixed contamination on each package of radioactive material must be kept as low as reasonably achievable and 

may not exceed the limits set forth in § 173.443(a), Table 9 (as shown above). 

 During transport, non-fixed contamination levels on packages transported as exclusive use by rail or highway may not exceed 10 times the limits in 

§ 173.443(a), Table 9 (as shown above).

Provisions for Non-fixed (Removable) Contamination on Excepted and Empty Radioactive Material Packages

 The non-fixed radioactive surface contamination on the external surface of excepted and empty packages shall not exceed the limits specified in 

§ 173.443(a), Table 9 (as shown above). 

 The internal contamination of an empty package must not exceed 100 times the limits in § 173.443(a), Table 9 (as shown above).

Provisions for Non-fixed (Removable) Contamination on Packages and in Rail and Road Vehicles 

used for Exclusive Use Shipments of Radioactive Material

 The levels of non-fixed radioactive contamination on the packages (a) at the beginning of transport, may not exceed the levels prescribed in the above 

table, and (b) at any time during transport, may not exceed ten times the levels prescribed in § 173.443(a), Table 9 (as shown above) [see 

§ 173.443(b)]. 

 Each conveyance, overpack, freight container, or tank used for transporting Class 7 (radioactive) material as an exclusive use shipment that utilizes the 

provisions of § 173.443(b) must be surveyed with appropriate radiation detection instruments after each exclusive use transport. If contamination 

values exceed acceptable levels, the transport vehicle may not be returned to exclusive use transport service, and then only for subsequent exclusive 

use shipment, unless the radiation dose rate at each accessible surface is demonstrated to be 0.005 mSv/h (0.5 mrem/h) or less, and that there is no 

significant non-fixed radioactive surface contamination as specified in § 173.443(a), Table 9 (as shown above) [see § 173.443(c)].

Provisions for Non-fixed (Removable) Contamination in Closed Rail and Road Vehicles 

that are used Solely for the Transportation of Radioactive Material (§ 173.443(d)) 

 The contamination levels must not exceed 10 times the levels prescribed in § 173.443(a), Table 9 (as shown above). 

 Each vehicle is marked with the words “For Radioactive Materials Use Only” in letters at least 76 mm (3 in) high in a conspicuous place on both sides 

of the exterior of the vehicle. 

 The vehicle must meet the placard requirements of Subpart F of Part 172. 

 A survey of the interior surfaces of the empty closed vehicle must show that the radiation dose rate at any point does not exceed 0.1 mSv/h (10 

mrem/h) at the surface or 0.02 mSv/h (2 mrem/h) at 1 m (3.3 feet) from the surfaces. 

 Each vehicle shall be kept closed except for loading or unloading.

Provisions for Quality Control Prior to Each Shipment of Radioactive Material (§ 173.475) 

 Before each shipment of any radioactive materials package, the offeror must ensure, by examination or appropriate tests, that:

(a) the packaging is proper for the contents to be shipped; 

(b) the packaging is in unimpaired physical condition, except for superficial marks; 

(c) each closure device of the packaging, including any required gasket, is properly installed, secured, and free of defects; 

(d) for fissile material, each moderator and neutron absorber, if required, is present and in proper condition; 

(e) each special instruction for filling, closing, and preparation of the packaging for shipment has been followed; 

(f) each closure, valve, or other opening of the containment system is properly closed and sealed; 

(g) each packaging containing liquid in excess of an A2 quantity and intended for air shipment has been tested to show that it will not leak under an 

ambient atmospheric pressure of not more than 25 kPa, absolute (3.6 psia), where the test must be conducted on the entire containment system, 

or on any receptacle or vessel within the containment system, to determine compliance with this requirement; 

(h) the internal pressure of the containment system will not exceed the design pressure during transportation; and 

(i) the external radiation and contamination levels are within the allowable limits specified in §§ 173.441 and 173.443. 
N–7 

4. Hazard Communications for Class 7 (Radioactive) Materials: Shipping Papers (49 CFR 172, Subpart C)

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements.

NOTE: IAEA, IATA/ICAO, and IMO may require additional hazard communication information. [1]

Shipping Paper Entries 

Always Required Sometimes Required Optional Entries

Basic description (in sequence): 

 UN Identification number 

 Proper Shipping Name 

 Hazard Class (7) 

 Maximum activity contained in each 

package in SI units (e.g., Bq, TBq), 

or in both SI and customary units (e.g., 

Ci, mCi) with customary units in 

parentheses following the SI units 

 Number and type of packages 

Additional description: 

 Name of each radionuclide[2]

 Description of physical and 

chemical form (unless special form) 

 “Special form” when not in the 

proper shipping name

 Category of label used 

 Transport index (TI) of each package 

bearing a Yellow-II or Yellow-III label 

Additional entry requirements: 

 24 hour emergency telephone 

number

 Shipper’s Certification shall be 

provided by each person offering 

radioactive material for 

transportation[3]

 Proper page numbering (e.g., Page 1 

of 4) 

Materials-based Requirements: 

 The criticality safety index (CSI) or “Fissile Excepted” for 

fissile material 

 “Highway route controlled quantity” or “HRCQ” for highway 

route controlled quantities 

 The letters “RQ” entered either before or after the basic 

description for each hazardous substance [see § 171.8] 

 Enter applicable subsidiary hazard class(es) in parentheses 

immediately following the primary hazard class when a 

subsidiary hazard label is required 

 A hazardous waste manifest and the word “Waste” 

preceding the proper shipping name is required for 

radioactive material that is hazardous waste 

Package-based Requirements: 

 The applicable DOE or NRC package approval identification 

marking for each Type B(U), Type B(M), or fissile material 

package 

 The International Atomic Energy Agency (IAEA) Certificate of 

Competent Authority identification marking for export 

shipment or shipment in a foreign made package 

Shipment- and Administrative-based Requirements: 

 Specify “exclusive use shipment” as required 

 Specify instructions for maintaining exclusive use controls for 

shipments of LSA material or SCO under exclusive use 

 Specify the notation “DOT–SP” followed by the special 

permit number for a special permit shipment 

 The weight in grams or 

kilograms may be 

inserted instead of 

activity units for fissile 

radionuclides, except for 

Pu-239 and Pu-241 

 The weight in grams of 

Pu-239 and Pu-241 may 

be inserted in addition to 

the activity units 

 Other information is 

permitted provided it 

does not confuse or 

detract from the proper 

shipping name or other 

required information 

Special Considerations/Exceptions for Shipping Papers 

 For shipments of multiple cargo types, any HAZMAT entries must appear as the first entries on the shipping papers, or be entered in a 

color that readily contrasts with any description on the shipping papers or highlighted on the shipping papers in a contrasting color, or be 

designated by an “X” (or “RQ” if appropriate). 

 Emergency response information consistent with §§ 172.600 – 172.606 shall be readily available on the transport vehicle. 

 Shipments of excepted radioactive material in excepted packages, under UN2908, UN2909, UN2910, and UN2911, are excepted from 

shipping paper requirements if (a) the material is not a hazardous substance or hazardous waste and (b) the package does not contain 

fissile material or contain fissile material that is excepted by § 173.453. 

 For road transport, the shipping papers shall be (a) readily available to authorities in the event of accident or inspection, (b) stored within 

the driver’s immediate reach while he is restrained by the lap belt, (c) readily visible to a person entering the driver's compartment or in a 

holder which is mounted to the inside of the door on the driver's side of the vehicle, and (d) either in a holder mounted to the inside of the 

door on the driver's side of the vehicle or on the driver's seat [see § 177.817(e)]. 

[1] International Atomic Energy Agency (IAEA); International Air Transportation Association (IATA); International Civil Aviation 

Organization (ICAO); International Maritime Organization (IMO). 

[2] For mixtures of radionuclides, the radionuclides to be shown must be determined in accordance with § 173.433(g), which is commonly 

known as the 95% rule; abbreviations (symbols) are authorized. 

[3] The Shipper’s certification shall satisfy the requirements of § 172.204. 
N–8 

5. Hazard Communication for Class 7 (Radioactive) Materials: Marking of Packages: 

(49 CFR 172, Subpart D; and 49 CFR 173.471, 178.3 and 178.350) 

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements.

NOTE: IAEA, IATA/ICAO, and IMO may require additional hazard communication information.

Markings on Packages 

Markings Always Required 

Unless Excepted[1] Additional Markings Sometimes Required Optional Markings 

For Non-bulk Packages: 

 Proper shipping name 

 Identification number 

(preceded by “UN” or “NA,” as 

appropriate) 

 Name and address of 

consignor or consignee, unless 

the package is: 

 highway only and no motor 

carrier transfers; or 

 part of a rail carload or 

truckload lot or freight 

container load, and entire 

contents of railcar, truck, or 

freight container are shipped 

from one consignor to one 

consignee 

For Bulk Packages:

 Identification number on 

orange panel or white 

square-on-point display [see 

§§ 172.332 or 172.336]: 

 on each side and each end, 

if the packaging has a 

capacity of 3,785 L 

(1,000 gallons) or more[2], or

 on two opposing sides, if the 

packaging has a capacity of 

less than 3,785 L 

(1,000 gallons)[2] 

Package-based marking requirements:

 Gross mass, including the unit of measurement (which may be 

abbreviated) for each package with gross mass greater than 50 kg 

(110 lb) 

 Package type as appropriate, i.e., “TYPE IP–1,” “TYPE 

IP–2,” “TYPE IP–3,” “TYPE A,” “TYPE B(U)” or “TYPE B(M)”[1]

 Marked with international vehicle registration code of country of origin 

for IP–1, IP–2, IP–3 or Type A package design (e.g., “USA”) 

 Both the name and 

address of consignor 

and consignee is 

recommended. 

 Other markings on 

packages such as 

advertising are 

permitted, but must be 

located away from 

required markings and 

labeling. 

For marking exceptions for 

LSA material and SCO, 

[see 

§ 173.427(a)(6)(vi)] (e.g., 

RADIOACTIVE-LSA, 

RADIOACTIVE-SCO, or 

RQ, as appropriate). 

For an overpack, the 

marking “OVERPACK” in 

lettering 12 mm 

(0.5 inches) high. This 

marking is not required if 

the package type 

contained in the overpack 

is visible from the outside

[see § 173.25].

 Radiation (trefoil) symbol[3] on outside of outermost 

receptacle of each Type B(U) or Type B(M) 

packaging design 

 Each NRC-approved package (e.g., Type AF, Type B(U), Type B(M), 

Type B(U)F, and Type B(M)F) must be marked with the identification 

marking indicated in the package approval

 For Specification 7A packaging, mark on the outside with “USA DOT 7A 

Type A”, and the name and address or symbol of the manufacturer 

satisfying §§ 178.3 and 178.350

Materials-based requirements:

 For a non-bulk IP–1 package containing a liquid, use 

underlined double arrow symbol indicating upright 

orientation[4], where the symbol is placed on two 

opposite sides of the packaging [see § 172.312]

 For a non-bulk package containing a hazardous substance, mark the 

outside of each package with the letters “RQ” in association with the 

proper shipping name 

Administrative-based requirements:

 For each Type B(U), Type B(M) or fissile material package destined for 

export shipment, mark “USA” in conjunction with specification marking, 

or certificate identification; and package identification indicated in the 

U.S. Competent Authority Certificate 

 Mark “DOT–SP” followed by the special permit number assigned for 

each package authorized by special permit 

 Competent authority identification marking and revalidation for foreign 

made Type B(U), Type B(M), Type H(U), Type H(M), or fissile material 

package for which a Competent Authority Certificate is required 

Special Considerations for Marking Requirements 

 All markings are to be (a) on the outside of each package, (b) durable and legible, (c) in English, (d) printed on or affixed to the surface of a 

package or on a label, tag, or sign, (e) displayed on a background of sharply contrasting color, and (f) unobscured by labels or 

attachments. 

 When an overpack is used, see §§ 173.25 and 173.448(g) for marking requirements. 

[1] Some marking exceptions exist for excepted packages, as specified in §§ 173.421, 173.422, 173.424, 173.426 and 173.428. 

[2] If the identification number marking on a bulk package is not visible, the transport vehicle or freight container must be marked on each 

side and each end [see § 172.331]. 

[3] The radiation symbol shall be resistant to the effects of fire and water, plainly marked by embossing, stamping or other means resistant 

to the effects of fire and water and conform to the size requirements of Appendix B to Part 172. 

[4] The arrows must be either black or red on white or other suitable contrasting background and commensurate with the size of the 

package; depicting a rectangular border around the arrows is optional. 
N–9 

6. Hazard Communications for Class 7 (Radioactive) Materials: 

Labeling of Packages (49 CFR 172.400-450)

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements.

NOTE: IAEA, IATA/ICAO, and IMO may require additional hazard communication information.

Requirements for Labels[1]

 Label each package, except for (a) excepted packages of radioactive material; and (b) Low Specific Activity (LSA) material and Surface 

Contaminated Objects (SCO), packaged or unpackaged, when transported under exclusive use controls domestically and when the 

material or object contains less than an A2 quantity. 

 Labels are required to be (a) printed or affixed to a surface other than the bottom of the package, (b) placed near the proper shipping 

name marking, (c) printed or affixed to a background of contrasting color or have a dotted or solid line outer border, (d) clearly visible, 

(e) not obscured by markings or other attachments, (f) representative of the hazardous material content, and (g) in conformance with 

the label specifications of § 172.407. 

 The appropriate radioactive label must be affixed to opposite sides or two ends (other than the bottom) of all non-bulk packages of 

radioactive material. 

Category of Radioactive Labels [3] Other Radioactive Labels[2]

 

White-I Yellow-II Yellow-III Fissile Empty 

Maximum Radiation Surface Level (RSL) Fissile labels required for 

each package containing 

fissile material, other than 

fissile-excepted material; 

and labels must be 

affixed adjacent to 

radioactive category 

labels.

Empty labels required 

for empty Class 7 

(radioactive) 

packages satisfying 

§ 173.428; and any 

previously-used 

labels must not be 

visible.

mSv/h RSL < 0.005 0.005 < RSL < 0.5 0.5 < RSL < 2[5]

mrem/h RSL < 0.5 0.5 < RSL < 50 50 < RSL < 200[5]

Transport Index (TI):[4]

TI = 0 0 < TI < 1 1 < TI < 10[5]

Contents on Labels 

 Each radioactive category label must contain: (a) Except for LSA-I material, the names of the radionuclides in the package where, for 

mixtures of radionuclides, the names listed must be in accordance with the 95% rule specified in 

§ 173.433(g); and, for LSA-I material, the term “LSA-I”; (b) maximum activity in appropriate SI units (e.g., Bq, TBq), or appropriate 

customary units (e.g., Ci, mCi) in parentheses following SI units; and (c) for Yellow-II or Yellow-III labels the Transport Index (TI). 

Abbreviations and symbols may be used. Except for Pu-239 and Pu-241, the weight in g or kg of fissile radionuclides may be inserted 

instead of activity units; for Pu-239 and Pu-241, the weight in g of fissile radionuclides may be inserted in addition to the activity units 

[see § 173.403 for fissile material definition]. 

 Each fissile label must contain the relevant Criticality Safety Index (CSI) [see § 172.403(e)]. 

[1] Additional labels may be required if the contents of a package contains material that also meets the definition of one or more other 

hazard class. See §§ 172.402 and 406(c) for details on additional labeling requirements. [See §§ 172.400a, 173.421 through 

173.427 for details when labels are not required, and see § 172.407 for details on label durability, design, size, color, form 

identification, exceptions, and the trefoil symbol size]. 

[2] A “Cargo Aircraft Only” label is required for each package containing a hazardous material which is authorized for cargo aircraft only 

[see § 172.402(c)]. 

[3] The category of the label must be the higher of the two values specified for RSL and TI [see § 172.403(b)]. 

[4] The TI is determined from the radiation level 1 meter from the package surface [see TI definition in § 173.403]. If the measured TI is 

not greater than 0.05, the value may be considered to be zero. When an overpack is used, it must be labeled in accordance with

§ 72.403(h). 

[5] Packages with a TI > 10 or an RSL > 2 mSv/h (200 mrem/h) must be transported under exclusive use provisions [see § 173.441(b)]. 

Any package containing a Highway Route Controlled Quantity (HRCQ) must be labelled as RADIOACTIVE YELLOW-III. 
N–10 

7. Hazard Communications for Class 7 (Radioactive) Materials: Placarding (49 CFR 172, Subpart F) 

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements.

NOTE: IAEA, IATA/ICAO, and IMO may require additional hazard communication information.

Conditions when Display of Placards is Required [§§ 172.504, 172.507(a), 172.508, and 172.512]

 Each bulk package, freight container, unit load device[1], transport vehicle, or rail car containing any quantity of hazardous material 

must be placarded on each side and each end with the placards specified in § 172.504(e). 

 Radioactive placards are required for: shipments that contain a package labeled as Radioactive Yellow-III; unpackaged LSA-I or 

SCO-I when transported under exclusive use provisions; shipments required by §§ 173.427, 173.441, and 173.457 to be operated 

under exclusive use; and closed vehicles marked “For Radioactive Materials Use Only” transported under § 173.443(d). 

 The Radioactive placard is placed on a square background on any motor vehicle used to transport a package containing a Highway

Route Controlled Quantity (HRCQ) Class 7 (radioactive) material[2]. 

Visibility and Display of Radioactive Placards [§ 172.516]

 Placards are required to: 

• be clearly visible, on a motor vehicle and rail car, from the direction they face, except from the direction of another transport 

vehicle or rail car to which the motor vehicle or rail car is coupled[3]

• be securely attached or affixed thereto or placed in a holder thereon 

• be located clear of appurtenances and devices such as ladders, pipes, doors, and tarpaulins 

• be located, so far as practical, so dirt or water is not directed to it from the transport vehicle wheels 

• be located at least 3 inches (76.0 mm) away from any marking (e.g. advertising) that could reduce its effectiveness 

• have “RADIOACTIVE” printed on it displayed horizontally, reading from left to right 

• be maintained by the carrier so format, legibility, color, and visibility of the placard will not be substantially reduced due to 

damage, deterioration, or obscurement by dirt or other matter 

• be affixed to a background of contrasting color, or have a dotted or solid line outer border which contrasts with the background 

color. 

Radioactive Placards 

PLACARD (FOR OTHER THAN HRCQ) PLACARD FOR HRCQ

White triangular background color in the lower portion 

with yellow triangle in the upper portion; trefoil symbol, 

text, class number and inner and outer borders in black. 

[see § 172.556 and Appendix B of Part 172]

Square background must consist of a white square surrounded by 

one-inch black border. The placard inside the square is identical to 

that for other than HRCQ. 

[see § 172.527]

General Specifications for Placards and Subsidiary Hazard Placarding 

 Placards must conform to the specifications in § 172.519. 

 A CORROSIVE placard is also required for each transport vehicle that contains 454 kg (1001 pounds) or more gross weight of nonfissile, fissile-excepted, or fissile uranium hexafluoride [see § 172.505(b)]. 

 Placards are also required for subsidiary hazards of POISON INHALATION HAZARD, POISON GAS, or DANGEROUS WHEN WET 

[see § 172.505]. 

[1] See § 172.512 for exceptions and variations to the placarding requirements for freight containers and aircraft unit load devices. 

[2] See § 173.403 for the definition of Highway Route Controlled Quantity (HRCQ). A package containing an HRCQ must be labeled 

with RADIOACTIVE Yellow-III labels [see §§ 172.403(c) and 172.507(a)]. 

[3] Required placarding of the front of a motor vehicle may be on the front of a truck-tractor instead of or in addition to the placarding 

on the front of the cargo body to which a truck-tractor is attached § 172.516(b). 
N–11 

8. Requirements/Guidance for Registration, Emergency Response and Action for Class 7 (Radioactive) Materials: 

(49 CFR 107, Subpart G; 49 CFR 171.15; 49 CFR 172, Subparts F and G)

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements. 

Provisions for Persons Who Offer or Transport Class 7 (Radioactive) Materials (49 CFR 107, Subpart G)

 Any person, other than those excepted by § 107.606, who offers for transportation, or transports, in foreign, interstate or intrastate commerce any of the 

following Class 7 (radioactive) materials must satisfy registration and fee requirements of Part 107, Subpart G:

o a highway route-controlled quantity of radioactive material;

o a shipment in a bulk packaging with a capacity > 13,248 L (3,500 gallons) for liquids or gases, or > 13.24 cubic meters (468 cubic feet) for solids; 

or

o any quantity of radioactive material that requires placarding, under provisions of Part 172, Subpart F.

 Any person required to register must submit a complete and accurate registration statement on DOT Form F 5800.2 by June 30th for each registration 

year, or in time to have on file a current Certificate of Registration in accordance with § 107.620. 

 Each registrant or designee must maintain for a period of 3 years from the date of issuance a copy of the registration statement and Certificate of 

Registration issued by PHMSA and must furnish its Certificate of Registration (or a copy thereof) and related records to an authorized representative or 

special agent of DOT upon request. 

 Each motor carrier subject to registration requirements of this subpart must carry a copy of its current Certificate of Registration or another document 

bearing the registration number on board each truck and truck tractor, and the Certificate of Registration or document must be made available, upon 

request, to enforcement personnel. 

 The amount of fees to be paid and procedures to be followed are found at §§ 107.612 and 107.616. 

Provisions for Providing and Maintaining Emergency Response Information (49 CFR 172, Subpart G)

 When shipping papers for the transportation of radioactive materials are required [see Part 172, Subpart C], emergency response information shall

o be provided and maintained during transportation and at facilities where materials are loaded for transportation, stored incidental to transportation, 

or otherwise handled during any phase of transportation; 

o be provided by persons who offer for transportation, accept for transportation, transfer or otherwise handle hazardous materials during 

transportation; 

o be immediately available for use at all times the hazardous material is present; and 

o include and make available the emergency response telephone number [see § 172.604] to any person, representing a Federal, State or local 

government agency, who responds to an incident involving the material or is conducting an investigation which involves the material.

 Emergency response information is information that can be used in mitigating an incident involving radioactive materials. It must contain at least the 

information specified in §§ 172.602 and 172.604; and includes an emergency response telephone number that is monitored at all times the material is 

in transportation by (a) knowledgeable person, or (b) a person who has immediate access to a knowledgeable person, or (c) an organization capable of 

accepting responsibility for providing the necessary detailed information concerning the material.

 Each carrier who transports or accepts for transportation radioactive material for which a shipping paper is required shall instruct, according to the 

requirements of § 172.606, the operator of a conveyance to contact the carrier in the event of an incident involving the material.

Actions to be Taken in the Event of Spillage, Breakage, or Suspected Contamination by Radioactive Material 

 If there is evidence of a leaking package or conveyance, access to the package or conveyance must be restricted, the area impacted and the extent of 

the contamination must be determined, and appropriate measures must be taken to minimize impact to persons and the environment [see 

§ 173.443(e)]. 

 Except for a road vehicle used solely for transporting Class 7 (radioactive) material [see § 173.443(d)], each aircraft used routinely, and each motor 

vehicle used for transporting radioactive materials under exclusive use, must be (a) periodically checked for radioactive contamination, (b) taken out of 

service if contamination levels are above acceptable limits, and (c) remain out of service until the radiation dose rates at accessible surfaces are less 

than 0.005 mSv/h (0.5 mrem/h) and non-fixed radioactive surface contamination levels are below the limits in §§ 173.443(a), Table 9; and 173.443(c) 

for exclusive use vehicle provisions [see Chart 3]. 

 Following any breakage, spillage, release or suspected radioactive contamination incident, any rail or air carrier shall notify, as soon as possible, the 

offeror (i.e. the consignor); special provisions apply for buildings, areas, and equipment that might become contaminated during rail transport. 

Alternative provisions may apply for motor vehicles transporting radioactive materials under exclusive use [see §§ 174.750(a), 175.705(e), and

177.843(b)].

Provisions for Immediate Notification for Reportable Incidents Involving Radioactive Materials (§§ 171.15 and 171.16)

 Each person in physical possession of radioactive material must provide notice in the event of a reportable incident (see § 171.15(b)) as soon as 

practical, but no later than 12 hours after the occurrence of the reportable incident, to the National Response Center (NRC) by telephone at 800–424–

8802 (toll free) or 202–267–2675 (toll call) or online at http://www.nrc.uscg.mil. 

 Each notice must include the information specified in § 171.15(a)(1) – (a)(7). 

 A detailed incident report must also submitted as required by § 171.16.

Guidance on Responding to Emergencies (Emergency Response Guidebook)

 The DOT issues guidance to aid first responders in quickly identifying the hazards of the dangerous goods involved in an accident or 

incident, and for protecting themselves and the general public during the initial response to the accident or incident. For each proper 

shipping name or UN ID Number, the user is led to a specific guide that provides insight into potential hazards and steps to be taken 

for public safety and emergency response. 

 The current edition of the Emergency Response Guidebook is available at http://phmsa.dot.gov/hazmat/outreach-training/erg.
N–12 

9. Requirements for Training and Safety and Security Plans for Class 7 (Radioactive) Materials: 

(49 CFR 172, Subparts H and I, 49 CFR 173, and 10 CFR 37) 

These are basic reference charts; refer to current U.S. DOT and NRC regulations for complete requirements. 

Training (49 CFR 172, Subpart H)

 For any person who is employed by an employer or is self-employed, and who directly affects hazardous materials transportation

safety, a systematic program shall be established to ensure that the person: 

 has familiarity with the general provisions of Part 172, Subpart H;

 is able to recognize and identify radioactive materials;

 has knowledge of specific requirements of Part 172 that are applicable to functions performed by the employee;

 has knowledge of emergency response information, self-protection measures and accident prevention methods and procedures; 

and

 does not perform any function related to the requirements of Part 172 unless instructed in the requirements that apply to that 

function.

 The person shall be trained pursuant to the requirements of § 172.704(a) and (b), may be trained by the employer or by other public 

or private sources, and shall be tested by appropriate means. The training must include the following: 

(a) general awareness training providing familiarity with applicable regulatory requirements; 

(b) function-specific training applicable to functions the employee performs; 

(c) safety training concerning emergency response information, measures to protect the employee from hazards, and 

methods and procedures for avoiding accidents; 

(d) security awareness training providing awareness of security risks and methods designed to enhance transportation 

security; and 

(e) in-depth security training if a security plan is required for the shipment(s) involved. 

 Initial and recurrent training shall comply with the requirements of § 172.704(c). 

 Records of training shall be created and retained in compliance with the requirements of § 172.704(d). 

Security (49 CFR 172, Subpart I, 49 CFR 173, and 10 CFR 37)

 A security plan for hazardous materials that conforms to the requirements of Part 172, Subpart I must be developed and adhered to 

by each person who offers for transportation in commerce or transports in commerce in a motor vehicle, rail car, or freight container 

any of the following radioactive materials: 

(a) IAEA Code of Conduct Category 1 and 2 materials (see §§ 172.800(b)(15) and 10 CFR 37); 

(b) a highway route controlled quantity (HRCQ) of radioactive material as defined in § 173.403 [see § 172.800(b)(15)]; 

(c) known radionuclides in forms listed as radioactive material quantities of concern (RAM–QC) by the NRC [see §§ 172.800(b)(15) 

and 10 CFR 37]; or 

(d) a quantity of uranium hexafluoride requiring placarding under § 172.505(b) [see § 172.800(b)(14)]. 

 The security plan must include an assessment of possible transportation security risks and appropriate measures to address the

assessed risks. 

 Specific measures put into place by the plan may vary commensurate with the level of threat at a particular time. 

 At a minimum, a security plan must address personnel security, unauthorized access, and enroute security. 

 The security plan must be 

(a) in writing; 

(b) retained for as long as it remains in effect; 

(c) available as copies or portions thereof to the employees who are responsible for implementing it, consistent with personnel

security clearance or background investigation restrictions and a demonstrated need to know; 

(d) revised and updated as necessary to reflect changing circumstances; and 

(e) maintained (all copies) as of the date of the most recent revision, when it is updated or revised. 

 Security plans that conform to regulations, standards, protocols, or guidelines issued by other Federal agencies, international 

organizations, or industry organizations may be used to satisfy the requirements in Part 172, provided such security plans address 

the requirements specified in Part 172, Subpart I. 

 Additional security planning requirements may apply for rail transport of a highway route controlled quantity of radioactive material 

[see §§ 172.820 and 173.403]. 
APPENDIX O 

MODEL WASTE MANAGEMENT PROCEDURES 

O–1 

Model Waste Management Procedures 

General Guidelines 

 All radioactivity labels must be defaced or removed from containers and packages prior 

to disposal into ordinary “non-radioactive” waste streams. If waste is compacted, all 

labels that are visible in the compacted mass must be defaced or removed. 

 Remind workers that non-radioactive waste such as leftover reagents, boxes, and 

packaging material should not be mixed with radioactive waste. 

 Occasionally monitor all procedures to ensure that radioactive waste is not created 

unnecessarily. Review all new procedures to ensure that waste is handled in a manner 

consistent with established procedures. 

 In all cases, consider the entire effect of various available disposal routes. Consider 

occupational and public exposure to radiation, other hazards associated with the 

material and routes of disposal (e.g., toxicity, carcinogenicity, pathogenicity, 

flammability), and costs. 

 The waste management program should include waste handling procedures for the 

users within their laboratories or assigned areas, and for waste handlers who may 

collect waste from areas of use to bring to the storage area for eventual disposal. 

 Housekeeping staff should be provided adequate training to avoid the possibility of 

unauthorized disposal or exposure of these individuals to radioactive materials or to 

radiation. 

 A waste generator, collector, or processor who transports, or offers for transportation, 

low-level radioactive waste intended for ultimate disposal at a licensed low-level 

radioactive waste land disposal facility must prepare a Manifest in accordance with 

Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Appendix G, 

“Requirements for Transfers of Low-Level Radioactive Waste Intended for Disposal at 

Licensed Land Disposal Facilities and Manifests.” 

Model Procedure for Disposal by Decay-in-Storage 

Applicants should ensure that adequate space and facilities are available for the storage of 

waste for decay-in-storage (DIS). Licensees can minimize the need for storage space if the 

waste is segregated according to physical half-life. 

 Only short-lived waste (physical half-life of less than or equal to 120 days) may be 

disposed of by DIS. 

 Short-lived waste should be segregated from long-lived waste. 

 Waste should be stored in suitable well-marked containers, and the containers should 

provide adequate shielding. 

 Liquid and solid wastes should be stored separately. 
O–2 

 When the container is full, it should be sealed. The sealed container should be identified 

with a label affixed or attached to it. 

 The identification label should include the date when the container was sealed, the 

longest-lived radionuclide in the container, total activity, and the name of the individual 

who sealed the container. The container may be transferred to the DIS area. When 

large quantities are held for DIS, sufficient quantities may be present even after many 

half-lives and persons performing surveys should be aware of the potential for 

measurable radiation. 

 The contents of the container should be allowed to decay for a period of time after which 

it is expected that the radiation levels would not be distinguishable from background. 

The period of time depends on both the half-life of the radionuclide(s) and the original 

amount present. 

 Prior to disposal as ordinary trash, each container should be monitored with an 

appropriate radiation detection instrument, on the lowest setting, as follows: 

— Check the radiation survey meter for proper operation. 

— Survey the contents of each container in a low background area. 

— Remove any shielding from around the container. 

— Monitor all surfaces of the container. 

— Discard the contents as ordinary trash only if the surveys of the contents indicate 

no residual radioactivity (i.e., surface readings are indistinguishable from 

background). 

— If the surveys indicate residual radioactivity, return the container to DIS area and 

contact the radiation safety officer for further instructions. 

— If the surveys indicate no residual radioactivity, record the date when the 

container was sealed, the disposal date, type of waste (e.g., used or unused 

material, gloves, etc.), survey instrument used, and the name of the individual 

performing surveys and disposing of the waste. 

In accordance with 10 CFR 20.1904(b), all radiation labels must be defaced or removed from 

containers and packages prior to disposal as ordinary trash. Syringes/needles placed into 

sealed waste containers for decay do not need the labels removed, provided that the following 

is done: waste barrels are sealed prior to delivery to the waste disposal firm and delivered 

directly from the licensee’s facility; labels are removed from the waste barrels/containers; waste 

is incinerated, not placed in a landfill; and the waste disposal firm is cautioned not to open the 

container prior to incineration. 

Model Procedure for Disposal of Liquids into Sanitary Sewerage 

 Confirm that the sewer system is a public system, not a private sanitary sewer, septic 

system or leach field. 
O–3 

 Confirm that the liquid waste being discharged is soluble (or is biological material that is 

readily dispersible) in water. 

 Calculate the amount of each radionuclide that can be discharged by using the 

information from prior, similar discharges and the information in 10 CFR Part 20, 

Appendix B. 

 Make sure that the amount of each radionuclide does not exceed the monthly and 

annual discharge limits specified in 10 CFR 20.2003(a)(4) and 10 CFR Part 20, 

Appendix B, Table 3. 

 If more than one radionuclide is released, the sum of the ratios of the average monthly 

discharge of a radionuclide to the corresponding limit in 10 CFR Part 20, Appendix B, 

Table 3 must not exceed unity. 

 Confirm that the total quantity of licensed material released into the sanitary sewerage 

system in a year does not exceed 185 gigabecquerel (GBq) [5 Curies (Ci)] of H-3 

(tritium), 37 GBq [1 Ci] of C-14, and 37 GBq [1 Ci] of all other radionuclides combined. 

 Record the date, radionuclide(s), estimated activity of each radionuclide, location where 

the material is discharged, and the name of the individual discharging the waste. 

 Liquid waste should be discharged only via designated sinks, toilets, or other release 

points. 

 Discharge liquid waste slowly to minimize splashing with water running, to be sure that 

the material moves out of the sink and into the sewer system. 

 Survey the sink and surrounding work surfaces to confirm that no residual material or 

contamination remained in the sink or on work surfaces. 

 Decontaminate all areas or surfaces if found to be contaminated. 

 Maintain records of releases of licensed material to the sanitary sewer system. These 

records should include, for each release, the date, radionuclide(s), estimated activity of 

each radionuclide, location where the material is discharged, and the name of the 

individual discharging the waste. For the licensed facility as a whole, records should be 

maintained of the quantity and concentration of radionuclides that are released into the 

sewer system that demonstrate compliance with the regulatory limits for total quantity 

released and concentrations released by the licensed facility. 

Model Procedure for Incineration 

These guidelines apply to non-commercial waste disposal, i.e., incineration of a licensee’s own 

waste. An applicant or licensee does not need specific U.S. Nuclear Regulatory Commission 

(NRC) approval to incinerate certain categories of radioactive waste. For example, 10 CFR 

20.2005 provides that tritium and carbon-14 in low level concentrations in liquid scintillation 

media and animal tissue may be disposed of without regard to radioactivity. After a licensee 

reviews their program and confirms that they have waste that requires specific NRC approval for 

incineration, please provide the following information: 
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 Describe the training and experience of the person who will be responsible for the 

on-site and day-to-day supervision of incinerator operations. 

 Describe the waste that is proposed to be incinerated, to include: the chemical and/or 

physical form of the waste containing licensed material and a description of how the 

waste is segregated, packaged, and labeled for transfer from the generation site to the 

incinerator; the name of the radionuclide; concentration of radioactivity averaged over 

the weight of the material to be incinerated (microcuries per gram of waste medium) for 

each isotope to be incinerated; and the total radioactivity of each isotope per burn and 

the total number of burns per year. Describe procedures for ensuring that these 

frequencies and activities will not be exceeded. 

 Describe the procedures for packaging, handling, securing, and monitoring of waste to 

prevent contamination and/or unnecessary exposure to personnel or property during the 

waste life cycle. 

 Describe the method for measuring or estimating the concentration of radioactive 

material remaining in the ash residue. Describe the procedures for collection, handling, 

and disposal of the ash residue. 

 Describe the recordkeeping procedures for the waste incineration program. Records 

must be adequate to document all receipts, incinerations, environmental releases of 

effluents, and any disposals of ash generated in the incineration process. These records 

must be maintained in the same units as applicable regulations. 

 Describe the characteristics of the incinerator and site location, including: height of the 

stack, rated air flow (cubic feet per hour or similar units), proximity of the stack or other 

discharge to occupied areas (e.g., residences, school, hospital), and distance to the 

nearest air intake ducts of adjacent buildings. Describe any scrubbers, filters, or air 

cleaning equipment that is present. 

 State how the concentration of radionuclides released, both as airborne effluent and as 

any liquid effluent from scrubbers, condensers, or associated systems, will be measured 

or otherwise determined. Describe any stack monitoring that is planned. 

 Provide a copy of the written safety analysis that demonstrates the applicant will be able 

to incinerate the types and quantities of radioactivity specified in the application without 

exceeding the environmental release limits specified in 10 CFR Part 20. 

 Provide a written commitment that the applicant or licensee has coordinated with 

appropriate State and local authorities and that they have obtained such permits and 

other authorizations as may be necessary. 

 Provide a copy of the radiation safety procedures for monitoring personnel involved in 

incineration operations and for monitoring all effluent generated by the incineration 

process. The procedures must ensure that regulatory limits for environmental releases 

of radioactivity will not be exceeded. The applicant should describe the disposal method 

for any ash generated that exceeds regulatory limits. 
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Model Procedure for Compaction 

The following information should be provided by licensees that propose to compact waste. 

 Describe the compactor to demonstrate that it is adequately designed and manufactured 

to safely compact the type and quantity of waste generated during licensed operations. 

Provide manufacturer’s specifications, annotated sketches or photographs, and other 

information about the compactor design. 

 Describe the type, quantities, and concentrations of waste to be compacted. 

 Provide an analysis of the potential for airborne release of radioactive material during 

compaction activities. 

 Provide the location of the compactor(s) within the waste processing area(s), as well as 

a description of the ventilation and filtration systems used in conjunction with the 

compactors. Include a description of the procedures for monitoring filter blockage and 

exchange. 

 Discuss the methods used to monitor worker breathing zones and/or exhaust systems. 

 Discuss the types and frequencies of surveys that will be performed for contamination 

control in the compactor area. 

 Discuss the instruction provided to compactor operators, including instructions for 

protective clothing, checks for proper functioning of equipment, method of handling 

uncompacted waste, and examining containers for defects. 

APPENDIX P 

COMMENCEMENT OF CONSTRUCTION AT EXISTING AND 

PROPOSED BYPRODUCT (INCLUDING IRRADIATORS), 

SOURCE, AND SPECIAL NUCLEAR MATERIAL FACILITIES WITH 

SIGNIFICANT ENVIRONMENTAL IMPACTS

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Commencement of Construction at Existing and Proposed Byproduct 

(Including Irradiators), Source, and Special Nuclear Material Facilities With 

Significant Environmental Impacts

PURPOSE AND SCOPE 

This appendix provides information concerning the definition of “construction,” and activities that 

may be performed by materials license applicants, potential applicants (hereinafter collectively 

referred to as “applicants”), and licensees before the U.S. Nuclear Regulatory Commission 

(NRC) staff has concluded its environmental review of the proposed licensing action. 

This appendix applies to all Title 10 of the Code of Federal Regulations (10 CFR) Parts 30 

(byproduct material), 36 (irradiators), 40 (source material) and 70 (special nuclear material 

[SNM]) facilities except for uranium recovery facilities, uranium hexafluoride conversion 

facilities, enrichment facilities, fuel fabrication facilities, uranium hexafluoride deconversion 

facilities, and facilities for SNM of greater than critical mass. 

Except as noted in the previous paragraph, this appendix applies to licensing actions related to 

the receipt and possession of licensable byproduct, source, and SNM for the conduct of any 

activity which the NRC determines will significantly affect the quality of the environment. Except 

as noted in the previous paragraph, the information in this appendix should be reviewed by all 

holders of operating licenses for byproduct (including irradiators), source, and SNM facilities, 

and all persons who have submitted applications to construct byproduct, source, and SNM 

facilities, or have submitted letters of intent to submit such applications under 10 CFR Parts 30, 

36, 40, and 70. 

If a licensing action initiated pursuant to 10 CFR Parts 30, 40, or 70 meets any of the criteria in 

10 CFR 51.20 or 51.21, then commencement of construction of a facility before the NRC staff 

has completed its environmental review process is grounds for denial of the license application, 

in accordance with 10 CFR 30.33(a)(5), 40.32(e), and 70.23(a)(7). However, if the licensing 

action meets the criteria in 10 CFR 51.22(c) for a categorical exclusion, and the NRC has not 

determined that an environmental assessment or an environmental impact statement is required 

in accordance with 10 CFR 51.22(b), then commencement of construction before the NRC staff 

concludes the environmental process should not be the sole basis for denial of the license 

application, as the NRC has already determined that this category of actions does not have a 

significant impact on the environment. In accordance with 10 CFR 36.15, commencement of 

construction of an irradiator will only be grounds for denial if the licensee or applicant has not 

submitted both an application and the requisite licensing fee. 

BACKGROUND 

The NRC amended its regulations in September 2011, by revising certain provisions applicable 

to the licensing and approval processes for byproduct (including irradiators), source, and SNM 

licenses in the final rule, “Licenses, Certifications, and Approvals for Materials Licensees” 

(76 FR 56951; September 15, 2011) (Material Licenses Construction Rule). The revisions 

contained in the Material Licenses Construction Rule revised the definitions of “construction” 

and “commencement of construction” for materials licensing actions. 
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The definitions of “commencement of construction” in 10 CFR 30.4, 36.2, 40.4, and 70.4 are 

identical. 

Commencement of construction means taking any action defined as “construction” or any other 

activity at the site of a facility subject to the regulations in this part that has a reasonable nexus 

to: 

1. radiological health and safety; or 

2. common defense and security. 

The definitions of “construction” in 10 CFR 30.4, 36.2, and 70.4 are identical. 

Construction is defined as the installation of foundations, or in-place assembly, erection, 

fabrication, or testing for any structure, system, or component of a facility or activity subject to 

the regulations in this part that are related to radiological safety or security. The term 

“construction” does not include: 

(1) changes for temporary use of the land for public recreational purposes; 

(2) site exploration, including necessary borings to determine foundation conditions or other 

preconstruction monitoring to establish background information related to the suitability 

of the site, the environmental impacts of construction or operation, or the protection of 

environmental values; 

(3) preparation of the site for construction of the facility, including clearing of the site, 

grading, installation of drainage, erosion and other environmental mitigation measures, 

and construction of temporary roads and borrow areas; 

(4) erection of fences and other access control measures that are not related to the safe use 

of, or security of, radiological materials subject to this part; 

(5) excavation; 

(6) erection of support buildings (e.g., construction equipment storage sheds, warehouse 

and shop facilities, utilities, concrete mixing plants, docking and unloading facilities, and 

office buildings) for use in connection with the construction of the facility; 

(7) building of service facilities (e.g., paved roads, parking lots, railroad spurs, exterior utility 

and lighting systems, potable water systems, sanitary sewerage treatment facilities, and 

transmission lines); 

(8) procurement or fabrication of components or portions of the proposed facility occurring 

at other than the final, in-place location at the facility; or 

(9) taking any other action that has no reasonable nexus to: 

(i) radiological health and safety, or 

(ii) common defense and security. 
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“Construction,” as defined in 10 CFR 40.4, also includes the installation of wells associated with 

radiological operations (e.g., production, injection, or monitoring well networks associated with 

in-situ recovery or other facilities). 

The Atomic Energy Act of 1954, as amended, expressly limits the NRC’s regulatory authority to 

matters concerning the radiological public health and safety or common defense and security 

and non-radiological hazards to the extent such hazards result from the actual processing of byproduct material. The NRC has determined that this authority does not extend to site 

preparation activities that do not have a nexus to radiological health and safety or common 

defense and security. 

This appendix should be used to evaluate whether a particular construction activity has a nexus 

to radiological health and safety, and thus falls under the jurisdiction of the NRC for licensing 

purposes. An activity or action has a reasonable nexus to radiological health and safety or the 

common defense and security if that activity or action has a rational, direct link to ensuring that 

a materials facility is operating, or will operate, in accordance with the NRC’s regulations and in 

a manner that protects the public health and safety or the common defense and security from 

radiological hazards. The definitions of construction in 10 CFR 30.4, 36.2, 40.4, and 70.4 

150.31 list activities that are not considered “construction.” This appendix provides examples of 

activities that fall under each of the excepted activities that do not constitute construction. Site 

preparation activities that are not considered “construction,” while not under NRC jurisdiction, 

may be subject to the regulatory authority of another Federal, State, or local agency which may 

require National Environmental Policy Act or state environmental review. NRC’s responsibilities 

under the National Historic Preservation Act of 1966, as amended (NHPA), must also be 

satisfied before a license is issued. Specifically, as noted in the statements by the NRC 

published with the final Material Licenses Construction Rule, under certain circumstances the 

NRC may be required to deny a license application if the NRC determines that the applicant 

intentionally significantly adversely affected, or allowed to be affected, a historic property with 

intent to avoid the requirements of §106 of the NHPA. 

DISCUSSION OF EXAMPLES 

In addition to the background discussion provided above, the following examples clarify the 

delineation of site preparation activities and construction activities. It is important to recognize 

that the NRC may have regulatory authority over activities that can occur before construction 

begins, such as procurement of basic components as defined in 10 CFR Part 21, the process of 

dedicating commercial grade items or basic components, or procurement of items relied on for 

safety (IROFS) as defined in 10 CFR Part 70. It should also be noted that, while site 

preparation activities may not require prior NRC approval, the approval of local, State, or other 

Federal agencies may be required. 

BYPRODUCT MATERIAL (10 CFR PART 30) 

Prior to the conclusion of the environmental review process, applicants for byproduct material 

licenses or license amendments should not perform construction activities that have a nexus to 

radiological health and safety or the common defense and security. An activity or action has a 

reasonable nexus to radiological health and safety or the common defense and security if that 

activity or action has a rational, direct link to ensuring that a licensed materials facility is 

operating, or will operate, in accordance with the NRC’s regulations and in a manner that 

protects the public health and safety or the common defense and security from 

radiological hazards. 
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Installation of foundations or in-place assembly, erection, fabrication, or testing for any structure, 

system, or component of a facility or activity subject to 10 CFR Part 30 that are related to 

radiological health and safety or common defense and security should not be performed prior to 

the conclusion of the environmental review of a license application or amendment. Byproduct 

material license applicants subject to 10 CFR Part 30 may perform those site preparation 

activities identified in revised 10 CFR 30.4 before the NRC has completed its environmental 

review of the license application. 

Excavation and other site preparation activities that do not have a reasonable nexus to 

radiological public health and safety or common defense and security, whether permanent or 

temporary, are not “construction” activities. For example, piles driven to support the erection of 

a bridge for a temporary or permanent access road to a new facility would not be considered as 

construction and may be performed prior to the NRC staff concluding its environmental review 

of a proposed action. 

The installation of a temporary feature within an excavation for a building in which materials 

license activities will be conducted and that will be removed during construction is a site 

preparation activity. Such features include retaining walls, dewatering systems, ramps, and 

other structures that will have no physical presence following construction. 

Construction includes installation of the foundation, including soil compaction; the installation of 

permanent drainage systems and geofabric; the placement of backfill, concrete (e.g., mudmats), 

or other materials that will not be removed before placement of the foundation of a structure; the 

placement and compaction of a subbase; the installation of reinforcing bars to be incorporated 

into the foundation of the structure; the erection of concrete forms for the foundations that will 

remain in place permanently (even if non-structural); and the placement of concrete or other 

material constituting the foundation of any safety-related feature. 

The term “permanent” in this context includes anything that will exist in its final, in-place facility 

location after commencement of operations with licensed material. Construction also includes 

the “onsite, in-place” fabrication, erection, integration, or testing activities for any in-scope 

safety-related equipment. The terms “onsite, in place, fabrication, erection, integration, or 

testing” describe the process of constructing a facility in its final, onsite plant location, where 

components or modules are integrated into the final, in-plant location. The fabrication, 

assembly, and testing of components and modules in a shop building, warehouse, or laydown 

area, even if located onsite, is not construction. However, the installation or integration of the 

safety-related equipment into its final plant location is construction. 

Construction also includes driving piles for safety-related equipment. Hence, an applicant must 

obtain a license before driving piles for safety-related equipment. However, driving piles that do 

not ensure the structural stability or integrity of a safety-related structure (e.g., piles driven to 

support the erection of a bridge for a temporary or permanent access road) is not construction; 

therefore, those piles may be driven prior to the NRC staff concluding its environmental review 

of a proposed action. 

IRRADIATORS (10 CFR PART 36) 

An applicant for a new irradiator license under 10 CFR Part 36 may perform the nonconstruction activities identified in revised 10 CFR 36.2 at any time. However, installation of 

foundations or in-place assembly, erection, fabrication, or testing for any structure, system, or 

component of a facility or activity subject to 10 CFR Part 36 that has a reasonable nexus to 
P–5 

radiological safety or security should not be performed prior to the submission of an application 

for a license and the fee required by 10 CFR 170.31. An activity or action has a reasonable 

nexus to radiological health and safety or the common defense and security if that activity or 

action has a rational, direct link to ensuring that a licensed materials facility is operating, or will 

operate, in accordance with the NRC’s regulations and in a manner that protects the public 

health and safety or the common defense and security from radiological hazards. Activities that 

have a reasonable nexus to radiological health and safety or common defense and security 

include, but are not limited to, construction of systems subject to 10 CFR Part 36, Subpart C, 

and the following: 

 earthwork 

 pool excavation 

 footings and foundation for pool 

 irradiator foundations and walls 

 backfill pool 

 install pool liner 

 mechanical rough-in 

 electrical rough-in 

 shoring for roof 

 form and place roof 

 slab on grade 

Subpart C of 10 CFR Part 36 lists the systems that have a nexus to radiological health and 

safety and defines the related engineering and safety concerns associated with each system: 

 access control: adequacy of access control systems using interlocks and radiation 

monitors to prevent inadvertent entry to areas where radiation sources are unshielded; 

to provide emergency exits; and to ensure compliance with all the requirements of 

10 CFR 36.23; (For computer-controlled access-control systems, licensing staff should 

consider expert evaluation of the software/system logic before operational testing.) 

 site: potential need for protection against flooding and earth slides 

 base (soil, rock) for the pool and shielding structures: strength, settlement, liquefaction, 

ground water, and soil compaction 

 footers and foundations for the pool and shielding structures: strength and 

reinforcement, alignment with pool and shielding structures 

 pool and shielding structures: strength and reinforcement, proper density of shielding 

materials, correct dimensions, minimization of voids in concrete or other shielding 

 pool liner: contact with pool structure, penetrations in the liner, leak-tight welds 

 pool plumbing: makeup water system; water cleanup system; effect of construction 

materials on pool-water chemistry; drainage system (potentially contaminated spilled 

water should flow into the pool); siphon breakers; radiation detection and alarm systems 

 penetrations through shielding: any significant effect on structural strength, shielding, 

or both 
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 source rack protection (if the product to be irradiated moves on a product conveyor 

system): source rack and the mechanism that moves the rack must be protected by a 

barrier or guides to prevent products and product carriers from hitting or touching the 

rack or mechanism 

 source-rack mechanical positioning system: strength and stiffness of the rack and 

positioning cables or chains, source shroud will not interfere with source positioning, 

adequacy of motive power, potential for jamming 

 source-rack movement and position-sensing system: structural attachments for 

electrical and mechanical transducers, adequacy of transducers for interacting with the 

source-rack control system 

 source-rack electrical control system: adequacy of the design of logistical and 

operational electrical circuitry and electromechanical components, to ensure 

unambiguous response of the system, which includes programmable controllers or 

computers and their interaction with operations, interlocks, doors, signals, and alarms 

 source-leak detection: adequacy of systems for detecting and isolating leaking sources 

 hard wiring: adequacy of wire gauge and insulation to safely carry design currents and 

to withstand radiation and ozone damage if exposed; locating and attaching wiring to 

prevent fretting, wear, and exposure to potential fire hazards; accessibility to wiring for 

inspection and repair 

 uninterruptable electrical power supply: adequate and reliable power capability to 

operate all electrical systems that are important to safety (including backup power 

sources); compatibility of the power supply with the electrical system 

 fire protection system: adequacy to detect fire and smoke and to be manually as well as 

automatically initiated; must ensure that raised sources are immediately lowered into 

the pool 

 emergency systems for returning an up-stuck source rack to the pool: capability of the 

electrical control system to sense and signal the occurrence of an up-stuck source-rack; 

adequacy of mechanical or electrical means for personnel to safely release and lower 

the rack; need for, and adequacy of, a system to cool the source-rack until it can be 

released and lowered 

 ozone ventilation system: capability of the system to be properly initiated and to provide 

adequate volume flow rate of air to protect personnel and components 

 system for transferring sources from and to transport vehicles: adequately sized 

openings in the shield-structure roof if sources are roof-loaded; structural adequacy of 

the roof-shield plug and its supports for its removal and replacement; structural and 

mechanical adequacy of systems for moving shipping containers into and out of the 

pool area 

 
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ACTIVITIES WHICH HAVE NO REASONABLE NEXUS TO RADIOLOGICAL SAFETY OR 

SECURITY 

The NRC has determined that, in general, the following activities at byproduct (including 

irradiators), source, and SNM facilities listed in 10 CFR 30.4, 36.2, 40.4, and 70.4, do not have 

a reasonable nexus to radiological health and safety and the common defense and security may 

be performed by a licensee or applicant at any time. Note that in some circumstances, based 

on the specific licensing proposal, any of these activities could be determined to have a 

reasonable nexus to radiological health and safety or common defense and security and, based 

on that determination, these activities would be construction: 

(1) changes for temporary use of the land for public recreational purposes; 

(2) site exploration, including necessary borings to determine foundation conditions or other 

preconstruction monitoring to establish background information related to the suitability 

of the site, the environmental impacts of construction or operation, or the protection of 

environmental values; 

(3) preparation of the site for construction of the facility, including clearing of the site, 

grading, installation of drainage, erosion and other environmental mitigation measures, 

and construction of temporary roads and borrow areas; 

(4) erection of fences and other access control measures that are not related to the safe use 

of, or security of, radiological materials subject to 10 CFR Parts 30, 36, 40, or 70; 

(5) excavation; 

(6) erection of support buildings (e.g., construction equipment storage sheds, warehouse 

and shop facilities, utilities, concrete mixing plants, docking and unloading facilities, and 

office buildings) for use in connection with the construction of the facility; 

(7) building of service facilities (e.g., paved roads, parking lots, railroad spurs, exterior utility 

and lighting systems, potable water systems, sanitary sewerage treatment facilities, and 

transmission lines); 

(8) procurement or fabrication of components or portions of the proposed facility occurring 

at other than the final, in-place location at the facility; or 

(9) taking any other action that has no reasonable nexus to: 

(i) radiological health and safety, or 

(ii) common defense and security. 

While the above site preparation activities may not require prior NRC approval, the approval of 

local, State, or other Federal agencies may be required. 

 
P–8 

REFERENCES 

 NUREG–1748, “Environmental Review Guidance for Licensing Actions Associated with 

Materials Facilities,” August 2003 

 DC/COL–ISG–4, “Interim Staff Guidance on the Definition of Construction and on 

Limited Work Authorizations,” February 9, 2009, ML082970729 
APPENDIX Q 

SAFETY CULTURE POLICY STATEMENT 

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Safety Culture 

The Safety Culture Policy Statement was published in the Federal Register (76 FR 34773) on 

June 14, 2011, and can be found at: http://www.gpo.gov/fdsys/pkg/FR-2011-06-14/pdf/2011-

14656.pdf. It is also posted in the U.S. Nuclear Regulatory Commission (NRC) Agencywide 

Documents Access and Management System Accession Number ML11146A047. 

Safety Culture Policy Statement 

The purpose of this Statement of Policy is to set forth the Commission’s expectation that 

individuals and organizations establish and maintain a positive safety culture commensurate 

with the safety and security significance of their activities and the nature and complexity of their 

organizations and functions. This includes all licensees, certificate holders, permit holders, 

authorization holders, holders of quality assurance program approvals, vendors and suppliers of 

safety-related components, and applicants for a license, certificate, permit, authorization, or 

quality assurance program approval, subject to NRC authority. The Commission encourages 

the Agreement States, Agreement State licensees and other organizations interested in nuclear 

safety to support the development and maintenance of a positive safety culture, as articulated in 

this Statement of Policy. 

Nuclear Safety Culture is defined as the core values and behaviors resulting from a collective 

commitment by leaders and individuals to emphasize safety over competing goals to ensure 

protection of people and the environment. Individuals and organizations performing regulated 

activities bear the primary responsibility for safety and security. The performance of individuals 

and organizations can be monitored and trended and, therefore, may be used to determine 

compliance with requirements and commitments and may serve as an indicator of possible 

problem areas in an organization’s safety culture. The NRC will not monitor or trend values. 

These will be the organization’s responsibility as part of its safety culture program. 

Organizations should ensure that personnel in the safety and security sectors have an 

appreciation for the importance of each, emphasizing the need for integration and balance to 

achieve both safety and security in their activities. Safety and security activities are closely 

intertwined. While many safety and security activities complement each other, there may be 

instances in which safety and security interests create competing goals. It is important that 

consideration of these activities be integrated so as not to diminish or adversely affect either; 

thus, mechanisms should be established to identify and resolve these differences. A safety 

culture that accomplishes this would include all nuclear safety and security issues associated 

with NRC regulated activities. 

Experience has shown that certain personal and organizational traits are present in a positive 

safety culture. A trait, in this case, is a pattern of thinking, feeling, and behaving that 

emphasizes safety, particularly in goal conflict situations, e.g., production, schedule, and the 

cost of the effort versus safety. It should be noted that although the term ‘‘security’’ is not 

expressly included in the following traits, safety and security are the primary pillars of the NRC’s 

regulatory mission. Consequently, consideration of both safety and security issues, 

commensurate with their significance, is an underlying principle of this Statement of Policy. 

 
Q–2 

The following are traits of a positive safety culture: 

(1) Leadership Safety Values and Actions—Leaders demonstrate a commitment to safety in 

their decisions and behaviors; 

(2) Problem Identification and Resolution—Issues potentially impacting safety are promptly 

identified, fully evaluated, and promptly addressed and corrected commensurate with 

their significance; 

(3) Personal Accountability—All individuals take personal responsibility for safety; 

(4) Work Processes—The process of planning and controlling work activities is implemented 

so that safety is maintained; 

(5) Continuous Learning—Opportunities to learn about ways to ensure safety are sought out 

and implemented; 

(6) Environment for Raising Concerns—A safety conscious work environment is maintained 

where personnel feel free to raise safety concerns without fear of retaliation, intimidation, 

harassment, or discrimination; 

(7) Effective Safety Communication—Communications maintain a focus on safety; 

(8) Respectful Work Environment—Trust and respect permeate the organization; and 

(9) Questioning Attitude—Individuals avoid complacency and continuously challenge 

existing conditions and activities in order to identify discrepancies that might result in 

error or inappropriate action. 

There may be traits not included in this Statement of Policy that are also important in a positive 

safety culture. It should be noted that these traits were not developed to be used for inspection 

purposes. 

It is the Commission’s expectation that all individuals and organizations, performing or 

overseeing regulated activities involving nuclear materials, should take the necessary steps to 

promote a positive safety culture by fostering these traits as they apply to their organizational 

environments. The Commission recognizes the diversity of these organizations and 

acknowledges that some organizations have already spent significant time and resources in the 

development of a positive safety culture. The Commission will take this into consideration as 

the regulated community addresses the Statement of Policy. 
APPENDIX R 

CHECKLIST FOR REQUESTS TO WITHHOLD PROPRIETARY 

INFORMATION FROM PUBLIC DISCLOSURE (UNDER 10 CFR 2.390)

R–1 

Checklist for Requests to Withhold Proprietary Information From Public 

Disclosure (Under 10 CFR 2.390) 

In order to request that the U.S. Nuclear Regulatory Commission (NRC) withhold information 

from public disclosure, the applicant or licensee must submit the information, including an 

affidavit, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, “Public 

Inspections, Exemptions, Requests for Withholding.” The applicant should submit all of the 

following: 

 A proprietary copy of the information. Brackets should be placed around the material 

considered to be proprietary. This copy should be marked as proprietary. 

 A non-proprietary copy of the information. Applicants should white out or black out the 

proprietary portions (i.e., those in the brackets), leaving the non-proprietary portions 

intact. This copy should not be marked as proprietary. 

 An affidavit that: 

 Is signed under oath and affirmation (notarization may suffice). 

 Clearly identifies (such as by name or title and date) the document to be withheld. 

 Clearly identifies the position of the person executing the affidavit. This person 

must be an officer or upper-level management official who has been delegated the 

function of reviewing the information the organization is seeking to withhold and is 

authorized to apply for withholding on behalf of the organization. 

 States that the organization submitting the information is the owner of the 

information or is required, by agreement with the owner of the information, to treat 

the information as proprietary. 

 Provides a rational basis for holding the information in confidence. 

 Fully addresses the following issues: 

 Is the information submitted to, and received by, the NRC in confidence? 

Provide details. 

 To the best of the applicant’s knowledge, is the information currently available in 

public sources? 

 Does the applicant customarily treat this information, or this type of information, 

as confidential? Explain why. 

 Would public disclosure of the information be likely to cause substantial harm to 

the competitive position of the applicant? If so, explain why in detail. The 

explanation should include the value of the information to your organization, the 

amount of effort or money expended in developing the information, and the 

ease or difficulty for others to acquire the information. 





NUREG-1556, Vol. 11

Revision 1, Final

Consolidated Guidance About Materials Licenses: Program-Specific 

Guidance About Licenses of Broad Scope