NUREG-2212, Dfc Standard Review Plan for Applications for 10 CFR Part 70 Licenses for Possession and Use of Special Nuclear Materials of Critical Mass But Not Subject to the Requirements in 10 CFR Part 70, Subpart H: Difference between revisions

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#REDIRECT [[NUREG-2212, January 18 2023 Public Meeting Slides (NUREG-2212)]]
| number = ML22335A087
| issue date = 12/31/2022
| title = NUREG-2212, Dfc Standard Review Plan for Applications for 10 CFR Part 70 Licenses for Possession and Use of Special Nuclear Materials of Critical Mass But Not Subject to the Requirements in 10 CFR Part 70, Subpart H
| author name = Ani S, Diaz-Maldonado M, Harris T, Kline K, Naquin T, Norris M, Siurano-Perez O, Trefethen J
| author affiliation = NRC/NMSS
| addressee name =
| addressee affiliation =
| docket =
| license number =
| contact person =
| document report number = NUREG-2212
| document type = NUREG
| page count = 1
| project =
| stage = Other
}}
 
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{{#Wiki_filter:NUREG-2212 Standard Review Plan for Applications for 10 CFR PART 70 Licenses for Possession and Use of Special Nuclear Materials of Critical Mass but not Subject to the Requirements in 10 CFR Part 70, Subpart H Draft Report for Comment Office of Nuclear Material Safety and Safeguards
 
AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material                                      Non-NRC Reference Material As of November 1999, you may electronically access          Documents available from public and special technical NUREG-series publications and other NRC records at the      libraries include all open literature items, such as books, NRCs Library at www.nrc.gov/reading-rm.html. Publicly      journal articles, transactions, Federal Register notices, released records include, to name a few, NUREG-series        Federal and State legislation, and congressional reports.
publications; Federal Register notices; applicant, licensee, Such documents as theses, dissertations, foreign reports and vendor documents and correspondence; NRC                and translations, and non-NRC conference proceedings correspondence and internal memoranda; bulletins and        may be purchased from their sponsoring organization.
information notices; inspection and investigative reports; licensee event reports; and Commission papers and their      Copies of industry codes and standards used in a attachments.                                                substantive manner in the NRC regulatory process are maintained at NRC publications in the NUREG series, NRC regulations,            The NRC Technical Library and Title 10, Energy, in the Code of Federal Regulations        Two White Flint North may also be purchased from one of these two sources:              11545 Rockville Pike Rockville, MD 20852-2738
: 1. The Superintendent of Documents U.S. Government Publishing Office                    These standards are available in the library for reference Washington, DC 20402-0001                            use by the public. Codes and standards are usually Internet: https://bookstore.gpo.gov/                  copyrighted and may be purchased from the originating Telephone: (202) 512-1800                            organization or, if they are American National Standards, Fax: (202) 512-2104                                  from American National Standards Institute
: 2. The National Technical Information Service                  11 West 42nd Street 5301 Shawnee Road                                          New York, NY 10036-8002 Alexandria, VA 22312-0002                                  Internet: https://www.ansi.org/
Internet: https://www.ntis.gov/                            (212) 642-4900 1-800-553-6847 or, locally, (703) 605-6000 Legally binding regulatory requirements are stated only in laws; NRC regulations; licenses, including technical A single copy of each NRC draft report for comment is        specifications; or orders, not in NUREG-series available free, to the extent of supply, upon written        publications. The views expressed in contractor prepared request as follows:                                          publications in this series are not necessarily those of the NRC.
Address: U.S. Nuclear Regulatory Commission                  The NUREG series comprises (1) technical and administrative reports and books prepared by the staff Office of Administration                          (NUREG-XXXX) or agency contractors (NUREG/CR-Digital Communications and Administrative          XXXX), (2) proceedings of conferences (NUREG/CP-Services Branch                                  XXXX),(3) reports resulting from international agreements Washington, DC 20555-0001                          (NUREG/IA-XXXX),(4) brochures (NUREG/BR-XXXX),(5) compilations of legal decisions and orders of the E-mail: distribution.resource@nrc.gov              Commission and the Atomic and Safety Licensing Boards Facsimile: (301) 415-2289                          and of Directors decisions under Section 2.206 of the NRCs regulations(NUREG-0750), and (6) Knowledge Management prepared by NRC staff or agency Some publications in the NUREG series that are posted        contractors.
at the NRCs Web site address www.nrc.gov/reading-rm/
doc-collections/nuregs are updated periodically and may      DISCLAIMER: This report was prepared as an account of work sponsored by an agency of the U.S. Government.
differ from the last printed version. Although references to  Neither the U.S. Government nor any agency thereof, nor material found on a Web site bear the date the material      any employee, makes any warranty, expressed or implied, was accessed, the material available on the date cited        or assumes any legal liability or responsibility for any third partys use, or the results of such use, of any information, may subsequently be removed from the site.                    apparatus, product, or process disclosed in this publication, or represents that its use by such third party would not infringe privately owned rights.
 
NUREG-2212 Standard Review Plan for Applications for 10 CFR PART 70 Licenses for Possession and Use of Special Nuclear Materials of Critical Mass but not Subject to the Requirements in 10 CFR Part 70, Subpart H Draft Report for Comment Manuscript Completed: September 2022 Date Published: December 2022 Prepared by:
S. Ani M. Díaz-Maldonado J. Downs T. Harris K. Kline J. Munson T. Naquin M. Norris O. Siurano-Pérez J. Trefethen O.Siurano-Pérez, R. Jervey, and J. Tobin, Project Managers Office of Nuclear Material Safety and Safeguards
 
COMMENTS ON DRAFT REPORT Any interested party may submit comments on this draft report for consideration by the U.S. Nuclear Regulatory Commission (NRC) staff. Comments may be accompanied by additional relevant information or supporting data. Please specify the report number, NUREG-2212, draft, in your comments, and send them by the end of the comment period specified in the Federal Register notice announcing the availability of this report.
Addresses: You may submit comments by any one of the following methods. Please include Docket ID NRC-2020-0018 in the subject line of your comments. Comments submitted in writing or in electronic form will be posted on the NRC website and on the Federal rulemaking website http://www.regulations.gov.
Federal Rulemaking Website: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-2020-0018.
Mail comments to: Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Program Management, Announcements and Editing Staff.
For any questions about the material in this report, please contact:
Mr. Osiris Siurano-Pérez; phone: (301) 415-7827; email: Osiris.Siurano-Perez@nrc.gov Mr. Richard Jervey; phone: (301) 415-6201; email: Richard.Jervey@nrc.gov Ms. Jennifer Tobin; phone: (301) 415-2328; email: Jennifer.Tobin@nrc.gov Please be aware that any comments that you submit to the NRC will be considered a public record and entered into the Agencywide Documents Access and Management System (ADAMS). Do not provide information you would not want to be publicly available.
PAPERWORK REDUCTION ACT STATEMENT This NUREG provides voluntary guidance for implementing the mandatory information collections in NRC Form 313 and 10 CFR Parts 20, 30, 37, 71 and 73 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), under control numbers 3150-0120, 3150 0014, 3150 0017, 3150-0214, 3150-0008 and 3150 0002, respectively. Send comments regarding this information collection to the FOIA, Library, and Information Collections Branch (T6-A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555 0001, or by e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202 (3150-0120, 3150 0014, 3150 0017, 3150-0214, 3150-0008 and 3150 0002), Office of Management and Budget, Washington, DC, 20503.
PUBLIC PROTECTION NOTIFICATION The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.
 
ABSTRACT This standard review plan (SRP) contains information intended to provide guidance for submitting applications under Title 10 of the Code of Federal Regulations (10 CFR) Part 70, Domestic Licensing of Special Nuclear Material, for possession and use of special nuclear materials (SNM) of critical mass that are not subject to the requirements in 10 CFR Part 70, Subpart H, Additional Requirements for Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material. Specifically, this NUREG describes the types of information required under 10 CFR 70.22, Contents of applications, for those that apply for a new (or the renewal or amendment of an existing) materials license for possession and use of special nuclear material (SNM) in quantities exceeding the thresholds for critical mass quantities in 10 CFR 150.11, Critical mass, a and that, due to the nature of their activities, are not subject to the requirements in 10 CFR Part 70, Subpart H.
The U.S. Nuclear Regulatory Commission (NRC) guidance in this document is intended to be used by the staff to conduct reviews of applications for activities other than those described in 10 CFR 70.60, Applicability, and that, due to the nature of the activities, do not create significant hazards associated with the use of SNM (such as criticality events) covered under the regulations of 10 CFR Part 70. As such, this subset of 10 CFR Part 70 applicants and licensees is not subject to the requirements in 10 CFR Part 70, Subpart H.
Prior to developing this document, specific guidance at a level appropriate to these applicants and licensees (commonly known as greater than critical mass applicants and licensees) did not exist. Therefore, the NRC prepared NUREG-2212 to assist these applicants, licensees, and facilities in preparing new license or renewal applications for the following activities that are not subject to the requirements in 10 CFR Part 70, Subpart H: (1) experiments using subcritical assemblies, (2) instrument calibration, (3) instruction in radiation detection and measurement, (4) experiments with uranium (U)-235 target foils, (5) low-enriched uranium sources for radiation detection testing, and (6) research and development in homeland security applications.
The SRP provides NRC staff reviewers with guidance that describes methods or approaches that the staff has found acceptable for meeting applicable NRC requirements in 10 CFR Part 70.
Implementation of the criteria and guidelines in the SRP by staff members in their review of applications provides assurance that a given design ensures adequate protection of the public health and safety and the environment. This NUREG is intended to improve industry and public stakeholder understanding of the staffs review process. It should be noted that the SRP is not a substitute for NRC regulations, and compliance with the SRP is not required. It also contains related information that applicants and licensees may find useful regarding NRC policy, such as the policy on safety culture (appendix G), the use of discretion on issuing notices of violations (section 2.2.7, Audit program), and the NRC Enforcement Policy (section 2.2.7).
a  For the purpose of this part, SNM in quantities not sufficient to form a critical mass means uranium enriched in the 235 isotope (U-235) in quantities not exceeding 350 grams of contained U-235; U-233 in quantities not exceeding 200 grams; plutonium (Pu) in quantities not exceeding 200 grams; or any combination of them in accordance with the following formula: (grams contained U-235/350) + (grams U-233)/200) + (grams Pu/200) < 1. For each kind of SNM, determine the ratio between the quantity of that SNM and the quantity specified above for the same kind of SNM. The sum of such ratios for all kinds of SNM in combination shall not exceed unity. For example, the following quantities in combination would not exceed the limitation and are within the formula, as follows: (175 (grams contained U-235/350) + (50 grams U-233)/200) + (50 grams Pu/200) = 1).
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TABLE OF CONTENTS ABSTRACT ................................................................................................................... iii TABLE OF CONTENTS.................................................................................................. v DEFINITIONS .................................................................................................................ix LIST OF APPLICABLE REGULATIONS..................................................................... ..xv ABBREVIATIONS AND ACRONYMS .........................................................................xvii INTRODUCTION .......................................................................................................... xix Purpose of the Standard Review Plan .................................................................................xix Format of this Guidance Document..................................................................................... xxi Uses of SNM Addressed by this Guidance .........................................................................xxii Special Considerations Regarding SNM Security Requirements ........................................xxii Applicable Regulations ......................................................................................................xxiii Management Responsibility ..............................................................................................xxiv How to File .........................................................................................................................x. xv Contents of Initial Applications for a 10 CFR Part 70 License ........................................... xxvi Required Fees.................................................................................................................. xxvi Identifying and Protecting Sensitive Information ...............................................................xxvii Application Signature ......................................................................................................xxviii Amendments and Renewals to a License........................................................................xxviii Applications for Exemptions from the NRCs Regulations .................................................xxix Transfer of Control of a License or Bankruptcy of a Licensee ......................................... ... xxx License Terminations ....................................................................................................... xxxi 1 GENERAL INFORMATION ....................................................................................... 1 1.1  Applicants Name and Contact Information .................................................................. 2 1.2  Location and Address of Material Use ......................................................................... 3 1.3  Activities for which the Material Is Requested .............................................................. 3 1.4  Period of Time for which the License Is Requested ..................................................... 4 1.5  Radioactive Material for which the License is Requested ............................................ 4 1.6  Facility Description and Process Overview .................................................................. 4 1.7  Organization and Administration .................................................................................. 4 2 RADIATION PROTECTION .......................................................................................7 2.1  Organizational and Personnel Qualifications ............................................................... 8 2.1.1 Organizational Management............................................................................ 8 2.1.2 Radiation Safety Officer................................................................................... 8 2.1.3 Authorized Users ............................................................................................ 9 2.1.4 Radiation Safety Training................................................................................11 2.2  Radiation Protection Program ......................................................................................12 2.2.1 Alara Program.................................................................................................13 2.2.2 Occupational Dose .........................................................................................14 2.2.3 Public Dose ....................................................................................................17 2.2.4 Minimization of Contamination ........................................................................19 2.2.5 Ventilation and Respiratory Protection ............................................................20 2.2.6 Waste Management .......................................................................................22 2.2.7 Audit Program ................................................................................................23 2.3  Facilities and Equipment..............................................................................................25 2.3.1 Radiation Monitoring Instruments ...................................................................26 2.3.2 Leak Tests ......................................................................................................28 v
 
2.3.3 Surveys .........................................................................................................28 2.4  Commitment to Written Procedures ............................................................................30 2.5  Operating and Emergency Procedures .......................................................................31 2.6  Transportation ............................................................................................................33 2.7  References .................................................................................................................35 3 NUCLEAR CRITICALITY SAFETY ..........................................................................39 3.1  Use of Industry Standards ..........................................................................................39 3.2  Criticality Accident Alarm System ...............................................................................40 3.3  Emergency Planning and Response ...........................................................................41 3.4  Subcriticality and Double Contingency Principle .........................................................42 3.5  Organization and Administration of the Nuclear Criticality Safety Program .................43 3.6  Nuclear Criticality Safety Program Management Measures ........................................45 3.7  Technical Practices for Nuclear Criticality Safety ........................................................45 3.8  Calculational Method Validation ..................................................................................46 3.9  Criticality Safety Evaluations.......................................................................................47 3.10 Evaluation and Implementation of Controlled Parameters ..........................................48 3.11 Additional Nuclear Criticality Safety Program Commitments .......................................52 3.12 Emergency Plan .........................................................................................................53 3.13 References .................................................................................................................53 4 CHEMICAL SAFETY ............................................................................................... 55 4.1  References .................................................................................................................57 5 FIRE SAFETY.......................................................................................................... 59 5.1  Facility Design ............................................................................................................59 5.2  Fire Protection ............................................................................................................60 5.2.1 Fire Protection Systems.................................................................................60 5.2.2 Employee Training.........................................................................................61 5.2.3 Emergency Response ...................................................................................62 5.3  Process Fire Safety ....................................................................................................62 5.4  Combustible Loading and Potential Fire Scenarios.....................................................63 5.5  Summary ....................................................................................................................63 5.6  References .................................................................................................................64 6 NATIONAL ENVIRONMENTAL POLICY ACT........................................................ 65 6.1  Classification of Licensing and Regulatory Actions .....................................................65 6.2  References .................................................................................................................69 7 MATERIAL CONTROL AND ACCOUNTING .......................................................... 71 7.1  Reports of Loss or Theft or Attempted Theft of Special Nuclear Material ....................73 7.2  Material Status Reports ..............................................................................................73 7.3  Nuclear Material Transaction Reports.........................................................................74 7.4  Recordkeeping ...........................................................................................................74 7.5  Written Material, Control, & Accounting Procedures ...................................................75 7.6  Physical Inventories ....................................................................................................76 7.7  Records Access and Storage .....................................................................................76 7.8  Additional Information .................................................................................................76 7.9  References .................................................................................................................76 8 DECOMMISSIONING AND FINANCIAL ASSURANCE ......................................... 79 8.1  Financial Assurance and Decommissioning Funding Plan ..........................................79 8.2  Recordkeeping ...........................................................................................................79 8.3  References .................................................................................................................80 vi
 
9 PHYSICAL SECURITY ............................................................................................ 81 9.1  Category II Special Nuclear Material Physical Protection Requirements .....................82 9.1.1 Applicability ...................................................................................................82 9.1.2 General Performance Objectives ...................................................................82 9.1.3 Physical Protection Requirements at Fixed Sites ...........................................83 9.1.4 Physical Protection Requirements in Transit..................................................84 9.2  Category III SNM Physical Protection Requirements ..................................................85 9.2.1 Applicability ...................................................................................................85 9.2.2 General .........................................................................................................86 9.2.3 Physical Protection Requirements at Fixed Sites ...........................................86 9.2.4 Physical Protection Requirements in Transit..................................................86 9.3  Plutonium-238 and Plutonium/Beryllium Sealed Sources ...........................................87 9.3.1 Applicability ...................................................................................................87 9.3.2 General .........................................................................................................88 9.3.3 Background Investigations and Access Authorization Program .....................88 9.3.4 Physical Protection Requirements for Use and Storage.................................89 9.3.5 Physical Protection during Transit .................................................................89 9.3.6 Additional Information ....................................................................................90 9.4  References .................................................................................................................91 10 EMERGENCY MANAGEMENT ............................................................................... 93 10.1 Emergency Plan .........................................................................................................94 10.2 References .................................................................................................................96 APPENDIX A      TYPICAL DUTIES AND RESPONSIBILITIES OF THE RADIATION SAFETY OFFICER ...................................................... A-1 APPENDIX B      FACILITIES AND EQUIPMENT CONSIDERATIONS...................... B-1 APPENDIX C      RADIATION MONITORING AND INSTRUMENT CALIBRATION PROGRAM ............................................................. C-1 APPENDIX D      GUIDANCE FOR MONITORING REQUIREMENTS ........................ D-1 APPENDIX E      PUBLIC DOSE ..................................................................................E-1 APPENDIX F      SURVEYS AND MONITORING ........................................................ F-1 APPENDIX G      SAFETY CULTURE ......................................................................... G-1 APPENDIX H      MAJOR U.S. DEPARTMENT OF TRANSPORTATION REGULATIONS ............................................................................... H-1 APPENDIX I      NUCLEAR CRITICALITY GLOSSARY.............................................. I-1 vii
 
DEFINITIONS The following definitions in Title 10 of the Code of Federal Regulations (10 CFR) 70.4, Definitions, are applicable to greater than critical mass facilities.
Act means the Atomic Energy Act of 1954 (68 Stat 919), including any amendments thereto.
Acute, as used in this part, means a single radiation dose or chemical exposure event or multiple radiation dose or chemical exposure events occurring within a short time (24 hours or less).
Agency (Government) means any executive department, commission, independent establishment, corporation, wholly or partly owned by the United States of America which is an instrumentality of the United States, or any board, bureau, division, service, office, officer, authority, administration, or other establishment in the executive branch of the Government.
Agreement State as designated in part 150 of this chapter means any State with which the Commission has entered into an effective agreement under subsection 274b. of the Act.
Non-Agreement State means any other State.
Alert means events may occur, are in progress, or have occurred that could lead to a release of radioactive material[s] but that the release is not expected to require a response by an offsite response organization to protect persons offsite.
Atomic energy means all forms of energy released in the course of nuclear fission or nuclear transformation.
Commission means the Nuclear Regulatory Commission or its duly authorized representatives.
Common defense and security means the common defense and security of the United States, Construction means the installation of foundations, or in-place assembly, erection, fabrication, or testing for any structure, system, or component of a facility or activity subject to the regulations in this part that are related to radiological safety or security. The term construction does not include:
(1) Changes for temporary use of the land for public recreational purposes.
(2) Site exploration, including necessary borings to determine foundation conditions or other preconstruction monitoring to establish background information related to the suitability of the site, the environmental impacts of construction or operation, or the protection of environmental values.
(3) Preparation of the site for construction of the facility, including clearing of the site, grading, installation of drainage, erosion and other environmental mitigation measures, and construction of temporary roads and borrow areas.
(4) Erection of fences and other access control measures that are not related to the safe use of, or security of, radiological materials subject to this part.
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(5) Excavation.
(6) Erection of support buildings (e.g., construction equipment storage sheds, warehouse and shop facilities, utilities, concrete mixing plants, docking and unloading facilities, and office buildings) for use in connection with the construction of the facility.
(7) Building of service facilities (e.g., paved roads, parking lots, railroad spurs, exterior utility and lighting systems, potable water systems, sanitary sewerage treatment facilities, and transmission lines).
(8) Procurement or fabrication of components or portions of the proposed facility occurring at other than the final, in-place location at the facility; or (9) Taking any other action that has no reasonable nexus to:
(i) Radiological health and safety, or (ii) Common defense and security.
Contiguous sites means licensee-controlled locations, deemed by the Commission to be in close enough proximity to each other, that the special nuclear material must be considered in the aggregate for the purpose of physical protection.
Critical mass has two definitions in 10 CFR. The definition in 10 CFR 70.4 is used to identify a limit of SNM above which requires the increased programmatic and regulatory controls in Subpart H for the use of the materials. The definition in 10 CFR 150.11, Critical mass, however, prescribes an upper limit, below which fewer programmatic controls are required, as stated below:
10 CFR 70.4. Critical mass of special nuclear material (SNM), as used in Subpart H, means special nuclear material in a quantity exceeding 700 grams of contained uranium-235; 520 grams of uranium-233; 450 grams of plutonium; 1500 grams of contained uranium-235, if no uranium enriched to more than 4 percent by weight of uranium-235 is present; 450 grams of any combination thereof; or one-half such quantities if massive moderators or reflectors made of graphite, heavy water, or beryllium may be present.
10 CFR 150.11, Critical Mass. For the purposes of this part, special nuclear material in quantities not sufficient to form a critical mass means uranium enriched in the isotope U-235 in quantities not exceeding 350 grams of contained U-235; uranium-233 in quantities not exceeding 200 grams; plutonium in quantities not exceeding 200 grams; or any combination of them in accordance with the following formula: For each kind of special nuclear material, determine the ratio between the quantity of that special nuclear material and the quantity specified above for the same kind of special nuclear material. The sum of such ratios for all kinds of special nuclear material in combination shall not exceed unity.
x
 
Decommission means to remove a facility or site safely from service and reduce residual radioactivity to a level that permits (1) Release of the property for unrestricted use and termination of the license; or (2) Release of the property under restricted conditions and termination of the license.
Department and Department of Energy means the Department of Energy Organization Act (Pub. L. 95-91, 91 Stat. 565, 42 U.S.C. 7101 et seq.), to the extent that the Department, or its duly authorized representatives, exercises functions formerly vested in the U.S. Atomic Energy Commission, its Chairman, members, officers and components and transferred to the U.S. Energy Research and Development Administration and to the Administrator thereof pursuant to sections 104(b), (c) and (d) of the Energy Reorganization Act of 1974 (Pub. L.
93-438, 88 Stat. 1233 at 1237, 42 U.S.C. 5814) and retransferred to the Secretary of Energy pursuant to section 301(a) of the Department of Energy Organization Act (Pub. L. 95-91.
91 Stat. 565 at 577-578, 42 U.S.C. 7151).
Double contingency principle means that process designs should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.
Effective dose equivalent means the sum of the products of the dose equivalent to the body organ or tissue and the weighting factors applicable to each of the body organs or tissues that are irradiated. Weighting factors are: 0.25 for gonads, 0.15 for breast, 0.12 for red bone marrow, 0.12 for lungs, 0.03 for thyroid, 0.03 for bone surface, and 0.06 for each of the other five organs receiving the highest dose equivalent.
Effective kilograms of special nuclear material means: (1) For plutonium and uranium-233 their weight in kilograms; (2) For uranium with an enrichment in the isotope U-235 of 0.01 (1%)
and above, its element weight in kilograms multiplied by the square of its enrichment expressed as a decimal weight fraction; and (3) For uranium with an enrichment in the isotope U-235 below 0.01 (1%), by its element weight in kilograms multiplied by 0.0001.
Formula quantity means strategic special nuclear material in any combination in a quantity of 5000 grams or more computed by the formula, grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium). This class of material is sometimes referred to as a Category I quantity of material.
Hazardous chemicals produced from licensed materials means substances having licensed material as precursor compound(s) or substances that physically or chemically interact with licensed materials; and that are toxic, explosive, flammable, corrosive, or reactive to the extent that they can endanger life or health if not adequately controlled. These include substances commingled with licensed material and include substances such as hydrogen fluoride that is produced by the reaction of uranium hexafluoride and water, but do not include substances prior to process addition to licensed material or after process separation from licensed material.
License, except where otherwise specified, means a license issued pursuant to the regulations in this part [10 CFR].
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Management measures mean the functions performed by the licensee, generally on a continuing basis, that are applied to items relied on for safety, to ensure the items are available and reliable to perform their functions when needed. Management measures include configuration management, maintenance, training and qualifications, procedures, audits and assessments, incident investigations, records management, and other quality assurance elements.
Person means (1) any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, Government agency other than the Commission or the Department, except that the Department shall be considered a person within the meaning of the regulations in this part to the extent that its facilities and activities are subject to the licensing and related regulatory authority of the Commission pursuant to section 202 of the Energy Reorganization Act of 1974 (88 Stat. 1244), any State or any political subdivision of or any political entity within a State, any foreign government or nation or any political subdivision of any such government or nation, or other entity; and (2) any legal successor, representative, agent, or agency of the foregoing.
Principal activities, as used in this part, means activities authorized by the license which are essential to achieving the purpose(s) for which the license was issued or amended. Storage during which no licensed material is accessed for use or disposal and activities incidental to decontamination or decommissioning are not principal activities.
Produce, when used in relation to special nuclear material, means (1) to manufacture, make, produce, or refine special nuclear material; (2) to separate special nuclear material from other substances in which such material may be contained; or (3) to make or to produce new special nuclear material.
Research and development means (1) theoretical analysis, exploration, or experimentation; or (2) the extension of investigative findings and theories of a scientific or technical nature into practical application for experimental and demonstration purposes, including the experimental production and testing of models, devices, equipment, materials, and processes.
Sealed source means any special nuclear material that is encased in a capsule designed to prevent leakage or escape of the special nuclear material.
Site Area emergency means events may occur, are in progress, or have occurred that could lead to a significant release of radioactive material and that could require a response by offsite response organizations to protect persons offsite.
Source material means source material as defined in section 11z. of the Act and in the regulations contained in Part 40 of this chapter [10 CFR].
Special nuclear material means (1) plutonium, uranium 233, uranium enriched in the isotope 233 or in the isotope 235, and any other material which the Commission, pursuant to the provisions of section 51 of the act, determines to be special nuclear material, but does not include source material; or (2) any material artificially enriched by any of the foregoing but does not include source material.
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Special nuclear material of low strategic significance means:
(1) Less than an amount of special nuclear material of moderate strategic significance as defined in paragraph (1) of the definition of strategic nuclear material of moderate strategic significance in this section, but more than 15 grams of uranium-235 (contained in uranium enriched to 20 percent or more in U-235 isotope) or 15 grams of uranium-233 or 15 grams of plutonium or the combination of 15 grams when computed by the equation, grams = (grams contained U-235) + (grams plutonium) + (grams U-233); or (2) Less than 10,000 grams but more than 1,000 grams of uranium-235 (contained in uranium enriched to 10 percent or more but less than 20 percent in the U-235 isotope);
or (3) 10,000 grams or more of uranium-235 (contained in uranium enriched above natural but less than 10 percent in the U-235 isotope).
This class of material is sometimes referred to as a Category III quantity of material.
Special nuclear material of moderate strategic significance means:
(1) Less than a formula quantity of strategic special nuclear material but more than 1,000 grams of uranium-235 (contained in uranium enriched to 20 percent or more in the U-235 isotope) or more than 500 grams of uranium-233 or plutonium, or in a combined quantity of more than 1,000 grams when computed by the equation, grams = (grams contained U-235) + 2 (grams U-233 + grams plutonium); or (2) 10,000 grams or more of uranium-235 (contained in uranium enriched to 10 percent or more but less than 20 percent in the U-235 isotope).
This class of material is sometimes referred to as a Category II quantity of material.
Special nuclear material scrap means the various forms of special nuclear material generated during chemical and mechanical processing, other than recycle material and normal process intermediates, which are unsuitable for use in their present form, but all or part of which will be used after further processing.
United States, when used in a geographical sense, includes Puerto Rico and all territories and possessions of the United States.
Uranium enrichment facility, means:
(1) Any facility used for separating the isotopes of uranium or enriching uranium in the isotope 235, except laboratory scale facilities designed or used for experimental or analytical purposes only; or (2) Any equipment or device, or important component part especially designed for such equipment or device, capable of separating the isotopes of uranium or enriching uranium in the isotope 235.
Worker means an individual who receives an occupational dose as defined in 10 CFR 20.1003.
xiii
 
1                          LIST OF APPLICABLE REGULATIONS 2
3 The following list includes the main requirements in Title 10 of the Code of Federal Regulations 4 (10 CFR) Part 70, Domestic Licensing of Special Nuclear Material, that are applicable to 5 licensees authorized (and applicants seeking authorization) to possess special nuclear 6 materials in critical mass quantities not subject to the 10 CFR Part 70, Subpart H, requirements 7 (Note: Other 10 CFR Part 70 requirements may apply, as determined by the Commission).
8 9 70.9    Completeness and accuracy of information 10 70.21  Filing, (f) 11 70.22  Contents of applications, (a)(1), (2), (3), (4), (6), (7), and (8) 12 70.23  Requirements for the approval of applications, (a)(1), (2), (3), (4), (5), (6), (9) and (10) 13 70.24  Criticality accident requirements 14 70.25  Financial assurance and recordkeeping for decommissioning 15 70.41  Authorized use of special nuclear material, (a) 16 70.42  Transfer of special nuclear material 17 70.50  Reporting requirements 18 70.51  Records requirements 19 70.52  Reports of accidental criticality 20 70.55  Inspections 21 70.56  Tests 22 70.81  Modification and revocation of licenses 23 70.91  Violations 24 70.92  Criminal penalties xv
 
1                  ABBREVIATIONS AND ACRONYMS 2
3 ADAMS Agencywide Documents Access and Management System 4 ALARA as low as is reasonably achievable 5 ALI  annual limits on intake 6 ANSI  American National Standards Institute 7 AU    authorized user 8
9 Be    beryllium 10 11 CAAS  criticality accident alarm system 12 CATEX categorical exclusion 13 CFR  Code of Federal Regulations 14 CSE  criticality safety evaluation 15 16 DFP  decommissioning funding plan 17 DOE  U.S. Department of Energy 18 DOT  U.S. Department of Transportation 19 20 GTCM  greater than critical mass 21 22 IN    information notice 23 ISO  International Organization for Standardization 24 25 MC&A  material control and accounting 26 mGy  milligray 27 mR    milliroentgen 28 mrem  millirem 29 mSv  millisievert 30 31 NCRP  National Council on Radiation Protection and Measurements 32 NCS  nuclear criticality safety 33 NEPA  National Environmental Policy Act 34 NFPA  National Fire Protection Association 35 NIST  National Institute of Standards and Technology 36 NMMSS Nuclear Material Management and Safeguards System 37 NMSS  Office of Nuclear Material Safety and Safeguards 38 NRC  U.S. Nuclear Regulatory Commission 39 NSTS  National Source Tracking System 40 NVLAP National Voluntary Laboratory Accreditation Program 41 42 OCFO  Office of the Chief Financial Officer 43 OCR  optical character reader 44 OMB  Office of Management and Budget 45 46 Pu    plutonium 47 xvii
 
1 RIS  reporting identification symbol 2 RPP  radiation protection program 3 RSO  radiation safety officer 4 RWP  radiation work permit 5
6 SNM  special nuclear material 7 SRP  standard review plan 8 SSD  sealed source and device 9 Sv  Sievert 10 11 TEDE total effective dose equivalent 12 13 U    uranium 14 15 ZnS  zinc sulfide 16 xviii
 
1                                                INTRODUCTION 2
3 Purpose of the Standard Review Plan 4
5 This standard review plan (SRP) describes the types of information required under Title 10 of 6 the Code of Federal Regulations (10 CFR) 70.22, Contents of applications, for applications 7 under 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, for receipt, 8 possession, use, and storage of special nuclear material (SNM) 1 in critical mass quantities but 9 not subject to 10 CFR Part 70, Subpart H, Additional Requirements for Certain Licensees 10 Authorized To Possess a Critical Mass of Special Nuclear Material. This SNM is intended to be 11 used for conducting activities other than those described in 10 CFR 70.60, Applicability, and 12 that, due to the nature of their activities, do not create significant hazards associated with the 13 use of SNM (such as criticality events) covered under the regulations of 10 CFR Part 70. As 14 such, this subset of 10 CFR Part 70 applicants and licensees is not subject to the requirements 15 in 10 CFR Part 70, Subpart H.
16 17 This group of applicants or licensees, currently referred to as greater than critical mass (GTCM) 18 applicants, licensees, or facilities, are not engaged in SNM enrichment, conversion, or 19 manufacturing activities and yet are similar enough to each other 2 and form a large enough 20 group to consider their own characteristics differently from other users of SNM who use other 21 guidance, such as NUREG-1556, Consolidated Guidance About Materials Licenses, 22 Volume 17, Program-Specific Guidance About Licenses for Special Nuclear Material of Less 23 than Critical Mass, or NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License 24 Applications, for completing applications for possession and use of SNM.
25 26 For the purposes of this NUREG and the associated regulatory analysis, GTCM is any quantity 27 of SNM exceeding (i.e., >) the definition of critical mass in 10 CFR 150.11, Critical mass:
28 29          (a) For the purposes of this part, special nuclear material 3 in quantities not 30                sufficient to form a critical mass means uranium enriched in the isotope 31                uranium (U) -235 in quantities not exceeding 350 grams of contained U-235; 32                uranium-233 not exceeding 200 grams; plutonium in quantities not exceeding 33                200 grams [in any form other than plutonium (Pu)/beryllium (Be) neutron 34                sources], or any combination of them in accordance with the following 35                formula 4:
36 1 The regulations in 10 CFR 70.4, Definitions, define SNM as (1) plutonium, uranium 233, uranium enriched in the isotope 233 or in the isotope 235, and any other material which the Commission, pursuant to the provisions of section 51 of the [Atomic Energy Act of 1954, as amended,] determines to be special nuclear material, but does not include source material; or (2) any material artificially enriched by any of the foregoing but does not include source material.
2 The safety and risks involved are more akin to those inherent with byproduct materials, which result from nuclear fission and are non-fissile. These cannot be used for atomic weapons, so the material is not at risk for proliferation.
Byproduct materials are non-fissile and are not subject to self-sustaining criticality or the risks that are associated with criticality. The U.S. Nuclear Regulatory Commission (NRC) developed this guidance for these specific SNM licensees because, although they use a certain quantity of SNM, they pose little risk of accidental criticality, diversion, or proliferation.
3 This document uses SNM, licensed material, and radioactive material interchangeably.
4 The sum of such ratios for all kinds of SNM, in combination, should not exceed unity. Note that the formula in 10 CFR 150.11 is just an example of specific SNM quantities that, altogether, meet the criteria (i.e., 175 grams of contained U-235 + 50 grams of U-233 + 50 grams of Pu). If the calculated number is greater than 1, then the amount of SNM for which the applicant seeks NRC authorization is a critical mass quantity of SNM.
xix
 
1                    (grams U-235)          +    (grams U-233)          +    (grams Pu)      < 1 2                        350                      200                      200 3
4 Applicants for a license to receive, possess, use, and store SNM must calculate the ratio among 5 their requested quantities of SNM using this formula. If the result is greater than 1 (i.e., exceeds 6 unity), then the SNM meets the criteria for a critical mass quantity defined in 10 CFR 150.11.
7 The formula determines the appropriate guidance for applicants (i.e., NUREG-1556, Volume 17, 8 for applications for licenses of lower than critical mass versus NUREG-2212 for applications for 9 licenses of GTCM). NOTE: Any quantity of SNM exceeding the GTCM threshold in 10 10 CFR 150.11 falls under NRC-exclusive jurisdiction.
11 12 Table 1-1 shows the threshold quantities for SNM to be considered GTCM under 10 CFR 13 Part 70. (NOTE: Materials meeting these criteria are not subject to the requirements in 14 10 CFR Part 70, Subpart H, when used for the specific activities listed under the Uses of SNM 15 addressed by this guidance section of this document).
16 17 In addition to the limits listed in Table 1-1, the NRC staff developed a limit for Pu/Be sources for 18 when the licensee requests authorization for possession and use of Pu/Be sources only. In 19 these cases, a quantity of more than 2,000 grams of plutonium is treated as GTCM, when the 20 Pu/Be ratio is more than 1,000. 5 21 Table 1-1 Radioactive Materials Considered SNM Special Nuclear Material            Covered by This                Threshold        Regulatory Authority NUREG                    Quantities Greater Than Critical Mass (g) 6 Plutonium                          Yes                      >200                NRC/NMSS U-235                            Yes                      350                NRC/NMSS U-233                            Yes                      200                NRC/NMSS 22 23 This document describes the type of information needed to develop an application for a specific 24 license for receipt, possession, use, and storage of SNM in GTCM. This guide is not intended to 25 address the following:
26 27
* Applications for possession of quantities of SNM equal to or lower than the 28          10 CFR 150.11 critical mass limits. (NOTE: Guidance for applying for a license to 29          possess and use SNM of less than critical mass, as defined in 10 CFR 150.11, is found 30          in NUREG-1556, Volume 17, or in the relevant Agreement State Program guidance.)
31 5  The technical basis for treating 2,000 grams of Pu 239 in the form of Pu/Be sealed sources as less than a critical mass is described in Technical Basis for 2000 Grams of Plutonium Limit in Pu/Be Neutron Source Mass Limit, issued June 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16190A294). The Pu/Be limit will not apply if the request includes possession and use of Pu/Be neutron sources in combination with other types of SNM. If so, then the limits in 10 CFR 150.11(a) apply.
6  Any quantities of SNM below the thresholds specified on this table fall below Agreement State or NRC regulatory authority under 10 CFR Part 70 in non-Agreement States.
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1
* Applications for possession of quantities of SNM in excess of formula quantities of SNM 2        and for authorization to manufacture and distribute SNM. (NOTE: These activities are 3        subject to the requirements in 10 CFR Part 70, Subpart H. The applicable guidance for 4        the review of applications for these activities is NUREG-1520.
5 6 This guidance is not intended to be used by the NRC staff for reviewing applications for 7 authorization to engage in enriched uranium processing, fabrication of uranium fuel or fuel 8 assemblies, uranium enrichment, enriched uranium hexafluoride conversion, plutonium 9 processing, fabrication of mixed oxide fuel or fuel assemblies, scrap recovery of SNM, or any 10 other activity that the Commission determines could significantly affect public health and safety.
11 These are subject to the requirements in 10 CFR Part 70, Subpart H.
12 13 Format of this Guidance Document 14 15 This guidance identifies information that the NRC staff expects to be included in applications for 16 new licenses, renewals, and amendments of specific licenses for receipt, possession, use, and 17 storage of SNM in critical mass quantities, as defined in 10 CFR 150.11, but that are not subject 18 to the requirements in 10 CFR Part 70, Subpart H. Each technical chapter of NUREG-2212 19 presents guidance in the following format:
20 21 (1)    Regulatory Requirementsreferences the applicable regulations 22 23 (2)    Regulatory Criteriaoutlines the criteria the NRC staff uses to evaluate the adequacy of 24        an application 25 26 (3)    Discussionprovides supplemental information on the technical topic 27 28 (4)    Regulatory Guidanceidentifies the current regulatory guidance that is applicable to the 29        GTCM licensees, not subject to Subpart H, that may be consulted by the staff and 30        applicants 31 32 (5)    Notesexplain regulatory requirements, identify other relevant NRC guidance, define 33        technical terms, clarify information, provide examples of precedents, or advise when 34        additional information on a topic is not required to be provided 35 36 (6)    Additional Informationidentifies relevant information that facilitates the staffs review of 37        an application and may be included in the license application 38 39 (7)    Referencesprovides full references to all documents used in the preparation of this 40        guidance report 41 42 The guidance in this document is intended to help the NRC staff standardize and streamline its 43 review, but the NRC anticipates applicants will use it to improve the efficiency of the license 44 application process. To facilitate the NRC staffs review, applicants should refer to the 45 requirements in 10 CFR 70.22, follow the guidance in this document (and its appendices), and 46 submit the license application to the NRC. Appendices A through G to this document provide 47 additional, specific guidance relevant to completing an application for a GTCM license.
48 49 50 51 xxi
 
1 Uses of SNM Addressed by this Guidance 2
3 The staff will use this guidance to standardize and streamline its evaluation of applications for 4 GTCM SNM for the following activities:
5 6
* research and development and for educational purposes 7
8
* experiments using subcritical assemblies 9
10
* instrument calibration 11 12
* student instruction in radiation detection and measurement 13 14
* U-235 target foils experiments 15 16
* the manufacture or fabrication of sealed and unsealed sources containing low-enriched 17        uranium (i.e., enriched to less than 20 percent by weight) that are used in radiation 18        detection testing equipment 19 20 Special Considerations Regarding SNM Security Requirements 21 22 The NRC categorizes SNM as formula quantity, moderate strategic significance material, and 23 low strategic significance SNM. Chapter 9 of this document provides an in-depth discussion of 24 the physical security requirements applicable to GTCM licensees, according to the category of 25 SNM.
26 27 Formula quantity of SNM 28 29 Pursuant to the regulations in 10 CFR 70.4, a formula quantity of SNM means the following:
30 31        strategic special nuclear material in any combination in a quantity of 5000 grams 32        or more computed by the formula:
33 34        grams = (grams contained U-235) + 2.5 (grams U-233 + grams plutonium) 35 36 Formula quantity SNM is sometimes referred to as a Category I quantity of material (see 37 10 CFR 73.2, Definitions).
38 39 SNM of moderate strategic significance 40 41 SNM of moderate strategic significance is defined in 10 CFR 70.4 as follows:
42 43        (1) Less than a formula quantity of strategic special nuclear material but more 44            than 1,000 grams of U-235 (contained in uranium enriched to 20 percent or 45            more in the U-235 isotope) or more than 500 grams of uranium-233 or 46            plutonium, or in a combined quantity of more than 1,000 grams when 47            computed by the equation, 48 49            grams = (grams contained U-235) + 2 (grams U233 + grams plutonium); or 50 xxii
 
1        (2) 10,000 grams or more of uranium-235 (contained in uranium enriched to 2            10 percent or more but less than 20 percent in the U-235 isotope).
3 4 Special nuclear material of moderate strategic significance is sometimes referred to as a 5 Category II quantity of material (see 10 CFR 73.2).
6 SNM low strategic significance 7
8 SNM of low strategic significance is defined in 10 CFR 70.4 as follows:
9 10        (1) Less than an amount of special nuclear material of moderate strategic 11            significancebut more than 15 grams of uranium-235 (contained in uranium 12            enriched to 20 percent or more in U-235 isotope) or 15 grams of uranium-233 13            or 15 grams of plutonium or the combination of 15 grams when computed by 14            the equation, grams = (grams contained U-235) + (grams plutonium) +
15            (grams U-233); or 16 17        (2) Less than 10,000 grams but more than 1,000 grams of uranium-235 18            (contained in uranium enriched to 10 percent or more but less than 19            20 percent in the U-235 isotope); or 20 21        (3) 10,000 grams or more of uranium-235 (contained in uranium enriched above 22            natural but less than 10 percent in the U-235 isotope).
23 24 Special nuclear material of low strategic significance is also referred to as a Category III quantity 25 of material (see 10 CFR 73.2).
26 27 The categories of SNM in 10 CFR 73.2 (i.e., material of formula quantity, of moderate strategic 28 significance, or of low strategic significance) determine which physical security and material 29 control and accounting requirements in 10 CFR Part 73, Physical Protection of Plants and 30 Materials, apply to licensees possessing, handling, and storing different categories of SNM.
31 Chapter 9 of this guidance document provides an in-depth discussion of the physical security 32 requirements applicable to GTCM licensees, according to the category of SNM. For these 33 reasons, NRC staff reviewers and applicants for a GTCM license should be familiar with 34 chapter 9.
35 36 Applicable Regulations 37 38 Applicants and licensees are responsible for obtaining up-to-date copies of applicable 39 regulations and understanding and complying with the requirements of the regulations 40 applicable to holders of licenses for the possession and use of SNM of GTCM.
41 42 The NRCs current regulations can be found at https://www.nrc.gov/reading-rm/doc-43 collections/cfr/index.html. The following is a list of relevant regulations:
44 45
* 10 CFR Part 2, Agency Rules of Practice and Procedure 46 47
* 10 CFR Part 19, Notices, Instructions and Reports to Workers: Inspection and 48        Investigations 49 50
* 10 CFR Part 20, Standards for Protection against Radiation xxiii
 
1
* 10 CFR Part 21, Reporting of Defects and Noncompliance 2
3
* 10 CFR Part 37, Physical Protection of Category 1 and Category 2 Quantities of 4          Radioactive Material 5
6
* 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and 7          Related Regulatory Functions 8
9
* 10 CFR Part 70, Domestic Licensing of Special Nuclear Material 10 11
* 10 CFR 70.9, Completeness and accuracy of information 12 13
* 10 CFR Part 71, Packaging and Transportation of Radioactive Material 14 15
* 10 CFR Part 73, Physical Protection of Plants and Materials 16 17
* 10 CFR Part 74, Material Control and Accounting of Special Nuclear Material 18 19
* 10 CFR 150.11, Critical mass 20 21
* 10 CFR Part 170, Fees for Facilities, Materials, Import and Export Licenses, and Other 22          Regulatory Services under the Atomic Energy Act of 1954, as Amended 23 24
* 10 CFR Part 171, Annual Fees for Reactor Operating Licenses, and Fuel Cycle 25          Licenses and Materials Licenses, Including Holders of Certificates of Compliance, 26          Registrations, and Quality Assurance Program Approvals and Government Agencies 27          Licensed by the NRC 28 29 Copies of the of the NRCs regulations may be obtained by calling the U.S. Government 30 Publishing Office order desk toll-free at (866) 512-8600; in Washington, DC, at (202) 512-1800; 31 or online at https://bookstore.gpo.gov/catalog/code-federal-regulations-cfrs-print.
32 33 A single copy of the above documents may be requested from the NRCs regional offices (see 34 https://www.nrc.gov/about-nrc/locations.html for addresses and telephone numbers). In addition, 35 10 CFR Parts 1 through 199 may be downloaded from the NRCs website at 36 http://www.nrc.gov/reading-rm/doc-collections/ under Regulations (10 CFR).
37 38 NRC regulations and amendments may also be accessed from the NRC Library link on the 39 NRCs public website at http://www.nrc.gov. The NRC publishes new and amended regulations 40 in the Federal Register.
41 42 Management Responsibility 43 44 The NRC recognizes that effective radiation safety program management is vital to achieving 45 safe and compliant operations. Compliance with the NRCs regulations provides reasonable 46 assurance that licensed activities will be conducted safely, and that effective management will 47 result in increased safety and compliance.
48 49 50 xxiv
 
1 How to File 2
3 Paper Applications 4
5 Applicants for an SNM license should do the following:
6 7
* Submit adequate information to satisfy the requirements in 10 CFR 70.22 and 8          10 CFR 70.23, Requirements for the approval of applications.
9 10
* Provide sufficient detail for the NRC to determine that equipment, facilities, training, 11          experience, the radiation safety program, and all other safety programs are adequate to 12          protect health and safety and minimize danger to life and property.
13 14
* Identify and cross-reference the chapter number and its topic on each separate sheet 15          submitted with the application.
16 17
* If proprietary information and other sensitive information (e.g., personal privacy and 18          security related) is submitted, the information should be clearly identified per 19          10 CFR 2.390, Public inspections, exemptions, requests for withholding (see 20          chapter 6).
21 22
* Ensure applications are signed in accordance with the requirements in 10 CFR 70.22(d).
23 24
* Retain one copy of the license application for future reference.
25 26 All license applications will be available for review by the general public in the NRCs Public 27 Document Room. If it is necessary to submit proprietary and other sensitive information, follow 28 the procedure in 10 CFR 2.390 for identifying and marking this information. Failure to follow this 29 procedure could result in disclosure of the proprietary or other sensitive information to the public 30 or in substantial delays in processing the application. Applicants should not submit personal 31 information of employees (e.g., home address, home telephone number, social security number, 32 date of birth, radiation dose information), unless the NRC specifically requests it.
33 34 The NRC scans paper applications through an optical character reader and converts them to an 35 electronic format. To ensure the smooth transfer of paper documents into an electronic format, 36 applicants should do the following:
37 38
* Submit all documents typed, on paper that will feed easily into a document scanner 39          (i.e., legal or letter size).
40 41
* Choose typeface designs that are sans serif, such as Arial, Helvetica, Futura.
42 43
* Use an 11-point or larger font.
44 45
* Avoid stylized characters, such as script or italics.
46 47
* Ensure that the print is clear and sharp.
48 49
* Ensure a high contrast between the ink and paper (black ink on white paper is best).
50 xxv
 
1 Electronic Applications 2
3 Applications may also be submitted in electronic form using the NRCs Electronic Information 4 Exchange by or CD-ROM. Detailed guidance on making electronic submissions can be obtained 5 by visiting the NRCs website at http://www.nrc.gov/site-help/e-submittals.html. The guidance 6 discusses, among other topics, the formats the NRC can accept, the use of electronic 7 signatures, and the treatment of nonpublic information.
8 9 Contents of Initial Applications for a 10 CFR Part 70 License 10 11 Note: This section addresses the required contents of a new application. Section IX discusses 12 the required contents of applications for amendments and renewals of licenses.
13 14 All items in an application should be in sufficient detail for the NRC staff to determine that the 15 proposed equipment, facilities, training and experience, radiation safety, and other safety 16 programs satisfy the NRCs regulatory requirements and are adequate to protect public health 17 and safety and minimize danger to life and property. There is no specific license application 18 form required for GTCM applicants. NRC Form 313, Application for Materials License, which is 19 used to apply for byproduct material licenses, may also be used to apply for a GTCM SNM 20 license, but it is not required. The NRC describes the information to be provided in a license 21 application in 10 CFR 70.22. The NRCs reviewers will ensure the application contains all 22 required information as part of its acceptance review. The NRC encourages applicants to 23 request preapplication meetings with the staff to discuss regulatory requirements and the 24 licensing review process. The NRC lists filing and other applicable fees in 10 CFR 70.21, 25 Filing.
26 27 Licensees should submit items described in 10 CFR 70.22(b)-(m) on separate sheets of paper.
28 This part of the license application provides the technical basis for the license request. The 29 information should describe the applicable safety programs that will provide adequate 30 assurance of the safety and security of the receipt, use, handling, and possession of SNM.
31 Applicants may organize the license application by using the chapter headings of this NUREG.
32 License applicants that follow the guidance and the model procedures in this NUREG will 33 facilitate the NRCs review.
34 35 The grant of an NRC license does not relieve a licensee from complying with other applicable 36 Federal, State, or local regulations (e.g., local zoning requirements, a local ordinance requiring 37 registration of a radiation-producing device).
38 39 All information submitted to the NRC during the licensing process may be incorporated as part 40 of the license and will be subject to review during scheduled NRC inspections.
41 42 Required Fees 43 44 The appropriate fee must accompany each application for which a fee is specified. Applicants 45 should refer to 10 CFR 170.31, Schedule of fees for materials licenses and other regulatory 46 services, including inspections, and import and export licenses, to determine the amount of the 47 fee. The regulations at 10 CFR 170.11, Exemptions, provide information on exemptions from 48 these fees. In accordance with the requirements in 10 CFR 170.12, Payment of fees, a 49 remittance for the full amount of the fee must accompany each application for which the NRC 50 prescribes a fee. The NRC will not issue a license until it receives the prescribed application fee.
51 xxvi
 
1 Fees will not be refunded after the NRC staff begins its technical review. The NRC will charge 2 application fees whether or not the Commission approves the application.
3 4 The regulations in 10 CFR 171.16, Annual fees: materials licensees, holders of certificates of 5 compliance, holders of sealed source and device registrations, holders of quality assurance 6 program approvals, and government agencies licensed by the NRC, provide the annual fee 7 requirements that are applicable to GTCM SNM licensees [see 10 CFR 171.16(d), Table 2 to 8 paragraph (d)Schedule of Materials Annual Fees and Fees for Government Agencies 9 Licensed by NRC, Category 1A(2)(a) (Facilities with limited operationsProgram Codes 10 21310 and 21320); Category 1F (Licenses for possession and use of special nuclear materials 11 greater than critical mass, as defined in 10 CFR 70.4 of this chapter, for development and 12 testing of commercial products, and other non-fuel-cycle activitiesProgram Code 22155]; and 13 the Office of Nuclear Materials Safety and Safeguards Program Code Descriptions and 14 Inspection Priorities (ML19322A311, Program Codes 21310, 21320, and 22155). The regulation 15 at 10 CFR 171.16(c) states that reduced annual fees are available for licensees that qualify as 16 small entities. This regulation describes the process for licensees to apply for certification as a 17 small entity. In 10 CFR 171.11, the NRC provides information on classes of licensees that are 18 not required to pay annual fees, explains which entities may seek exemptions from such fees, 19 and describes the bases on which the NRC can exempt licensees from those fees.
20 21 Applicants should direct all questions about the NRCs fees to the Office of the Chief Financial 22 Officer at NRC Headquarters in Rockville, Maryland, (301) 415-7554. Information about fees 23 may also be obtained by calling the NRCs toll-free number, (800) 368-5642, extension 24 415-7554. The email address is Fees.Resource@nrc.gov.
25 26 Identifying and Protecting Sensitive Information 27 28 All licensing application materials, except portions that contain sensitive information, will be 29 made available for review in the NRCs Public Document Room and electronically at the NRC 30 Library. For more information on the NRC Library, visit www.nrc.gov/reading-rm.html.
31 32 The licensee should identify and mark sensitive information that it believes should not be 33 disclosed to the public. Licensees submitting applications that contain sensitive information 34 should mark them, in accordance with 10 CFR 2.390, before submitting the information to the 35 NRC. The following are examples of sensitive information:
36 37
* Proprietary information and trade secrets: If it is necessary to submit proprietary 38          information or trade secrets, follow the procedure in 10 CFR 2.390(b). Failure to follow 39          this procedure could result in disclosure of the proprietary information to the public or 40          substantial delays in processing the application.
41 42
* Personal privacy information: Personal information about employees or other individuals 43          should not be submitted unless the NRC specifically requests it. Examples of private 44          information are social security number, home address, home telephone number, date of 45          birth, and radiation dose information. If private information is submitted, it should be 46          separated from the public portion of the application and clearly marked as follows:
47          Privacy Act InformationWithhold under 10 CFR 2.390. Further information is 48          available in Regulatory Issue Summary 2007-04, Personally Identifiable Information 49          Submitted to the U.S. Nuclear Regulatory Commission, dated March 9, 2007, which can 50          be found on the NRCs Generic Communications webpage under Regulatory Issue 51          Summaries: http://www.nrc.gov/reading-rm/doc-collections/gen-comm/.
xxvii
 
1 2
* Security-related information: Following the events of September 11, 2001, the NRC 3            changed its procedures to avoid the release of information terrorists could use to plan or 4            execute an attack against facilities or citizens in the United States. As a result, certain 5            types of information are no longer routinely released to the public and are treated as 6            sensitive unclassified information. This includes certain information about the quantities 7            and locations of radioactive material at licensed facilities, and associated security 8            measures. Therefore, sensitive security-related information in an application should be 9            marked: Security RelatedWithhold under 10 CFR 2.390. Further information is 10            available in Regulatory Information Summary 2005-31, Revision 1, Control of 11            Security-Related Sensitive Unclassified Nonsafeguards Information Handled by 12            Individuals, Firms, and Entities Subject to NRC Regulation of the Use of Source, 13            Byproduct, and Special Nuclear Material, dated December 26, 2017, which can be 14            found on the NRCs Generic Communications webpage under Regulatory Issue 15            Summaries: https://www.nrc.gov/reading-rm/doc-collections/gen-comm/reg-16            issues/2005/. Additional information on procedures and any updates are available at 17            http://www.nrc.gov/reading-rm/sensitive-info.html.
18 19 Application Signature 20 21 All applications and statements must be signed by the applicant or licensee or a corporate 22 officer, as stated in 10 CFR 70.22(d). To ensure adequate management involvement, a duly 23 authorized management representative must sign the submitted application acknowledging 24 managements commitments and responsibilities. Individuals acting in a private capacity are 25 required to date and sign the license application or letter transmitting the license application.
26 Otherwise, representatives of the corporation or legal entity filing the application should date 27 and sign the license application document. Representatives signing an application must be 28 authorized to make binding commitments and to sign official documents on behalf of the 29 applicant.
30 31 Amendments and Renewals to a License 32 33 For changes that require preapproval under 10 CFR 70.72(d)(1), the licensee shall submit a 34 license amendment request to the NRC in accordance with 10 CFR 70.34, Amendment of 35 licenses, and receive the NRCs prior approval before implementing changes to its NRC-36 regulated programs. 7 If information provided in the original application is to be modified or 37 changed, the licensee must submit an application for a license amendment in accordance with 38 10 CFR 70.34. An administrative change to the license that does not constitute a modification or 39 alteration to the facility or regulated activities also requires a license amendment. The NRC staff 40 reviews and determines whether to grant or deny an amendment request. The proposed change 41 is not in effect until the NRC staff has approved and issued the amendment. In addition, 42 10 CFR 70.33, Applications for renewal of licenses, sets out the requirements for the renewal 43 of a license and 10 CFR 70.38(a) provides that a licensee must apply for license renewal at 44 least 30 days before the license expires to continue the license beyond its expiration date.
45 46 Licensees are required to request and obtain an amendment to the license before making 47 changes to their radiation safety programs or any other safety programs included in the license 7  Licensees may make changes to the site, structures, processes, systems, equipment, components, computer programs, and activities of personnel, without prior Commission approval, if the change does not fall within the criteria described in 10 CFR 70.72(c)(1)-(3).
xxviii
 
1 application. The following are examples of changes that require an amendment:
2
* a change of responsible persons or radiation safety officer (RSO) 3 4
* changes in areas of use 5
6
* changes in the mailing address 7
8
* changes in the address(es) of storage locations 9
10
* changes in licensed material, including increases in the possession limit of SNM and the 11          addition of new types of devices 12 13 Applicants for a license amendment or license renewal should do the following:
14 15
* Submit a letter to the NRC requesting an amendment or renewal.
16 17
* Provide the license and docket numbers in the request.
18 19
* For renewals, if many outdated documents are referenced or there have been significant 20          changes in regulatory requirements, the NRCs guidance, the licensees organization, or 21          the licensees radiation protection program (RPP), provide a complete and up-to-date 22          application. Alternatively, clearly describe the exact nature of the changes, additions, 23          and deletions.
24 25 Applications for Exemptions from the NRCs Regulations 26 27 Licensees may request exemptions from the NRCs regulations. The exemption request must 28 demonstrate that the exemption is authorized by law; will not endanger life, property, or the 29 common defense and security; and is otherwise in the public interest.
30 31 Licensees may request specific exemptions from the 10 CFR Part 70 regulations under 32 10 CFR 70.17, Specific exemptions. Various sections of the NRCs regulations address 33 requests for exemptions from other regulations. The regulations concerning exemptions state 34 the NRC may grant an exemption, acting on its own initiative or upon an application from a 35 licensee.
36 37 Exemptions do not revise regulations, are not intended to apply to large classes of licensees, 38 and are limited to unique situations of limited duration. Exemption requests must be 39 accompanied by descriptions of the following:
40 41
* the exemption requested, basis, and justification for the requested exemption 42 43
* any proposed compensatory safety measures intended to provide a level of health and 44          safety equivalent to the regulation for which the exemption is being requested 45 46 Note: Unless the NRC has granted an exemption from a regulation in writing, full compliance 47 with all applicable regulations is expected and enforced.
48 49 50 xxix
 
1 2 Transfer of Control of a License or Bankruptcy of a Licensee 3
4 Transfer of Control of a License 5
6 The regulation at 10 CFR 70.36, Inalienability of licenses, states the following:
7 8          (a) No license granted under the regulations in this part and no right to possess 9              or utilize special nuclear material granted by any license issued pursuant to 10              the regulations in this part shall be transferred, assigned or in any manner 11              disposed of, either voluntarily or involuntarily, directly or indirectly, through 12              transfer of control of any license to any person unless the Commission shall 13              after securing full information, find that the transfer is in accordance with the 14              provisions of the Act, and shall give its consent in writing.
15 16          (b) An application for transfer of license must include:
17 18              (1) The identity, technical and financial qualifications of the proposed 19                    transferee; and 20 21              (2) Financial assurance for decommissioning information required by 22                    10 CFR 70.25.
23 24 Transferring control of a license may be the result of mergers, buyouts, or majority stock 25 transfers. The NRC does not interfere with business decisions of licensees.
26 27 NUREG-1556, Volume 15, Guidance About Changes of Control and About Bankruptcy 28 Involving Byproduct, Source, or Special Nuclear Materials Licenses, chapter 5, 8 discusses the 29 information that should be included in a request for NRC approval of a proposed transfer of 30 control of a license and the criteria the NRC staff uses to evaluate such a request.
31 32 Bankruptcy 33 34 The regulations in 10 CFR 70.32, Conditions of licenses, state the following:
35 36          (a) Each license shall contain and be subject to the following conditions:
37 38              (9)(i) Each licensee shall notify the appropriate NRC Regional Administrator, 39                      in writing, immediately following the filing of a voluntary or involuntary 40                      petition for bankruptcy under any Chapter of Title 11 (Bankruptcy) of the 41                      United States Code by or against:
42 43                      (A) The licensee.
44 45                      (B) An entity (as that term is defined in 11 U.S.C. 101(14)) controlling 46                          the licensee or listing the license or licensee as property of the 8  Applicants for an NRC license are not subject to the NRCs regulations on changes of control of a license unless a change of ownership or management control that would change the information under NRCs review impacts the basis for issuance of the license. In such cases, the applicant must immediately notify the NRC for appropriate action (e.g., withdrawal or rejection of application, termination of licensing review).
xxx
 
1                          estate; or 2                    (C) An affiliate (as that term is defined in 11 U.S.C. 101(a)) of the 3                          licensee.
4 5 A licensees (or applicants) financial condition could affect its ability to control licensed material.
6 Therefore, the NRC must be notified to ensure that appropriate measures to protect the public 7 health and safety are or will be taken. NUREG-1556, Volume 15, chapter 6, 9 includes guidance 8 on the information that should be included in a notification to the NRC of a bankruptcy.
9 NUREG-1556 also describes the criteria the NRC staff used to evaluate the notification of 10 bankruptcy.
11 12 License Terminations 13 14 In accordance with the requirements in 10 CFR 70.38, Expiration and termination of licenses 15 and decommissioning of sites and separate buildings or outdoor areas, licensees may request 16 a termination of the specific license.
17 18
* 10 CFR 70.38(a) states the following:
19 20          Each specific license expires at the end of the day on the expiration date stated 21          in the license unless the licensee has filed an application for renewal under &sect; 22          70.33 not less than 30 days before the expiration date stated in the existing 23          license. If an application for renewal has been filed at least 30 days before the 24          expiration date stated in the existing license, the existing license expires at the 25          end of the day on which the Commission makes a final determination to deny the 26          renewal application or, if the determination states an expiration date, the 27          expiration date stated in the determination.
28 29
* 10 CFR 70.38(b) states the following:
30 31          Each specific license revoked by the Commission expires at the end of the day 32          on the date of the Commission's final determination to revoke the license, or on 33          the expiration date stated in the determination, or as otherwise provided by 34          Commission Order.
35 36
* 10 CFR 70.38(c) states the following:
37 38          Each specific license continues in effect, beyond the expiration date, if 39          necessary, with respect to possession of special nuclear material until the 40          Commission notifies the licensee in writing that the license is terminated. During 41          this time, the licensee shall:
42 43          (1) Limit actions involving special nuclear material to those related to 44              decommissioning; and 45 9    Persons who are in the process of applying for an NRC license and who do not already hold one or more other NRC licenses are not required to provide notifications on bankruptcy to the NRC. However, applicants must advise the NRC of bankruptcy issues that result in changes to the financial qualifications information submitted to the NRC.
xxxi
 
1        (2) Continue to control entry to restricted areas until they are suitable for release 2              in accordance with NRC requirements.
3
* To request license termination under 10 CFR 70.38(d)(1)-(4), licensees must do the 4        following:
5 6        -    Notify the NRC, in writing, within 60 days of the occurrence of any of the following:
7 8              (1) expiration of its license 9
10              (2) a decision to cease licensed activities permanently at the entire site 11 12              (3) a decision to cease licensed activities permanently in any separate building or 13                  outdoor area that contains residual radioactivity such that the building or area is 14                  unsuitable for release according to NRC requirements 15 16              (4) no principal activities having been conducted at the entire site under the license 17                  for a period of 24 months 18 19              (5) no principal activities having been conducted for a period of 24 months in any 20                  separate building or outdoor area, if it contains residual radioactivity making it 21                  unsuitable for release according to NRC requirements.
22 23        -    Within 12 months of notification of any of the above, submit a decommissioning plan, 24              if required by 10 CFR 70.38(g).
25 26        -    Conduct decommissioning, as required by 10 CFR 70.38(h) and 10 CFR 70.38(j).
27 28        -    Submit to the appropriate NRC office a completed NRC Form 314, Certificate of 29              Disposition of Materials (or equivalent information), and a demonstration that the 30              premises are suitable for release for unrestricted use (e.g., results of final leak tests) 31              (see NUREG-1757, Consolidated Decommissioning Guidance: Decommissioning 32              Process for Materials Licensees, Volume 1).
33 34        -    Send records important to decommissioning to the appropriate NRC regional office 35              before a license is terminated (see section 3.3 of NUREG-1757, Volume 3, Financial 36              Assurance, Recordkeeping, and Timeliness). If licensed activities are transferred or 37              assigned in accordance with 10 CFR 70.42, Transfer of special nuclear material, 38              and 10 CFR 70.51(b), transfer records important to decommissioning to the new 39              licensee.
40 41 In accordance with the requirements in 10 CFR 70.38(k), in part, the Commission will terminate 42 a license following its determination that the following has occurred:
43 44              (1) Special nuclear material has been properly disposed.
45 46              (2) Reasonable effort has been made to eliminate residual radioactive 47                  contamination, if present; and 48 49              (3) (i) A radiation survey has been performed which demonstrates that the 50                  premises are suitable for release in accordance with the criteria for 51                  decommissioning in 10 CFR part 20, subpart E; or xxxii
 
1 2                    (ii) Other information submitted by the licensee is sufficient to 3                          demonstrate that the premises are suitable for release in accordance 4                          with the criteria for decommissioning in 10 CFR part 20, subpart E.
5 6              (4) Records required by &sect; 70.51(a) have been received.
7 8 Additional guidance for requesting termination of a license is found in NUREG-1757, 9 Volume 1. 10 However, licensees should also contact the NRC to ensure an understanding of 10 what actions should be taken to initiate and complete the license termination process on a 11 license- or facility-specific basis. The licensees obligations regarding license termination begin 12 at the time the licensee determines that it will cease operations, or on the date the license is set 13 to expire, whichever is earlier. These obligations are to undertake the necessary 14 decommissioning activities, to submit NRC Form 314 or equivalent information, and to take any 15 other actions summarized above. A license continues in effect beyond the expiration date, if 16 necessary, until the Commission notifies the licensee in writing that the license is terminated 17 (see 10 CFR 70.38(c). If a licensee abandons a site or refuses to decommission a site, the NRC 18 would consider the case for civil or criminal action, as warranted.
10  Facilities for which this guidance is intended generally fall under Groups 1 or 2, as described in NUREG-1757, Volume 1.
xxxiii
 
1                              1        GENERAL INFORMATION 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) 70.22, 4 Contents of applications [(a)(1)-(4), (6)-(8)], and 10 CFR 70.23, Requirements for the 5 approval of applications [(a)(1)-(6), (9), and (10)].
6 7 Regulatory Criteria: The regulations in 10 CFR 70.22 describe the information that must be 8 provided in an application for a license under 10 CFR Part 70, Domestic Licensing of Special 9 Nuclear Material.
10 11
* The regulations in 10 CFR 70.22(a)(1) require that an application include information 12        about the individual who can answer questions about the application and a telephone 13        number where the individual may be contacted, as well as the organizational groups that 14        are responsible for managing the proposed activities and the location and address 15        where the material will be handled and stored.
16 17
* The regulations in 10 CFR 70.22(a)(2) require that an application identify and describe 18        the activities for which the SNM license is requested, the place where the activities will 19        be conducted, and a facility description and overview.
20 21
* The regulations in 10 CFR 70.22(a)(3) require, in part, that an application identify the 22        period of time for which the license is requested.
23 24
* The regulations in 10 CFR 70.22(a)(4) require that an application identify the name, 25        amount, and specifications (including the chemical and physical form and, where 26        applicable, isotopic content) of the SNM that the applicant proposes to use.
27 28
* The regulations in 10 CFR 70.22(a)(6) require that the application contain the technical 29        qualifications, including training and experience, of the applicant and members of the 30        staff who will engage in the proposed activities.
31 32
* The regulations in 10 CFR 70.22(a)(7) require applicants to provide a description of 33        equipment and facilities that will be used by the applicant to protect health and minimize 34        danger to life or property (such as handling devices, working areas, shields, measuring 35        and monitoring instruments, devices for the disposal of radioactive effluents and wastes, 36        storage facilities, criticality accident alarm systems).
37 38
* The regulations in 10 CFR 70.22(a)(8) require applicants to discuss their proposed 39        procedures to protect health and minimize danger to life or property (such as procedures 40        to avoid accidental criticality, procedures for personnel monitoring and waste disposal, 41        and post-criticality accident emergency procedures).
42 43 The regulations in 10 CFR 70.23(a) state, in part, that an application for a license will be 44 approved if the Commission determines the following:
45 46 47 48 49 50 1
 
1          (1) The special nuclear material is to be used for the conduct of research or 2              development activities of a type specified in section 31 of the [Atomic Energy 3              Act of 1954, as amended (Act)], in activities licensed by the Commission 4              under section 103 or 104 of the Act, or for such other uses as the 5              Commission determines to be appropriate to carry out the purposes of the 6              Act.
7 8          (2) The applicant is qualified by reason of training and experience to use the 9              material for the purpose requested in accordance with the regulations in this 10              chapter.
11 12          (3) The applicants proposed equipment and facilities are adequate to protect 13              health and minimize danger to life or property; and 14 15 16          (4) The applicants proposed procedures to protect health and to minimize 17              danger to life or property are adequate.
18 19          (5) Where the nature of the proposed activities is such as to require 20              consideration by the Commission, that the applicant appears to be financially 21              qualified to engage in the proposed activities in accordance with the 22              regulations in this part.
23 24          (6) Where the applicant is required to submit a summary description of the 25              fundamental material controls provided in his procedures for the control of 26              and accounting for special nuclear material pursuant to &sect; 70.22(b), the 27              applicants proposed controls are adequate.
28 29 30          (9) Where the applicant is required to submit a plan for physical protection of 31              special nuclear material in transit pursuant to &sect; 70.22(g), of this chapter, the 32              applicants plan is adequate.
33 34          (10) Where the applicant is required to submit a physical security plan pursuant 35              to &sect; 70.22(h), the applicants proposed plan is adequate; 36 37 Discussion: The sections below discuss the regulatory requirements specific to license 38 applications to possess and use special nuclear material (SNM) that exceeds the thresholds in 39 10 CFR 150.11, Critical mass, and that are not subject to 10 CFR Part 70, Subpart H, 40 Additional Requirements for Certain Licensees Authorized To Possess a Critical Mass of 41 Special Nuclear Material. The application should provide adequate information identifying the 42 applicant and its characteristics, a general description of the proposed facility and proposed 43 activities, the quantity and form of the SNM, the applicants technical qualifications, and the time 44 period for which the license is requested. This information will be used by all reviewers, NRC 45 managers, and the general public to understand the proposed facility and its processes, and the 46 individuals responsible for safely operating it.
47 48 1.1      Applicants Name and Contact Information 49 50 Under 10 CFR 70.22(a)(1), if the applicant is an individual, the full name, address, age, 51 citizenship, and three personal references shall be provided. An individual may be designated 2
 
1 as the applicant only if the individual is acting in a private capacity and the use of the radioactive 2 material is not connected with employment in a corporation or other legal entity.
3 4 If the applicant is a corporation or institution, name the State where it was incorporated, 5 locations of principal offices, names, addresses, and citizenship of its principal officers. A 6 division or department within a legal entity may not be a licensee. Notify the U.S. Nuclear 7 Regulatory Commission (NRC) of changes in mailing address; these changes do not require a 8 fee.
9 10 Of significance to this part of the application is information regarding control of the organization 11 or ownership of any alien, foreign corporation, or government (see 10 CFR 70.36, Inalienability 12 of licenses). The NRC must be notified by formal amendment before control of the license is 13 transferred or when ownership changes. Section IX contains more details.
14 15 Note: If the applicant is controlled by a foreign entity, the NRC will need to consider foreign 16 ownership, control, and domination. The NRC will need to determine that the applicant, a 17 foreign-owned corporation or entity, will not be a threat to the common defense and security and 18 does not present an unreasonable risk to public health and safety, in accordance with the Act, 19 sections 57c, 63b and 82b. NUREG-1556, Consolidated Guidance About Materials Licenses, 20 Volume 15, Guidance About Changes of Control and About Bankruptcy Involving Byproduct, 21 Source, or Special Nuclear Materials Licenses, contains additional information.
22 23 1.2      Location and Address of Material Use 24 25 The regulations in 10 CFR 70.22(a)(1) require that an applicant specify the street address, city, 26 and State or other descriptive address (e.g., on Highway 28, 7 miles east of the intersection of 27 Highway 18 and State Route 160, Anytown, State) for each facility where the licensed material 28 will be used (see NRC Form 313, Application for Materials License). The descriptive address 29 should be sufficient to allow an NRC inspector to find the facility location. A post office box 30 address is not acceptable.
31 32 An NRC-approved license amendment is required before receiving, using, and storing licensed 33 material at an address or location not included with the application or not already listed on the 34 license.
35 36 To conduct operations at temporary job sites (i.e., locations where work is done for limited 37 periods of time), specify temporary job sites anywhere in the United States where the NRC 38 maintains jurisdiction. The NRC has additional requirements on temporary sites in 39 10 CFR Part 20, Standards for Protection against Radiation, Subpart I, Storage and Control of 40 Licensed Material, and 10 CFR Part 73, Physical Protection of Plants and Materials.
41 42 1.3      Activities for which the Material Is Requested 43 44 In accordance with the requirements in 10 CFR 70.23(a)(1), in part, an application for a license 45 will be approved if the Commission determines that the SNM will be used for the conduct of 46 activities licensed by the Commission under section 103 or 104 of the Atomic Energy Act of 47 1954, as amended, or for such other uses as the Commission determines to be appropriate to 48 carry out the purposes of the Act. In accordance with the requirements in 10 CFR 70.22(a)(2),
49 an applicant shall describe the activity for which the SNM is requested, or in which SNM will be 50 produced, the place at which the activity is to be performed, and the general plan for carrying 51 out the activity (see NRC Form 313). The described uses should contain sufficient information to 3
 
1 enable the reviewers to have a clear understanding of each use and determine the potential for 2 exposure of workers and members of the public to radiation and radioactive materials. The 3 specific uses of each radioisotope requested should be described. The application will include a 4 summary, nontechnical narrative description for each activity or process in which the applicant 5 proposes to acquire, deliver, receive, possess, produce, use, process, transfer, or store SNM.
6 The proposed uses of SNM for the facility must be for uses authorized under the Act.
7 8 1.4      Period of Time for which the License Is Requested 9
10 In accordance with the requirements in 10 CFR 70.22(a)(3), the applicant should state the 11 period for which licensing is requested. At the outset of applying, this period may not be known.
12 When the period of licensing is unknown, a period for licensing is established based upon the 13 activities of the applicant and the type of license. The applicant will have ample time before 14 licensing to discuss these needs with the NRC staff.
15 16 1.5      Radioactive Material for which the License Is Requested 17 18 In accordance with the requirements in 10 CFR 70.22(a)(4), the application must clearly state 19 the name, amount, and specifications (chemical and physical form, isotopic content where 20 applicable). The applicant should list each requested radioisotope by its element name and its 21 mass number (e.g., uranium (U)-233). It should be specified whether the material will be 22 acquired and used in unsealed or sealed form. The name of the specific chemical compound 23 that contains the radioisotope is not required. Applicants requesting an authorization to use 24 volatile radioactive material must provide appropriate facilities, engineering controls, and 25 radiation safety procedures for handling such material.
26 27 The anticipated possession limit in milligrams (mg) or grams (g) for each radioisotope should 28 also be specified. Possession limits must cover the total anticipated inventory, including 29 licensed material in storage and waste, and should be commensurate with the applicants needs 30 and facilities for safe handling. Applicants should review the requirements for submitting a 31 certification for financial assurance for decommissioning before specifying possession limits of 32 any radioisotope with a half-life greater than 120 days. Chapter 8 discusses these requirements.
33 34 1.6      Facility Description and Process Overview 35 36 In accordance with the requirements in 10 CFR 70.23(a)(3), an application for a license will be 37 approved if the Commission determines that the applicants proposed equipment and facilities 38 are adequate to protect health and minimize danger to life or property. In accordance with the 39 requirements in 10 CFR 70.22(a)(7), the application should describe equipment and facilities to 40 be used by the applicant to protect health and minimize danger to life or property. The applicant 41 should state the general facility and process descriptions involved in the activities for which the 42 license is requested. The application should include a facility layout description; an overview of 43 the different processes at the facility involving licensed material; and a site overview, including a 44 description of the proximity of facility buildings and nearby populations.
45 46 1.7      Organization and Administration 47 48 The regulations in 10 CFR 70.23(a)(2) state that the Commission will approve an application if it 49 determines that the applicant is qualified by reason of training and experience to use the 50 material for the purpose requested. The regulations in 10 CFR 70.22(a)(6) require an applicant 4
 
1 to provide the technical qualifications, including training and experience of the applicant and 2 members of its staff, to engage in the proposed regulated activities. The applicant should also 3 describe the qualification of its personnel for key management positions in terms of education 4 (i.e., degree and field), training, and experience. Qualification criteria should be described 5 generally, in terms of academic credentials, formal continuing education, and work experience 6 (e.g., bachelors degree in nuclear engineering or related scientific or engineering field, with 7 5 years of experience managing the radiation protection program in the Department of Nuclear 8 Engineering).
9 10 The regulations in 10 CFR 70.23(a)(4) state that the Commission will approve an application if 11 its proposed procedures to protect health and to minimize danger to life or property are 12 adequate. The regulations in 10 CFR 70.22(a)(8) require applicants to provide information 13 regarding their proposed procedures to protect health and minimize danger to life or property.
14 The applicant should describe the policies and procedures that it will use to manage and closely 15 monitor the facility design, engineering, construction, and modifications to adequately protect 16 public health and safety and the environment. The applicant should describe the relationships 17 among major facility safety functions and programs, such as the radiation safety, nuclear 18 criticality safety, fire safety, chemical safety, environmental monitoring, and emergency 19 planning. The application should discuss the procedures for communications and authority 20 among the organizational units involved in the engineering, health, safety, environmental, and 21 operations functions of the facility.
5
 
1                              2        RADIATION PROTECTION 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) Part 19, 4 Notices, Instructions and Reports to Workers: Inspection and Investigations; 10 CFR Part 20, 5 Standards for Protection against Radiation; 10 CFR Part 70, Domestic Licensing of Special 6 Nuclear Material; 10 CFR 73.67, Licensee fixed site and in-transit requirements for the 7 physical protection of special nuclear material of moderate and low strategic significance; 8 10 CFR 21.21, Notification of failure to comply or existence of a defect and its evaluation; and 9 10 CFR 71.5, Transportation of licensed material.
10 11 Regulatory Criteria: An application for a license to possess special nuclear material (SNM) in 12 critical mass quantities must provide the following radiation protection information, in 13 accordance with the requirements in 10 CFR 70.22(a)(6) and 10 CFR 20.1101, Radiation 14 protection programs:
15 16
* description of the organization and technical qualifications of the applicant and members 17          of his staff to engage in the applicants proposed activities 18 19
* commitment to maintain occupational exposures to radiation as low as is reasonably 20          achievable (ALARA) 21 22
* description of the facilities and equipment adequate to protect health and minimize 23          danger to life and property, including that licensees must do the following:
24 25          -  possess or have access to radiation monitoring instruments that are necessary to 26              protect health and minimize danger to life or property 27          -  conduct testing to determine whether there is any radioactive leakage from licensed 28              material 29          -  conduct surveys of potential radiological hazards in the workplace 30 31
* written radiation protection procedures and radiation work permits 32 33
* description of the operating and emergency procedures that include the following 34          provisions:
35 36          -  instructions to keep radiation doses to workers and members of public ALARA 37          -  instructions for maintaining security during storage and transportation 38          -  instructions to maintain accountability during use 39          -  use of personnel monitoring and radiation survey equipment 40          -  instructions for packaging and transporting licensed material 41          -  instructions on whom to contact when an emergency occurs 42 43
* description of how the applicant ensures the safe transportation of licensed material 44 45 Discussion: The applicant should provide the details of its radiation protection program (RPP) 46 to address the occupational radiation protection measures in 10 CFR Parts 19, 20, 70, and 71.
47 Specifically, applicants must develop, document, and implement an RPP in accordance with 48 10 CFR 20.1101 to ensure adequate protection for the radiological health and safety of workers.
49 50 7
 
1 The content and level of detail in this chapter are generally greater than in other chapters 2 because the criteria are needed to demonstrate compliance with 10 CFR Part 20 requirements.
3 4 2.1    Organizational and Personnel Qualifications 5
6 2.1.1  Organizational Management 7
8 The regulations in 10 CFR 70.22(a)(6) require that an applicant provide the technical 9 qualifications, including training and experience of the applicant, key staff members, and 10 managers. The information in the submittal should identify, by name, the individuals who will be 11 managing the activities described in the license application with sufficient description to convey 12 the training and experience of each of these individuals. Organizational relationships among the 13 individual positions of the staff should be defined and understandable to the reviewer.
14 15 The organizational structure and associated administrative program should include 16 administrative policies, procedures and management policies, and qualifications of key 17 management positions. The submittal should describe how these will provide reasonable 18 assurance that the health, safety, and environmental protection functions will be effective.
19 20 The applicant should also describe the qualifications regarding education (i.e., degree and field),
21 training, and experience for key management positions. Responsibilities or job descriptions for 22 decision-making individuals (i.e., management positions) should be described for the facility 23 manager, operations manager, shift supervisor, and managers for various safety and 24 environmental disciplines. Alternative named management positions may be proposed.
25 Qualifications should be described generally, in terms of academic credentials, formal 26 continuing education, and work experience.
27 28 2.1.2  Radiation Safety Officer 29 30 Regulatory Requirements: 10 CFR 20.1101, 10 CFR 70.22(a)(6), and 10 CFR 70.23(a)(2).
31 32 Regulatory Criteria: Radiation safety officers (RSOs) must have training and specific 33 experience appropriate for the types and quantities of licensed material to be authorized by the 34 license. The training should be commensurate with the discussion below:
35 36
* The applicant shall appoint a suitably educated, experienced, and trained RPP director 37        (typically referred to as the RSO. The RSO needs independent authority to stop 38        operations that he or she considers unsafe. The RSO must have sufficient time and 39        commitment from management to fulfill certain duties and responsibilities to ensure that 40        radioactive materials are used in a safe manner. The applicant should also provide the 41        NRC with the education and experience of the RPP directors staff and identify the 42        relationship between line managers and those responsible for the RPP.
43 44 Discussion: The RSO is the individual responsible for the RPP. The RSO should have 45 independent authority to stop operations that may be considered unsafe and should have 46 sufficient time and commitment from management to fulfill certain duties and responsibilities to 47 ensure that radioactive materials are used in a safe manner. Appendix A to this NUREG 48 describes typical RSO duties.
49 50 51 8
 
1 To demonstrate adequate training and experience, the RSO should have: (1) sufficient 2 knowledge of physical, chemical, biological sciences, or engineering; and (2) training and 3 experience commensurate with the scope of proposed licensed activities. Training should 4 include the following subjects:
5 6
* radiation protection principles 7
* characteristics of ionizing radiation 8
* units of radiation dose and quantities 9
* radiation detection instrumentation 10
* biological hazards of exposure to radiation (appropriate to types and forms of SNM to be 11        used) 12
* regulatory requirements and standards 13 14
* hands-on use of radioactive materials 15 16 The length of training and experience described will depend upon the type, form, quantity, and 17 proposed use of the licensed material requested. Ultimately, the proposed training and 18 experience should be sufficient to ensure that the RSO can identify and control the anticipated 19 radiation hazards. In addition to having direct access to the plant or facility manager, the RSO 20 should be skilled in the interpretation of data and regulations pertinent to radiation protection, be 21 familiar with the operation of the facility and radiation protection concerns of the site, participate 22 as a resource in radiation safety management decisions, and be responsible for establishing 23 and carrying out the RPP. Appendix A to NUREG-1556, Consolidated Guidance About 24 Materials Licenses, Volume 17, Program-Specific Guidance About Licenses for Special 25 Nuclear Material of Less than Critical Mass, describes some typical RSO duties. This is not an 26 all-encompassing list of radiation safety duties but does identify issues for consideration.
27 28 Applicants should provide information about the proposed RSOs training and experience 29 relative to the licensed material requested in the application. Applicants should not submit 30 extraneous information, such as unrelated lists of publications, research grants, and committee 31 and society memberships. Applicants are required to notify the NRC of changes in the 32 designation of the RSO. The applicant must submit to the NRC the name and qualifications of 33 the replacement RSO as part of an amendment request.
34 35 2.1.3  Authorized Users 36 37 Regulatory Requirements: 10 CFR 19.11, Posting of notices to workers; 10 CFR 19.12, 38 Instruction to workers; 10 CFR 19.13, Notifications and reports to individuals; 39 10 CFR 20.1101; 10 CFR 70.22(a)(6); and 10 CFR 70.23(a)(2).
40 41 Regulatory Criteria: Authorized users (AUs), including radiation workers, must have adequate 42 training and experience to safely possess and use radioactive materials.
43 44 45 9
 
1 Discussion: An AU is a category of staff that the licensee has selected to meet requirements 2 that the NRC has defined, established, and reviewed during the licensing process. An AU is a 3 person who uses or directly supervises the use of licensed material. The AUs primary 4 responsibility is to ensure that radioactive materials are used safely and in accordance with 5 regulatory requirements. The AU is also responsible for ensuring that procedures and 6 engineering controls are used to keep occupational doses and doses to members of the public 7 ALARA.
8 9 Some organizations may have multiple AUs or may have multiple uses of material on the 10 license. These individuals must have adequate training and experience with the types, 11 quantities of licensed materials, and activities approved by the license. An AU (also known as 12 principal investigator/researcher) is a person, other than the RSO, whose training and 13 experience has been reviewed and approved by the applicant or the NRC, or both, and who 14 uses or directly supervises the use of licensed material. The AU is also responsible for keeping 15 occupational doses and doses to members of the public ALARA.
16 17 The AUs must have adequate and appropriate training to provide reasonable assurance that the 18 applicant uses licensed material safely; adequately secures, and controls access to, licensed 19 materials; and responds appropriately to events or accidents involving licensed materials to 20 prevent the spread of contamination.
21 22 To demonstrate adequate training and experience, the AU should have (1) sufficient knowledge 23 of physical, chemical, and biological sciences, or engineering and (2) training and experience 24 commensurate with the scope of the proposed activities. Training should include the following 25 topics:
26 27
* radiation protection principles 28 29
* characteristics of ionizing radiation 30 31
* units of radiation dose and quantities 32 33
* radiation detection instrumentation 34 35
* biological hazards of exposure to radiation (appropriate to types and forms of SNM to be 36        used) 37 38
* regulatory requirements and standards 39 40
* hands-on use of radioactive materials 41 42 The type, form, quantity, and proposed use of the licensed material will determine the length of 43 training and experience required.
44 45 An AU supervises the use of radioactive materials when directing personnel during operations 46 involving licensed materials. Although the AU may delegate specific tasks to supervised users 47 (e.g., doing surveys, keeping records), the AU is responsible for the safe use of radioactive 48 materials to ensure areas are not contaminated.
49 50 10
 
1 Applicants must maintain records of AU training and experience, which must be commensurate 2 with the type and quantity of material the applicant proposes to use. For example, an individual 3 with training and experience only with sealed radioactive sources may not be qualified to use or 4 supervise the use of unsealed licensed material. In addition, an individual with experience using 5 only trace quantities may not understand the risks of working with much larger (e.g., 10 or more 6 times larger) quantities of the same substance. Applicants should pay particular attention to the 7 type of radiation involved. For example, an individual experienced with gamma emitters may not 8 have appropriate experience for high energy beta emitters.
9 10 2.1.4    Radiation Safety Training 11 12 Regulatory Requirements: 10 CFR 19.11; 10 CFR 19.12; 10 CFR 19.13; 10 CFR 37.43, 13 General security program requirements; 10 CFR 70.9, Completeness and accuracy of 14 information; 10 CFR 70.23(a)(2).
15 16 Regulatory Criteria: Individuals whose assigned duties involve exposure to radiation or 17 radioactive material (from both licensed and unlicensed sources) and who, during their 18 employment, are likely to receive an annual occupational dose of radiation greater than 19 1 millisievert (mSv) (100 millirem (mrem)), must receive instruction commensurate with their 20 duties and responsibilities, as required by 10 CFR 19.12. Also, any licensee that possesses an 21 aggregated Category I or Category II quantity of radioactive material (as defined in 22 10 CFR 37.5, Definitions) must provide a training program for those individuals implementing 23 the security program. Each individual should also receive periodic (i.e., annual) refresher 24 training.
25 26 Discussion: Before beginning work with or in the vicinity of licensed material, all individuals 27 who are likely to receive an occupational dose in excess of 1 mSv (100 mrem) per year must 28 receive radiation safety training commensurate with their assigned duties and specific to the 29 licensees radiation safety program.
30 31 Licensees should not assume that prior employment or academic training has adequately 32 covered safety instructions. Licensees should provide site-specific training for all individuals and 33 should discuss procedures for working with radioactive materials, such as waste processing or 34 transportation. Also, ancillary personnel (e.g., clerical, housekeeping, security), whose duties 35 may require them to work in the vicinity of radioactive material (whether escorted or not), need 36 to be informed about radiation hazards and the appropriate precautions. The licensee should 37 assess each individuals involvement with licensed material and cover each applicable subject 38 appropriately. Additionally, some form of familiarization training or briefing may be necessary for 39 visitors to the site.
40 41 Training may be in the form of lectures, demonstrations, videotape, or self-study and should 42 emphasize practical subjects important to the safe use of licensed material. The program should 43 consider both the topics pertinent for each group of workers and the method and frequency of 44 training. The person conducting the training should be a qualified individual (i.e., a person who 45 meets the qualifications for RSO or AU on the license and is familiar with the licensees 46 program).
47 48 Regulatory Guidance: The following regulatory guidance may be helpful in assessing the 49 qualifications of proposed personnel:
50 51 11
 
1
* American Nuclear Society/Health Physics Society (ANSI/HPS) N13.36, Ionizing 2        Radiation Safety Training for Workers, October 30, 2001 3
4
* American Society for Testing and Materials (ASTM) El 168-95, Radiological Protection 5        Training for Nuclear Facility Workers, reapproved in 2008 6
7
* National Council on Radiation Protection and Measurements (NCRP) Report No. 134, 8        Operational Radiation Safety Training, 2000 9
10
* Regulatory Guide 1.8, Qualification and Training of Personnel for Nuclear Power Plants 11 12
* Regulatory Guide 8.13, Instruction Concerning Prenatal Radiation Exposure 13 14
* Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation 15        Exposure 16 17 2.2    Radiation Protection Program 18 19 Regulatory Requirements: 10 CFR Part 19, 10 CFR Part 20, 10 CFR Part 70, 10 CFR 73.67, 20 10 CFR 21.21, 10 CFR 71.5.
21 22 Regulatory Criteria: An application for a license to possess a critical mass of SNM must 23 provide the following information:
24 25
* Outline the RPP structure and define the responsibilities of key program personnel.
26 27
* Staff the RPP with suitably trained people, provide sufficient resources, and carry out the 28        program.
29 30
* Commit to the radiation protection functions independence from the facilitys operations.
31 32
* Review, at least annually, the content and implementation of the RPP, in accordance 33        with 10 CFR 20.1101(c). The review should consider facility changes, new technologies, 34        and other process enhancements that could improve the effectiveness of the overall 35        program.
36 37
* Take steps to keep radiation exposures ALARA.
38 39
* Describe equipment and facilities adequate to protect personnel, the public, and the 40        environment.
41 42
* Ensure conduct of licensed activities by individuals qualified by training and experience.
43 44
* Write operating and emergency procedures to address all likely scenarios.
45 46
* Describe the organizational structure and the individuals responsible for ensuring 47        implementation of the radiation safety and security program.
48 49
* Manage records.
12
 
1
* Implement an audit program to ensure that the licensee reviews the radiation safety and 2          security programs at least annually.
3 4 Discussion: The NRC recognizes that effective radiation safety program management is vital to 5 achieving safe and compliant operations. Compliance with NRC regulations provides 6 reasonable assurance that licensed activities will take place safely. Effective management will 7 result in increased safety and compliance.
8 9 The RPP must address the occupational radiation protection measures in 10 CFR Parts 19, 20, 10 and 70. The intent of a satisfactory RPP is to safely control the receipt, possession, use, 11 transfer, and disposal of licensed material such that the total dose to an individual does not 12 exceed the standards for protection against radiation prescribed in the regulations. Specifically, 13 licensees must develop, document, and conduct an RPP in accordance with 10 CFR 20.1101.
14 15 Regulatory Guidance: The following regulatory guidance discusses methods and procedures 16 that the NRC staff considers acceptable for maintaining radiation exposures to occupational 17 workers and the public ALARA:
18 19
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational and Public 20          Radiation Exposures As Low As Is Reasonably Achievable 21 22 2.2.1    ALARA Program 23 24 Regulatory Requirements: 10 CFR 20.1101(b); 10 CFR 20.1101(d); 10 CFR 20.1402, 25 Radiological criteria for unrestricted use; 10 CFR 20.1702(a); 10 CFR 20.2002(d).
26 27 Regulatory Criteria: Pursuant to 10 CFR 20.1101, applicants and licensees shall establish a 28 written and comprehensive ALARA program, preparing procedures and policies to ensure that 29 occupational and public radiation exposures, as well as releases to the environment, are 30 maintained ALARA.
31 32 Discussion: A central condition to an effective RPP is maintaining exposures to radiation 33 ALARA. There are regulatory limits for occupational and public exposures to radiation and 34 radioactive materials. An ALARA program is established to ensure that the licensee designs its 35 processes and procedures to keep exposures as low as possible and lower than any regulatory 36 limit where possible.
37 38 Licensees should make every reasonable effort to maintain radiation exposures as far below the 39 limits specified in that part as practicable. Two basic conditions are considered necessary in any 40 program for keeping occupational exposures as far below the specified limits as is reasonably 41 achievable. The management of the licensed facility should be committed to maintaining 42 exposures ALARA, and the personnel responsible for radiation protection should be continually 43 vigilant in seeking means to reduce exposures. The commitment made by licensee 44 management to minimize exposures should provide clearly defined radiation protection 45 responsibilities and an environment in which the radiation protection staff can do its job properly.
46 Elements of an ALARA program include the following:
47 48 49 50 51 13
 
1
* establishing a written, comprehensive, and effective ALARA program 2
3
* preparing policies and procedures to ensure that occupational radiation exposures are 4        maintained ALARA 5
6
* outlining specific ALARA program goals and establishing an ALARA program 7        organization and structure, including written procedures for its implementation in plant 8        design and operations 9
10
* maintaining applicant staff awareness of the commitment of management to ALARA 11 12
* ensuring regular assessment by management to understand and evaluate exposures to 13        staff and the public 14 15
* overseeing and supervising work to evaluate process improvements resulting in 16        exposure reductions 17 18
* training staff in ALARA concepts to engage staff participation 19 20
* giving the RSO and health physics staff independent authority to stop work to prevent 21        unnecessary personnel exposures 22 23
* regularly reviewing work processes by management to seek improvements 24 25 Regulatory Guidance: The following regulatory guidance may be helpful for implementing and 26 maintaining radiation and RPP that ensures exposures to occupational workers and the public 27 are ALARA:
28 29
* Regulatory Guide 8.2, Administrative Practices in Radiation Surveys and Monitoring 30 31
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation 32        Exposures as Low as Is Reasonably Achievable 33 34
* Regulatory Guide 8.13, Instructions Concerning Prenatal Radiation Exposure 35 36
* Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection 37 38
* Regulatory Guide 8.29, Instructions Concerning Risks from Occupational Radiation 39        Exposure 40 41 2.2.2  Occupational Dose 42 43 Regulatory Requirements: 10 CFR Part 20, Subpart C, Occupational Dose Limits.
44 Regulatory Criteria: In accordance with 10 CFR 20.1101, licensees shall establish a written 45 and comprehensive ALARA program and prepare procedures and policies to ensure that 46 occupational and public radiation exposures, as well as releases to the environment, are 47 maintained ALARA.
48 14
 
1 The regulations in 10 CFR 20.1201, Occupational dose limits for adults, require licensees to 2 control occupational exposure to adults who are likely to receive an annual dose in excess of 3 any of the following (each evaluated separately):
4 5
* 5 rem (0.05 Sv) total effective dose equivalent (TEDE) 6 7
* 15 rem (0.15 Sv) eye dose equivalent 8
9
* 50 rem (0.5 Sv) shallow-dose equivalent to the skin 10 11
* 50 rem (0.5 Sv) shallow-dose equivalent to any extremity 12 13 In accordance with 10 CFR 20.1207, Occupational dose limits for minors, the annual 14 occupational limit for minors is 10 percent of the annual dose limits specified for adult workers:
15 16
* 0.5 rem (0.005 Sv) TEDE 17 18
* 1.5 rem (0.015 Sv) eye dose equivalent 19 20
* 0.5 rem (0.05 Sv) shallow-dose equivalent to the skin 21 22
* 0.5 rem (0.05 Sv) shallow-dose equivalent to any extremity 23 24 In accordance with 10 CFR 20.1208, Dose equivalent to an embryo/fetus, for declared 25 pregnant women the dose limit to the embryo/fetus is 0.5 rem (5 mSv) for the entire gestation 26 period.
27 28 In accordance with 10 CFR 20.1502, Conditions requiring individual monitoring of external and 29 internal occupational dose, each licensee shall monitor occupational exposure to radiation from 30 licensed and unlicensed radiation sources under their control and shall supply and require the 31 use of individual monitoring devices by:
32 33
* individuals entering a high or very high radiation area 34 35
* adults likely to receive in 1 year an intake in excess of 10 percent of the applicable 36          annual limits on intake (ALIs) for ingestion and inhalation 37 38
* minors and declared pregnant women likely to receive in 1 year a committed effective 39          dose equivalent in excess of 0.1 rem (1.0 mSv) 40 41 Discussion: In accordance with 10 CFR 20.1502, if an adult (individual) is likely to receive in 42 1 year a dose greater than 10 percent of any applicable limit, monitoring for occupational 43 exposure is required. Employees who regularly work with radioactive materials and have the 44 potential to receive an occupational exposure should be monitored with approved devices as 45 part of their employment. Not every employee on site needs to be monitored. There is typically a 46 variety of work that does not require monitoring, perhaps due to its administrative nature or to 47 the fact that the employee has no access to licensed materials and associated exposure.
48 Monitoring should only be conducted when the work environment calls for it. The licensee may 49 evaluate the dose the individual is likely to receive before allowing the individual to receive the 50 dose, to understand whether this meets the threshold requirement for monitoring. This 15
 
1 documented evaluation need not be made for every individual; evaluations can be made for 2 employees with similar job functions or work areas. Appendix D to this document and 3 Regulatory Guide 8.34, Monitoring Criteria and Methods To Calculate Occupational Radiation 4 Doses, contain further guidance on evaluating the need to provide monitoring. This evaluation 5 must be documented and reflect the work processes ongoing. If work processes change, the 6 licensee should consider a reevaluation and, if not done, document the reasons for not doing 7 the evaluation and make them available for review.
8 9 If this prospective evaluation shows that an individuals dose is not likely to exceed 10 percent 10 of any applicable regulatory limit, there are no recordkeeping or reporting requirements 11 regarding the individuals exposure. For individuals who have received doses at other facilities 12 in the current year, the licensee need not consider the previous dose in this prospective 13 evaluation. When determining the need for monitoring and associated recordkeeping and 14 reporting, only the dose that could be received at the facilities of the applicant or licensee 15 performing the evaluation needs to be considered. If it were determined that monitoring was not 16 required, and a subsequent evaluation shows that the 10 percent regulatory threshold has been 17 or will be exceeded, the dose received by an individual when monitoring was not provided 18 should be estimated, recorded, and reported (if required). These estimates can be based on any 19 combination of work location radiation monitoring, survey results, monitoring results of 20 individuals in similar work situations, or other estimates to produce a best estimate of the 21 actual dose received.
22 23 Licensees should use NRC Form 4, Cumulative Occupational Dose History, and NRC Form 5, 24 Occupational Dose Record for a Monitoring Period, to record the individual dose. If monitoring 25 is not required to demonstrate compliance with all limits but is required relative to one or more 26 specific limits, the licensee should enter N/A for not applicable in the blocks on NRC Form 4 27 and NRC Form 5 to indicate the areas for which monitoring was not required (e.g., extremity or 28 skin doses). Where monitoring was provided but not measurable, the licensee should enter 29 ND for not detectable.
30 31 Licensees should also perform prospective evaluations of the doses that may be received by 32 occupationally exposed minors and declared pregnant women. As with individual adult workers, 33 licensees must supply and require the use of individual monitoring devices to monitor external 34 exposures and the occupational intake of radioactive material when the results of prospective 35 dose evaluations exceed the doses specified in 10 CFR 20.1502.
36 37 When personnel dosimeters that require processing to determine the radiation dose are used to 38 comply with the individual monitoring requirement for external doses in 10 CFR 20.1502(a),
39 dosimeters must be processed by a National Voluntary Laboratory Accreditation Program 40 (NVLAP)-accredited processor (10 CFR 20.1501(d)). The exchange frequency for dosimeters is 41 typically monthly or quarterly. Applicants should consult with their NVLAP-accredited processor 42 for its recommendations for exchange frequency and proper use of the dosimeter.
43 44 Guidance about methodologies to determine internal occupational dose and summation of 45 occupational dose appears in Regulatory Guide 8.7, Instructions for Recording and Reporting 46 Occupational Radiation Exposure Data; Regulatory Guide 8.34; and Regulatory Guide 8.9, 47 Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program.
48 49 For all employees for whom it is determined monitoring is required, dosimetry used will be 50 accredited by the NVLAP. In addition, the applicant must maintain records of the occupational 51 exposures of employees working under the license. Documentation may require determination 16
 
1 of any occupational dose received for work under a different license. Certain occupations 2 require annual reporting to the NRC.
3 4 Regulatory Guidance:
5 6
* American National Standards Institute (ANSI) N13.15-1985, Radiation Detectors 7          Personnel Thermoluminescence Dosimetry SystemsPerformance 8
9
* ANSI N13.11-2001, Personnel Dosimetry PerformanceCriteria for Testing 10 11
* ANSI N13.22-1995, Bioassay Program for Uranium 12 13
* ANSI N13.27-1981, Performance Requirements for Pocket-Sized Alarm Dosimeters 14          and Alarm Ratemeters 15 16
* ANSI N13.30-1996, Performance Criteria for Radiobioassay 17 18
* Regulatory Guide 8.4, Personnel Monitoring DeviceDirect-Reading Pocket 19          Dosimeters 20 21
* Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation 22          Exposure Data 23 24
* Regulatory Guide 8.28, Audible-Alarm Dosimeters 25 26
* Regulatory Guide 8.34, Monitoring Criteria and Methods to Calculate Occupational 27          Radiation Doses 28 29
* Regulatory Guide 8.35, Planned Special Exposure 30 31 2.2.3    Public Dose 32 33 Regulatory Requirements: 10 CFR 20.1003, Definitions; 10 CFR 20.1101; 10 CFR 20.1301, 34 Dose limits for individual members of the public; 10 CFR 20.1302, Compliance with dose 35 limits for individual members of the public; 10 CFR 20.2107, Records of dose to individual 36 members of the public.
37 38 Regulatory Criteria: Licensees must ensure that licensed material will be used, transported, 39 stored, and disposed of in such a way that members of the public will not receive more than 40 1 mSv (100 mrem) in a year; that air emissions of radioactive material to the environment, 41 excluding radon-222 and its daughters, will not result in exposures to individual members of the 42 public in excess of 0.1 mSv (10 mrem) TEDE in a year from those emissions; and the dose in 43 any unrestricted area will not exceed 0.02 mSv (2 mrem) in any 1 hour, from licensed 44 operations. In addition, licensees must strive to maintain doses to members of the public that 45 are ALARA.
46 47 Discussion: Public dose is defined in 10 CFR Part 20 as the dose received by a member of 48 the public from exposure to radiation and/or radioactive material released by a licensee, or to 49 any other source of radiation under the control of a licensee. Public dose excludes doses 50 received from background radiation and from medical procedures. Whether the dose to an 17
 
1 individual is an occupational dose or a public dose depends on the individuals assigned duties.
2 It does not depend on the area (restricted, controlled, or unrestricted) the individual is in when 3 the dose is received.
4 5 To the extent practical, procedures and engineering controls based upon sound radiation 6 protection principles are carried out to achieve and maintain doses to members of the public 7 ALARA and in accordance with the following 10 CFR Part 20 requirements:
8 9
* The radiation dose received by individual members of the public does not exceed 1 mSv 10          (100 mrem) in 1 calendar year resulting from the licensees possession or use of 11          licensed materials.
12 13
* The radiation dose in unrestricted areas does not exceed 0.02 mSv (2 mrem) in any 14          1 hour.
15 16
* The air emissions of radioactive materials do not result in doses greater than the 17          constraint limit of 0.1 mSv (10 mrem).
18 19 Members of the public include persons who live, work, study, or may be near locations where 20 licensed material is used or stored, and employees whose assigned duties do not include the 21 use of licensed material but may work in the vicinity where such materials are used or stored.
22 23 Licensees must ensure that licensed material will be used, transported, stored, and disposed of 24 in such a way that members of the public will not receive more than 1 mSv (100 mrem) in 25 1 year, and the dose in any unrestricted area will not exceed 0.02 mSv (2 mrem) in any 1 hour, 26 from licensed operations. In addition, licensees must strive to maintain doses to members of the 27 public that are ALARA.
28 29 Typical unrestricted areas may include offices, shops, laboratories (where licensed material is 30 not used or stored), areas outside buildings, property, and storage areas. The licensee does not 31 control access to these areas for purposes of controlling exposure to radiation or radioactive 32 materials, but the licensee may control access to these areas for other reasons, such as 33 security.
34 35 Pursuant to 10 CFR 20.1302, an applicant is required to demonstrate compliance with the dose 36 restrictions. Appendix E to this document and Regulatory Guide 8.34 contain information on 37 methods to evaluate conditions and demonstrate compliance with the article. This must be a 38 documented, reviewable evaluation. If work processes change such that the potential dose to 39 the public might change, it should be reevaluated.
40 41 There are many possible dose pathways that contribute to the TEDE. The TEDE can, however, 42 be broken down into the following three major dose pathway groups:
43 44 (1)      airborne radioactive material (e.g., inhalation) 45 46 (2)      waterborne radioactive material (e.g., ingestion) 47 48 (3)      external radioactive exposure (e.g., source) 49 50 18
 
1 The licensee should review these major pathways and decide which are applicable to its 2 operations.
3 4 Licensees should design a monitoring program to ensure compliance with 10 CFR 20.1302(b).
5 The extent and frequency of monitoring will depend upon each licensees needs. Additional 6 guidance regarding monitoring of effluents appears in section 2.3.3.
7 Under 10 CFR 20.2107, licensees must maintain records sufficient to demonstrate compliance 8 with the dose limits for members of the public until the Commission terminates the license.
9 Appendix E to this document contains additional guidance regarding compliance with the 10 recordkeeping requirements.
11 12 Regulatory Guidance:
13 14
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation 15        Exposures as Low as Is Reasonably Achievable 16
* Regulatory Guide 8.34, Monitoring Criteria and Methods to Calculate Occupational 17        Radiation Doses 18 19 2.2.4  Minimization of Contamination 20 21 Regulatory Requirements: 10 CFR 20.1406, Minimization of contamination.
22 23 Regulatory Criteria: Part of an effective ALARA program is planning and consideration for 24 work processes to minimize the spread of contamination. Applicants must describe how facility 25 design and procedures for operation will minimize, to the extent practicable, contamination of 26 the facility and the environment; facilitate eventual decommissioning; and minimize, to the 27 extent practicable, the generation of radioactive waste.
28 29 When designing facilities and developing procedures for their safe use, applicants should think 30 ahead and consider how to minimize radioactive contamination during operation, 31 decontamination and decommissioning efforts, and radioactive waste generation. When 32 submitting new applications, applicants should consider the following:
33 34
* implementation of and adherence to good health physics practices in operations 35 36
* minimization of areas, to the extent practicable, where licensed materials are used and 37        stored 38 39
* maximization of the frequency of surveys, within reason, to minimize spread of 40        contamination in the event of a spill 41 42
* choice of isotope to be used, whenever practical, in consideration of half-life and 43        chemical composition 44 45
* appropriate filtration of effluent streams 46 47
* use of nonporous materials, for example, for laboratory bench tops and flooring 48 49 19
 
1
* ventilation stacks and ductwork with minimal lengths and minimal abrupt changes in 2          direction 3
4
* use of appropriate plumbing materials with minimal pipe lengths and traps 5
6
* minimization of the number of disposal sites (sinks) used for liquid waste disposal 7
8 Discussion: Sealed sources and devices (SSDs) that are approved by the NRC or an 9 Agreement State and located and used according to their SSD registration certificates usually 10 pose little risk of contamination. Leak tests performed as specified in the SSD registration 11 certificate should identify defective sources. Leaking sources must be immediately withdrawn 12 from use and decontaminated, repaired, or disposed of according to NRC requirements. These 13 steps minimize the spread of contamination and reduce radioactive waste associated with 14 decontamination efforts. Other efforts to minimize radioactive waste do not apply to programs 15 using only SSDs that have not leaked.
16 17 This is not an item for an application per se but is addressed here as part of the applicants 18 commitment to written procedures. It is an integral part of an ALARA program and part of an 19 effective RPP.
20 21 Regulatory Guidance:
22 23
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation 24          Exposures as Low as Is Reasonably Achievable 25 26
* Regulatory Guide 8.24, Health Physics Surveys During Enriched Uranium-235 27          Processing and Fuel Fabrication 28 29 2.2.5    Ventilation and Respiratory Protection 30 31 Regulatory Requirements: 10 CFR 20.1101(b), 10 CFR 20.1201(a)(ii), 10 CFR 20.1202(b),
32 10 CFR 20.1204(a), 10 CFR 20.1502(b), 10 CFR 20.1702(a), 10 CFR 20.1703(a),
33 10 CFR 20.1703(c).
34 35 Regulatory Criteria: Applicants must describe the potential for internal exposure and the 36 measures that will be taken to minimize internal dose. The regulations in 10 CFR 20.1701 37 require licensees to use process or engineering controls to the extent practical to control the 38 concentration of radioactive material in air. Licensees should use respiratory protection devices 39 only after considering other measures to limit intake.
40 41 Discussion: Another consideration for any RPP is the need for respiratory protection or 42 additional ventilation. Work with unsealed material or processes that can cause radioactive 43 materials to become airborne require assessment. Surveys will be part of an effective ALARA 44 program if there is a potential for airborne contamination (see appendix F). Proper engineering 45 controls are carried out for all research involving radioactive materials. Hazard assessments are 46 made and documented, to include dose and dose rates, contamination controls, effluent 47 releases, and an evaluation of the need for ventilation controls. Licensees can use health 48 physics surveys of airborne radioactivity concentration to evaluate process and engineering 49 controls, conduct increased surveillance, determine exposure time limits for workers, and 50 support a program for the use of respiratory protective equipment.
20
 
1 It should be noted that engineering controls are the preferred means of preventing airborne 2 contamination and that thorough justification will be required if an applicant chooses to use 3 respiratory protective equipment to control internal exposure. Respiratory protection will require 4 the need for other occupational safety considerations, such as heat stress. Uncontrolled areas 5 inside a plant may require a survey periodically to ensure there is no effect on the general 6 public. The frequency of routine surveys should be commensurate with the nature of the work; 7 the quantities of material being processed; and the specific protective facilities, equipment, and 8 procedures used to protect workers.
9 10 The regulations in 10 CFR 20.1902(d) describe posting requirements for such areas. The use of 11 other precautionary procedures should be considered, such as increased surveillance and 12 limitation of exposure times. The applicant that requires a respiratory protection program needs 13 to address the following:
14 15
* Install appropriately sized ventilation and containment systems in areas of the plant 16          identified as having potential airborne concentrations of radionuclides that could exceed 17          the occupationally derived air concentration values specified in 10 CFR Part 20, 18          Appendix B, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of 19          Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for 20          Release to Sewerage, during normal operations.
21 22
* Describe management measures, including preventive and corrective maintenance and 23          performance testing, to ensure that the ventilation and containment systems operate 24          when required and are within their design specifications.
25 26
* Describe the criteria for the ventilation and containment systems, including minimum 27          flow velocity at openings in these systems, maximum differential pressure across filters, 28          and types of filters to be used.
29 30
* Describe the frequency and types of tests to measure the performance of ventilation and 31          containment systems, the acceptance criteria, and the actions to be taken when the 32          acceptance criteria are not satisfied.
33 34
* Establish a respiratory protection program that meets the requirements of 35          10 CFR Part 20, Subpart H, Respiratory Protection and Controls to Restrict Internal 36          Exposure in Restricted Areas.
37 38
* Prepare written procedures for the selection, fitting, issuance, maintenance, testing, 39          training of personnel, monitoring, and recordkeeping for individual respiratory protection 40          equipment and for specifying when such equipment is to be used.
41 42
* Revise the written procedures for the use of individual respiratory protection equipment, 43          as applicable, when making changes to processing, facility, or equipment.
44 45
* Maintain records of the respiratory protection program, including training in respirator 46          use.
47 Appendix B to this document contains further elaboration on respiratory protection.
48 49 21
 
1 Regulatory Guidance:
2 3
* American Conference of Governmental Industrial Hygienists 2095, Industrial Ventilation:
4        A Manual of Recommended Practice for Design, 2007 5
6
* ANSI Z88.2-1992, Practices for Respiratory Protection 7
8
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation 9        Exposures as Low as Is Reasonably Achievable 10 11
* Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection 12 13
* Regulatory Guide 8.24, Health Physics Survey for Enriched Uranium-235 Processing 14        and Fuel Fabrication 15 16
* Regulatory Guide 8.25, Air Sampling in the Workplace 17 18 2.2.6  Waste Management 19 20 Regulatory Requirements: 10 CFR 20.1501, General; 10 CFR 20.1904, Labeling 21 containers; 10 CFR 20.2001, General requirements; 10 CFR 20.2002, Method for obtaining 22 approval of proposed disposal procedures; 10 CFR 20.2003, Disposal by release into sanitary 23 sewerage; 10 CFR 20.2004, Treatment or disposal by incineration; 10 CFR 20.2005, 24 Disposal of specific wastes; 10 CFR 20.2006, Transfer for disposal and manifests; 25 10 CFR 20.2007, Compliance with environmental and health protection regulations; 26 10 CFR 20.2108, Disposal of certain byproduct material; 10 CFR 37.11(c); 10 CFR 70.51, 27 Records requirements; 10 CFR 71.43, Lifting and tie-down standards for all packaging.
28 29 Regulatory Criteria: Radioactive waste must be disposed of in accordance with regulatory 30 requirements and license conditions. Appropriate records of waste disposal must be maintained.
31 32 Discussion: Radioactive waste is normally generated when performing licensed activities. Such 33 waste may include used or unused radioactive material and unusable items contaminated with 34 radioactive material (e.g., absorbent paper, gloves). Licensees may not receive radioactive 35 waste from other licensees for processing, storage, or disposal, unless the NRC specifically 36 authorizes them to do so.
37 38 All radioactive waste must be stored in appropriate containers until its disposal, and the integrity 39 of the waste containers must be assured. Radioactive waste containers must be appropriately 40 labeled. All radioactive waste must be secured against unauthorized access or removal. The 41 NRC requires licensees to manage radioactive waste generated at their facilities by one or more 42 of the following methods:
43 44
* transfer to an authorized recipient 45 46
* release into sanitary sewerage 47 48
* extended interim storage 49 50
* prior approval by the NRC of any alternate method 22
 
1
* release in effluents to unrestricted areas, other than into sanitary sewerage 2
3
* incineration 4
5 Licensees may choose any one or more of these methods to dispose of their radioactive waste.
6 It has been the NRCs experience that many of the facilities dispose of SNM by the first method.
7 Applicants should describe their program for management and disposal of radioactive waste.
8 The program should include procedures for handling of waste, safe and secure storage, 9 characterization, minimization, and disposal of radioactive waste. Appropriate training should be 10 provided to waste handlers. Regulations require licensees to maintain all appropriate records of 11 disposal of radioactive waste.
12 13 Regulatory Guidance:
14 15
* Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste 16        Generation: Life-Cycle Planning 17 18 2.2.7  Audit Program 19 20 Regulatory Requirements: 10 CFR 20.1101; 10 CFR 20.2102, Records of radiation protection 21 programs; 10 CFR 21.21(a); 10 CFR 37.33, Access authorization program review; 22 10 CFR 37.55, Security program review.
23 24 Regulatory Criteria: Licensees must review the content and implementation of their RPPs at 25 least annually to ensure the following:
26 27
* Compliance with NRC and U.S. Department of Transportation (DOT) regulations (as 28        applicable) is assured, and the terms and conditions of the license are met.
29 30
* Occupational doses and doses to members of the public are ALARA (10 CFR 20.1101).
31 32
* Records of audits and other reviews of program content are maintained for 3 years.
33 34
* The access authorization program is effective and complies with 10 CFR Part 37, 35        Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material, if 36        required.
37 38
* The security program is effective and complies with 10 CFR Part 37, Subpart C, 39        Physical Protection Requirements During Use, if required.
40 41
* Records of audits and other reviews of program content are maintained for 3 years after 42        the record is made.
43 44 Discussion: If an audit identifies violations of NRC requirements, the licensee should first 45 evaluate the safety significance of each violation to set priorities and identify resources to 46 correct these violations. Information Notice (IN) 96-28, Suggested Guidance Relating to 47 Development and Implementation of Corrective Action, dated May 1, 1996, provides guidance 48 on this subject. Certain identified problems or potential violations may require notification or a 49 report to the NRC. Licensees are encouraged to contact the NRC for guidance if there is any 50 uncertainty regarding a reporting requirement. The NRC routinely reviews a licensees records 23
 
1 to verify if appropriate corrective actions were taken promptly to prevent recurrence. It is in the 2 best interest of the licensee to identify potential violations of regulatory requirements and take 3 the necessary steps to correct them. The NRC can exercise discretion and may elect not to cite 4 the licensee for these violations if it takes prompt and effective corrective actions.
5 NUREG-1600, General Statement of Policy and Procedure for NRC Enforcement Actions, 6 issued May 2000, contains information on the NRCs use of discretion on issuing a notice of 7 violation.
8 9 Additionally, 10 CFR 20.2102 requires licensees to keep records of the RPP, including a 10 description of the program components, audits, and other aspects of program implementation 11 for 3 years from the date of the record. Records of these audits should include the date of audit, 12 name of person(s) who conducted the audit, persons contacted by the auditor(s), areas audited, 13 audit findings, corrective actions, and follow-up. Licensees must maintain these records for NRC 14 inspections. NUREG-1556, Volume 17, Appendix B, Suggested Format for Providing 15 Information Requested in Items 5 Through 11 of U.S. Nuclear Regulatory Commission 16 Form 313, lists considerations for the audit of an RPP.
17 18 The NRC encourages licensee management to conduct performance-based reviews by 19 observing work in progress, interviewing staff, and spot-checking required records. As a part of 20 the audit program, licensees should consider including unannounced audits of users of SNM of 21 less than critical mass to determine whether radiation safety procedures are being followed.
22 Licensees should consider providing specialized audit training if staff other than the RSO are 23 used to conduct audits of the performance of licensed materials users. Appendix E to this 24 document contains applicable audit program areas for potential specialized training. It is 25 essential that, once identified, problems are corrected comprehensively and in a timely manner.
26 IN 96-28 provides guidance on this subject. The NRCs Enforcement Policy may be found online 27 at https://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html. The NRC 28 Enforcement Manual may be found online at https://www.nrc.gov/about-29 nrc/regulatory/enforcement/guidance.html. For examples of the NRCs use of discretion in 30 issuing a notice of violation, refer to the most recent version of NRCs enforcement documents 31 at https://www.nrc.gov/reading-rm/doc-collections/enforcement/.
32 33 Licensees must maintain records of audits and other reviews of program content and 34 implementation for 3 years from the date of the record, in accordance with 10 CFR 20.2102.
35 The NRC has found acceptable those audit records that contain the following information:
36 37
* date of audit 38 39
* name of person or persons who conducted the audit 40 41
* names of persons contacted by the auditor or auditors 42 43
* areas audited 44 45
* audit findings and corrective actions 46 47
* follow-up 48 49 24
 
1 2.3      Facilities and Equipment 2
3 Regulatory Requirements: 10 CFR 20.1101; 10 CFR 20.1406; 10 CFR Part 37; 10 CFR 37.5; 4 10 CFR 37.49, Monitoring, detection, and assessment; 10 CFR 37.53, Requirements for 5 mobile devices; 10 CFR 70.22(a)(7); 10 CFR 70.23(a)(3); 10 CFR 70.25(g); 10 CFR 70.41(a).
6 7 Regulatory Criteria: Facilities and equipment must be adequate to protect health and minimize 8 danger to life or property. Facilities and equipment must also provide enhanced physical 9 protection of aggregated Category 1 and Category 2 quantities of radioactive material, as 10 defined in 10 CFR 37.5. Applicants must describe how facility design and procedures for 11 operation will minimize, to the extent practicable, contamination of the facility and the 12 environment; facilitate eventual decommissioning; and minimize, to the extent practicable, the 13 generation of radioactive waste.
14 15 Discussion: Consistent with 10 CFR 70.22(a)(7), facilities and equipment must be adequate to 16 protect health and minimize danger to life and property. Applicants must demonstrate that their 17 facilities and equipment provide sufficient engineering controls and barriers to protect the health 18 and safety of the public and its employees, keep exposures to radiation and radioactive 19 materials ALARA, and minimize the danger to life and property from the types and quantities of 20 radioactive materials to be used. The application should contain detailed descriptions and 21 diagrams of the facilities, including information about the shielding properties of the construction 22 materials used. Restricted areas are defined as areas to which access is limited by the licensee 23 to protect individuals against undue risks from exposure to radiation and radioactive materials.
24 These should be identified, as well as radioactive material storage areas.
25 26 Licensees should include a description of the area(s) assigned for the receipt, storage, security, 27 preparation, and measurement of radioactive materials. They should submit a diagram showing 28 the locations of the shielding, the proximity of radiation sources to unrestricted areas, and other 29 items related to radiation safety. When applicable to facilities where radioactive materials may 30 become airborne, the diagrams should contain schematic descriptions of the ventilation 31 systems, with pertinent airflow rates, pressures, filtration equipment, and monitoring systems.
32 Appendix B to this document provides a variety of considerations for facility design and 33 equipment. Not all of these would apply to every licensee but may be helpful in planning.
34 35 Modifications to operating and maintenance procedures and to plant equipment and facilities 36 should be made where they will substantially reduce exposures at a reasonable cost.
37 38 Applicants may delay completing facilities and acquiring equipment until after the application 39 review is completed in case changes are required. This also ensures the adequacy of the 40 facilities and equipment before the applicant makes a significant financial commitment. In all 41 cases, the applicant cannot possess or use licensed material until after the facilities are 42 approved and completed, equipment is procured and ready for use, and the license is issued.
43 44 Section 2.4 of this document includes additional information regarding radiation monitoring, 45 respiratory protection, dosimetry, and bioassay.
46 47 Applicants are reminded that records important to decommissioning include the following:
48 49
* as-built drawings and modifications of structures and equipment in restricted areas 50 51 25
 
1
* as-built drawings and modifications of locations of possible inaccessible contamination, 2        such as buried pipes, that may be subject to contamination 3
4
* records of spills and unusual occurrences that may result in contamination of the facility 5        or site 6
7 These records are required to be maintained in an identifiable location. Facilities are required to 8 meet NRC criteria before release. Therefore, careful facility design is important to prevent 9 contamination, facilitate decontamination, and reduce the costs needed for decommissioning.
10 11 Regulatory Guidance:
12 13
* Regulatory Guide 4.7, General Site Suitability Criteria for Nuclear Power Stations 14 15 2.3.1  Radiation Monitoring Instruments 16 17 Regulatory Requirements: 10 CFR 20.1501, 10 CFR 20.2103(a), 10 CFR 70.22(a)(7).
18 19 Regulatory Criteria: Pursuant to 10 CFR 20.1501, licensees must possess, or have access to, 20 radiation monitoring instruments that are necessary to protect health and minimize danger to life 21 or property. Instruments used for quantitative radiation measurements must be calibrated 22 periodically for the radiation measured.
23 24 Discussion: Monitoring of personnel and working areas are essential to an effective ALARA 25 program. Radiation surveys are done for two purposes: (1) to ascertain radiation levels, 26 concentrations of radioactive material, and potential radiological hazards that could be present 27 in the facility, and (2) to detect releases of radioactive material from plant equipment and 28 operations. Radiation surveys will focus on those areas of the plant necessary to show 29 compliance with the dose limits and monitoring requirements of 10 CFR Part 20, Subpart C; 30 Subpart D; and Subpart F.
31 32 Licensees shall possess, or have access to, calibrated radiation detection and measurement 33 instruments or licensed services to perform, as necessary, the following:
34 35
* package surveys 36 37
* contamination surveys 38 39
* sealed source leak tests 40 41
* air sampling measurements 42 43
* bioassay measurements 44 45
* effluent release measurements 46 47
* unrestricted area dose rate measurements 48 49 The choice of instrument should be appropriate for the type of radiation to be measured and for 50 the type of measurement to be taken (e.g., count rate, dose rate). Most of the radioactive 26
 
1 emissions from SNM are alpha emissions. Specific gamma or x-ray emissions are associated 2 with alpha decay, and a standard Geiger-Mueller detector will give positive confirmation of the 3 presence or absence of loose material. When more information is required for nuclide 4 identification, the applicant should include instrumentation capable of detecting alpha emissions, 5 such as zinc sulfide (ZnS) detectors. Applications should describe the instrumentation available 6 for use and may include instrumentation applicants intend to purchase before starting licensed 7 activities. The description should include the type of instrument and probe and the instruments 8 intended purpose.
9 10 Instruments used for qualitative surveys are only intended to detect contamination in the 11 laboratory. Such instruments should be checked for operational response with an appropriate 12 check source containing radioactive material and can be calibrated with an electronic pulser 13 instead of a radioactive source. However, these instruments cannot be used for measurement 14 of surface contamination or radiation levels without calibration with appropriate radioactive 15 sources.
16 17 For the purposes of this document, survey instruments are defined as any device used to 18 measure the radiological conditions at a licensed facility. Some of the survey instruments that 19 may be used to perform the above functions include the following:
20 21
* portable or stationary count rate meters 22 23
* portable or stationary dose rate or exposure rate meters 24 25
* single or multichannel analyzers 26 27
* liquid scintillation counters 28 29
* gamma counters 30 31
* proportional counters 32 33
* ZnS detectors 34 35
* neutron detectors 36 37
* solid state detectors 38 39 The NRC requires that calibrations be made by the instruments manufacturer or a person 40 specifically authorized by the NRC or an Agreement State unless the applicant specifically 41 requests this authorization. Applicants seeking authorization to do survey instrument 42 calibrations shall submit procedures for review. Records of calibrations shall be retained in 43 accordance with regulatory guidance. Appendix C to this document provides additional 44 guidance.
45 46 Regulatory Guidance:
47 48
* Regulatory Guide 8.2, Administrative Practices in Radiation Surveys and Monitoring 49 50 27
 
1
* Regulatory Guide 8.4, Personnel Monitoring DeviceDirect-Reading Pocket 2        Dosimeters 3
4
* Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation 5        Exposure Data 6
7
* Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a 8        Bioassay Program 9
10
* Regulatory Guide 8.24, Health Physics Surveys During Enriched Uranium-235 11        Processing and Fuel Fabrication 12 13 2.3.2  Leak Tests 14 15 Regulatory Requirements: 10 CFR 20.1501; 10 CFR 20.2103, Records of surveys; 16 10 CFR 70.56, Tests.
17 18 Regulatory Criteria: The NRC requires testing to detect radioactive leakage from the sealed 19 sources. Records of test results must be maintained in accordance with license conditions and 20 NRC regulations. The measurement of the leak-test sample is a quantitative analysis requiring 21 that instrumentation used to analyze the sample be capable of detecting 185 becquerel 22 (0.005 microcurie) of radioactivity.
23 24 Discussion: Manufacturers, consultants, and other organizations may be authorized by the 25 NRC or an Agreement State either to do the entire leak test sequence for other licensees or to 26 provide leak test kits to licensees. In the latter case, the licensee is expected to take the leak 27 test sample according to the kit suppliers instructions and return it to the kit supplier for 28 evaluation and reporting results. Leak test samples should be collected at the most accessible 29 area where contamination would accumulate if the sealed source were leaking. Licensees may 30 also be authorized to do the entire leak test sequence themselves. If an applicant chooses to do 31 leak testing, it will submit procedures for review and approval.
32 33 Regulatory Guidance:
34 35
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation 36        Exposures as Low as Is Reasonably Achievable 37 38
* Regulatory Guide 8.24, Health Physics Surveys During Enriched Uranium-235 39        Processing and Fuel Fabrication 40 41 2.3.3  Surveys 42 43 Regulatory Requirements: 10 CFR 20.1101; 10 CFR 20.1501; 10 CFR 20.1906, Procedures 44 for receiving and opening packages; 10 CFR 20.2103.
45 46 Regulatory Criteria: The regulations in 10 CFR 20.1501 and 10 CFR 20.2103 contain general 47 survey and survey recordkeeping requirements. The regulations in 10 CFR 20.1906 contain 48 general survey requirements for receiving and opening packages.
49 50 28
 
1 Discussion: The selection and proper use of radiation detection instruments is one of the most 2 important factors in ensuring that surveys accurately assess the radiological conditions. The 3 regulations in 10 CFR 20.1003 define a survey as an evaluation of the radiological conditions 4 and potential hazards incident to the production, use, transfer, release, disposal, or presence of 5 radioactive material or other sources of radiation. These evaluations may be measurements 6 (e.g., radiation levels measured with a survey instrument or results of wipe tests for 7 contamination), calculation, or a combination of measurements and calculations.
8 9 The regulations in 10 CFR 70.23(a)(4) state that the NRC will approve an application if the 10 Commission determines that the applicants proposed procedures to protect health and to 11 minimize danger to life or property are adequate. The regulations in 10 CFR 20.1101 require, 12 in general, that applicants develop, document, and implement an RPP commensurate with the 13 scope and extent of licensed activities and sufficient to ensure compliance with the provisions 14 of 10 CFR Part 20. Therefore, 10 CFR 70.22(a)(7) requires applicants to describe the 15 equipment and facilities that they will use to protect health and minimize danger to life or 16 property, such as handling devices, working areas, shields, measuring and monitoring 17 instruments, devices for the disposal of radioactive effluents and wastes, and storage facilities.
18 In addition, 10 CFR 70.22(a)(8) requires applicants to describe these proposed procedures to 19 protect health and minimize danger to life or property. To meet regulatory requirements for 20 surveying in 10 CFR 20.1501, measurements of radiological quantities should be understood in 21 terms of their properties (i.e., alpha, beta, gamma, and neutrons) and compared to the 22 appropriate limits. Licensees should also use surveys to plan work in areas where licensed 23 material or radiation exists and to evaluate doses to workers and individual members of the 24 public. In certain cases, the NRC may require environmental monitoring to demonstrate 25 compliance with 10 CFR Part 20.
26 27 The selection and proper use of appropriate instruments is one of the most important factors in 28 ensuring that surveys accurately assess radiological conditions. Radiation surveys detect and 29 evaluate contamination of the following:
30 31
* facilities 32 33
* equipment 34 35
* personnel (during use, transfer, or disposal of licensed material) 36 37
* restricted and unrestricted areas 38 39 The regulations in 10 CFR 20.1501 state that surveys are required when it is reasonable, under 40 the circumstances, to evaluate a radiological hazard and when necessary for the licensee to 41 comply with regulations. Many different types of surveys may need to be done due to the 42 particular use of licensed materials. The most important are as follows:
43 44
* surveys for radioactive contamination that could be present on surfaces of floors, walls, 45          laboratory furniture, equipment, and packages of radioactive material received or 46          prepared for shipment 47 48
* measurements of radioactive material concentrations in air for areas where radioactive 49          materials are handled or processed in unsealed form and where operations could 50 29
 
1          expose workers to the inhalation of radioactive material or where licensed material is or 2          could be released to unrestricted areas 3
4
* measurements of radioactive material concentrations in water that is released to the 5          environment or to the sanitary sewer 6
7
* bioassays to determine the kinds, quantities, or concentration, and, in some cases, the 8          location of radioactive material in the human body, which can be made by direct 9          measurement (in vivo counting) or by analysis and evaluation of material excreted or 10          removed from the human body 11 12
* surveys of external radiation exposure levels in both restricted and unrestricted areas 13 14 Appendix F to this document provides additional information. The frequency of routine surveys 15 depends on the nature, quantity, and use of radioactive materials, as well as the specific 16 protective facilities, equipment, and procedures that are designed to protect the worker from 17 external and internal exposure. Also, the frequency of the survey depends on the type of survey.
18 Not all instruments can measure a given type of radiation. The presence of other radiation may 19 interfere with a detector's ability to measure the radiation of interest. Correct use of radiation 20 detection and measurements is an important aspect of any radiation safety program. The 21 regulations in 10 CFR Part 20 do not specify limits for surface contamination. Each applicant 22 should propose and justify what removable surface contamination limits will be allowable before 23 decontamination will be carried out in each work area.
24 25 Regulatory Guidance:
26 27
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation 28          Exposures as Low as Is Reasonably Achievable 29 30
* Regulatory Guide 8.24, Health Physics Survey During Enriched Uranium-235 31          Processing and Fuel Fabrication 32 33
* Branch Technical Position, Guidelines for Decontamination of Facilities and Equipment 34          Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, 35          or Special Nuclear Materials, April 1993 36 37 2.4      Commitment to Written Procedures 38 39 Regulatory Requirements: 10 CFR 20.1101(b), 10 CFR 70.22(a)(8).
40 41 Regulatory Criteria: As required by regulations, the licensee shall use, to the extent practical, 42 procedures and engineering controls based on sound radiation protection principles to achieve 43 occupational doses to members of the public that are ALARA.
44 45 Discussion: Written, approved procedures and engineering controls are required to carry out 46 activities involving material on the application. Processes should be evaluated for workable 47 reductions in exposure through application of ALARA concepts, such as time, distance, and 48 shielding. Use of mockups of work to be performed and prestaging of materials are other 49 ALARA principles that should be considered. The applicant will establish a process for 50 procedure generation, modification, authorization, distribution, and training, such that changes 30
 
1 in technology or practices are communicated effectively and promptly. Part of the process 2 should provide for revising procedures, as necessary, to incorporate facility or operational 3 changes. The RSO, or an individual who has the qualifications of the RSO, should approve all 4 procedures related to radiation protection. The radiation work permits (RWPs) should define the 5 authorized activities, the level of approval required (a radiation specialist, as a minimum),
6 information requirements, period of validity, expiration and termination times, and recordkeeping 7 requirements.
8 9 The applicants commitment to preparing written radiation protection procedures and RWPs is 10 acceptable if the license application provides data and information that meet each of the 11 following commitments:
12 13
* Prepare written, approved procedures to carry out activities related to the RPP.
14        Procedures should address applicable radiation protection requirements found in 15        10 CFR Parts 19, 20, 70, and 71 and any other applicable regulations.
16 17
* Establish a process for procedure generation or modification, authorization, distribution, 18        and training, such that changes in technology or practices are communicated effectively 19        and in a timely manner. Review and revise procedures, as necessary, to incorporate any 20        facility or operational changes. The RSO, or an individual who has the qualifications of 21        the RSO, should approve all procedures related to radiation protection.
22 23
* Specify written, approved RWPs for activities involving licensed material that are not 24        covered by written radiation protection procedures. The RWPs should define the 25        authorized activities, the level of approval required (a radiation specialist, as a 26        minimum), information requirements, period of validity, expiration and termination times, 27        and recordkeeping requirements.
28 29 Regulatory Guidance:
30 31
* Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation 32        Exposures as Low as Is Reasonably Achievable 33 34 2.5    Operating and Emergency Procedures 35 36 Regulatory Requirements: 10 CFR 19.11(a)(3); 10 CFR 20.1101; 10 CFR 20.1406; 37 10 CFR 20.1801, Security of stored material; 10 CFR 20.1802, Control of material not in 38 storage; 10 CFR 20.2201, Reports of theft or loss of licensed material; 10 CFR 20.2202, 39 Notification of incidents; 10 CFR 20.2203, Reports of exposures, radiation levels, and 40 concentrations of radioactive material exceeding the constraints or limits; 10 CFR 21.21; 41 10 CFR Part 37, Subpart B, Background Investigations and Access Authorization Programs; 42 10 CFR 37.21(a); 10 CFR 37.45, LLEA Coordination; 10 CFR 37.49; 10 CFR 70.22(a);
43 10 CFR 70.23(a)(4); 10 CFR 70.50, Reporting requirements.
44 45 Regulatory Criteria: Each licensee should develop, implement, and maintain operating and 46 emergency procedures that include the following provisions:
47 48
* instructions to keep radiation doses to workers and members of the public ALARA 49 50 31
 
1
* instructions for conducting operations to minimize the introduction of residual 2          radioactivity into the site 3
4
* instructions for maintaining security during storage and transportation 5
6
* instructions for maintaining accountability during use 7
8
* instructions for properly storing and disposing of radioactive waste 9
10
* use of personnel monitoring and radiation survey equipment 11 12
* instructions for packaging and transporting licensed material 13 14
* instructions on how to respond and whom to contact when an emergency occurs 15 16
* instructions for identifying and reporting to the NRC defects and noncompliance, as 17          required by 10 CFR 21.21 18 19
* instructions for maintaining records, in accordance with the NRC regulations and any 20          license conditions 21 22 Discussion: There are two important aspects to operating and emergency procedures for 23 applicants. One is for the day-to-day safe use and handling of materials as they apply to the 24 application in intended uses and outcomes. An application must describe procedures for 25 processes that address routine radiation protection and safety matters. A licensee will train its 26 staff in these procedures on a regular basis. Every employee needs to have some level of 27 safety training in this area. The other important aspect in this area is planning for the safety of 28 the public and the environment. An applicant needs to plan its responses if the business of the 29 applicant causes inadvertent effects offsite of the working facility.
30 31 The emergency management plan requirements in 10 CFR 70.22(i) state that licenses 32 authorizing radioactive material exceeding certain thresholds must either (1) submit an 33 emergency plan, or (2) submit an evaluation showing that the maximum dose to a person offsite 34 due to a release would not exceed 1 rem effective dose equivalent, or an intake of 2 milligrams 35 of soluble uranium.
36 37 Licensees should develop, implement, and maintain operating and emergency procedures to 38 verify that they use all licensed materials in accordance with licensed activities, maintain control 39 and accountability, and ensure radiation doses received by occupational workers and members 40 of the public are ALARA. The operating procedures should describe the operations involving 41 SNM and include a general plan for carrying out the activity. The written procedures should 42 provide reasonable assurance that only appropriately trained personnel will handle and use 43 licensed material without undue hazard to workers or members of the public. Each licensee 44 must develop, implement, and maintain operating and emergency procedures containing the 45 following elements:
46 47
* contamination controls 48 49
* personnel and area monitoring (including frequency and limits) 50 32
 
1
* use of protective clothing and equipment 2
3
* recording requirements 4
5
* reporting requirements 6
7
* waste disposal practices 8
9 Licensees should post a copy of the operating and emergency procedures in all laboratory or 10 work areas where radioactive materials are used. If posting of procedures is not practicable, the 11 licensee may post a notice that describes the documents and states where they may be 12 examined. Licensees should also provide copies of operating and emergency procedures to all 13 authorized users. These instructions should describe immediate action to be taken in case of an 14 emergency to prevent release of radioactive material or further contamination of work areas and 15 personnel. Examples of emergency procedures include turning off the ventilation systems, 16 evacuating the area, reentering, and containing spills. The instructions should specifically state 17 the names and telephone numbers of responsible persons to be notified.
18 19 The licensee must notify the NRC when licensed material is lost or stolen, or other related 20 conditions occur. The RSO must be proactive in evaluating whether NRC notification is 21 required. Appendix F and the regulations (10 CFR 20.2201-20.2203; 10 CFR 70.50; and 22 10 CFR 21.21) describe when and where notifications are required.
23 24 Regulatory Guidance:
25 26
* NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and 27          Other Radioactive Material Licensees, January 1988 28 29
* Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste 30          Generation: Life-Cycle Planning 31 32 2.6      Transportation 33 34 Regulatory Requirements: NRC: 10 CFR 20.1101; 10 CFR Part 37, Subpart D, Physical 35 Protection in Transit; 10 CFR 70.42, Transfer of special nuclear material; 10 CFR 70.51; 36 10 CFR 71.5; 10 CFR 71.14, Exemption for low-level materials; 10 CFR 71.22, General 37 license: fissile material; 10 CFR 71.37, Quality assurance; 10 CFR 71.38, Renewal of a 38 certificate of compliance; 10 CFR 71.47, External radiation standards for all packages; 39 10 CFR 71.63, Special requirement for plutonium shipments;10 CFR 71.64, Special 40 requirements for plutonium air shipments; 10 CFR Part 71, Subpart G, Operating Controls and 41 Procedures (except 10 CFR 71.93, 71.99, and 71.100); 10 CFR Part 71, Subpart H, Quality 42 Assurance. DOT: 49 CFR Parts 171-180 and 49 CFR Parts 390-397.
43 44 Regulatory Criteria: The regulations in 10 CFR 71.5(a) require, in part, that each licensee who 45 transports licensed material outside the site of usage, as specified in the NRC license, or where 46 transport is on public highways, or who delivers licensed material to a carrier for transport, shall 47 comply with the applicable requirements of the DOT regulations in 49 CFR Parts 107, 171 48 through 180, and 390 through 397, appropriate to the mode of transport. The NRC requires 49 licensees and applicants seeking authorization to transport or ship licensed material, including 50 radioactive waste, to develop, implement, and maintain safety programs for the transport of 33
 
1 radioactive material. An application must describe the transportation program to ensure 2 adequate protection of workers and members of the public from unnecessary exposures to 3 radiation during the transportation of the material and verify compliance with NRC and DOT 4 regulations.
5 6 If the NRC authorizes a licensee to possess and use aggregated Category 1 or Category 2, or 7 both, quantities of radioactive materials and to transport or deliver licensed material to a carrier 8 for transporting these materials outside the site of usage, it must comply with the regulations in 9 10 CFR Part 37. The regulations in 10 CFR 37.75(a) state that each licensee that plans to 10 transport, or deliver to a carrier for transport, licensed material that is a Category 1 quantity of 11 radioactive material outside the confines of the licensees facility or other place of use or storage 12 shall do the following:
13 14        (1) Preplan and coordinate shipment arrival and departure times with the 15              receiving licensee.
16 17        (2) Preplan and coordinate shipment information with the governor or the 18              governors designee of any State through which the shipment will pass to:
19 20              (i) Discuss the State's intention to provide law enforcement escorts; and 21 22              (ii) Identify safe havens; and 23 24        (3) Document the preplanning and coordination activities.
25 26 The regulations in 10 CFR 37.75(b) require that each licensee that plans to transport, or deliver 27 to a carrier for transport, licensed material that is a Category 2 quantity of radioactive material 28 outside the confines of the licensees facility or other place of use or storage shall coordinate the 29 shipment no-later-than arrival time and the expected shipment arrival with the receiving 30 licensee. The licensee shall document the coordination activities.
31 32 The regulations in 10 CFR 37.75(c) require that each licensee who receives a shipment of a 33 Category 2 quantity of radioactive material shall confirm receipt of the shipment with the 34 originator. If the shipment has not arrived by the no-later-than arrival time, the receiving licensee 35 shall notify the originator.
36 37 The regulations in 10 CFR 37.75(d) require that each licensee who transports or plans to 38 transport a shipment of a Category 2 quantity of radioactive material and determines that the 39 shipment will arrive after the no-later-than arrival time provided pursuant to 10 CFR 37.75(b),
40 shall promptly notify the receiving licensee of the new no-later-than arrival time.
41 42 The regulations in 10 CFR 37.75(e) require that licensees, as a record, retain for 3 years a copy 43 of the documentation for preplanning and coordination and any revision thereof.
44 45 Discussion: The regulations in 10 CFR Part 71 establish the requirements for packaging, 46 preparing for shipment, and transporting licensed material. The regulations in 10 CFR Part 20 47 describe the requirements for controlling the handling of radioactive material. The regulations in 48 10 CFR Part 73, Physical Protection of Plants and Materials, establish the requirements for the 49 physical protection of SNM. Applicants who will transport or ship licensed material, including 50 radioactive waste, must develop, carry out, and maintain safety programs for the transport of 51 radioactive material to ensure compliance with NRC and DOT regulations.
34
 
1 Licensees should consider the safety of all individuals who may handle or come into contact 2 with the packages containing licensed material. Therefore, the primary considerations in 3 packaging licensed material should be to ensure that the package integrity is not compromised 4 during transport and that the radiation levels (including removable contamination levels) at the 5 package surfaces not only meet the regulatory requirements of 10 CFR 71.47 but are ALARA.
6 The general performance objectives of 10 CFR 73.67(a) require that each licensee possessing, 7 using, or transporting SNM of moderate or low strategic significance shall establish and 8 maintain a physical protection system that will (1) minimize the possibilities for unauthorized 9 removal of SNM consistent with the potential consequences of such actions, and (2) facilitate 10 the location and recovery of missing SNM. Chapter 10 contains more information on physical 11 security requirements. Specific requirements are dependent on the category of SNM 12 possessed, as defined in 10 CFR 70.4, Definitions.
13 14 Shipments of hazardous materials, including radioactive materials and the regulations that 15 govern them, may be complex, regulated by multiple agencies (e.g., the NRC and DOT), and 16 varied in nature. Organizations with little or no experience in transporting or providing 17 radioactive material to a carrier for transport should consult a specialist to address factors such 18 as vehicle restrictions; nuclear safety; package selection; contamination control; and, if needed, 19 physical protection of the material during transit to ensure that materials are shipped safely and 20 securely and in compliance with applicable NRC and DOT regulations.
21 22 Appendix H to this document identifies many of the regulatory requirements that should be 23 reviewed in constructing an adequate transportation program.
24 25 Regulatory Guidance:
26 27
* RAMREG-002/NUREG-1660, U.S. Specific Schedules of Requirements for Transport of 28        Specified Types of Radioactive Material Consignments, January 1999 29 30
* Regulatory Guide 7.7, Administrative Guide for Verifying Compliance with Packaging 31        Requirements for Shipping and Receiving of Radioactive Material 32 33
* Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used 34        in Transport of Radioactive Material 35 36 2.7    References 37 38 American Conference of Governmental Industrial Hygienists 2095, Industrial Ventilation: A 39 Manual of Recommended Practice for Design, 2007.
40 41 American National Standards Institute (ANSI), Personnel Dosimetry PerformanceCriteria for 42 Testing, ANSI N13.11-2001.
43 44 ANSI, Radiation Detectors - Personnel Thermoluminescence Dosimetry Systems 45 Performance, ANSI Ni 3.15-1985.
46 47 ANSI, Bioassay Program for Uranium, ANSI N13.22-1995.
48 49 ANSI, Performance Requirements for Pocket-Sized Alarm Dosimeters and Alarm Ratemeters, 50 ANSI N13.27-1981.
35
 
1 ANSI, Performance Criteria for Radiobioassay, ANSI N13.30-1996.
2 3 ANSI/Health Physics Society (HPS), Ionizing Radiation Safety Training for Workers, 4 ANSI/HPS N13.36, October 30, 2001.
5 6 ANSI, Practices for Respiratory Protection, ANSI Z88.2-1992.
7 8 American Society for Testing and Materials, Radiological Protection Training for Nuclear 9 Facility Workers, ASTM El 168-95, reapproved in 2008.
10 11 Code of Federal Regulations, Chapter I, Title 10 (10 CFR), Energy, Part 19, Notices, 12 Instructions and Reports to Workers: Inspection and Investigations.
13 14 10 CFR Part 20, Standards for Protection Against Radiation.
15 16 10 CFR Part 21, Reporting of Defects and Non-compliance.
17 18 10 CFR Part 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive 19 Material.
20 21 10 CFR Part 70, Domestic Licensing of Special Nuclear Material.
22 23 10 CFR Part 71, Packaging and Transportation of Radioactive Material.
24 25 10 CFR Part 73, Physical Protection of Plants and Materials.
26 27 Code of Federal Regulations, Title 49 (49 CFR), Transportation, Parts 171-180 28 49 CFR Parts 390-397.
29 30 National Council on Radiation Protection and Measurements (NCRP), Operational Radiation 31 Safety Training, Report No. 134, Bethesda, Maryland, 2000.
32 33 U.S. Nuclear Regulatory Commission (NRC), Branch Technical Position, Guidelines for 34 Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or 35 Termination of Licenses for Byproduct, Source, or Special Nuclear Materials, April 1993 36 (Agencywide Documents Access and Management System Accession No. ML103620647).
37 38 NRC, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other 39 Radioactive Material Licensees, NUREG-1140, January 1988.
40 41 NRC, Consolidated Guidance About Materials Licenses, Program-Specific Guidance About 42 Licenses for Special Nuclear Material of Less than Critical Mass, NUREG-1556, Volume 17.
43 44 NRC, General Statement of Policy and Procedure for NRC Enforcement Actions, 45 NUREG-1600, May 2000.
46 47 NRC, U.S. Specific Schedules of Requirements for Transport of Specified Types of Radioactive 48 Material Consignments, RAMREG-002/NUREG-1660, January 1999.
49 50 NRC, Qualification and Training of Personnel for Nuclear Power Plants, Regulatory Guide 1.8.
51 36
 
1 NRC, General Site Suitability Criteria for Nuclear Power Stations, Regulatory Guide 4.7.
2 3 NRC, Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning, 4 Regulatory Guide 4.21.
5 6 NRC, Administrative Guide for Verifying Compliance with Packaging Requirements for 7 Shipments of Radioactive Material, Regulatory Guide 7.7.
8 9 NRC, Establishing Quality Assurance Programs for Packaging Used in Transport of 10 Radioactive Material, Regulatory Guide 7.10, Washington, DC.
11 12 NRC, Administrative Practices in Radiation Surveys and Monitoring, Regulatory Guide 8.2.
13 NRC, Personnel Monitoring DeviceDirect-Reading Pocket Dosimeters, Regulatory Guide 14 8.4.
15 16 NRC, Instructions for Recording and Reporting Occupational Radiation Exposure Data, 17 Regulatory Guide 8.7.
18 19 NRC, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program, 20 Regulatory Guide 8.9.
21 22 NRC, Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as Is 23 Reasonably Achievable, Regulatory Guide 8.10.
24 25 NRC, Instruction Concerning Prenatal Radiation Exposure, Regulatory Guide 8.13.
26 27 NRC, Acceptable Programs for Respiratory Protection, Regulatory Guide 8.15.
28 29 NRC, Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication, 30 Regulatory Guide 8.24.
31 32 NRC, Air Sampling in the Workplace, Regulatory Guide 8.25.
33 34 NRC, Audible Alarm Dosimeters, Regulatory Guide 8.28.
35 36 NRC, Instruction Concerning Risks from Occupational Radiation Exposure, Regulatory 37 Guide 8.29.
38 39 NRC, Monitoring Criteria and Methods to Calculate Occupational Radiation Doses, Regulatory 40 Guide 8.34.
41 42 NRC, Planned Special Exposure, Regulatory Guide 8.35.
43 44 NRC Enforcement Manual, available online at https://www.nrc.gov/about-45 nrc/regulatory/enforcement/guidance.html#manual.
46 47 NRC, Suggested Guidance Relating to Development and Implementation of Corrective Action, 48 Information Notice 96-28, May 1, 1996.
37
 
1                          3      NUCLEAR CRITICALITY SAFETY 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) 4 70.22(a)(4-8); 10 CFR 70.23(a)(2-4); 10 CFR 70.24, Criticality accident requirements; 5 10 CFR 70.50, Reporting requirements; 10 CFR 70.52, Reports of accidental criticality.
6 7 Regulatory Criteria: The license application should address the following areas of information 8 to expedite the review of the U.S. Nuclear Regulatory Commission (NRC):
9 10
* Evaluate the applicability of Regulatory Guide 3.71, Nuclear Criticality Safety Standards 11          for Nuclear Materials Outside Reactor Cores.
12 13
* Ensure that the application describes the commitments to the criticality accident alarm 14          system (CAAS) requirements in 10 CFR 70.24.
15 16
* Make commitments to ensure appropriate emergency planning and response.
17 18
* Provide information on the methods it will use to demonstrate that all nuclear processes 19          will be subcritical under normal and credible abnormal conditions with an approved 20          margin of subcriticality for safety.
21 22
* Put in place and maintain a nuclear criticality safety (NCS) program and include the 23          commitments associated with the program.
24 25 Discussion: As specified in 10 CFR 70.22(2)(8), applicants must describe, in their license 26 application, procedures to protect health and minimize danger to life or property, including 27 procedures to avoid accidental criticality. These requirements are implemented by 28 establishment of an NCS program as described in this chapter. This chapter discusses specific 29 technical and industry terms. Appendix I to this document contains definitions of the terms.
30 31 3.1      Use of Industry Standards 32 33 Regulatory Guide 3.71 endorses American National Standards Institute (ANSI)/American 34 Nuclear Society (ANS)-8 national standards, with some exceptions and qualifications. The NRC 35 endorsement of these standards means that they provide methods and practices generally 36 acceptable to the NRC staff for the prevention and mitigation of criticality accidents. However, 37 application of a standard is not a substitute for detailed NCS analyses for specific operations.
38 39 If an applicant desires to undertake activities to which an NRC-endorsed standard applies, it 40 should address the subjects covered by the standard and the applicable acceptance criteria 41 relevant to the standard as described below.
42 43 The license application should contain a commitment to follow the requirements (i.e., shall 44 statements) of the standard, subject to any NRC exceptions and qualifications. The application 45 should clearly specify the version of the standard and the specific provisions to which the 46 applicant is committing.
47 48 If there are requirements of a standard to which the applicant does not desire to commit, it 49 should provide sufficient information for the staff to determine that the requirement is not 39
 
1 relevant to the applicants activities, or that the license application contains other commitments 2 that achieve an equivalent safety purpose.
3 4 The applicant may also choose to demonstrate compliance with regulatory requirements by 5 committing to following the recommendations (i.e., should statements) of a standard, although 6 committing to the recommendations is not required for compliance with a standard.
7 8 The applicant should clarify its intent in committing to requirements expressed only as general 9 principles in the standards by making more specific commitments and describing how its license 10 application implements those principles. An applicant should generally use the most current 11 revision of the NRC-endorsed standards in the version of Regulatory Guide 3.71 in effect when 12 it submits the license application. If the applicant commits to a standard or a version of a 13 standard that the NRC has not endorsed, is not the most current endorsed version, or commits 14 to unendorsed portions of an otherwise endorsed standard, the license application should justify 15 the acceptability of these commitments. The use of other than ANS standards 16 (e.g., International Organization for Standardization (ISO) 1709, Principles of Criticality Safety 17 in Storing, Handling and Processing; ISO 7753, Performance and Testing Requirements for 18 Criticality Detection and Alarm Systems) may be acceptable if suitably justified.
19 20 The applicant should consult the most current version of Regulatory Guide 3.71 and consider 21 whether newer versions of these standards or new standards should be endorsed and should 22 make note of any exceptions of qualifications. The NRC staff believes that greater than critical 23 mass license applicants should, at a minimum, commit to complying with the base 24 programmatic elements of ANS 8.1, Nuclear Criticality Safety in Operations with Fissionable 25 Materials Outside Reactors, and ANS 8.19, Administrative Practices for Nuclear Criticality 26 Safety, as indicated in Regulatory Guide 3.71. Other standards that should be considered for 27 applicability include ANS-8.3, 8.5, 8.6, 8.7, 8.10, 8.12, 8.14, 8.15, 8.17, 8.19, 8.20, 8.21, 8.22, 28 8.23, 8.24, and 8.26.
29 30 3.2      Criticality Accident Alarm System 31 32 The applicants commitments to 10 CFR 70.24 are acceptable if the applicant has met the 33 following criteria, or if the application has identified and justified an alternative.
34 35 The applicant should describe a facility CAAS, or the use of portable instruments and 36 administrative procedures that meet the requirements of 10 CFR 70.24.
37 38 The applicant may commit to the current NRC-endorsed version of ANSI/ANS-8.3, Criticality 39 Accident Alarm System, with exceptions, as noted in Regulatory Guide 3.71, or may propose 40 an acceptable alternative (e.g., ISO 7753) with justification.
41 42 An applicant that commits to the specific requirements of ANSI/ANS-8.3 should provide 43 additional details on the CAAS:
44 45
* The applicant should describe a CAAS appropriate to the facility for the type of radiation 46          detected, intervening shielding, and magnitude of the minimum accident of concern.
47 48
* The applicants description of its CAAS should include the type of radiation detector and 49          alarm; the detection threshold and minimum accident of concern; the detector logic used 50          to provide dual alarm coverage, minimize false alarms, and detect failure; and the 40
 
1        methods used to determine radius of coverage, placement of alarms, and actions for 2        maintaining and calibrating the system.
3 4
* The applicant should commit to designing a CAAS to be resistant to damage from 5        anticipated adverse events such as a fire, explosion, corrosive atmosphere, seismic 6        shock equivalent to the site-specific design-basis earthquake or equivalent value 7        specified by the Uniform Building Code, or other adverse conditions that do not result in 8        the evacuation of the entire facility.
9 10
* The applicant should commit to rendering operations safe, by shutdown, evacuation, and 11        quarantine, if necessary, in any area where CAAS coverage has been lost and not 12        restored within a specified number of hours. The number of hours may be determined on 13        a process-by-process basis, because shutting down certain processes, even to make 14        them safe, may carry a larger risk than being without a CAAS for a short time. The 15        applicant should commit to compensatory measures (e.g., limiting access to only 16        personnel with alarming dosimetry compliant with 10 CFR 70.24(a)(1)), halting 17        movement of special nuclear material (SNM) when the CAAS is not functional.
18 19
* The applicant should commit to providing emergency power for the CAAS or provides 20        justification for the use of continuous monitoring with portable instruments.
21 22 Applicants for a greater than critical mass (GTCM) license often request an exemption from the 23 requirement in 10 CFR 70.24 for a CAAS (e.g., areas where the credible upset, quantities, and 24 forms of SNM make criticality incredible). Exemptions have been justified in the past by 25 commitments to limit activities to storage only in a subcritical configuration or to segregate and 26 control materials so that less than a critical mass would be present in any single isolated 27 building or area. Justification may require acceptable controls to assure handling will only use 28 subcritical mass/geometry or that the specific materials are of a form and geometry such that 29 they are highly stable and subcritical in all credible configurations. If handling or use of materials 30 does occur in areas without a fixed facility CAAS, it may require sufficient alarming portable 31 instrumentation that would serve the same function as a CAAS (i.e., meeting 32 10 CFR 70.24(a)(1) or its intent) for those individuals in the potentially affected area. Any 33 instrumentation would have to be discussed in sufficient detail that it is clear the instruments are 34 suitable for that function. Any exemption request should use conservative assumptions in the 35 evaluation and meet the requirements of 10 CFR 70.17, Specific exemptions.
36 37 3.3    Emergency Planning and Response 38 39 The applicant should ensure appropriate emergency planning and response by meeting the 40 following criteria or justifying acceptable alternatives:
41 42
* The applicant commits to the requirements in ANSI/ANS-8.23, Criticality Accident 43        Emergency Planning and Response, as endorsed by Regulatory Guide 3.71.
44 45
* The applicant has an emergency plan or satisfies the alternative requirements in 46        10 CFR 70.22(h)(1)(i).
47 48
* The applicant commits to personnel accident dosimeters, or other means of satisfying 49        the requirements of 10 CFR 70.24(b)(1), in areas that require a CAAS.
50 41
 
1
* These dosimeters, or other radiation instruments, should be readily available to 2        personnel responding to an emergency in accordance with 10 CFR 70.24(a)(3), and 3        there should be a method for prompt onsite dosimeter readout.
4 5 The NRC staff review considers that the requirement for an emergency plan for criticality safety 6 is similar to the need for a CAAS. If an exemption from a CAAS is justified, the staff may find 7 that the criticality safety portion of the emergency plan is likewise satisfied with minimal 8 additional detail.
9 10 3.4    Subcriticality and Double Contingency Principle 11 12 In accordance with the requirements in 10 CFR 70.22(a)(8), an applicant shall provide proposed 13 procedures to protect health and minimize danger to life or property (e.g., procedures to avoid 14 accidental criticality, procedures for personnel monitoring and waste disposal, post-criticality 15 accident emergency procedures). An applicants commitments to procedures that limit the risk 16 of creating an accidental criticality should describe the method or methods that will be used to 17 demonstrate that all nuclear processes will be subcritical under normal and credible abnormal 18 conditions with an approved margin of subcriticality for safety by meeting the criteria described 19 below or justifying acceptable alternatives.
20 21 The applicant should commit to one or more of the following methods for determining subcritical 22 limits on controlled parameters under normal conditions or subcritical values under abnormal 23 conditions:
24 25
* using the subcritical values in a currently endorsed standard (e.g., ANSI/ANS-8.1, 8.5, 26        8.7, 8.12, or 8.15) 27 28
* using the subcritical or critical values, with appropriate margin, from widely accepted 29        industry handbooks (e.g., LA-10860-MS, Critical Dimensions of Systems Containing 30        235 U, 239Pu, and 233U; Atlantic Richfield Hanford (ARH)-600, Criticality Handbook, 31        issued June 1968), experimental data, or peer-reviewed publications 32 33
* using industry-accepted hand calculation methods (e.g., areal density, solid angle 34        technique), subject to the limitations of those methods 35 36
* using deterministic or probabilistic (e.g., discrete ordinates, Monte Carlo) computer 37          codes to calculate k eff and validating these calculational methods in accordance with the 38        requirements of ANSI/ANS 8.24, Validation of Neutron Transport Methods for Nuclear 39        Criticality Safety Calculations 40 41 For each method used to demonstrate subcriticality, the applicant should commit to using the 42 method consistent with any limitations, with an appropriate margin of subcriticality and within its 43 area of applicability. The license application should describe the margin of subcriticality and 44 area of applicability.
45 46 The applicant should commit to determining safety limits based on one of the methods 47 discussed above. The applicant should commit to evaluating controlled parameters at their 48 safety limits, or more conservatively, and to evaluating parameters that are not controlled at 49 their most reactive credible values.
50 42
 
1 The applicant should describe a program that ensures compliance with the double contingency 2 principle, where practicable. This principle, as stated in ANSI/ANS-8.1 (and in 10 CFR 70.4, 3 Definitions), is as follows:
4 5        Process designs should incorporate sufficient factors of safety to require at least 6        two unlikely, independent, and concurrent changes in process conditions before 7        a criticality accident is possible.
8 9 Note: For GTCM licensees covered by this guidance, the term process refers to a discrete 10 system for which a criticality safety analysis is performed; the term process conditions means 11 changes in the physical characteristics of such a system capable of resulting in a more reactive 12 system configuration.
13 14 Each process that has credible normal or abnormal conditions leading to criticality should have 15 sufficient engineered and administrative controls in place to ensure double contingency 16 protection where applicable.
17 18 Additional guidance pertaining to compliance with the double contingency principle appears in 19 appendix 5-A to NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License 20 Applications. Note that the requirements of Subpart H, Additional Requirements for Certain 21 Licensees Authorized To Possess a Critical Mass of Special Nuclear Material, of 22 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, do not apply to applicants 23 covered by this guidance. However, the expectation is to demonstrate subcriticality under 24 normal and credible abnormal conditions with an approved margin of subcriticality for safety, a 25 fundamental principle of NCS applicable to all 10 CFR Part 70 applicants, along with the double 26 contingency principle, as discussed in ANSI/ANS-8.1. GTCM applicants subject to this guidance 27 should comply with those portions of appendix 5-A pertaining to the subcriticality requirement 28 (also stated in 10 CFR 70.61(d)) and the double contingency principle.
29 30 3.5    Organization and Administration of the Nuclear Criticality Safety Program 31 32 The applicants management of the NCS program should meet the following criteria or identify 33 and justify an alternative approach:
34 35
* Describing and committing to carrying out and maintaining an NCS program to meet the 36        regulatory requirements in 10 CFR Part 70 and to ensuring adequate protection against 37        the occurrence of accidental criticality. The primary means of doing this should be 38        prevention (i.e., ensuring that processes will be subcritical under normal and credible 39        abnormal conditions).
40 41
* Describing NCS program objectives. The applicants NCS programs may be graded 42        commensurate with the complexity and scope of planned activities but should address 43        all of the program elements listed below in some fashion. For example, applicants that 44        are designated storage only may not need handling procedures, postings, or frequent 45        operational audits. The description should not merely list the program objectives but 46        should describe how the applicant will meet those objectives.
47 48
* Performing and documenting criticality safety evaluations (CSEs) for new or changed 49        processes, establishing safety limits and controls as necessary to ensure that processes 50        will remain subcritical under normal and credible abnormal conditions.
43
 
1
* Establishing, as practicable, double contingency protection and defense-in-depth 2        measures and ensuring sufficient margins of safety and subcriticality to provide 3        additional assurance that the likelihood of criticality will be acceptably low.
4 5
* Establishing and maintaining a CAAS system and emergency response procedures to 6        protect health and safety in the event criticality occurs.
7 8
* Providing technical support to emergency response personnel in responding to and 9        recovering from abnormal conditions and emergencies, up to and including a criticality 10        accident.
11 12
* Periodically verifying the adequacy of criticality controls, including observation of 13        operations and verification of equipment configuration.
14 15
* Ensuring the adequacy of CSEs through peer reviews, self-assessments, and validation 16        and verification of calculational methods.
17 18
* Training and otherwise supporting operations in procedures to ensure the safe handling 19        of SNM.
20 21
* Supporting regulatory compliance regarding event reporting (10 CFR 70.50 and 22        applicable portions of 10 CFR Part 70, Appendix A).
23 24 The applicant should outline an NCS program structure that is consistent, to the extent practical, 25 with current industry practices (e.g., ANSI/ANS-8.1 and ANSI/ANS-8.19), including establishing 26 the roles and responsibilities of key program personnel (e.g., NCS manager, NCS senior 27 engineers, NCS engineers). The program structure should include the following:
28 29
* The applicant should describe the training and qualification of key NCS program 30        personnel.
31 32
* The applicant should commit to establishing and maintaining NCS safety limits and 33        operating limits and agree to maintain management measures to ensure their continued 34        reliability and availability.
35 36
* The applicant should commit to supporting operations personnel through development of 37        training, preparation of NCS postings, and other appropriate operator aids for key 38        administrative controls and review procedures and operations to ensure instructions and 39        limits or controls are unambiguous, easily understood, and readily achievable.
40 41
* The applicant should commit to developing postings and other markings, as well as 42        criticality alarm signals that are distinctive.
43 44
* The applicant should evaluate modifications to the facility or safety program to ascertain 45        their effect on criticality safety.
46 47
* The applicant should describe an organizational structure in which the NCS organization 48        is independent of operations, to the extent practical.
49 44
 
1
* Personnel should work in accordance with written, approved procedures when the 2        activity could affect NCS.
3 4
* The applicant should establish management policies that reinforce operators stop-work 5        authority and encourage reporting defective conditions.
6 7 Applicants should establish and describe an adequate organization and staffing, in accordance 8 with the requirements in 10 CFR 70.22(a)(6) and 10 CFR 70.23(a)(2), with sufficient staff, 9 resources, and clear responsibilities to allow independent technical reviews of CSEs, periodic 10 assessments of those operations and controls important to criticality safety, and the issuance of 11 stop-work directives separate from operations should an unsafe condition be encountered. In 12 evaluating applicant programs, the NRC staff generally requires that, at a minimum, two 13 individuals who are knowledgeable and qualified in NCS be on an applicants staff with 14 commensurate responsibilities.
15 16 3.6    Nuclear Criticality Safety Program Management Measures 17 18 As required by 10 CFR 70.22(a)(6), an applicant must provide information on the technical 19 qualifications, including training and experience of the applicant and members of the staff, to 20 engage in the proposed activities. As such, the application should describe the following 21 management measures associated with the NCS program:
22 23
* the requirements of ANSI/ANS-8.19 and ANSI/ANS-8.20, Nuclear Criticality Safety 24        Training, as they pertain to training, procedures, audits, and assessments 25 26
* regarding training 27 28        -  a commitment to training personnel in the areas discussed in section 7 of 29            ANSI/ANS-8.20 30 31        -  a commitment to training personnel regarding procedural compliance, stop-32            work authority, response to alarms, and reporting of defective conditions 33 34
* regarding periodic assessments, a commitment to conducting and documenting 35        walkthroughs (i.e., the observation of operations to verify compliance with criticality limits 36        and implementation of criticality controls) for all areas where SNM is used, stored, or 37        handled, such that all areas will be reviewed at some specified frequency 38 39
* a graded approach used to justify the schedule for doing walkthroughs, conducting 40        audits, or taking corrective action 41 42 3.7    Technical Practices for Nuclear Criticality Safety 43 44 The application should describe the following technical practices used to determine and 45 implement NCS safety limits:
46 47
* conducting CSEs using industry-accepted and peer-reviewed methods (e.g., in 48        ANSI/ANS-8.1, 8.19, 8.24) 49 50 45
 
1
* validating calculational methods used to develop NCS safety limits, in accordance with 2            ANSI/ANS-8.24 3
4
* using the code within its validated area(s) of applicability, or, if there are insufficient 5            benchmark experiments over the needed range of variables, ensuring that the area of 6            applicability is extended by making use of trends in the bias, taking the uncertainty due 7            to extrapolation appropriately into account 8
9
* demonstrating an adequate margin of subcriticality for safety, by ensuring that the 10            margin is large compared to the uncertainty in the calculated value of k eff 11 12 The minimum margin of subcriticality is an allowance for unknown or unquantified uncertainties 13 that have not been accounted for in the validation and is a measure of the degree of confidence 14 that systems calculated to be subcritical are actually subcritical. The minimum margin of 15 subcriticality may be used to define a maximum allowable value of k eff that is considered 16 subcritical, referred to as the upper subcritical limit, as follows:
17 18                                    k calc  1 -  -  - k AOA - k m 19 20 where  = the calculational bias,  = uncertainty in the bias, k AOA = margin due to extending 21 the area of applicability beyond the experimental data, and k m = minimum margin of 22 subcriticality.
23 24 The applicants overall margin of subcriticality, which includes the minimum margin of 25 subcriticality and other factors that provide conservatism in the calculation of k eff , should be 26 sufficient to provide reasonable assurance that processes evaluated to be subcritical are 27 actually subcritical. NUREG-1520, appendix 5-B, includes additional guidance on the margin of 28 subcriticality.
29 30 3.8        Calculational Method Validation 31 32 The applicant should include a summary description of a documented, reviewed, and approved 33 (by the applicants NCS function and management) validation report for each method that will be 34 used to conduct an NCS analysis. For methods such as experimental data, handbooks, industry 35 standards, and hand calculations, the validation may consist of a demonstration of the methods 36 applicability to the applicants processes, including specification of any limitations or 37 assumptions needed to ensure their validity. For methods that rely on the explicit calculation of 38 k eff , the validation should evaluate critical benchmark experiments similar in geometry, material 39 composition, and neutron energy spectrum to the systems to be evaluated.
40 41 For computer calculation methods, the applicants criticality code validation should provide 42 reasonable assurance that processes evaluated to be subcritical are actually subcritical. This 43 should include the applicants selection of benchmark experiments, statistical methodology, and 44 results (determination of the area(s) of applicability and upper subcritical limit(s)). The 45 applicants validation may be graded commensurate with the amount of subcritical margin, given 46 the forms and quantities of material present. Where normal or abnormal conditions are deeply 47 subcritical, a correspondingly less rigorous validation may be employed. (Additional guidance 48 appears in appendix 5-B to NUREG-1520.) Where there is little benchmark data available to 49 validate an applicants calculations, additional margin may be employed. However, even with a 50 large margin to criticality, comparison against experimental data is essential. If no experimental 46
 
1 data covering its processes are available, the applicant must use one of the methods other than 2 computer calculation listed in section 3.4.
3 4 The applicants summary description of its validation in the license application should include 5 the following:
6 7
* a description of the methodology, including the method used to select benchmark 8          experiments, determining the bias and bias uncertainty, and calculating an upper 9          subcritical limit (including extending the area(s) of applicability, if needed) 10 11
* a general description of the physical systems and area(s) of applicability covered by the 12          validation 13 14
* a brief description of the benchmark experiment sets used 15 16
* the minimum margin of subcriticality and its technical justification 17 18
* the calculational system (hardware and software, including nuclear data sets) used 19 20
* any limitations on the use of the method 21 22 The applicants commitments to performing validation should include the following:
23 24
* a commitment to perform validation consistent with ANSI/ANS-8.24, with any exceptions 25          and qualifications, as stated in Regulatory Guide 3.71, or other widely accepted industry 26          practices, with justification 27 28
* a verification process, including verification upon installation and at specified periods; a 29          commitment to evaluate methods for the need for reverification upon changes to the 30          calculational system 31 32
* a commitment to assess and document the applicability of the validation to each new 33          calculation 34 35 Additional information regarding validation may be found in the current version of NUREG-1520.
36 37 3.9      Criticality Safety Evaluations 38 39 The application should discuss the commitments regarding doing CSEs:
40 41
* The application should describe the written procedures to perform CSEs that incorporate 42          the following principles:
43 44            The application will establish NCS safety limits based on analyses assuming 45              optimum or the most reactive credible values of parameters (e.g., most reactive 46              conditions physically possible, bounding values limited by regulatory requirements) 47              unless specified controls limit parameters to a particular range of values. If less than 48              the optimum values are used, and corresponding controls are not identified, the CSE 49              will justify the basis.
47
 
1            The applicant may establish NCS operating limits to ensure that safety limits are 2              unlikely to be exceeded. The application should consider process variability and 3              uncertainty in determining operating limits. Additional conservatism may be applied.
4 5            The application will describe specific controls and management measures necessary 6              to enforce NCS safety limits and will specify operating limits.
7 8
* The applicant should commit to providing the technical basis demonstrating subcriticality 9          under normal and credible abnormal conditions and compliance with the double 10          contingency principle in the CSEs.
11 12 3.10    Evaluation and Implementation of Controlled Parameters 13 14 Parameters available for NCS control are mass, geometry, density, enrichment (or isotopics),
15 reflection, moderation, concentration, interaction (or spacing), neutron absorption (or poison),
16 volume, heterogeneity, physicochemical form, and process variables. The number and names 17 of specific parameters will vary from one applicant to another.
18 19 The applicants commitments to technical practices associated with evaluating and carrying out 20 controlled parameters should either meet the following acceptance criteria or identify and justify 21 acceptable alternatives:
22 23
* The applicant should state that the use of a single NCS control to maintain the values of 24          two or more controlled parameters constitutes only one component necessary to meet 25          the double contingency principle.
26 27
* The applicant should commit to the following preferred hierarchy of controls:
28 29            the preferred use of passive engineered controls; in particular, passive engineered 30              geometry control 31 32            the following order of preference for NCS control: (1) passive engineered, (2) active 33              engineered, (3) enhanced administrative, and (4) simple administrative controls 34 35            a preference for designating explicit NCS controls over reliance on the natural and 36              credible course of events 37 38          The applicant should commit to a preference for control of two or more parameters over 39          multiple controls on a single parameter. If relying on two or more controls on a single 40          parameter, the applicant commits to a preference for diverse over redundant means of 41          control. These commitments do not mean an applicant will follow the preferred hierarchy 42          of controls in every case. However, an applicants sets of controls should follow this 43          preference in the majority of cases. In general, for example, where passive controls are 44          readily available, they should be used rather than administrative controls. An applicant 45          should demonstrate how it is meeting these commitments, such as by providing 46          justification when deviating from these criteria.
47 48
* The use of each controlled parameter should meet the following general criteria, in 49          addition to the specific criteria for the following individual parameters listed below:
50 48
 
1      When a single-parameter limit is used (e.g., minimum critical mass, favorable 2        geometry limit, always-safe concentration), all other parameters are evaluated at 3        their optimum or most reactive credible values. In determining single-parameter 4        limits, it is permissible to specify a particular physicochemical form and isotopic 5        composition.
6 7      Examples are (1) minimum critical mass, based on spherical geometry, optimum 8        moderation, and full water reflection and (2) favorable geometry, based on having 9        equipment filled with optimally moderated material and full water reflection. This is for 10        a specific form of material (e.g., low-enriched uranium dioxide fuel).
11 12      When measurement of a parameter is needed, instrumentation is calibrated.
13 14      When criticality control is based on measuring a single parameter, independent 15        means of measurement (e.g., redundant in-line monitoring, dual independent 16        sampling) are used.
17 18      Safety limits on controlled parameters are established, taking any tolerances and 19        uncertainty into account.
20 21      Controlled dimensions and material properties are verified upon installation, following 22        system changes, and at periodic intervals.
23 24
* Criteria for the use of a controlled parameter should include the following (additional 25  guidance may be found in the current version of NUREG-1520. If the applicant is not 26  using a particular parameter, it should state it in the application; in such cases, the 27  following commitments are not needed):
28 29  o    Criteria for the use of mass control should include:
30 31            Compliance with mass limits is verified by either assuming the entire weight of 32            material is SNM or by measurement to verify the actual weight percent of SNM.
33 34            Over batching beyond double batching is considered unless it requires multiple 35            independent failures or is precluded by, for example, equipment capacity or 36            quantity of material available.
37 38  o    Criteria for the use of geometry control should include the following:
39 40            All credible means of losing geometry control (e.g., corrosion, leakage, bulging, 41            transfer to unfavorable geometry, changes to a more reactive material form) are 42            evaluated for their impact on reactivity.
43 44            Neutron interaction with other SNM is considered as part of the demonstration of 45            subcriticality unless meeting criteria for being neutronically isolated.
46 47  o    Criteria for the use of density control should include the following:
48 49            Controls are established to limit materials to less than theoretical density, and 50            density is confirmed by experimentation or measurement.
51 49
 
1 o Criteria for the use of enrichment or isotopics control (including abundance of 2  individual fissionable isotopes as well as ratios of fissionable elements (e.g., uranium 3  (U)-235 weight percent, plutonium (Pu)-239 weight percent, ratio of Pu/U) should 4  include the following:
5 6      The facility-wide limit on enrichment or isotopics should be used in calculations, 7      or reliable methods of labeling and segregating such materials should be used.
8 9 o Criteria for the use of reflection control should include the following:
10 11      Modeled reflection conditions account for wall thickness, structural materials, and 12      any transient or permanent adjacent reflecting materials. Criteria are established 13      for determining when such materials may be neglected.
14 15      When reflection is not controlled, reflection equivalent to 30 centimeters 16      (12 inches) of tight-fitting water, or 60 centimeters (24 inches) of tight-fitting 17      concrete are assumed.
18 19      When reflection is controlled, a minimum equivalent to 2.54 centimeters (1 inch) 20      tight-fitting water reflector is assumed to account for transient incidental 21      reflectors.
22 23      In the presence of special moderators and reflectors (e.g., deuterium, beryllium, 24      graphite, or high-density hydrocarbons), the adequacy of modeled reflections is 25      justified.
26 27      When reflection is controlled, the means of limiting personnel intrusion around 28      the individual units is established, preferably by means of rigid barriers.
29 30      In evaluating arrays of units, the above reflection conditions are applied to the 31      exterior of the array, while interstitial moderation is evaluated as below.
32 33 o Criteria for the use of moderation control should include the following:
34 35      A commitment is made to follow ANSI/ANS-8.22, Nuclear Criticality Safety 36      Based on Limiting and Controlling Moderators, with any exceptions and 37      qualifications as described in Regulatory Guide 3.71.
38 39      The preference is for physical barriers to preclude moderator intrusion into SNM 40      (e.g., double roof, double-sleeved piping, exclusion of sprinklers, raised or sloped 41      floors).
42 43      Moderator controlled areas are conspicuously marked and controls established 44      to preclude the introduction of moderating materials.
45 46 47      Firefighting procedures in moderation-controlled areas are evaluated for their 48      impact on reactivity. Any restrictions are included in procedures and training.
49 50 51 50
 
1 o Criteria for the use of concentration control should include the following:
2 3      Controls are established to preclude the inadvertent introduction of precipitating 4      agents (e.g., locking tanks). Credible means of concentrating or precipitating 5      SNM are evaluated.
6 7      Transfers from favorable to unfavorable geometry are precluded consistent with 8      the double contingency principle; this includes use of dual independent sampling 9      or in-line monitoring, or both.
10 11      Homogeneity of solutions is justified, with controls established as needed.
12 13 o Criteria for the use of interaction control should include the following:
14 15      The preference is for the use of physical barriers (e.g., fixed spacers, cages). If 16      such engineered controls are not feasible, administrative controls with visual 17      aids, such as painted lines and postings, may be used. In all such cases, multiple 18      spacing upsets are required before criticality is possible.
19 20      The structural integrity of, for example, spacers and storage racks, is sufficient to 21      withstand credible abnormal conditions, including seismic events.
22 23      Moveable engineered barriers and containers (e.g., birdcage drums, 55-gallon 24      drums) are inspected periodically for deformation.
25 26 o Criteria for the use of neutron absorber control should include the following:
27 28      The preference is for the use of fixed absorbers meeting ANSI/ANS-8.21, 29      regardless of whether they are existing structural materials or added specifically 30      for criticality control, with any exceptions and qualifications as described in 31      Regulatory Guide 3.71.
32 33      When using borosilicate glass Raschig rings, the applicant meets ANSI/ANS-8.5, 34      Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of 35      Fissile Material, with any exceptions and qualifications as described in 36      Regulatory Guide 3.71.
37 38      When using soluble neutron absorbers, the applicant meets ANSI/ANS-8.14, 39      Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors, with 40      any exceptions and qualifications as described in Regulatory Guide 3.71.
41 42      The effective of neutron spectra on absorber effectiveness is considered.
43 44 o Criteria for the use of volume control should include the following:
45 46      The preference is for the use of fixed geometry to limit accumulations of SNM to 47      less than a subcritical volume. Irrespective of the shape of such containers, 48      volume control should be based on the most reactive geometry (normally a 49      sphere).
50 51 51
 
1        o    Criteria for the use of heterogeneity control (normally only significant for low-2              enriched uranium) should include the following:
3 4                  Methods of causing fissile material to become inhomogeneous are evaluated and 5                  controls established as needed. If calculations for heterogeneous materials are 6                  performed, they should be validated using heterogeneous benchmarks.
7 8                  Assumptions that can affect the physical scale of heterogeneity are justified, and 9                  process variables that can affect the scale are controlled.
10 11        o    Criteria for the use of physicochemical form control should include the following:
12 13                  The most reactive credible material form is used in calculations, or controls are 14                  established to limit the material to a less reactive form.
15 16                  Abnormal conditions evaluated include both in-situ changes and migration from 17                  one area to another.
18 19        o    Criteria for the use of process variable control (e.g., temperature, pressure, acid 20              molarity, electrical load, conductivity, radiation) should include the following:
21 22                  Controls are established to limit process variables within their allowed ranges, 23                  and the correlation of the variable to its associated indirectly controlled parameter 24                  is established by experiment or measurement.
25 26 3.11    Additional Nuclear Criticality Safety Program Commitments 27 28 The applicants description of additional commitments in its NCS program should include the 29 following:
30 31
* The applicant should commit to periodically assessing the adequacy of engineered and 32        administrative criticality controls, to promptly detecting any NCS deficiencies, and to 33        taking prompt and effective corrective action to prevent recurrence.
34 35
* The applicant should commit to suspending operations or otherwise rendering processes 36        safe upon loss of double contingency protection, until such protection can be restored, 37        and to assessing the adequacy of the affected controls.
38 39
* The applicant should commit to retaining records of NCS deficiencies and documenting 40        any corrective actions taken.
41 42
* The applicant should commit to identifying all equipment and procedures needed for 43        criticality controls to serve their safety function (i.e., ensure their effectiveness to 44        maintain controlled parameters within subcritical limits) and to maintaining such 45        equipment and procedures as part of its facility management measures, including audits 46        and inspections.
47 48
* The applicant should commit to evaluating facility changes for their impact on NCS 49        before implementing them and describe its change review and approval process.
50 52
 
1
* The applicant should include the criteria applicable to the description of its measures to 2          carry out the event reporting requirements in 10 CFR 70.50 and 10 CFR 70.52.
3 4 3.12    Emergency Plan 5
6 Prevention is the primary means of protection against the consequences of accidental criticality, 7 but there is still a risk that it will occur. For facilities that require a CAAS in accordance with 8 10 CFR 70.24, an applicant must either submit an emergency plan or an evaluation 9 demonstrating that an emergency plan is not required, in accordance with the provisions in 10 10 CFR 70.22(i)(1)(i) and (ii).
11 12 A facility CAAS is required if the applicant is authorized to possess a quantity exceeding the 13 mass-based criteria in 10 CFR 70.24(a). If the applicant is required to have a CAAS and 14 submits a request for exemption from the CAAS requirements of 10 CFR 70.24, the exemption 15 should be justified. As stated in ANSI/ANS-8.3, installation of an alarm system implies a 16 nontrivial risk of criticality. Whenever it is determined that criticality in a given area is credible, 17 and that it has the potential to adversely affect workers or the public, a criticality alarm should be 18 required, unless an exemption is granted pursuant to 10 CFR 70.17, Specific exemptions.
19 False alarms and evacuations could pose a risk of injury to workers, and maintenance may 20 expose workers who would not otherwise be exposed to occupational or potential criticality 21 doses, so consideration may be given to whether installation of a CAAS results in a net risk 22 benefit.
23 24 If a facility CAAS is required, an applicant must establish emergency procedures in response to 25 activation of the alarm, in accordance with 10 CFR 70.24(a). In addition, an applicant must 26 submit either an emergency plan or an evaluation demonstrating that an emergency plan is not 27 required, in accordance with 10 CFR 70.22(i)(1)(i) and (ii). Criteria pertaining to an applicants 28 emergency procedures and emergency plan, if required, should follow current standards, such 29 as ANSI/ANS-8.3 and 8.23.
30 31 3.13    References 32 33 American Nuclear Society (ANS) 8.1, Nuclear Criticality Safety in Operations with Fissionable 34 Materials Outside Reactors.
35 36 ANS 8.19, Administrative Practices for Nuclear Criticality Safety.
37 38 American National Standards Institute (ANSI)/ANS-8.3, Criticality Accident Alarm System.
39 40 ANSI/ANS-8.5, Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of 41 Fissile Material.
42 43 ANSI/ANS-8.14, Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors.
44 45 ANSI/ANS-8.20, Nuclear Criticality Safety Training.
46 47 ANSI/ANS-8.22, Nuclear Criticality Safety Based on Limiting and Controlling Moderators.
48 49 ANSI/ANS-8.23, Criticality Accident Emergency Planning and Response.
50 53
 
1 ANSI/ANS 8.24, Validation of Neutron Transport Methods for Nuclear Criticality Safety 2 Calculations.
3 4 Atlantic Richfield Hanford (ARH)-600, Criticality Handbook, June 1968.
5 Code of Federal Regulations, Chapter I, Title 10, Energy, Part 70, Domestic Licensing of 6 Special Nuclear Material.
7 8 International Organization for Standardization (ISO) 1709, Principles of Criticality Safety in 9 Storing, Handling and Processing; ISO 7753, Performance and Testing Requirements for 10 Criticality Detection and Alarm Systems.
11 12 Paxton, H.C., and N.L. Pruvost, Critical Dimensions of Systems Containing 235U, 239Pu, and 13 233 U, LA-10860-MS, Los Alamos National Laboratory, 1987.
14 15 U.S. Nuclear Regulatory Commission (NRC), Nuclear Criticality Safety Standards for Nuclear 16 Materials Outside Reactor Cores, Regulatory Guide 3.71.
17 18 NRC, Standard Review Plan for Fuel Cycle Facilities License Applications, NUREG-1520.
54
 
1                                  4      CHEMICAL SAFETY 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) 4 70.22(a)(6)-(8), 10 CFR 70.23(a)(2)-(4).
5 6 Regulatory Criteria: The regulations in 10 CFR 70.22(a)(7) require each application for a 7 license to contain proposed procedures to protect health and minimize danger to life or property 8 (e.g., procedures to avoid accidental criticality, procedures for personnel monitoring and waste 9 disposal, post-criticality accident emergency procedures). Licensees must give reasonable 10 assurance that their facility provides adequate protection against chemical hazards that could 11 affect licensed materials and thus present an increased radiological or chemical risk.
12 Licensees must describe equipment, facilities, and procedures in sufficient detail to reasonably 13 assure the U.S. Nuclear Regulatory Commission (NRC) that it will provide adequate fire safety.
14 To allow the NRC staff to make determinations that the requirements of 10 CFR 70.23(a)(2) 15 through (4), as they relate to chemical safety, are met, the applicant must submit the following, 16 in adequate detail:
17 18
* Describe the proposed activities as required by 10 CFR 70.22(a)(2), including the 19          identification of any potential chemical hazards associated with the proposed activities 20          that are within the NRCs regulatory authority.
21 22
* Describe the technical qualifications of the applicant and staff as required by 23          10 CFR 70.22(a)(6). If the proposed operations involve chemical hazards within the 24          NRCs regulatory authority, this description must address the licensees ability to identify 25          and manage such hazards.
26 27
* Describe the equipment and facilities that the applicant or licensee will use to protect 28          health and minimize danger to life or property, as required by 10 CFR 70.22(a)(7). If the 29          proposed operations involve chemical hazards within the NRCs regulatory authority, this 30          description must address the licensees equipment and facilities that would be used to 31          control such hazards.
32 33
* Describe the proposed procedures that would be used to protect health and minimize 34          danger to life or property as required by 10 CFR 70.22(a)(8). If the proposed operations 35          involve chemical hazards within the NRCs regulatory authority, this description must 36          address the licensees proposed procedures for managing such hazards.
37 38 Discussion: Consistent with its statutory authority under the Atomic Energy Act of 1954, as 39 amended (Act), and other legislation, the NRC is responsible for licensing and regulating the 40 nations civilian use of byproduct, source, and special nuclear material (SNM) to assure the 41 adequate protection of public health and safety, promote the common defense and security, and 42 protect the environment.
43 44 This broad statutory authority authorizes the NRC to protect workers and the general public 45 from radiation hazards produced by radioactive materials, chemical hazards produced by 46 radioactive materials, and facility conditions that affect the safety of radioactive materials and 47 thus present an increased radiation risk to workers and the general public. The NRC has a 48 memorandum of understanding with the U.S. Occupational Safety and Health Administration 49 that delineates each agencys regulatory responsibility over chemical hazards (Agencywide 50 Documents Access and Management System (ADAMS) Accession No. ML11354A432).
55
 
1 In accordance with the requirements in 10 CFR 70.4, Definitions, hazardous chemicals 2 produced from licensed materials means substances having licensed material as precursor 3 compound(s) or substances that physically or chemically interact with licensed materials. The 4 NRC staff evaluates chemical safety issues for greater than critical mass facilities pursuant to 5 the requirements in 10 CFR 70.22(a)(8). Applicants (or licensees) for a license to possess and 6 use SNM in critical mass quantities should provide information that allows the NRC staff to 7 determine if the proposed activities involve chemical hazards that are directly associated with 8 NRC-licensed radioactive material. Generally, applicants for a license to possess and use 9 critical mass quantities of SNM do not store or possess large quantities of hazardous chemicals.
10 If large quantities are on site, the applicant needs to demonstrate that there are no chemical 11 hazards associated with the proposed operations or include an analysis that demonstrates that 12 no chemical hazards under the NRCs regulatory authority pose a threat to the safe handling of 13 licensed material. If the applicant possesses hazardous chemicals or intends to handle or mix 14 licensed material with a chemical solution, the application must demonstrate that the chemical 15 safety program is adequately defined and ensures that chemical hazards that are within the 16 NRCs regulatory charter will not threaten the licensees ability to safely handle the licensed 17 material.
18 19 In accordance with the requirements in 10 CFR 70.22(a)(7) and (8), and 10 CFR 70.23(a)(3) 20 and (4), the application must describe the elements of the safety program intended to protect 21 the health and safety of the workers and the public from any chemical hazards. For this reason, 22 the application must include chemical safety-related information including, if appropriate, a 23 description of the chemical safety program. Specific information to be supplied includes the 24 following:
25 26
* The description of the proposed activities must include information on any chemicals that 27          would be directly associated with the proposed activities, would be near the proposed 28          activities, or might be produced by the proposed activities. Any restrictions on the 29          chemicals in or near the proposed activities that are used to limit chemical safety 30          hazards must be identified.
31 32
* If the proposed activities include hazardous chemicals or the potential to generate 33          hazardous chemicals, the application should present an analysis that demonstrates that 34          accidents involving or producing hazardous or toxic chemicals would not jeopardize the 35          licensees ability to safely manage the licensed material.
36 37
* If there is no analysis or commitments to demonstrate that hazardous or toxic chemicals 38          will not affect the licensees management of the licensed material, then the description of 39          the technical qualification of the applicant and members of the staff must include the 40          qualifications of personnel who would be responsible for the analysis of chemical 41          hazards of proposed activities, would review and approve such chemical hazards 42          analysis, and would conduct the proposed activities.
43 44
* If there is no analysis or commitments to demonstrate that hazardous or toxic chemicals 45          will not affect the licensees management of the licensed material, the description of the 46          equipment and facilities that will be used to protect health and minimize danger to life or 47          property must include the equipment and facility features that would be used to manage 48          chemical hazards or mitigate the consequences of chemical hazards associated with the 49          proposed activities.
50 56
 
1
* If no analysis or commitments demonstrate that hazardous or toxic chemicals will not 2        affect the licensees management of the licensed material, the description should include 3        any procedures used to identify and analyze chemical hazards that are within the NRCs 4        regulatory authority, as well as any procedures used to manage these hazards. The 5        description of proposed procedures should include the following:
6 7            Describe the method (or methods) used to evaluate chemical hazards associated 8            with the proposed activities and within the NRCs regulatory jurisdiction. The specific 9            method must be appropriate for the nature of the chemical hazard (i.e., chemical 10            toxicity, chemical reactivity). If the proposed activities are defined, the results of the 11            chemical hazards evaluation must be provided.
12 13            Describe the procedures used for an independent review and acceptance of 14            chemical hazard evaluations.
15 16            Describe the procedures used to control any chemical hazards associated with the 17            proposed activities identified in the chemical hazards analysis.
18 19 The license application should identify any technical, industry, or professional standards used in 20 the hazards analyses. Appropriate standards include Prudent Practices in the Laboratory:
21 Handling and Management of Chemical Hazards, Updated Version, by the National Research 22 Council; Laboratory Safety Guidance by the Occupational Safety and Health Administration; 23 and Safety in Academic Chemistry Laboratories by the American Chemical Society. These 24 standards do not address all aspects of a chemical safety program, particularly if the proposed 25 operations involve the potential for energetic reactions that could release and disperse licensed 26 material. When selecting and using standards, consideration should be given to the findings and 27 recommendations in reports such as the U.S. Chemical Safety and Hazard Investigation Board 28 Case Study, Texas Tech University Laboratory Explosion, issued 2011.
29 30 4.1    References 31 32 Occupational Safety and Health Administration, Laboratory Safety Guidance, 33 (OSHA) 3404-11R, U.S. Department of Labor, Washington, DC, 2011.
34 35 Memorandum of Understanding between the U.S. Nuclear Regulatory Commission and the 36 Occupational Safety and Health Administration, September 6, 2013 (ML11354A432).
37 38 National Research Council of the National Academies, Prudent Practices in the Laboratory:
39 Handling and Management of Chemical Hazards, Updated Version, The National Academies 40 Press, 2011.
41 42 American Chemical Society, Safety in Academic Chemistry Laboratories, Volume 1, Accident 43 Prevention for College and University Students, 7th Edition, American Chemical Society Joint 44 Board-Council Committee on Chemical Safety, 2003.
45 46 U.S. Chemical Safety and Hazard Investigation Board, Texas Tech University Laboratory 47 Explosion, Case Study by the U.S. Chemical Safety and Hazard Investigation Board, 48 No 2010-05-I-TX, 2011.
49 57
 
1                                            5        FIRE SAFETY 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) 4 70.22(a)(6)-(8) and 10 CFR 70.23(a)(3)-(4).
5 6 Regulatory Criteria: Applicants must provide reasonable assurance that they will protect health 7 and minimize danger to life or property, which includes providing adequate protection against 8 fires and explosions that could affect licensed materials and thus present an increased 9 radiological or chemical risk. In accordance with the requirements of 10 CFR 70.22(a)(7)-(8),
10 the application must describe equipment, facilities, and procedures in sufficient detail to 11 reasonably assure the U.S. Nuclear Regulatory Commission (NRC) that it will provide adequate 12 protection of health and minimize danger to life or property, which includes fire safety.
13 14 Discussion: The application should address all major aspects of a fire safety program, 15 including facility design, fire protection systems, employee training, emergency response, 16 process fire safety, combustible loading, and potential fire scenarios.
17 18 5.1        Facility Design 19 20 In accordance with the requirements of 10 CFR 70.22(a)(7), an adequate application describes 21 equipment and facilities that the applicant will use to protect health and minimize danger to life 22 or property. Although 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, does 23 not require a separate fire safety program, an application should document commitments 24 pertaining to the fire safety of the equipment and facilities that the applicant will use to protect 25 health and minimize danger to life or property, including a description of building construction, 26 fire area determination, fire-rated walls or opening protection, electrical installation, lightning 27 protection, emergency lighting, life safety and egress, ventilation, and fire water drainageall 28 facility design characteristics that affect fire safety.
29 30 Compliance with applicable National Fire Protection Association (NFPA) standards 1 or the 31 building code may be sufficient to address these areas in the application. The NRC staff 32 recognizes NFPA 801, Standard for Fire Protection for Facilities Handling Radioactive 33 Materials, as one standard that specifies acceptable facility fire safety design criteria; however, 34 the application may document other nationally recognized codes and standards, if appropriate 35 (e.g., NFPA 45, Standard on Fire Protection for Laboratories Using Chemicals). The 36 application should list any relevant code(s) and describe the facilitys degree of compliance with 37 that code.
38 39 The application should document information on the facilitys construction materials and any 40 building codes used during its construction. The NRC staff recognizes NFPA 220, Standard on 41 Types of Building Construction, as another standard that specifies acceptable facility 42 construction material design criteria. Some building materials (e.g., reinforced concrete and 43 protected steel) are inherently more resistant to fires, while other construction materials 44 (e.g., wood frame) are not as fire resistant. Within the structure, fire barriers may be constructed 45 to divide the building into separate fire areas. Fire barriers are made of fire-resistant materials 46 and are designed to contain a fire to one area until it can be extinguished. Fire barriers are 1  This guidance document includes references to the NFPA codes because these are standards the NRC staff uses in its licensing review. These various NFPA codes are also referenced throughout chapter 7 of NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications. For example, section 7.4.2 of NUREG-1520 specifically references NFPA 801 as an NRC-accepted standard.
59
 
1 typically walls, although some floor or ceiling assemblies have a fire resistance rating and may 2 be credited as fire barriers.
3 4 In a fire barrier, all components of the assembly (i.e., wall, doors, windows, dampers) have a 5 certified fire rating. For openings within the assembly, the rating is meant to indicate the amount 6 of time that the opening protection can resist the fire from penetrating the assembly. All 7 penetrations (e.g., piping, cables, air ducts) should be sealed with a fire-resistant seal or 8 damper. The application should include a description and fire rating of any opening protection, 9 penetration seals, or fire dampers.
10 11 The application should describe the facilitys electrical system. Nearly all electrical installations 12 in the United States are required by local jurisdictions to comply with NFPA 70, National 13 Electrical Code. It should confirm this compliance and briefly describe any high voltage 14 equipment in areas where licensed material is present. Furthermore, the application should 15 describe any lightning protection installed at the facility.
16 17 Occupant egress (i.e., travel pathway to safety) is an important aspect of life safety that the 18 application should discuss. The NRC staff recognizes NFPA 101, Life Safety Code, as a 19 standard that specifies acceptable occupant egress design criteria. Maintaining adequate 20 clearances and design of exit access pathways and corridors is a key component of a timely exit 21 strategy. The facility should also provide sufficient emergency lighting and exit signage so that, 22 in the event of a power failure, occupants can still reach a safe area or exit the building.
23 24 In accordance with the requirements of 10 CFR 70.22(a)(7), an adequate application describes 25 equipment and facilities that will be used to protect health and minimize danger to life or 26 property, including nuclear safety and environmental issues related to the facilitys fire protection 27 design. For example, if a sprinkler system is being used where licensed material is present, the 28 quantity and enrichment level of special nuclear material may dictate the need to address a 29 potential criticality accident. Additionally, if dispersible licensed material is present, water used 30 for firefighting may become contaminated and should be monitored and treated before being 31 released into the sewer system. The same principle applies to ventilation systems, especially 32 when attached to fume hoods or glove boxes where dispersible licensed material is handled.
33 These concerns may not be applicable to all facilities, depending on the type of equipment and 34 nuclear material present.
35 36 5.2      Fire Protection 37 38 In accordance with the requirements of 10 CFR 70.22(a)(6)-(8), an adequate application 39 describes the technical qualifications, including training and experience, and the equipment, 40 facilities, and proposed procedures that protect health and minimize danger to life or property.
41 Therefore, the application should document the fire protection systems, employee training, and 42 emergency response organizations available for the facilities it includes.
43 44 5.2.1    Fire Protection Systems 45 46 In accordance with the requirements of 10 CFR 70.22(a)(7), an adequate application describes 47 equipment and facilities used to protect health and minimize danger to life or property, including 48 the fire protection systems that provide adequate protection against fires and explosions in the 49 areas in which licensed material is present. For these areas, the application should describe any 50 fire suppression, detection, or alarm systems, as well as all inspecting, testing, and 51 60
 
1 maintenance conducted on these systems. Compliance with applicable NFPA standards or the 2 building code may be sufficient to address many of these areas in the application.
3 4 A fire suppression system is typically composed of water- or chemical-based agents that are 5 used to put out a fire or contain it until emergency response personnel can arrive. Sprinklers are 6 the most commonly used form of fire suppression. Sprinkler systems are typically designed and 7 installed in accordance with NFPA 13, Standard for the Installation of Sprinkler Systems and 8 inspected, tested, and maintained in accordance with NFPA 25, Standard for the Inspection, 9 Testing, and Maintenance of Water-Based Fire Protection Systems. Documenting NFPA code 10 compliance is an acceptable way to describe the design, installation, inspection, testing, and 11 maintenance of sprinkler systems.
12 13 It is possible that other forms of fire suppression (i.e., foam or gaseous systems) may be used.
14 In this case, the application should document compliance with any relevant NFPA codes or 15 provide details about the installation of the suppression system. The applicant should provide 16 hand-held fire extinguishers, as required, and train employees who may handle licensed 17 material in how to operate them. The application should indicate the type and placement of 18 these extinguishers, along with all relevant testing, inspections, and training.
19 20 Fire detection and alarm systems typically work together to sense a fire in its incipient stages 21 and notify occupants and the fire department. These systems are typically installed, maintained, 22 and tested in accordance with NFPA 72, National Fire Alarm and Signaling Code. Some 23 detection systems may be incorporated directly into the ventilation system for when rapid 24 detection is especially important, given the nature of the materials being used. The application 25 should describe the detector and alarm spacing, the type, and the testing and maintenance 26 schedule.
27 28 In accordance with the requirements of 10 CFR 70.22(a)(8), an adequate application describes 29 the proposed procedures to protect health and minimize danger to life or property, which includes 30 the inspection, testing, and maintenance of fire protection systems. The NRC assesses the 31 adequacy of an applicants fire protection program based on the fire protection features that are 32 available and reliable when called upon. Features that are not properly inspected and 33 maintained can impact the safety of the facility. All fire protection systems should be maintained 34 in accordance with the applicable NFPA or building code, and the application should explain any 35 deviations from the prescribed maintenance plan.
36 37 5.2.2    Employee Training 38 39 In accordance with the requirements of 10 CFR 70.22(a)(6), an adequate application describes 40 the technical qualifications, including training and experience of the applicant and members of its 41 staff to engage in the proposed activities, including fire safety training for employees that work in 42 areas where licensed material is present. For these areas, the application should describe any 43 emergency training provided to the occupants of the facility. Compliance with applicable NFPA 44 standards may be sufficient to address this area. The application should describe and document 45 examples of employee training such as evacuation exercises (i.e., fire drills), the safe shutdown 46 process, and the use of fire extinguishers, listing all relevant codes and describing the facilitys 47 degree of compliance.
48 49 50 51 61
 
1 5.2.3    Emergency Response 2
3 In accordance with the requirements of 10 CFR 70.22(a)(8), an adequate application describes 4 the proposed procedures to protect health and minimize danger to life or property, which includes 5 the emergency response capabilities for the facility. An adequate application describes the 6 emergency response for areas in which licensed material is present.
7 8 The application should describe the emergency organization(s) that would respond to the facility 9 in the event of a fire. This may be a dedicated onsite brigade or the local fire department. If an 10 offsite fire department would respond to a fire, periodic training may be necessary to familiarize 11 the emergency responders with facility access procedures, layout, and pre-fire plans. The 12 application should describe the training and experience that the responders have in handling 13 fires relating to nuclear material. The application should also describe the water supply available 14 for firefighting operations. The application should list any relevant code(s) in describing the 15 response organizations qualifications and capabilities. The applicant and the responding fire 16 department(s) should have a written plan or agreement to define the emergency response.
17 18 5.3      Process Fire Safety 19 20 In accordance with the requirements of 10 CFR 70.22(a)(7), an adequate application describes 21 equipment and facilities that will protect health and minimize danger to life or property, including 22 the process fire safety within areas where licensed material is present. Process fire safety 23 involves design considerations to prevent an accident or to mitigate the consequences of an 24 accident resulting from the use of process chemicals, combustible metals, flammable and 25 combustible liquids and gases, high-temperature equipment, hot cells and glove boxes, and 26 laboratories. For these areas, the application should identify fire and explosion hazards, along 27 with the fire and explosion parameters of hazardous materials. Compliance with applicable 28 NFPA standards or the building code may be sufficient to address many of these areas 29 (e.g., NFPA 30, Flammable and Combustible Liquids Code; NFPA 69, Standard on Explosion 30 Prevention Systems; and NFPA 86, Standard for Ovens and Furnaces). The application 31 should list any relevant code(s) and describe the facilitys degree of compliance.
32 33 Hazardous chemicals and processes involving licensed material can contribute to the fire 34 hazard. In areas that have fire hazards that may threaten licensed material, the application 35 should identify any hazardous chemicals, processes, and design standards used to ensure fire 36 safety. The NRC staff recognizes NFPA 801 as one standard that provides acceptable design 37 criteria for radiological process areas that may contain hazardous material, laboratories, 38 high-temperature equipment, hot cells, and glove boxes. However, the applicant may use other 39 nationally recognized codes and standards, if appropriate.
40 41 The following are some of the more common material hazards found in laboratories:
42 43
* Fluorine reacts violently with organic material or metal powders and water vapor.
44 45
* Hydrogen is an explosive and flammable gas used in reduction processes.
46 47
* Hydrogen peroxide off-gases hydrogen and oxygen and is incompatible with some 48          extinguishers.
49 62
 
1
* Nitric acid nitrates organic material, which lowers the ignition temperature of 2        combustibles.
3 4
* Sulfuric acid absorbs water from organic material in an exothermic reaction, thereby 5        causing ignition.
6 7
* Zirconium is a combustible metal that burns at elevated temperatures.
8 9 5.4    Combustible Loading and Potential Fire Scenarios 10 11 In accordance with the requirements of 10 CFR 70.22(a)(7), an adequate application describes 12 equipment and facilities to protect health and minimize danger to life or property, including 13 potential fire scenarios for areas in which licensed material is present. For these areas, the 14 application should identify possible ignition sources and the measures that are in place to 15 sufficiently demonstrate reasonable assurance that the facility is protected from fires and 16 explosions. Compliance with applicable NFPA standards or the building code may be sufficient 17 to address many of these areas. The application should identify relevant code(s) and describe 18 any deviations from code compliance.
19 20 The application should describe potential combustible loading in each fire area. Combustibles 21 may include materials like chemicals, wood pallets, plastics, paper, or trash. This description 22 should include not only the combustibles that are present in the facility at present but also any 23 that may reasonably be assumed to be present in the future. If there are significant 24 combustibles, the application should consider the credible fire scenarios and document the 25 safety features to prevent potential fires from creating a radiological hazard. Examples of safety 26 features include sprinklers, fire extinguishers, fire barriers, and employee training.
27 28 5.5    Summary 29 30 In summary, in accordance with the requirements of 10 CFR 70.22(a)(6)-(8), an application 31 should document the following information about the applicants fire safety program:
32 33
* a list of all nuclear material and its location, quantity, and form, as well as a description 34        of any physical barriers separating the material with applicable fire ratings 35 36
* a description of the facilitys building construction, fire area determination, fire rated walls 37        and opening protection, electrical installation, lightning protection, emergency lighting, 38        life safety and egress, ventilation systems, and fire water drainage 39 40
* a description of the facilitys fire protection systems (i.e., suppression, alarm, and 41        detection) 42 43
* a description of the protection for laboratory equipment that may handle nuclear material 44        (e.g., glove boxes, hot cells, fume hoods) 45 46
* a description of the inspection, testing, and maintenance of fire protection systems 47 48
* a description of all fire safety training given to employees (e.g., use of fire extinguishers, 49        safe shutdown, evacuation) 50 63
 
1
* a description of the emergency response capabilities for the facility, including the 2        proximity, qualifications, and training of the responding fire department, including any 3        pre-fire coordination with the fire department (e.g., fire drills, hazardous materials 4        training), and the water supply for firefighting 5
6
* a list of all hazardous chemicals or processes that may present a fire hazard in 7        radiological areas 8
9
* a description, by fire area, of any potential combustible loading, possible fire scenarios, 10        and safety controls 11 12 Compliance with any relevant NFPA or building codes (i.e., NFPA 45 or NFPA 801) may 13 contribute to the discussion of the above items.
14 15 5.6    References 16 17 Code of Federal Regulations, Chapter I, Title 10, Energy, Part 70, Domestic Licensing of 18 Special Nuclear Material.
19 20 NFPA, Standard for the Installation of Sprinkler Systems, NFPA 13, 2019.
21 22 NFPA, Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection 23 Systems, NFPA 25, 2020.
24 25 NFPA, National Fire Alarm and Signaling Code, NFPA 72, 2019.
26 27 NFPA, Standard on Explosion Prevention Systems, NFPA 69, 2019.
28 29 NFPA, National Electrical Code, NFPA 70, 2020.
30 31 NFPA, Standard for Fire Protection for Facilities Handling Radioactive Materials, NFPA 801, 32 2020.
33 34 NFPA, Flammable and Combustible Liquids Code, NFPA 30, 2021.
35 36 NFPA, Standard on Fire Protection for Laboratories Using Chemicals, NFPA 45, 2019.
37 38 NFPA, Standard for Ovens and Furnaces, NFPA 86, 2019.
39 40 NFPA, Life Safety Code, NFPA 101, 2021.
41 42 NFPA, Standard on Types of Building Construction, NFPA 220, 2021.
43 44 U.S. Nuclear Regulatory Commission, Standard Review Plan for Fuel Cycle Facilities License 45 Applications, NUREG-1520.
64
 
1                  6      NATIONAL ENVIRONMENTAL POLICY ACT 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) Part 51, 4 Environmental Protection Regulations for Domestic Licensing and Related Regulatory 5 Functions; 10 CFR 51.22, Criterion for categorical exclusion; identification of licensing and 6 regulatory actions eligible for categorical exclusion or otherwise not requiring environmental 7 review; and 10 CFR 70.21(f).
8 9 Regulatory Criteria: An application for a license to possess a critical mass quantity of special 10 nuclear material (SNM) for the authorized activities listed in 10 CFR 51.22 will be acceptable if it 11 meets the following criteria:
12 13
* The application accurately describes the activity for which the license is requested.
14 15
* The activity for which the license is requested meets any of the criteria for the 16          categorical exclusion (CATEX) defined in 10 CFR 51.22.
17 18 Discussion: The National Environmental Policy Act of 1969, as amended (NEPA), directs that, 19 to the fullest extent possible, (1) the policies, regulations, and public laws of the United States 20 shall be interpreted and administered in accordance with the policies set forth in NEPA, and 21 (2) all agencies of the Federal Government shall comply with the procedures in section 102(2) 22 of NEPA except where compliance would be inconsistent with other statutory requirements.
23 The regulations in 10 CFR Part 51 implement section 102(2) of NEPA in a manner consistent 24 with the domestic licensing and related regulatory authority of the U.S. Nuclear Regulatory 25 Commission (NRC) under the Atomic Energy Act of 1954, as amended (the Act); the Energy 26 Reorganization Act of 1974, as amended; and the Uranium Mill Tailings Radiation Control Act of 27 1978, and reflect the Commissions announced policy to take account of the regulations of the 28 Council on Environmental Quality published in the Federal Register (FR) on November 29, 1978 29 (43 FR 55978-56007), voluntarily, subject to certain conditions.
30 31 The regulations in 10 CFR Part 51 also address the limitations imposed on the NRCs authority 32 and responsibility under NEPA by the Federal Water Pollution Control Act Amendments of 33 1972, Pub. L. 92-500, 86 Stat. 816 et seq. (33 U.S.C. 1251 et seq.). In accordance with 34 section 511(c)(2) of the Federal Water Pollution Control Act (86 Stat. 893, 33 U.S.C 1371(c)(2)),
35 the NRC recognizes that responsibility for Federal regulation of non-radiological pollutant 36 discharges into receiving waters rests by statute with the U.S. Environmental Protection 37 Agency.
38 39 6.1      Classification of Licensing and Regulatory Actions 40 41 The NRC licensing actions fall into one of three classifications with respect to NEPA: (1) actions 42 requiring an environmental impact statement as listed in 10 CFR 51.20, Criteria for and 43 identification of licensing and regulatory actions requiring environmental impact statements, 44 (2) actions requiring an environmental assessment subject to 10 CFR 51.21, Criteria for and 45 identification of licensing and regulatory actions requiring environmental assessments, or 46 (3) actions that do not require an environmental assessment because they fall under a CATEX 47 in 10 CFR 51.22.
48 49 50 65
 
1 Facilities authorized to use and possess critical mass quantities of SNM (also known as greater 2 than critical mass (GTCM) facilities), which use SNM for research and development purposes, 3 are categorically excluded under 10 CFR 51.22(c)(14)(v). Section 1 gives details on these 4 facilities. Therefore, these facilities are not subject to the requirements to prepare environmental 5 assessments under 10 CFR 51.20 or 10 CFR 51.21.
6 7 The NRC anticipates that agency actions associated with GTCM licenses fall into one or more 8 of the categories of actions that are CATEX (mainly under 10 CFR 51.22(c)(14)(v)use of 9 radioactive materials for research and development and for educational purposes. If an 10 application for a license requests authorization for any of the types of actions described in 11 10 CFR 51.60, Environmental reportmaterials licenses, or 10 CFR 70.21(f), then the 12 applicant must follow the guidance in NUREG-1748, Environmental Review Guidance for 13 Licensing Actions Associated with NMSS Programs.
14 15 The regulations in 10 CFR 51.22 state that licensing, regulatory, and administrative actions 16 eligible for CATEX shall meet the following criterion: the action belongs to a category of actions 17 which the Commission, by rule or regulation, has declared to be a CATEX, after first finding that 18 the category of actions does not individually or cumulatively have a significant effect on the 19 human environment. The regulations in 10 CFR 51.22(c) list the categories of actions that are 20 CATEX, including, in part, the following:
21 22    (12)    issuance of an amendment to a license under parts 50, 52, 60, 61, 63, 70, 72, 23              or 75 of this chapter relating solely to safeguards matters (i.e., protection 24              against sabotage or loss or diversion of special nuclear material) or issuance 25              of an approval of a safeguards plan submitted under parts 50, 52, 70, 72, and 26              73 of this chapter, provided that the amendment or approval does not involve 27              any significant construction effects. These amendments and approvals are 28              confined to 29 30              (i) Organizational and procedural matters; 31 32              (ii) Modifications to systems used for security and/or materials accountability; 33 34              (iii) Administrative changes; and 35 36              (iv) Review and approval of transportation routes under 10 CFR 73.37.
37 38    (14)    Issuance, amendment, or renewal of materials licenses issued under 39              10 CFR Parts 30, 31, 32, 33, 34, 35, 36, 39, 40 or part 70 authorizing the 40              following types of activities:
41 42              (v) Use of radioactive materials for research and development and for 43                    educational purpose.
44 45              (xvi) Any use of source, byproduct, or special nuclear material not listed 46                    above which involves quantities and forms of source, byproduct, or 47                    special nuclear material similar to those listed in paragraphs (c)(14) (i) 48                    through (xv) of this section [category 14].
49 66
 
1 (15) Issuance, amendment, or renewal of licenses for import of nuclear facilities 2      and materials pursuant to part 110 of this chapter, except for import of spent 3      power reactor fuel.
4 5 (17) Issuance of an amendment to a permit or license under parts 30, 40, 50, 52, 6      or part 70 of this chapter which deletes any limiting condition of operation or 7      monitoring requirement based on or applicable to any matter subject to the 8      provisions of the Federal Water Pollution Control Act.
9 10 (18) Issuance of amendments or orders authorizing licensees of production or 11      utilization facilities to resume operation, provided the basis for the 12      authorization rests solely on a determination or redetermination by the 13      Commission that applicable emergency planning requirements are met.
14 15 (20) Decommissioning of sites where licensed operations have been limited to the 16      use of 17 18      (i) Small quantities of short-lived radioactive materials; 19 20      (ii) Radioactive materials in sealed sources, provided there is no evidence of 21            leakage of radioactive material from these sealed sources; or 22 23      (iii) Radioactive materials in such a manner that a decommissioning plan is 24            not required by 10 CFR 30.36(g)(1), 40.42(g)(1), or 70.38(g)(1), and the 25            NRC has determined that the facility meets the radiological criteria for 26            unrestricted use in 10 CFR 20.1402 without further remediation or 27            analysis.
28 29 (21) Approvals of direct or indirect transfers of any license issued by NRC and any 30      associated amendments of license required to reflect the approval of a direct 31      or indirect transfer of an NRC license.
32 33 (25) Granting of an exemption from the requirements of any regulation of this 34      chapter, provided that 35 36      (i) There is no significant hazards consideration; 37 38      (ii) There is no significant change in the types or significant increase in the 39            amounts of any effluents that may be released offsite; 40 41      (iii) There is no significant increase in individual or cumulative public or 42            occupational radiation exposure; 43 44      (iv) There is no significant construction impact; 45 46      (v) There is no significant increase in the potential for or consequences from 47            radiological accidents; and 48 49 50 51 67
 
1              (vi) The requirements from which an exemption is sought involve:
2 3                  (A) Recordkeeping requirements; 4
5                  (B) Reporting requirements; 6
7                  (C) Inspection or surveillance requirements; 8
9                  (D) Equipment servicing or maintenance scheduling requirements; 10 11                  (E) Education, training, experience, qualification, requalification or other 12                      employment suitability requirements; 13 14                  (F) Safeguard plans, and materials control and accounting inventory 15                      scheduling requirements; 16 17                  (G) Scheduling requirements; 18 19                  (H) Surety, insurance, or indemnity requirements; or 20 21                  (I) Other requirements of an administrative, managerial, or organizational 22                      nature.
23 24 The license or amendment application should do the following:
25 26
* Describe the activities for which the applicant is requesting authorization.
27 28
* Identify the category or categories of actions into which the authorized activities would 29        fall (i.e., environmental impact statement, environmental assessment, or CATEX).
30 31
* Demonstrate that the requested authorization is for activities categorically excluded by 32        addressing the licensing action descriptions for CATEX, as discussed in section 2 of 33        NUREG-1748, Environmental Review Guidance for Licensing Actions Associated with 34        NMSS Programs, issued August 2003. This guidance provides general procedures for 35        the environmental review of licensing actions regulated by the Office of Nuclear Material 36        Safety and Safeguards (NMSS).
37 38 To facilitate staff review, applicants and licensees making a proposed CATEX determination 39 should consider the following five review questions listed in appendix B to NUREG-1748:
40 41
* Is the action consistent with the Statement of Considerations for the CATEX chosen?
42 43
* Is the action likely to significantly affect any aspect of the natural environment?
44 45
* Is the action likely to significantly affect any aspect of the cultural environment, including 46        those that might be related to environmental justice?
47 48
* Is the action likely to generate a great deal of public interest about any environmental 49        issue?
50 68
 
1
* Is there a high level of uncertainty about the actions environmental effects?
2 3 6.2    References 4
5 Code of Federal Regulations, Chapter 1, Title 10 (10 CFR), Energy, Part 51, Environmental 6 Protection Regulations for Domestic Licensing and Related Regulatory Functions.
7 8 10 CFR Part 70, Domestic Licensing of Special Nuclear Material.
9 10 U.S. Nuclear Regulatory Commission, Environmental Review Guidance for Licensing Actions 11 Associated with NMSS Programs, NUREG-1748, August 2003.
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1                7        MATERIAL CONTROL AND ACCOUNTING 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) 70.22(b);
4 10 CFR Part 74, Material Control and Accounting of Special Nuclear Material.
5 6 Regulatory Criteria: Applicants or licensees that possess special nuclear material (SNM) must 7 ensure the control and accounting of licensed material.
8 9 Discussion: The material control and accounting (MC&A) regulations in 10 CFR Part 74 10 contain the requirements for the control and accounting of SNM at fixed sites and for 11 documenting the transfer of SNM. The following subparts of 10 CFR Part 74 distinguish among 12 licensees authorized to possess different types and quantities of SNM:
13 14
* Subpart B, General Reporting and Recordkeeping Requirements, applies to each 15        licensee who possesses, receives, or transfers SNM in a quantity of 1 gram or more of 16        contained uranium (U)-235, U-233, or plutonium.
17 18
* Subpart C, Special Nuclear Material of Low Strategic Significance, applies to each 19        licensee authorized to possess and use SNM of low strategic significance, as defined in 20        10 CFR 74.4, Definitions, also known as a Category III quantity of SNM (10 CFR 74.31, 21        Nuclear material control and accounting for special nuclear material of low strategic 22        significance;10 CFR 74.33, Nuclear material control and accounting for uranium 23        enrichment facilities authorized to produce special nuclear material of low strategic 24        significance).
25 26
* Subpart D, Special Nuclear Material of Moderate Strategic Significance, applies to 27        each licensee authorized to possess and use SNM of moderate strategic significance, as 28        defined in 10 CFR 74.4, also known as a Category II quantity of SNM (10 CFR 74.41, 29        Nuclear material control and accounting for special nuclear material of moderate 30        strategic significance).
31 32
* Subpart E, Formula Quantities of Strategic Special Nuclear Material, applies to each 33        licensee who is authorized to possess five or more formula kilograms of strategic SNM, 34        as defined in 10 CFR 74.4, also known as a Category I quantity of SNM (10 CFR 74.51, 35        Nuclear material control and accounting for strategic special nuclear material).
36 37 The regulations in 10 CFR 70.22(b), in part, require that each application for a license to 38 possess and use SNM in a quantity exceeding 1 effective kilogram contain a full description of 39 the program for control and accounting of such SNM as will be in the applicants possession and 40 how compliance with the applicable requirements in 10 CFR 74.31, 10 CFR 74.33, 41 10 CFR 74.41, and 10 CFR 74.51, as applicable, will be accomplished. However, the 42 requirements in 10 CFR 70.22(b) contain an exclusion for licensees governed by 43 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, including 44 nonpower reactors (research and test reactors), for uses of SNM as sealed sources, and 45 operations involved in waste disposal. The same exclusions are contained in the MC&A 46 requirements in 10 CFR 74.31, 10 CFR 74.33, 10 CFR 74.41, and 10 CFR 74.51.
47 48 Applicants for a license to possess greater than critical mass (GTCM) quantities of SNM, as 49 described in chapter 1 of this document, as sealed sources are subject to the general reporting 50 and recordkeeping requirements under Subpart B but are not subject to the more extensive 71
 
1 MC&A requirements under Subparts C, D, or E of 10 CFR Part 74, which apply to licensees as 2 described above. Applicants for licenses to possess SNM of low or moderate strategic 3 significance, or formula quantities of strategic SNM, as defined in 10 CFR 74.4 that do not meet 4 the exceptions for sealed sources, or the other conditions, as described above, may be subject 5 to the further MC&A requirements of Subparts C, D, or E of 10 CFR Part 74. Applicants and 6 licensees should review the types and quantities of SNM to determine if the more extensive 7 MC&A requirements of Subpart C or D apply.
8 9 Guidance on the MC&A requirements for applicants or licensees subject to 10 CFR 74.31 10 appears in NUREG-1065, Acceptable Standard Format and Content for the Fundamental 11 Nuclear Material Control (FNMC) Plan Required for Low-Enriched Uranium Facilities. Guidance 12 on the MC&A requirements for applicants or licensees subject to 10 CFR 74.33 is found in 13 NUREG/CR-5734, Recommendations to the NRC on Acceptable Standard Format and Content 14 for the Fundamental Nuclear Material (FNMC) Plan Required for Low-Enriched Uranium 15 Enrichment Facilities, issued November 1991. NUREG-2159, Acceptable Standard Format 16 and Content for the Material Control and Accounting Plan Required for Special Nuclear Material 17 of Moderate Strategic Significance, includes guidance for the MC&A requirements in 18 10 CFR 74.41.
19 20 All licensees that possess, receive, or transfer SNM in any form in a quantity of 1 gram or more 21 of contained U-235, U-233, or plutonium are subject to the general reporting and recordkeeping 22 requirements of 10 CFR Part 74, Subpart B. The application should describe how it will 23 accomplish the general reporting and recordkeeping requirements of Subpart B.
24 25 The following requirements of Subpart B apply to GTCM applicants:
26 27
* 10 CFR 74.11, Reports of loss or theft or attempted theft or unauthorized production of 28          special nuclear material, requires the licensee to notify the U.S. Nuclear Regulatory 29          Commission (NRC) Operations Center in the event of any lost, stolen, or unlawfully 30          diverted SNM, including attempts, within 1 hour of discovery.
31 32
* 10 CFR 74.13, Material status reports, requires the licensee to prepare material 33          balance reports concerning SNM that the licensee has received, produced, possessed, 34          transferred, consumed, disposed of, or lost.
35 36
* 10 CFR 74.15, Nuclear material transaction reports, requires the licensee who 37          transfers or receives SNM in certain quantities or who adjusts its inventory of SNM to 38          submit a nuclear material transaction report.
39 40
* 10 CFR 74.19, Recordkeeping, requires the licensee to maintain and retain records of 41          the receipt, inventory, acquisition, transfer, and disposal of all SNM. This section also 42          requires certain licensees to establish, maintain, and follow written MC&A procedures 43          that are sufficient to enable the licensee to account for the SNM in its possession under 44          license. This section also requires a licensee possessing certain quantities to take a 45          physical inventory of all SNM in its possession at intervals not to exceed 12 months.
46 47 48 49 50 51 72
 
1 7.1    Reports of Loss or Theft or Attempted Theft of Special Nuclear Material 2
3 The regulations in 10 CFR 74.11 require the licensee to notify the NRC Operations Center in the 4 event of any lost, stolen, or unlawfully diverted SNM, including attempts, within 1 hour of 5 discovery.
6 7 The applicant should describe how indicators of a possible loss, theft, or diversion of SNM, 8 whether arising from errors or deliberate actions, will be investigated and resolved. The 9 applicant should have well-defined procedures for promptly investigating and resolving 10 indications of possible missing SNM and procedures for promptly determining whether an actual 11 loss of SNM has occurred. Resolving a loss indicator means that the licensee has determined 12 that loss, including possible diversion or theft, has not occurred and is not occurring. Any 13 investigation of an indication of a loss or theft should provide, whenever possible, (1) an 14 estimate of the quantity of SNM involved, (2) the material type of physical form of the material, 15 (3) the type of unauthorized activity or event detected, (4) the time frame within which the loss 16 or activity could have occurred, (5) the most probable cause(s), and (6) recommendations for 17 precluding reoccurrence.
18 19 For indications that a loss or theft may have occurred, the resolution process should include 20 (1) thoroughly checking the accountability records and source information, (2) locating the 21 source of the problem, (3) isolating the exact reason for the problem within the area, 22 (4) determining the amounts of SNM involved, and (5) determining that the indication is or is not 23 resolved. If an investigation of an indicator results in a conclusion that the indication is true, 24 such conclusion must be reported to the NRC within 1 hour of its determination in accordance 25 with 10 CFR 74.11. Procedures should identify all documentation requirements associated with 26 the methods for the reporting, investigation, and resolution of missing SNM indicators.
27 28 7.2    Material Status Reports 29 30 The regulations in 10 CFR 74.13 require licensees to prepare material balance reports and 31 physical inventory listings concerning SNM that the licensee has received, produced, 32 possessed, transferred, consumed, disposed of, or lost. U.S. Department of Energy (DOE)/NRC 33 Form 742, Material Balance Report, and DOE/NRC Form 742C, Physical Inventory Listing, 34 are the means for submitting reports of material balance and physical inventory listing data to 35 the Nuclear Materials Management and Safeguards System (NMMSS), which is the national 36 database used for tracking certain nuclear material. DOE/Form 742 is used to report a summary 37 of activity for a specified material within a material balance reporting period, as specified in 38 10 CFR 74.13. The report conveys beginning and ending inventory balances; activities such as 39 shipment and receipts involving other facilities; decay, transmutation; and production 40 calculations. DOE/NRC Form 742C is used to report a facilitys physical inventory listing as of a 41 specified date.
42 43 The applicant should generally describe how material status reports are prepared and submitted 44 to NMMSS. Reports must be submitted for each reporting identification symbol (RIS), which can 45 only be obtained after the NRC license is issued. Once the license is issued, the licensee 46 should contact the NRCs Office of Nuclear Material Safety and Safeguards, Division of Fuel 47 Management, to request an RIS. Processing the request for an RIS will require the NRC license 48 number, the address where the material will be used and stored, the business address of the 49 licensee, and the name and telephone number of a contact person.
50 The applicant should have well-defined procedures for preparing and submitting reports in a 51 computer-readable format in accordance with the detailed instructions contained in 73
 
1 NUREG/BR-0007, Instructions for the Preparation and Distribution of Material Status Reports 2 (DOE/NRC Forms 742 and 742C), and in NMMSS Report D-24, Personal Computer Data 3 Input for Nuclear Regulatory Commission Licensees. The procedures should ensure that 4 reports are made and filed within the required time periods, as defined in 10 CFR 74.13. If it will 5 possess U.S. Government-owned material, the applicant should also have procedures in place 6 to ensure that it will meet the DOE-reporting requirements for all receipts, transfers, and 7 inventories of U.S. Government-owned, loaned, or leased material, as specified in 8 NUREG/BR-0007 as well.
9 10 If it has materials that are nationally tracked sources, the applicant should have procedures in 11 place to ensure reporting to the National Source Tracking System (NSTS), which is a secure 12 user-friendly web-based database designed to track Category I and II radioactive sources 13 regulated by the NRC and the Agreement States. Applicants that have less than a critical mass 14 and have plutonium sources (less than 16 curies) or plutonium/beryllium sources should report 15 them to the NSTS. Information on NSTS can be found on the NRC public website at 16 http://www.nrc.gov/security/byproduct/ismp/nsts.html.
17 18 If the applicant will be subject to the requirements in 10 CFR Part 75, Safeguards on Nuclear 19 MaterialImplementation of Safeguards Agreements Between the United States and the 20 International Atomic Energy Agency, it should describe how it will submit the required material 21 status reports in accordance with 10 CFR 75.35, Material status reports.
22 23 7.3      Nuclear Material Transaction Reports 24 25 The regulations in 10 CFR 74.15 require a licensee who transfers or receives SNM in certain 26 quantities or who adjusts its inventory of SNM to submit that information to NMMSS. DOE/NRC 27 Form 741, Nuclear Material Transaction Report, is the means by which licensees submit 28 transaction data to NMMSS. DOE/NRC Form 741 is used to report physical transfers of nuclear 29 materials between facilities and to report exchanges of foreign obligations on material between 30 facilities even when no physical transfer occurs. The form is also used to report onsite 31 transactions such as inventory corrections that otherwise increase or decrease foreign 32 obligation balances or nuclear material categories within a facility.
33 34 The applicant should generally describe how it will track licensed materials from receipt to 35 disposal to ensure accurate accounting records and that possession limits listed on the license 36 are not exceeded. The applicant should describe how it prepares nuclear material transaction 37 reports and submits them to NMMSS. The applicant should have well-defined procedures for 38 preparing and submitting reports in a computer-readable format, in accordance with the detailed 39 instructions contained in NUREG/BR-0006, Instructions for Completing Nuclear Material 40 Transaction Reports (DOE/NRC Forms 741 and 740M), and in NMMSS Report D-24.If it will 41 possess U.S. Government-owned material, the applicant should also have procedures in place 42 to ensure that it will meet the DOE-reporting requirements for all receipts, transfers, and 43 inventories of U.S. Government-owned, loaned, or leased material, as specified in 44 NUREG/BR-0006 as well. If the applicant will be subject to the requirements in 10 CFR Part 75, 45 it should describe how it will submit the required inventory change reports in accordance with 46 10 CFR 75.34, Inventory change reports.
47 48 7.4      Recordkeeping 49 50 The regulation in 10 CFR 74.19(a) states that licensees not subject to 10 CFR 74.31, 51 10 CFR 74.33, 10 CFR 74.41. or 10 CFR 74.51 are subject to the recordkeeping requirements 74
 
1 in 10 CFR 74.19(a)(1)-(4), which require a licensee to maintain records of receipt, inventory, 2 acquisition, transfer, and disposal of all SNM in its possession. Each record relating to MC&A 3 that is required by this regulation or by license condition is to be maintained and retained in 4 accordance with the appropriate regulation or license condition. If a retention period is not 5 otherwise specified, the licensee shall retain the record until the Commission terminates the 6 license.
7 8 The applicant should generally describe the recordkeeping system used to maintain records of 9 receipt, use, transfer, and disposal (as waste) of all licensed material. Table 7-1 lists each type 10 of record and how long the record must be maintained. Other records, such as transfer records, 11 could be linked to radioactive material inventory records. Receipt records should also document 12 cases where the licensee found excessive radiation levels or radioactive contamination on 13 packages or containers of material received and describe the action taken.
14 15 Table 7-1 Records Maintenance Type of Record                          How Long Record Must Be Maintained Receipt, Acquisition, or      For as long as the material is possessed until 3 years after transfer Physical Inventory            or disposal Transfer                      For 3 years after transfer Disposal                      Until the NRC terminates the license 16 17 Receipt, transfer, and disposal records typically contain the following information:
18 19
* radionuclide, quantity, and date of measurement of SNM 20 21
* for each sealed source, manufacturer, model number, location, and, if needed for 22          identification, serial number and as appropriate, manufacturer and model number of the 23          device containing the sealed source 24 25
* date of the transfer and name and license number of the recipient, and description of the 26          affected radioactive material (e.g., radionuclide, quantity, manufacturers name and 27          model number, serial number) 28 29
* for licensed materials disposed of as waste, the radionuclide, quantity, date of disposal, 30          and method of disposal (e.g., decay, sewer) 31 32 7.5      Written Material, Control, & Accounting Procedures 33 34 The regulation in 10 CFR 74.19(b) states that each licensee authorized to possess SNM in a 35 quantity exceeding 1 effective kilogram shall establish, maintain, and follow written MC&A 36 procedures that are sufficient to enable the licensee to account for the SNM in its possession 37 under license. If the applicant will possess greater than 1 effective kilogram of SNM, it should 38 generally describe the written procedures established to ensure all the applicable MC&A 39 requirements are met.
40 41 75
 
1 7.6      Physical Inventories 2
3 The regulation in 10 CFR 74.19(c) states that each licensee not subject to 10 CFR 74.31, 4 10 CFR 74.33, 10 CFR 74.41, or 10 CFR 74.51 and authorized to possess SNM in a quantity 5 greater than 350 grams of contained U-235, U-233, or plutonium, or any combination thereof, 6 shall make a physical inventory of all SNM in its possession under license at intervals not to 7 exceed 12 months. The applicant should generally describe how it performs physical inventories 8 of its SNM. The applicant should have well-defined procedures for the planning, conducting, 9 assessing, and reporting of the physical inventories. Licensees are required to submit reports 10 regarding the physical inventory in accordance with the requirements in 10 CFR 74.13. The 11 applicant should describe how it maintains and retains inventory records in accordance with 12 10 CFR 74.19.
13 14 7.7      Records Access and Storage 15 16 The regulation in 10 CFR 74.19(d) requires licensees to ensure that the recordkeeping system 17 can produce clear and legible copies of records after storage for the period specified by the 18 regulations. The section also states that the licensee shall maintain adequate safeguards 19 against tampering with and loss of records. The applicant should generally describe how it 20 stores records and how it controls its access to records to meet the requirement in 21 10 CFR 74.19(d).
22 23 7.8      Additional Information 24 25 As previously noted, GTCM applicants who intend to possess certain amounts and types of 26 SNM that are not in sealed sources may be subject to additional MC&A requirements in 27 10 CFR Part 74 other than those in Subpart B. A license to possess SNM of low strategic 28 significance (Category III), or SNM of moderate strategic significance (Category II), that is not in 29 sealed sources may be subject to requirements in 10 CFR Part 74, Subparts C and D, 30 respectively. Applicants for licenses to possess such material should contact the NRC for further 31 guidance.
32 33 7.9      References 34 35 Code of Federal Regulations, Chapter 1, Title 10 (10 CFR), Energy, Part 70, Domestic 36 Licensing of Special Nuclear Material.
37 38 10 CFR Part 74, Material Control and Accounting of Special Nuclear Material.
39 40 Nuclear Material Management and Safeguards System Report D-24, Personal Computer Data 41 Input for Nuclear Regulatory Commission Licensees 42 (http://nnsa.energy.gov/aboutus/ourprograms/nuclearsecurity/nmmsshome/nmmssinfo).
43 44 U.S. Nuclear Regulatory Commission (NRC), Acceptable Standard Format and Content for the 45 Material Control and Accounting Plan Required for Special Nuclear Material of Moderate 46 Strategic Significance, NUREG-2159.
47 48 NRC, Instructions for Completing Nuclear Material Transaction Reports (DOE/NRC Forms 741 49 and 740M), NUREG/BR-0006.
50 76
 
1 NRC, Instructions for the Preparation and Distribution of Material Status Reports (DOE/NRC 2 Forms 742 and 742C), NUREG/BR-0007.
3 4 NRC, Acceptable Standard Format and Content for the Fundamental Material Control (FNMC) 5 Plan Required for Low-Enriched Uranium Facilities, NUREG-1065.
6 7 NRC, Recommendations to the NRC on Acceptable Standard Format and Content for the 8 Fundamental Nuclear Material Control (FNMC) Plan Required for Low-Enriched Uranium 9 Enrichment Facilities, NUREG/CR-5734, November 1991.
77
 
1            8        DECOMMISSIONING AND FINANCIAL ASSURANCE 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) 70.22(a)(9);
4 10 CFR 70.25, Financial assurance and recordkeeping for decommissioning; 10 CFR 70.51, 5 Records requirements.
6 7 Regulatory Criteria: Pursuant to 10 CFR 70.25(b), a licensee authorized to possess licensed 8 material in excess of the limits specified in 10 CFR 70.25(d) must submit a decommissioning 9 funding plan (DFP) in accordance with the requirements in 10 CFR 70.25(b)(1) or provide a 10 certification of financial assurance for decommissioning in accordance with the requirements in 11 10 CFR 70.25(b)(2). The requirements in 10 CFR 70.25(e) cover preparing a DFP. Financial 12 instruments must meet the requirements in 10 CFR 70.25(f). In addition, in accordance with the 13 requirements in 10 CFR 70.25(g), all licensees are required to maintain records of information 14 important to the decommissioning of the facility in an identified location until the site is released 15 for unrestricted use. Licensees must transfer these records either to the new licensee before 16 licensed activities are transferred or assigned in accordance with 10 CFR 70.51(b)(6), or to the 17 appropriate U.S. Nuclear Regulatory Commission (NRC) regional office before the license is 18 terminated.
19 20 Discussion: The NRC regulations requiring certification of financial assurance for 21 decommissioning or a DFP are designed to provide reasonable assurance that the technical 22 and environmental components of decommissioning are carried out and that decommissioning 23 of the facilities is possible at the conclusion or termination of licensed activities. These 24 requirements specify that a licensee either set aside funds for decommissioning activities or 25 provide a guarantee that funds will be available. Criteria for determining whether an applicant is 26 required to submit a DFP or certification of financial assurance under 10 CFR 70.25(b) are set 27 forth in 10 CFR 70.25(d). Applicants should consider encapsulated sources and certain custom 28 sources as unsealed material requiring financial assurance in accordance with 29 10 CFR 70.25(b).
30 31 8.1      Financial Assurance and Decommissioning Funding Plan 32 33 The applicant must ensure that decommissioning activities will be carried out in accordance with 34 applicable NRC regulations and that they will be carried out with minimal impact on public and 35 occupational health and safety and the environment (see Federal Register (FR), Notice of Final 36 Rule, General Requirements for Decommissioning Nuclear Facilities, 53 FR 24018, 37 June 27, 1988). There are two parts to 10 CFR 70.25: financial assurance that applies to some 38 licensees (10 CFR 70.25(a)-(d) and (f)), and recordkeeping that applies to all licensees 39 (10 CFR 70.25(g)).
40 41 The NRC provides guidance on developing a site-specific DFP and approved wording for each 42 mechanism authorized by the regulation to guarantee or secure funds in NUREG-1757, 43 Consolidated NMSS Decommissioning Guidance, Volume 3, Financial Assurance, 44 Recordkeeping, and Timeliness, issued February 2012.
45 46 8.2      Recordkeeping 47 48 The requirements for maintaining records important to decommissioning, including the type of 49 information required, are stated in 10 CFR 70.25(g). Licensees need to maintain permanent 50 records of where licensed material is used or stored while the license is active. This is important 79
 
1 for making future determinations about the release of these locations for unrestricted use 2 (e.g., before the license is terminated). Acceptable records are sketches or written descriptions 3 of the specific locations or room numbers where SNM is used or stored and any records of spills 4 in or around the licensees facilities or information relevant to damaged devices or leak tests of 5 radioactive sources. The NRC requires all licensees to maintain these records in an identified 6 location until the site is released for unrestricted use. If the licensed activities are transferred to 7 another person or entity, these records shall be transferred to the new licensee before transfer 8 of the licensed activities. The new licensee is responsible for maintaining these records until the 9 license is terminated. When the license is terminated, these records shall be transferred to the 10 NRC.
11 12 Note: No response is needed from most applicants requesting only sealed sources. If financial 13 assurance or a DFP is required, applicants should submit the required documents as described 14 in NUREG-1757, Volume 3.
15 16 8.3      References 17 18 Code of Federal Regulations, Chapter 1, Title 10, Energy, Part 70, Domestic Licensing of 19 Special Nuclear Material.
20 21 U.S. Nuclear Regulatory Commission (NRC), Notice of Final Rule, General Requirements for 22 Decommissioning Nuclear Facilities, 53 FR 24018, June 27, 1988.
23 24 NRC, Consolidated NMSS Decommissioning GuidanceFinancial Assurance, Recordkeeping, 25 and Timeliness, NUREG-1757, Volume 3.
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1                                        9          PHYSICAL SECURITY 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) Part 20, 4 Standards for Protection against Radiation, Subpart I, Storage and Control of Licensed 5 Material; 10 CFR 20.2207, Reports of transactions involving nationally tracked sources; 6 10 CFR 70.22(4)(k); 10 CFR Part 73, Physical Protection of Plants and Materials; 7 10 CFR Part 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive 8 Material.
9 10 Regulatory Criteria: Applicants or licensees that possess special nuclear material (SNM) must 11 ensure the physical security and control 2 of licensed material.
12 13 Discussion: The regulations requiring control, security, and physical protection of SNM are 14 graded based on the risk of the SNM to be used for malicious purposes. Control of all licensed 15 material is covered in 10 CFR Part 20, Subpart I. These regulations require licensees to secure 16 licensed materials in storage from unauthorized removal or access and to control and maintain 17 constant surveillance of licensed material that is in a controlled or unrestricted area and that is 18 not in storage. These basic requirements are largely intended to prevent unintended or 19 inadvertent exposure of the public from gaining access to licensed material during use and 20 storage or from licensed material being lost or uncontrolled in public and are not intended to 21 cover physical protection of licensed material designed to prevent or minimize adversaries from 22 acquiring and using the material maliciously.
23 24 The SNM physical protection regulatory requirements designed to prevent or minimize 25 adversaries from acquiring and using the material maliciously at fixed sites and in transit are 26 graded using a material categorization approach. Depending on the type, quantity (i.e., mass) 27 and enrichment of the SNM, the material categorization approach places uranium and plutonium 28 into one of three risk categories: (1) Category I (i.e., formula quantity of strategic SNM),
29 (2) Category II (i.e., SNM of moderate strategic significance), or (3) Category III (i.e., SNM of 30 low strategic significance). The regulations in 10 CFR Part 73 provide SNM physical protection 31 requirements within these categories, which are based primarily on the ability of an adversary to 32 create an improvised nuclear device using SNM. However, the regulations contain exemptions 33 and exceptions to the three-category approach. With respect to 10 CFR Part 73, this guidance 34 addresses Category II and Category III physical protection requirements. In addition to the 35 regulations, following the events of September 11, 2001, the U.S. Nuclear Regulatory 36 Commission (NRC) issued security orders to certain licensees that required additional security 37 measures. New applicants should contact the NRC to determine whether existing security 38 orders are applicable to their facilities. In accordance with 10 CFR 70.22(k), each application for 39 a license to possess or use SNM of moderate strategic significance or 10 kilograms or more of 40 low strategic significance must include a physical security plan that demonstrates how the 41 applicant plans to meet the requirements of 10 CFR 73.67(d), (e), (f), and (g), as appropriate.
42 43 In addition to the SNM physical protection requirements in 10 CFR Part 73, certain isotopes and 44 forms of plutonium, as discussed below, must be protected in accordance with the requirements 45 in 10 CFR Part 37, which establish security requirements for the use and transport of 46 Category 1 and Category 2 quantities of radioactive material. These materials are widely used 47 in the United States by industrial, medical, and academic institutions and their theft or diversion 2  Control in this section is used as it relates to 10 CFR Part 20. The use of control in this section does not pertain to material control and accountability discussed in chapter 7 of this document.
81
 
1 could lead to their use in a radiological dispersal device or a radiological exposure device. The 2 security of plutonium (Pu)-238 material and plutonium/beryllium encapsulated, or sealed 3 sources, is addressed by 10 CFR Part 37. However, 10 CFR Part 37 does not include other 4 plutonium isotopes (such as Pu-240, Pu-241, or Pu-242) or unsealed materials.
5 6 Applicants and licensees should review its types, forms, and quantities of SNM to determine 7 which security or physical protection requirements apply. The discussion below discusses this 8 subject in greater detail.
9 10 9.1    Category II Special Nuclear Material Physical Protection Requirements 11 12 Regulations in 10 CFR 73.67, Licensee fixed site and in-transit requirements for the physical 13 protection of special nuclear material of moderate and low strategic significance, and 14 10 CFR Part 73, Appendix E, Levels of Physical Protection To Be Applied in International 15 Transport of Nuclear Materials, include the physical protection requirements specific to SNM of 16 moderate strategic significance (also known as Category II quantity of materialsee 17 10 CFR 73.2, Definitions).
18 19 9.1.1  Applicability 20 21 The definition of SNM of moderate strategic significance in 10 CFR 70.4, Definitions, is as 22 follows:
23 24        (1) Less than a formula quantity of strategic special nuclear material but more 25              than 1,000 grams of uranium-235 (contained in uranium enriched to 26              20 percent or more in the U-235 isotope) or more than 500 grams of 27              uranium-233 or plutonium, or in a combined quantity of more than 28              1,000 grams when computed by the equation, grams = (grams contained 29              U-235) + 2 (grams U-233 + grams plutonium), or 30 31        (2) 10,000 grams or more of uranium-235 (contained in uranium enriched to 32              10 percent or more but less than 20 percent in the U-235 isotope).
33 34 In accordance with 10 CFR 73.67(b)(1), licensees are exempt from the physical protection 35 requirements 10 CFR 73.67 to the extent they possess, use, or transport the following:
36 37        (i) Special nuclear material which is not readily separable from other radioactive 38              material and which has a total external radiation dose rate in excess of 1 gray 39              (100 rad) [100 rem] per hour at a distance of 1 meter (3.3 feet) from any 40              accessible surface without intervening shielding, or 41 42        (ii) Sealed plutonium-beryllium neutron sources totaling 500 grams or less 43              contained plutonium at any one site or contiguous sites, or 44 45        (iii) Plutonium with an isotopic concentration exceeding 80 percent in 46              plutonium-238.
47 48 9.1.2  General Performance Objectives 49 50 The general performance objectives in 10 CFR 73.67(a), applicable to fixed sites and transit of 51 SNM, specify that facilities with Category II quantities of SNM must minimize the possibilities for 82
 
1 unauthorized removal of SNM and facilitate the location and recovery of missing SNM. To 2 achieve these objectives, the physical security measures shall provide for (1) early detection 3 and assessment of unauthorized access or activities by an external adversary within the 4 controlled access area containing SNM, (2) early detection of removal of SNM by an external 5 adversary from a controlled access area, (3) assurance of proper placement and transfer of 6 custody of SNM, and (4) response to indications of an unauthorized removal of SNM and then 7 notification of the appropriate response forces of its removal to facilitate its recovery.
8 9 In accordance with 10 CFR 70.22(k), each applicant planning to possess or use Category II 10 quantities of SNM must include a physical security plan that demonstrates how it meets the 11 requirements in 10 CFR 73.67(d) and (e), as appropriate. Licensees shall retain a copy of the 12 physical security plan as a record for the period during which the licensee possesses 13 Category II quantities of SNM, and if any portion of the plan is superseded, retain that 14 superseded portion of the plan for 3 years after the effective date of the change. The regulations 15 in 10 CFR 73.67(c) require the licensee to submit the security plan to the NRC for review and 16 approval and then to implement the approved security plan.
17 18 Regulatory Guide 5.59, Standard Format and Content for a Licensee Physical Security Plan for 19 the Protection of Special Nuclear Material of Moderate or Low Strategic Significance, contains 20 additional guidance on developing a Category II physical protection program.
21 22 9.1.3  Physical Protection Requirements at Fixed Sites 23 24 In accordance with 10 CFR 73.67(d), applicable to fixed sites, licensees shall only use the 25 Category II quantities of SNM in a controlled access area that is illuminated sufficiently to allow 26 detection and surveillance of unauthorized penetration or activities (10 CFR 73.67(d)(1)). A 27 controlled access area is any temporarily or permanently established area that is clearly 28 demarcated, typically with a barrier (e.g., fence or wall), that affords isolation of the material or 29 persons within the controlled access area. The controlled access area shall be monitored with 30 an intrusion alarm or other device or procedures to detect unauthorized penetration or activities 31 (10 CFR 73.67(d)(3)).
32 33 Admittance to the controlled access area is controlled to allow only authorized persons 34 (e.g., persons with unescorted access or escorted visitors) who require such access to perform 35 their duties (10 CFR 73.67(d)(6)). Licensees shall screen individuals before granting them 36 unescorted access to the controlled access area to obtain information on which to base a 37 decision to permit such access (10 CFR 73.67(d)(4)). Licensees shall develop and maintain a 38 controlled badging and lock system to identify and limit access to the controlled access areas to 39 authorized individuals (10 CFR 73.67(d)(5)). Licensees shall assure that all visitors to the 40 controlled access areas are under the constant escort of an individual who has been authorized 41 unescorted access to the area (10 CFR 73.67(d)(7)). Vehicles and packages leaving the 42 controlled access area shall be searched on a random basis (10 CFR 73.67(d)(10)). The 43 purpose of the search is to detect unauthorized removal of SNM from the controlled access 44 area.
45 46 Within the controlled access area, Category II quantities of SNM shall be stored in a vault-type 47 room, approved security cabinet or equivalent. A vault-type room shall be capable of being 48 locked and shall be protected by an intrusion alarm, which creates an alarm upon the entry of a 49 person anywhere into the room and upon exit from the room or upon movement of an individual 50 within the room. The storage area(s) shall be sufficiently illuminated to allow detection and 51 surveillance of unauthorized penetration or activities (10 CFR 73.67(d)(2)).
83
 
1 Licensees shall establish a security organization to consist of at least one watchman per shift 2 able to assess and respond to any unauthorized penetrations or activities in the controlled 3 access areas (10 CFR 73.67(d)(8)). The security organization shall have the capability to 4 communicate with the appropriate response force, typically local law enforcement 5 (10 CFR 73.67(d)(9)). Licensees shall establish and maintain written response procedures for 6 dealing with threats of thefts or thefts of SNM (10 CFR 73.67(d)(11)). The licensee shall retain a 7 copy of the response procedures for 3 years thereafter. Copies of superseded material must be 8 retained for 3 years after each change.
9 10 9.1.4    Physical Protection Requirements in Transit 11 12 In accordance with 10 CFR 73.67(e), applicable to transit of SNM, licensees that transport, or 13 deliver to a carrier for transport, Category II quantities of SNM shall provide advance notification 14 to the receiver of any planned shipments specifying the mode of transport, estimated time of 15 arrival, location of the nuclear material transfer point, name of carrier and transport 16 identification. Before beginning the shipment, licensees shall receive confirmation from the 17 receiver that the receiver will be ready to accept the shipment at the planned time and location 18 and acknowledges the specified mode of transport. Before shipment, licensees shall check the 19 integrity of the container and locks or seals (10 CFR 73.67(e)(1)).
20 21 In-transit physical protection of Category II quantities of SNM must be arranged by either the 22 shipper or the receiver. The licensee that transports, exports, or delivers to a carrier for 23 transport Category II quantities of SNM shall arrange for the in-transit physical protection of the 24 materials unless the receiver is a licensee and has agreed in writing to arrange for the in-transit 25 physical protection (10 CFR 73.67(e)(1)). The receiver shall arrange for the in-transit physical 26 protection unless the shipper is a licensee and has agreed in writing to arrange for the in-transit 27 physical protection (10 CFR 73.67(e)(2))
28 29 In-transit physical protection shall include arranging for telephone or radio communications 30 between the transport and the licensee or its designee. The purpose of this communication is to 31 periodically confirm the status of the shipment, to notify parties of any delays in the scheduled 32 shipment, and to request appropriate local law enforcement agency response in the event of an 33 emergency. Planned communications shall include immediate notification of the arrival of the 34 shipment at its destination, or if the shipment is lost or unaccounted for after the estimated time 35 of arrival at its destination (10 CFR 73.67(e)(3)). Licensees receiving Category II quantities of 36 SNM shall check the integrity of the containers and seals upon receipt of the shipment and 37 notify the shipper of receipt of the material (10 CFR 73.67(e)(2)).
38 39 For in-transit physical protection considerations, the routing and transport shall minimize the 40 time that the material is in transit by reducing the number and duration of transfers and by 41 routing the material in the most safe and direct manner. Licensees shall screen all licensee 42 employees involved in the transportation of Category II quantities of SNM to obtain information 43 on which to base a decision to permit them control over the material (10 CFR 73.67(e)(3)).
44 Licensees shall make all shipments of Category II quantities of SNM either in dedicated 45 transports with no intermediate stops to load or unload other cargo and with no carrier or vehicle 46 transfers or temporary storage in-transit, or under arrangements whereby the custody of the 47 shipment and all custody transfers are acknowledged by signature. Licensees shall also 48 maintain Category II SNM under lock or under the control of an individual who has 49 acknowledged acceptance of custody by signature (10 CFR 73.67(e)(4)).
50 84
 
1 Licensees shall establish and maintain written response procedures for dealing with threats of 2 thefts or thefts of SNM. Response procedures shall include immediate initiation of a trace 3 investigation of any shipment that is determined to be lost or unaccounted for after a reasonable 4 time beyond the estimated arrival time. Response procedures shall also include notification of 5 the NRC Operations Center within 1 hour after the discovery of the loss of the shipment and 6 within 1 hour after recovery of or accounting for such lost shipment (10 CFR 73.67(e)(3)).
7 8 The licensee shall retain a copy of required documentation as a record for 3 years after the 9 close of the period for which the licensee possesses the SNM under each license for which the 10 original procedures were developed, and copies of superseded material must be retained for 11 3 years after each change (10 CFR 73.67(e)(4)).
12 13 9.2    Category III SNM Physical Protection Requirements 14 15 Regulations in 10 CFR 73.67 and Appendix E to 10 CFR Part 73 include physical protection 16 requirements specific to SNM of low strategic significance (also known as Category III quantity 17 of materialsee 10 CFR 73.2).
18 19 9.2.1  Applicability 20 21 The definition of SNM of low strategic significance (Category III) in 10 CFR 70.4 is as follows:
22 23        (1) Less than an amount of special nuclear material of moderate strategic 24              significance as defined in paragraph (1) of the definition of strategic nuclear 25              material of moderate strategic significance  but more than 15 grams of 26              uranium-235 (contained in uranium enriched to 20 percent) or 15 grams of 27              uranium-233 or 15 grams of plutonium or the combination of 15 grams when 28              computed by the equation, grams = (grams contained U-235) + (grams 29              plutonium) + (grams U-233); or 30 31        (2) Less than 10,000 grams but more than 1,000 grams of U-235 (contained in 32              uranium enriched to 10 percent or more but less than 20 percent in the U-235 33              isotope); or 34 35        (3) 10,000 grams or more of U-235 (contained in uranium enriched above natural 36              but less than 10 percent in the U-235 isotope).
37 38 In accordance with 10 CFR 73.67(b)(1), licensees are exempt from the requirements stated 39 below in 10 CFR 73.67 to that extent that they possess, use, or transport:
40 41        (iv) Special nuclear material which is not readily separable from other radioactive 42              material and which has a total external radiation dose rate in excess of 100 rems per 43              hour at a distance of 3 feet from any accessible surface without intervening shielding, 44              or 45 46        (v) Sealed plutonium-beryllium neutron sources totaling 500 grams or less contained 47              plutonium at any one site or contiguous sites, or 48 49        (vi) Plutonium with an isotopic concentration exceeding 80 percent in plutonium-238.
50 51 85
 
1 9.2.2    General 2
3 The general performance objectives in 10 CFR 73.67(a) specify that facilities with Category III 4 quantities of SNM must minimize the possibilities for unauthorized removal of SNM and facilitate 5 the location and recovery of missing SNM. To achieve these objectives, the physical security 6 measures shall provide for (1) early detection and assessment of unauthorized access or 7 activities by an external adversary within the controlled access area containing SNM, (2) early 8 detection of removal of SNM by an external adversary from a controlled access area, 9 (3) assurance of proper placement and transfer of custody of SNM, and (4) response to 10 indications of an unauthorized removal of SNM and then notification to the appropriate response 11 forces of its removal to facilitate its recovery.
12 13 In accordance with 10 CFR 70.22(k), each applicant planning to possess or use more than 14 10 kilograms of Category III quantities of SNM must include a physical security plan that 15 demonstrates how the applicant plans to meet the requirements in 10 CFR 73.67(f) and (g), as 16 appropriate. Applicants or licensees shall retain a copy of the physical security plan as a record 17 for the period during which the licensee possesses Category III quantities of SNM, and if any 18 portion of the plan is superseded, retain that superseded portion of the plan for 3 years after the 19 effective date of the change. The regulations in 10 CFR 73.67(c) require the licensee to submit 20 its security plan to the NRC for review and approval and to carry out the approved security plan.
21 Applicants planning to possess or use less than 10 kilograms of Category III quantities of SNM 22 are still required to carry out the security measures in 10 CFR 73.67(f) and (g) but do not need 23 to develop and submit a security plan to the NRC. In such cases, the applicant may choose to 24 prepare a security plan or use procedures, or both, to carry out the required security measures.
25 26 Regulatory Guide 5.59 contains additional guidance on developing a Category III physical 27 protection program.
28 29 9.2.3    Physical Protection Requirements at Fixed Sites 30 31 In accordance with 10 CFR 73.67(f), applicable to fixed sites, licensees shall only use and store 32 Category III quantities of SNM in a controlled access area (10 CFR 73.67(f)(1)). A controlled 33 access area is any temporarily or permanently established area that is clearly demarcated, 34 typically with a barrier (e.g., fence or wall), which affords isolation of the material or persons 35 within the controlled access area. The controlled access area shall be monitored with an 36 intrusion alarm or other device or procedures to detect unauthorized penetration or activities 37 (10 CFR 73.67(f)(2)).
38 39 Licensees shall establish a security organization to consist of at least one watchman per shift 40 able to assess and respond to any unauthorized penetrations or activities in the controlled 41 access areas (10 CFR 73.67(f)(3)). Licensees shall establish and maintain written response 42 procedures for dealing with threats of thefts or thefts of Category III quantities of SNM 43 (10 CFR 73.67(f)(4)). The licensee shall retain a copy of the response procedures for 3 years 44 thereafter. Copies of superseded material must be retained for 3 years after each change.
45 46 9.2.4    Physical Protection Requirements in Transit 47 48 In accordance with 10 CFR 73.67(g), applicable to transit, licensees that transport or deliver to a 49 carrier for transport Category III quantities of SNM shall provide advance notification to the 50 receiver of any planned shipments, specifying the mode of transport, estimated time of arrival, 51 location of the nuclear material transfer point, name of carrier, and transport identification.
86
 
1 Before beginning the shipment, licensees shall receive confirmation from the receiver that the 2 receiver will be ready to accept the shipment at the planned time and location and 3 acknowledges the specified mode of transport. Before shipment, licensees shall check the 4 integrity of the container and locks or seals. Transport of Category III quantities of SNM shall be 5 in a tamper-indicated sealed container (10 CFR 73.67(g)(1)). Licensees receiving Category III 6 quantities of special nuclear material shall check the integrity of the containers and seals upon 7 receipt of the shipment and notify the shipper of receipt of the material (10 CFR 73.67(g)(2)).
8 9 In-transit physical protection of Category III quantities of SNM must be arranged by either the 10 shipper or the receiver. A licensee that transports, exports, or delivers to a carrier for transport 11 of Category III quantities of SNM shall arrange for the in-transit physical protection of the 12 materials unless the receiver is a licensee and has agreed in writing to arrange for the in-transit 13 physical protection (10 CFR 73.67(g)(1)). The receiver shall arrange for the in-transit physical 14 protection unless the shipper is a licensee and has agreed in writing to arrange for the in-transit 15 physical protection (10 CFR 73.67(g)(2)). Arrangements shall include immediate notification of 16 the arrival of the shipment at its destination, or if the shipment is lost or unaccounted for after 17 the estimated time of arrival at its destination (10 CFR 73.67(g)(3)).
18 19 Licensees shall establish and maintain written response procedures for dealing with threats of 20 thefts or thefts of SNM. Response procedures shall include immediate initiation of a trace 21 investigation of any shipment that is determined to be lost or unaccounted for after a reasonable 22 time beyond the estimated arrival time. Response procedures shall also include notification of 23 the NRC Operations Center within 1 hour after the discovery of the loss of the shipment and 24 within 1 hour after recovery of or accounting for such lost shipment (10 CFR 73.67(g)(3)). The 25 licensee shall retain a copy of required documentation as a record for 3 years after the close of 26 the period for which the licensee possesses the SNM material under each license for which the 27 original procedures were developed, and copies of superseded material must be retained for 28 3 years after each change (10 CFR 73.67(g)(3)).
29 30 9.3      Plutonium-238 and Plutonium/Beryllium Sealed Sources 31 32 9.3.1    Applicability 33 34 The regulations in 10 CFR 37.5, Definitions, define a Category 1 and a Category 2 quantity of 35 radioactive material as follows:
36 37          Category 1 quantity of radioactive material means a quantity of radioactive 38          material meeting or exceeding the category 1 threshold in Table 1 of Appendix A 39          to this part. This is determined by calculating the ratio of the total activity of each 40          radionuclide to the category 1 threshold for that radionuclide and adding the 41          ratios together. If the sum is equal to or exceeds 1, the quantity would be 42          considered a category 1 quantity. Category 1 quantities of radioactive material do 43          not include the radioactive material contained in any fuel assembly, 44          subassembly, fuel rod, or fuel pellet.
45 46          Category 2 quantity of radioactive material means a quantity of radioactive 47          material meeting or exceeding the category 2 threshold but less than the 48          category 1 threshold in Table 1 of Appendix A to this part. This is determined by 49          calculating the ratio of the total activity of each radionuclide to the category 2 50          threshold for that radionuclide and adding the ratios together. If the sum is equal 51          to or exceeds 1, the quantity would be considered a category 2 quantity.
87
 
1        Category 2 quantities of radioactive material do not include the radioactive 2        material contained in any fuel assembly, subassembly, fuel rod, or fuel pellet.
3 4 The P-238 quantity in 10 CFR 150.11 (and Pu-238 sources in quantities greater than 5 16.2 curies), and Pu-239/beryllium sources greater than 16.2 curies, as well, are subject to the 6 requirements in 10 CFR Part 37. These requirements may also apply to applicants or licensees 7 requesting a new license (or renewal of an existing license) to possess byproduct material in 8 addition to SNM.
9 10 9.3.2  General 11 12 In accordance with 10 CFR Part 37, licensees authorized to possess Category 1 or Category 2 13 quantities of radioactive material listed in Appendix A, Category 1 and Category 2 Threshold, 14 to 10 CFR Part 37 (Appendix A lists the radionuclides and the threshold quantities necessary to 15 be subject to 10 CFR Part 37 under Category 1 or 2). The 10 CFR Part 37 regulations require 16 licensees to establish, carry out, and maintain enhanced security procedures that meet 17 increased control requirements to ensure physical protection of the radioactive material.
18 19 The Category 1 and 2 radioactive material covered by 10 CFR Part 37 are also subject to the 20 regulations in 10 CFR 20.2207, Reports of transactions involving nationally tracked sources, 21 which require that each licensee that manufactures, transfers, receives, disassembles, or 22 disposes of a nationally tracked source shall complete and submit a National Source Tracking 23 System (NSTS) Transaction Report. The NSTS is a secure, accessible, and easy-to-use 24 computer system that tracks high-risk radioactive sources from the time they are manufactured 25 or imported through the time of their disposal or export, or until they decay enough to no longer 26 be of concern.
27 28 The NRC provides additional guidance on the development of a 10 CFR Part 37 security 29 program in NUREG-2155, Implementation Guidance for 10 CFR Part 37, Physical Protection of 30 Category 1 and Category 2 Quantities of Radioactive Material.
31 32 9.3.3  Background Investigations and Access Authorization Program 33 34 In accordance with 10 CFR Part 37, Subpart B, Background Investigations and Access 35 Authorization Program, licensees must establish an access authorization to ensure that 36 individuals who have unescorted access to Category 1 and 2 quantities of radioactive material 37 are trustworthy and reliable. The program must include the following elements:
38 39
* Designate reviewing officials after making background investigations, in accordance with 40        10 CFR 37.23(b)(2), and determine their trustworthiness and reliability, in accordance 41        with 10 CFR 37.23(b)(1). A reviewing official is defined in 10 CFR 37.5 as the individual 42        who shall make the trustworthiness and reliability determination of an individual to 43        decide whether the individual may have, or continue to have, unescorted access to the 44        Category 1 or Category 2 quantities of radioactive materials in the licensees 45        possession.
46 47
* Conduct background investigations of persons to be granted unescorted access, 48        including verification of identity, education, and employment history; criminal records; 49        and fingerprint checks, and document the conclusions, in accordance with 50        10 CFR 37.23(a).
51 88
 
1
* Maintain lists of persons who are approved for unescorted access, in accordance with 2        10 CFR 37.23(e)(5).
3 4
* Provide security training, in accordance with 10 CFR 37.23(a)(2).
5 6
* Protect personal information against unauthorized disclosure, in accordance with 7        10 CFR 37.31(a).
8 9 9.3.4  Physical Protection Requirements for Use and Storage 10 11 In accordance with 10 CFR Part 37, Subpart C, Physical Protection Requirements During Use, 12 licensees must establish a physical protection program to monitor and immediately detect and 13 respond to any actual or attempted unauthorized access to radioactive material in use or 14 storage. The program must include the following elements:
15 16
* a written security plan in accordance with 10 CFR 37.43(a) 17 18
* implementing procedures in accordance with 10 CFR 37.43(b) 19 20
* security training in accordance with 10 CFR 37.43(c) 21 22
* protection of security-related information against unauthorized disclosure in accordance 23        with 10 CFR 37.43(d) 24 25
* coordination with local law enforcement authorities in accordance with 10 CFR 37.45 26 27
* establishment of security zones that define the area for the physical protection of 28        Category 1 or Category 2 quantities of radioactive material in accordance with 29        10 CFR 37.47 30 31
* establishment of continuous monitoring capability to promptly detect unauthorized entry 32        into security zones in accordance with 10 CFR 37.49(a) 33 34
* procedures for prompt assessment and immediate response to any attempted or actual 35        unauthorized entry, including requesting a response from the local law enforcement 36        agency in accordance with 10 CFR 37.43(b) 37 38
* notification of local law enforcement agencies and the NRC of security events in 39        accordance with 10 CFR 37.57 40 41
* maintenance, testing, and calibration of alarm and communication systems in 42        accordance with 10 CFR 37.51 43 44
* physical controls to secure mobile devices against unauthorized removal in accordance 45        with 10 CFR 37.53 46 47 9.3.5  Physical Protection during Transit 48 49 In accordance with 10 CFR Part 37, Subpart D, Physical Protection in Transit, licensees must 50 provide for physical protection of radioactive material in transit. The requirements apply to 89
 
1 material delivered to a carrier for transport, as well as cases in which the licensee transports 2 material. Normally, the shipping licensee has primary responsibility, but certain requirements 3 involve coordination with the recipient. The physical protection requirements for transport of 4 Category 2 quantities include the following:
5 6
* In accordance with 10 CFR 37.71(b), the licensee must verify the validity of the 7          recipients license authorization, using a license verification system, before transfer.
8 9
* In accordance with 10 CFR 37.75(a), the expected arrival times must be verified 10          between the shipper and recipient, and the recipient must notify the shipper when the 11          shipment arrives. In accordance with 10 CFR 37.79(c), the licensee must immediately 12          investigate lost or unaccounted-for shipments.
13 14
* In accordance with 10 CFR 37.79(a)(2), a licensee that transports material shall maintain 15          constant control or surveillance over the material and must have the capability for 16          immediate communication to summon a response or assistance.
17 18
* In accordance with 10 CFR 37.79(a)(3), if a licensee delivers material to a carrier for 19          transport, the carrier must have an established package tracking system, require an 20          authorized signature for delivery or return, and maintain constant control or surveillance 21          over shipments, with the capability for immediate communication to summon a response 22          or assistance.
23 24
* In accordance with the requirements in 10 CFR 37.81, Reportable safeguards events, 25          the shipping licensee must promptly notify local law enforcement authorities and the 26          NRC in cases of lost shipment or attempted or actual theft or diversion of shipments.
27 28 9.3.6    Additional Information 29 30 Physical security requirements applicable to SNM vary depending on the category of the SNM 31 to be possessed by an applicant or already possessed by a licensee. Physical security 32 requirements applicable to Category I SNM are not necessarily applicable to Category II or to 33 Category III (i.e., additional requirements apply as the category of SNM increases from III to I).
34 Because of the complexity of determining the physical security requirements that may apply to 35 different applicants requesting an SNM license, an example is provided to demonstrate how an 36 applicant may determine the physical requirements needed to protect the licensed material.
37 38 As an illustrative example, assume a greater than critical mass licensee has 800 grams of 39 U-235 enriched to 19.5 percent and 20 grams of Pu-239.
40 41 Because the uranium is enriched to less than 20 percent and has a total mass of less than 42 1 kilogram, it would be below the Category III limits, and thus, would not need additional 43 physical protection beyond the control of licensed material required by 10 CFR Part 20, 44 Subpart I. If the uranium were enriched to greater than 20 percent, it would need to be protected 45 in accordance with 10 CFR 73.67(f).
46 47 In this example, 20 grams of plutonium-239 would be a Category III quantity of SNM and would 48 need to be protected in accordance with 10 CFR 73.67(f). As such, the Pu-239 would need to 49 be used and stored in a controlled access area that is monitored with an intrusion alarm or other 50 device or procedures to detect unauthorized penetration or activities. Although a licensee would 90
 
1 not need to submit a security plan to the NRC for approval, it would need to develop, maintain, 2 and carry out response procedures. These procedures should describe the security 3 organization, which would include at least one individual to assess and respond to any 4 unauthorized activities. In general, response efforts would be supplemented by and coordinated 5 with local law enforcement. If the plutonium were Pu-238 (20 grams of Pu-238 equates to 6 346 curies, which is above the Category II threshold), it would need to be protected in 7 accordance with 10 CFR Part 37.
8 9 9.4    References 10 11 Code of Federal Regulations, Title 10 (10 CFR), Energy. Chapter I, Part 37, Physical 12 Protection of Category 1 and Category 2 Quantities of Radioactive Material.
13 14 10 CFR, Part 73, Physical Protection of Plants and Materials.
15 16 U.S. Nuclear Regulatory Commission (NRC), Standard Format and Content for a Licensee 17 Physical Security Plan for the Protection of Special Nuclear Material of Moderate or Low 18 Strategic Significance, Regulatory Guide 5.59.
19 20 NRC, Implementation Guidance for 10 CFR Part 37, Physical Protection of Category 1 and 21 Category 2 Quantities of Radioactive Material, NUREG-2155, March 2022.
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1                            10      EMERGENCY MANAGEMENT 2
3 Regulatory Requirements: Title 10 of the Code of Federal Regulations (10 CFR) 70.22(i);
4 10 CFR 70.24, Criticality accident requirements.
5 6 Regulatory Criteria: Applications to possess enriched uranium or plutonium above certain 7 thresholds must contain either (1) an evaluation showing that public doses would not exceed 8 specified limits, or (2) an emergency plan.
9 10 Discussion: The regulations requiring emergency planning with respect to possession of 11 special nuclear material (SNM) are contained in 10 CFR Part 70, Domestic Licensing of Special 12 Nuclear Material.
13 14 In accordance with 10 CFR 70.22(i)(1), each application to possess one of the following 15 materials must address this emergency management requirement:
16 17
* enriched uranium or plutonium for which a criticality accident alarm is required 18        (10 CFR 70.24(a))
19 20
* uranium hexafluoride in excess of 50 kilograms in a single container or 1,000 kilograms 21        total 22 23
* more than 2 curies of plutonium in unsealed form or on foils or plated sources 24 25 If one of these possession limit thresholds is exceeded, in accordance with the requirements in 26 10 CFR 70.22(i)(1), the application must contain one of the following:
27 28
* an evaluation showing that the maximum dose to a member of the public offsite due to a 29        release of radioactive materials would not exceed 1 rem effective dose equivalent, or an 30        intake of 2 milligrams of soluble uranium, in accordance with the requirements in 31        10 CFR 70.22(i)(1(i) 32 33
* an emergency plan for responding to the radiological hazards of an accidental release of 34        special nuclear material and to any associated chemical hazards directly incident 35        thereto, in accordance with the requirements in 10 CFR 70.22(i)(1)(ii); if subject to the 36        requirements in 10 CFR 70.24, Criticality accident requirements, the emergency plan 37        requirements in 10 CFR 70.24(a)(3) also apply 38 39 In accordance with the requirements in 10 CFR 70.22(i)(2)(i) through (vi), one or more of the 40 factors stated below may be used to support an evaluation of offsite consequences:
41 42        (i) The radioactive material is physically separated so that only a portion could 43              be involved in an accident; 44 45        (ii) All or part of the radioactive material is not subject to release during an 46              accident or to criticality because of the way it is stored or packaged; 47 48        (iii) In the case of fires or explosions, the release fraction would be lower than 49              0.001 due to the chemical or physical form of the material; 50 93
 
1        (iv) The solubility of the material released would reduce the dose received; 2
3        (v) The facility design or engineered safety features in the facility would cause 4              the release fraction to be lower than 0.001; 5
6        (vi) Operating restrictions or procedures would prevent a release large enough to 7              cause a member of the public offsite to receive a dose exceeding 1 rem 8              effective dose equivalent; or, 9
10        (vii) Other factors appropriate for the specific facility.
11 12 10.1  Emergency Plan 13 14 Regulatory Guide 3.67, Standard Format and Content for Emergency Plans for Fuel Cycle and 15 Materials Facilities, includes guidance if 10 CFR 70.22(i)(1)(ii) requires an emergency plan.
16 The regulations in 10 CFR 70.22(i)(3) state that an emergency plan must contain the following 17 information:
18 19        (i)      Facility description. A brief description of the applicants/licensees facility 20                  and area near the site.
21 22        (ii)      Types of accidents. An identification of each type of radioactive materials 23                  accident for which protective actions may be needed.
24 25        (iii)    Classification of accidents. A classification system for classifying 26                  accidents as alerts or site area emergencies.
27 28        (iv)      Detection of accidents. Identification of the means of detecting each type 29                  of accident promptly.
30 31        (v)      Mitigation of consequences. A brief description of the means and 32                  equipment for mitigating the consequences of each type of accident, 33                  including those provided to protect workers onsite, and a description of 34                  the program for maintaining the equipment.
35 36        (vi)      Assessment of releases. A brief description of the methods and 37                  equipment to assess releases of radioactive materials.
38 39        (vii)    Responsibilities. A brief description of the responsibilities of the 40                  applicants personnel should an accident occur, including identification of 41                  personnel responsible for promptly notifying offsite response 42                  organizations and the NRC [U.S. Nuclear Regulatory Commission]; also, 43                  responsibilities for developing, maintaining, and updating the plan.
44 45        (viii)    Notification and coordination. A commitment to and a brief description of 46                  the means to promptly notify offsite response organizations and request 47                  offsite assistance, including medical assistance for the treatment of 48                  contaminated injured onsite workers when appropriate. A control point 49                  must be established. The notification and coordination must be planned 50                  so that unavailability of some personnel, parts of the facility, and some 51                  equipment will not prevent the notification and coordination. The 94
 
1              applicant/licensee shall also commit to notify the NRC operations center 2              immediately after notification of the appropriate offsite response 3              organizations and not later than 1 hour after the licensee declares an 4              emergency.1 5
6        (ix)  Information to be communicated. A brief description of the types of 7              information on facility status, radioactive releases, and recommended 8              protective actions, if necessary, to be given to offsite response 9              organizations and to the NRC.
10 11        (x)    Training. A brief description of the frequency, performance objectives and 12              plans for the training that the applicant/licensee will provide workers on 13              how to respond to an emergency including any special instructions and 14              orientation tours the licensee would offer to fire, police, medical and other 15              emergency personnel. The training shall familiarize personnel with site-16              specific emergency procedures. Also, the training shall thoroughly 17              prepare site personnel for their responsibilities in the event of accident 18              scenarios postulated as most probable for the specific site, including the 19              use of team training for such scenarios.
20 21        (xi)  Safe shutdown. A brief description of the means of restoring the facility to 22              a safe condition after an accident.
23 24        (xii)  Exercises. Make provisions for quarterly communications checks with 25              offsite response organizations and biennial onsite exercises to test 26              response to simulated emergencies. Quarterly communications checks 27              with offsite response organizations must include the check and update of 28              all necessary telephone numbers. The licensee shall invite offsite 29              response organizations to participate in the biennial exercises.
30              Participation of offsite response organizations in biennial exercises 31              although recommended is not required. Exercises must use accident 32              scenarios postulated as most probable for the specific site and the 33              scenarios shall not be known to most exercise participants. The licensee 34              shall critique each exercise using individuals not having direct 35              implementation responsibility for the plan. Critiques of exercises must 36              evaluate the appropriateness of the plan, emergency procedures, 37              facilities, equipment, training of personnel, and overall effectiveness of 38              the response. Deficiencies found by the critiques must be corrected.
39 40        (xiii) Hazardous chemicals. A certification that the applicant has met its 41              responsibilities under the Emergency Planning and Community Right-to-42              Know Act of 1986, Title III, Pub. L. 99-499, if applicable to the applicant's 43              activities at the proposed place of use of the SNM.
44 45 In accordance with the regulations in 10 CFR 70.22(i)(4), an applicant shall allow the offsite 46 response organizations, expected to respond in case of an accident, 60 days to comment on the 47 licensees emergency plan before submitting it to the NRC. The licensee shall provide any 48 comments received within the 60 days to the NRC with the emergency plan.
49 50 51 95
 
1 In accordance with 10 CFR 70.32(i), a licensee may change an approved emergency plan 2 without prior Commission approval if the changes do not decrease the effectiveness of the plan 3 and the changes are provided to the Commission within 6 months. Applications including an 4 emergency plan should describe how the applicant will review proposed changes to the 5 emergency plan and the level of management approval the applicant requires before making 6 changes without prior Commission approval.
7 8 Chapter 8 of NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License 9 Applications, or Regulatory Guide 3.67 provide additional guidance for the development and 10 the review of emergency plans that the regulations in 10 CFR 70.22(i)(1) would require.
11 12 10.2    References 13 14 Code of Federal Regulations, Chapter I, Title 10, Energy, Part 70, Domestic Licensing of 15 Special Nuclear Material.
16 17 U.S. Nuclear Regulatory Commission (NRC), Standard Review Plan for Fuel Cycle Facilities 18 License Applications, NUREG-1520.
19 20 NRC, Standard Format and Content for Emergency Plans for Fuel Cycle and Materials 21 Facilities, Regulatory Guide 3.67.
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1                                              APPENDIX A 2        TYPICAL DUTIES AND RESPONSIBILITIES OF THE RADIATION 3                                        SAFETY OFFICER 4
5 The radiation safety officer (RSO) is responsible for the overall radiation safety at a licensed 6 facility or when the licensee otherwise engages in a licensed activity. Regulatory Guide 8.10, 7 Operating Philosophy for Maintaining Occupational and Public Radiation Exposures As Low As 8 Is Reasonably Achievable, discusses methods and procedures that the U.S. Nuclear 9 Regulatory Commission (NRC) staff considers acceptable for maintaining radiation exposures to 10 occupational workers and the public as low as is reasonably achievable (ALARA). This 11 regulatory guide also discusses the applicable regulations in Title 10 of the Code of Federal 12 Regulations (10 CFR) Part 20, Standards for Protection against Radiation, and regulatory 13 guidance for implementing an adequate safety program to achieve occupational doses that are 14 ALARA and for adequately protecting the public and the environment from radiation hazards.
15 The regulatory guide also discusses the main responsibilities of the RSO for ensuring 16 radiological safety and compliance with NRC regulations and the conditions of the license.
17 18 As discussed in Regulatory Guide 8.10, the RSOs general responsibilities should include the 19 following, among others:
20 21
* developing and planning radiological work activities and preparing procedures and other 22          written documentation (e.g., radiation work permits) for achieving ALARA goals and 23          ensuring such procedures are properly followed 24 25
* assuring that the proper radiation protection instrumentation, equipment, and supplies 26          are available at workplaces, are in good working order, and are used properly, and that 27          procedures for the use of instrumentation and equipment are available and properly 28          followed 29 30
* performing annual audits to review the effectiveness of the ALARA program [NOTE: this 31          audit may be performed as an integral part of the reviews performed to meet the 32          requirements in 10 CFR 20.1101(c) for a periodic (at least annual) review of the overall 33          radiation protection program]
34 35
* requiring, where practical, modifications to standard operating procedures, equipment, 36          and facilities that will substantially reduce occupational and public exposures 37 38
* ensuring that proper focus is given to the source of licensee radiation exposures in the 39          facility or other licensed activity by location, operation, and job category and maintaining 40          awareness of trends in occupational and public exposures 41 42
* investigating unexpected exposures to determine the causes, taking steps to reduce the 43          likelihood of similar occurrences in the future, and documenting conclusions and 44          corrective actions 45 46
* assessing management of radiological work controls if the planned controls 47          (e.g., radiation work permits, ALARA plans, work order instructions, radiological hold 48          points, and stop work criteria) are not being implemented properly 49 A-1
 
1
* routinely reviewing ALARA plans to ensure that they are effective in maintaining both 2  occupational and public doses ALARA A-2
 
1                                            APPENDIX B 2                  FACILITIES AND EQUIPMENT CONSIDERATIONS 3
4 This appendix lists the topics that should be considered when developing a description of the 5 facilities and equipment that a licensee will use, or otherwise have available. Not every applicant 6 will need to address each topic in its application.
7 8
* Restricted areas are defined as areas to which access is limited by the licensee to 9          protect individuals against undue risks from exposure to radiation and radioactive 10          materials. The application should contain detailed descriptions and diagrams of the 11          facilities, including information about the shielding properties of the construction 12          materials used. Scaled drawings and sketches should be submitted showing the 13          relationship between restricted areas and unrestricted areas and the location of all 14          pertinent safety-related equipment.
15 16
* Bench top or open work areas may be used for sealed sources, for small quantities of 17          solid materials in a form not likely to become airborne or dispersed, and for small 18          quantities of liquids of such low volatility as not to cause airborne contamination or 19          toxicity problems. Trays and absorbent surface covers to catch and retain spilled liquids 20          should be used on these open work surfaces and inside closed systems discussed 21          below. Surfaces should be smooth and nonporous to facilitate decontamination.
22 23
* Radioactive materials that are handled or used in unsealed forms should be confined to 24          control the release of material and to prevent the spread of contamination. Gaseous, 25          volatile, and fine-particulate solid materials should be handled in closed or isolated 26          systems such as fume hoods or glove boxes with controlled, and possibly filtered, 27          exhaust systems.
28 29
* Chemical-type fume hoods provide a working area with controlled inward airflow from 30          the room to the hood exhaust system. Hoods are used for gases, for unsealed volatile 31          licensed materials, and for processes such as evaporation that may release gases and 32          vapors. Fume hoods provide emergency ventilation and exhaust for unplanned releases, 33          such as accidental spills and ruptures, as well as routine exhaust of effluents. Filters 34          may be required in the exhaust stream unless monitoring or calculations, or both, 35          demonstrate that any planned or likely effluent will be in accordance with the limits found 36          in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for 37          Protection against Radiation, Appendix B, Annual Limits on Intake (ALIs) and Derived 38          Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent 39          Concentrations; Concentrations for Release to Sewerage.
40 41
* Glove boxes are sealed boxes with transparent viewing windows, sealable ports, or 42          doors for transferring materials and equipment, and gloves sealed to the box through 43          which licensed materials are handled. Glove boxes are used for the containment during 44          storage and the use of liquids and solids that can become airborne particulates or 45          aerosols. Glove boxes can be closed or exhausted, with filtration systems if appropriate, 46          to prevent contamination.
47 48
* Sink faucets should be designed, where possible, for operation by foot, knee, or elbow 49          rather than by hand.
50 B-1
 
1
* Plumbing and ductwork should be designed to avoid radioactive contamination buildup.
2  This buildup of contamination can create external radiation exposure hazards and 3  problems for decommissioning.
4 5
* Shielding consisting of lead or other high-density material in the form of bricks, panels, 6  L-shields, storage containers, or other shapes may be used on bench tops, in fume 7  hoods, or in glove boxes to reduce radiation exposure from gamma-emitting radioactive 8  materials. Similarly, shielding of low atomic number material, such as high-density 9  plastic, may be used to reduce the exposure from high-energy beta-emitting materials.
10  Shielded shipping containers are frequently used for continued storage after receipt of 11  materials.
12 13
* A particular sink should be designated for disposal of liquid radioactive waste to the 14  sanitary sewerage system. In some cases, depending on the number of users and 15  distance between areas of use, more than one sink may need to be designated.
16 17
* Labeled waste containers should be used. These containers may be shielded as 18  necessary, placed near the waste-generating areas and away from areas frequently 19  occupied by personnel. Additionally, these containers should be effectively enclosed to 20  prevent airborne contamination from radioactive materials deposited.
21 22
* Remote handling tools, such as forceps or extension handles, should be used to provide 23  distance in the handling of radioactive materials (as low as is reasonably achievable). In 24  addition, shielded handling devices, such as shielded syringes, can be used to protect 25  workers from materials that cannot be handled remotely. Pipetting should be done using 26  only appropriate devices. Pipetting by mouth should be strictly forbidden.
27 28
* Where appropriate, ventilation systems should be designed such that, in the event of an 29  accident, they can be shut down to prevent the spread of radioactivity.
30 31
* Designated areas should be provided for coats and personal belongings, to avoid 32  contamination.
33 34
* Areas with background radiation levels should be designated for personnel dosimetry 35  storage when not in use.
36 37
* Areas of use should be well lighted to avoid spills and other accidents that could result in 38  contamination buildup.
39 40
* Observation of activities done behind shielding with remote tools (or with extended arms 41  and hands, within limits consistent with permissible occupational exposures) can be 42  accomplished by mirrors, through shielded (e.g., leaded glass) windows, through 43  transparent plastic beta shields, or by remote video monitoring.
44 45
* The combination of containment, shielding, and handling devices proposed for any use 46  of radioactive materials should be appropriate to the type and quantity of materials to be 47  used and to the type and duration of the designated operations.
48 B-2
 
1
* If respiratory protective equipment will be used to limit inhalation of airborne licensed 2  material, the provisions of 10 CFR Part 20, Subpart H, Respiratory Protection and 3  Controls to Restrict Internal Exposure in Restricted Areas, need to be followed.
4 5
* If waste is compacted, the applicant needs to ensure that facilities are adequate to 6  ventilate the area of compaction and that air sampling for internal exposures is available, 7  if needed, in accordance with 10 CFR 20.1204, Determination of internal exposure.
B-3
 
1                                              APPENDIX C 2        RADIATION MONITORING AND INSTRUMENT CALIBRATION 3                                                PROGRAM 4
5 The specifications in Table C-1 will help applicants and licensees choose the proper radiation 6 detection equipment for monitoring the radiological conditions at their facility or facilities.
7 8 Table C-1 Typical Survey Instruments1 9              (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,          R-R                                            N/A X-ray Count Rate Meters GM                            Alpha          All energies (dependent on window                Moderate thickness)
Beta            All energies (dependent on window                Moderate thickness)
Neutron                      Neutron Gamma          All energies                                    < 1%
NaI Scintillator              Gamma          All energies (dependent on crystal              Moderate thickness)
ZnS                          Alpha          All energies                                    Moderate Plastic Scintillator          Beta            C-14 or higher (dependent on window              Moderate thickness)
Stationary Instruments Used To Measure Wipe, Bioassay, and Effluent Samples Detectors                Radiation                    Energy Range                      Efficiency Liquid Scintillation          Alpha          All energies                                    High Counting*
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%
10 1 The Health Physics & Radiological Health Handbook, 4th Edition, edited by Thomas E. Johnson and Brian K. Birky, 11  issued 2012 (except for
* items).
C-1
 
1 Calibration 2
3
* Calibration must produce readings within +/-20 percent of the actual values over the range 4        of the instrument.
5 6
* Calibration of liquid scintillation counters will include quench correction.
7 8 Calibration Records 9
10 Calibration records, for all survey instruments, should state the procedure used and the data 11 obtained. The description of the calibration should include the following:
12 13
* the owner or user of the instrument 14 15
* a description of the instrument, including the manufacturers name, model number, serial 16        number, and type of detector 17 18
* a description of the calibration source, including the exposure rate at a specified 19        distance or activity on a specified date 20 21
* for each calibration point, the calculated exposure rate or count rate, the indicated 22        exposure rate or count rate, the deduced correction factor (the calculated exposure rate 23        or count rate divided by the indicated exposure rate or count rate), and the scale 24        selected on the instrument 25 26
* for instruments with external detectors, the angle between the radiation flux field and the 27        detector (i.e., parallel or perpendicular) 28 29
* for instruments with internal detectors, the angle between radiation flux field and a 30        specified surface of the instrument 31 32
* for detectors with removable shielding, an indication whether the shielding was in place 33        or removed during the calibration procedure 34 35
* the exposure rate or count rate from a check source, if used 36 37
* the name of the person who did the calibration and the date it was done 38 39 The following information should be attached to the instrument as a calibration sticker or tag:
40 41
* for exposure rate meters, the source isotope used to calibrate the instrument (with 42        correction factors) for each scale 43 44
* the efficiency of the instrument, for each isotope the instrument will be used to measure 45        (if efficiency is not calculated before each use) 46 47
* for each scale or decade not calibrated, an indication that the scale or decade was 48        checked only for function but not calibrated 49 50 C-2
 
1
* the date of calibration and the next calibration due date 2
3
* the apparent exposure rate or count rate from the check source, if used 4
5 Air Sampler Calibration 6
7 To assess accurately the air concentration of radioactive materials in a given location, the 8 volume of air sampled and the quantity of contaminant in the sample must be determined.
9 Accurate determination of the volume of air sampled requires standard, reproducible, and 10 periodic calibration of the air metering devices that are used with air sampling instruments.
11 12 The American Conference of Governmental Industrial Hygienists Air Sampling Instruments for 13 Evaluation of Atmospheric Contaminants, 9th Edition, issued 2001, provides guidance on total 14 air sample volume calibration methods acceptable to the U.S. Nuclear Regulatory Commission 15 staff, as supplemented below.
16 17 Frequency of Calibration 18 19
* A licensee committed to a routine or emergency air sampling program should do an 20          acceptable calibration of all airflow or volume metering devices at least annually (see 21          Regulatory Guide 8.25, Air Sampling in the Workplace).
22 23
* Special calibrations should be done at any time there is reason to believe that the 24          operating characteristics of a metering device have been changed, by repair or 25          alteration, or whenever system performance is observed to have changed significantly.
26 27
* Routine instrument maintenance should be done as recommended by the manufacturer.
28 29
* Primary or secondary standard instruments used to calibrate air sampling instruments 30          should be inspected frequently for consistency of performance.
C-3
 
1                                            APPENDIX D 2                  GUIDANCE FOR MONITORING REQUIREMENTS 3
4 In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 20.1502, Conditions 5 requiring individual monitoring of external and internal occupational dose, licensees are 6 required to monitor exposures to radiation and radioactive materials at levels sufficient to 7 demonstrate compliance with the occupational dose limits of 10 CFR Part 20, Standards for 8 Protection against Radiation. According to 10 CFR 20.1502, as a minimum, monitoring is 9 required for individuals likely to receive in 1 year from sources external to the body, a dose in 10 excess of 10 percent of the applicable regulatory limits in 10 CFR 20.1201, Occupational dose 11 limits for adults. To demonstrate that dosimetry is not required, a licensee needs to do a 12 prospective evaluation to demonstrate that its workers are not likely to exceed 10 percent of the 13 applicable annual limits. Regulatory Guide 8.34, Monitoring Criteria and Methods To Calculate 14 Occupational Radiation Doses, provides guidance for prospectively evaluating the need to 15 monitor an individuals occupational radiation dose. If the prospective evaluation shows that an 16 individual is not likely to receive in a year a dose that exceeds the criteria in 10 CFR 20.1502, 17 then monitoring is not required, and the recordkeeping requirement in 10 CFR 20.2106, 18 Records of individual monitoring results, and the reporting requirements in 10 CFR 20.2206, 19 Reports of individual monitoring, are not applicable. However, for the individual that already 20 has received greater than the dose criteria in 10 CFR 20.1502 from prior employment in the 21 current monitoring year, the NRC requires subsequent employers to monitor any additional 22 radiation exposure.
23 24 The most common way that individuals might exceed 10 percent of the applicable limits is by 25 being involved in the processing of sealed sources or unsealed material, or both (e.g., assembly 26 lines, manufacturing processes and quality control activities). This could include internal 27 radioactive uptake, as well as external radiation exposure. However, for many processes, even 28 these activities result in the individual receiving minimal doses. Before allowing workers to do 29 these tasks, a licensee will need to evaluate the doses that its workers might receive to assess 30 whether dosimetry is required. This is known as a prospective evaluation.
31 32 Example 33 34 To demonstrate compliance with the occupational dose limits in 10 CFR 20.1201, Occupational 35 dose limits for adults, a university has estimated the doses to the extremities and whole body of 36 a person doing foil activation experiments using plutonium (Pu)-238/beryllium (Be) sources.
37 Each Pu-238/Be source is authorized to contain up to 352 grams of plutonium. The university 38 based its estimate on observations of individuals doing the recommended procedure according 39 to accepted radiation safety practices. The university had the following types of information:
40 41
* time needed to do the entire procedure (e.g., 15 minutes) 42 43
* expected dose rate received by the whole body of the individual, associated with the 44          shielded source, and determined using measured or manufacturer-determined data 45          (e.g., 0.02 millisievert (mSv)/h (2 millirem (mrem)/h) at 46 centimeters (18.1 inch) from 46          the shield) 47 48
* time the hands were exposed to the shielded source (e.g., 6 minutes) 49 D-1
 
1
* expected dose rate received by the extremities of the individual, associated with the 2          shielded source, and determined using measured or manufacturer-determined data on 3          contact with the shield (e.g., 0.15 mSv/h (15 mrem/h)).
4 5 From this information, the university estimated that the individual doing each foil activation 6 experiment could receive both of the following:
7 8
* less than 0.005 mSv (0.5 mrem) total effective dose equivalent (whole body) 9 10
* 0.015 mSv (1.5 mrem) to the hands 11 12 In accordance with 10 CFR 20.1201(a)(1)(i), the total effective dose equivalent (whole body) 13 limit for individual adults is equal to 50 mSv (5 rem) per year. Ten percent of that value is 5 mSv 14 (500 millirem) per year. If one of these procedures delivers 0.005 mSv (0.5 mrem), then an 15 individual could do 1,000 of these experiments each year and remain within 10 percent of the 16 applicable limit.
17 18 The applicable shallow does equivalent (extremities) is 500 mSv (50 rem) per year and 19 10 percent of that value is 50 mSv (5 rem or 5,000 millirem) per year. If one of these 20 experiments delivers 0.015 mSv (1.5 mrem), then an individual could do 3,333 of these 21 procedures each year and remain within 10 percent of the applicable limit.
22 23 Based on the above specific situation, no dosimetry is required if an individual does fewer than 24 1,000 foil activation experiments per year since the individuals radiation dose will not exceed 25 the 10 percent of the applicable limit in accordance with the criteria in 10 CFR 20.1502(a)(1).
26 27 Guidance to Licensees 28 29 Licensees that wish to demonstrate that they are not required to provide dosimetry to their 30 workers need to do prospective evaluations similar to that shown in the example above. The 31 expected dose rates, times, and distances used in the above example may not be appropriate 32 for all licensees. In their evaluations, licensees need to use information appropriate to the 33 type(s) of processes they intend to use.
34 35 Table D-1 may be helpful in doing a prospective evaluation.
36 37 Licensees should review evaluations periodically and revise them as needed. Licensees need to 38 check assumptions used in their evaluations to ensure that they are up to date and accurate.
39 For example, if workers become lax in following good radiation safety practices, perform the 40 task more slowly than estimated, use a new sealed source containing sources of different 41 activities or radionuclides, or use modified procedures, the licensee would need to make a new 42 evaluation.
43 44 45 46 47 48 49 50 51 D-2
 
1 Table D-1 Dosimetry Evaluation Dosimetry Evaluation for___________
A. Time needed to perform the entire routine procedure, _________ minutes/60 = ____ hours B. Expected whole body dose rate received by the individual, ________________ mrem/h C. Time the hands were exposed to the radiation source, ________ minutes/60 = ____ hours D. Expected extremity dose rate received by the individual, determined using the measured or manufacturer-provided data for the radiation source at the typical distance from the hands to the source, ________________ mrem/h Formula: (____ # hours in Row A) x (____ mrem/h in Row B) = (____ mrem per routine procedure) x (____ # of routine procedures each year) = ___ mrem *Whole Body Dose Formula: (____ # hours in Row C) x (____ mrem/h in Row D) = (____ mrem per routine procedure) x (____# of routine procedures each year) = ____ mrem **Extremity Dose
    *Expected whole body doses less than 500 mrem does not require dosimetry.
    **Expected extremity doses less than 5,000 mrem does not require dosimetry.
2 D-3
 
1                                            APPENDIX E 2                                          PUBLIC DOSE 3
4 This appendix describes different methods for determining radiation doses to members of the 5 public.
6 7 To achieve compliance with the requirements in Title 10 of the Code of Federal Regulations 8 (10 CFR) 20.1301, Dose limits for individual members of the public, licensees must conduct 9 operations to ensure the following:
10 11
* To the extent practical, procedures and engineering controls based upon sound radiation 12          protection principles are put in place to achieve doses to members of the public that are 13          as low as reasonably achievable (ALARA), in accordance with 10 CFR 20.1101(b).
14 15
* The radiation dose received by individual members of the public does not exceed 16          1 millisievert (mSv) (100 millirem (mrem)) in 1 calendar year resulting from the licensees 17          possession or use, or both, of licensed materials, in accordance with 10 CFR 1301(a)(1).
18 19
* The radiation dose in unrestricted areas does not exceed 0.02 mSv (2 mrem) in any 20          1 hour, in accordance with 10 CFR 1301(a)(2).
21 22
* The air emissions of radioactive materials do not result in doses greater than the 23          constraint limit of 0.1 mSv (10 mrem), in accordance with 10 CFR 1101(d).
24 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 licensed material but may work in the vicinity where such materials are used or stored.
25 Doses to Members of the Public INCLUDES doses from:                              DOES NOT INCLUDE doses from:
* Radiation and/or radioactive material
* Sanitary sewerage discharges from released by a licensee                            licensees
* Sources of radiation under the control of a
* Natural background radiation licensee
* Medical administration of radioactive
* Air effluents from sources of licensed              material radioactive materials
* Voluntary participation in medical research 26 Unrestricted areas, as defined in 10 CFR 20.1003, Definitions, are areas to which access is neither limited nor controlled by the licensee. These may include offices, shops, laboratories (where licensed material is not used or stored), areas outside buildings, property, and storage areas. The licensee is not required to control access to these areas to control the exposure of workers to radiation or radioactive materials. The regulations in 10 CFR 20.1003 also define Controlled Area, in part, as an area to which access can be limited (i.e., controlled) by the licensee for any reason (e.g., security).
27 E-1
 
1 Simple Method 2
3 The licensee may use any acceptable method from either 10 CFR 20.1302(b)(1) or 4 10 CFR 20.1302(b)(2) for determining compliance with the annual dose limits in 5 10 CFR 20.1301, Dose limits for individual members of the public. The method described in 6 10 CFR 20.1302(b)(2) limits the worst-case dose to 1 mSv (0.1 rem). Under this method, the 7 maximum concentration of radionuclides permitted at the unrestricted area boundary would 8 result in a dose to an individual of 0.5 mSv (0.05 rem), and the dose from external radiation to 9 an individual who is always present to receive the dose would be 0.5 mSv (0.05 rem). This 10 method provides licensees with an easy way to demonstrate compliance without considering 11 occupancy factors or summing of internal and external doses.
12 13 Calculation Methods 14 15 The licensee may also show compliance with 10 CFR 20.1301 by using the method described in 16 10 CFR 20.1302(b)(1). This method allows the licensee to take into account the length of time 17 an individual is receiving the dose from external sources and the concentration of radionuclides 18 in gaseous and liquid releases at the location of the individual. In practice, the licensee may 19 wish to make conservative assumptions to simplify the dose calculation. For example, if the 20 actual dose to an individual from external sources plus the maximum possible dose from 21 gaseous and liquid releases is below the public dose limit, the licensee may choose to use this 22 theoretical value rather than try to evaluate the exact concentration of radionuclides at the 23 individuals location. If the former calculation is over the limit, the licensee can show compliance 24 by doing the more difficult calculation. If appropriate, the licensee also has the option, with prior 25 U.S. Nuclear Regulatory Commission (NRC) approval, of adjusting the dose to take into account 26 the physical and chemical characteristics of its effluent.
27 28 Because, in accordance with 10 CFR 20.1302(b)(1), licensees shall demonstrate by 29 measurement or calculation that the total effective dose equivalent (TEDE) limit to the individual 30 who is likely to receive the highest dose from the licensed operation does not exceed the annual 31 dose limit, dose calculations need to take into account the worst case scenario in which an 32 individual would be located at the areas of greatest internal and external exposures. The 33 calculation could use an occupancy factor of 1, which means the individual was continuously 34 present 24 hours a day, 365 days a year. If this result shows that the limit is not exceeded, then 35 there is no need for further calculations. If the limit is exceeded with an occupancy factor of 1, 36 the licensee would have to make realistic assumptions about how many hours in a 24-hour day 37 an individual would be present at the points of highest internal and external exposures. To 38 obtain the occupancy factor, the licensee would take the number of hours of predicted 39 occupancy in a day and divide it by 24 hours, (e.g., 8 hours of predicted occupancy divided by 40 24 hours would give an occupancy factor of 8/24, or 1/3). Some licensees will require in-depth 41 reviews of their use of radioactive material to obtain appropriate assumptions for the external 42 (and internal) dose rates and occupancy factors.
43 44 Survey and Monitoring 45 46 In accordance with 10 CFR 60.132(c)(2), in part, effluent monitoring programs shall be designed 47 to measure the amount and concentration of radionuclides released. Dose rates vary, 48 depending upon the nature of the source and its condition. It may be necessary to provide 49 continuously operating dose-rate meters, or if the source is predictable, it may be adequate to 50 integrate dose rates over periods of time. The mode of discharge of airborne and waterborne 51 effluents will vary from facility to facility. Airborne effluents are most frequently discharged E-2
 
1 continuously during operations, but the operation itself may be discontinuous, whereas liquid 2 effluents may be discharged continuously or may be stored and subsequently discharged on a 3 batch basis. For each type of source and for each route of potential exposure, consider the 4 location of measurement points, whether continuous monitoring is required, the frequency of 5 sampling and measurement, and any additional information. For discharges of radionuclides, for 6 example, it may be necessary to obtain information on the chemical form, concentration, and 7 flowrate of the discharge as well as meteorological and hydrological data and other information 8 relating to the receiving environment.
9 10 Records 11 12 The regulations in 10 CFR 20.2107, Records of dose to individual members of the public, 13 require licensees to maintain records sufficient to demonstrate compliance with the dose limit for 14 individual members of the public until the NRC terminates the license. The regulations in 15 10 CFR 20.2102, Records of radiation protection programs, and 10 CFR 20.2103, Records of 16 surveys, contain additional requirements for maintaining records.
17 18 Records related to the radiation protection program must be appropriate to each licensed 19 operation or activity. Therefore, the reviewer and the applicant need to become familiar with the 20 requirements in 10 CFR Part 20, Standards for Protection against Radiation, Subpart L, 21 Records, that would be relevant to the proposed or licensed facility operations.
22 23 Regulatory Guide 8.21, Health Physics Surveys for Byproduct Material at NRC Licensed 24 Processing and Manufacturing Plants, and Appendix I, Radiation Safety Survey Topics, to 25 NUREG-1556, Consolidated Guidance about Materials Licenses, Volume 17, 26 Program-Specific Guidance About Licenses for Special Nuclear Material of Less than Critical 27 Mass, provide guidance on survey records required under 10 CFR 20.2103.
28 29 In general, records of radiation survey results should include the identification (make, model, 30 and serial number) of the instrument(s) used, date of the survey, name of the surveyors, specific 31 location where the survey was performed, items and equipment surveyed, reason for the 32 survey, background levels, and survey results (i.e., contamination levels with appropriate units).
33 The unique identification of the survey instrument allows it to be linked to its calibration and 34 maintenance records. Radiation survey records should also include a description or diagram of 35 the area surveyed, administrative action levels for controlling exposures and evaluations of the 36 effect of measured radiation levels, and documentation of any corrective action taken. The 37 radioactive material release records required pursuant to 10 CFR 20.2103(b)(4) should contain 38 adequate information to allow reasonable offsite dose assessments to be made in the future. In 39 addition, records should include information on quantities of specific radionuclides and the 40 estimated uncertainty in activity release values, if known.
41 42 The regulations in 10 CFR 20.1302, Compliance with dose limits for individual members of the 43 public, require surveys of radioactive materials in effluents released to unrestricted and 44 controlled areas to demonstrate compliance with the dose limits for individual members of the 45 public (i.e., 1 millisievert per year (mSv/yr) (100 millirems per year (mrem/yr))). Furthermore, 46 10 CFR 20.2103(b)(4) requires that records of measurement results and calculations used to 47 evaluate the release of radioactive effluents be maintained until the NRC terminates the license 48 requiring the record. Radionuclide releases may be determined by effluent monitors or by 49 effluent sampling and subsequent laboratory analysis. For those facilities that use sampling as 50 the primary quantification method, appropriate records for laboratory counting systems should 51 be kept (e.g., calibration, minimum detectable activity determinations, chi square, resolution). In E-3
 
1 summary, the information maintained in the records discussed in this section should include the 2 following, as applicable:
3 4
* type and energy of radiation involved 5
6
* physical, chemical, and isotopic description of the radioactive material (i.e., gaseous, 7        liquid or particulate), the chemical compound and particle size distribution, if known 8
9
* total activity released and the method of determination 10 11
* time and date of the start and end of the release 12 13
* dilution volume of the effluent stream 14 15
* identification of the release point (i.e., roof vent, facility stack, drain to the sewer system) 16 17
* location, identification, and function of the facility surveyed 18 19
* time and date of surveys 20 21
* location at which measurements were made, or where liquid and air samples were 22        obtained, either by written description or by sketches 23 24
* methods and instruments used in measurements (i.e., the method of obtaining the air or 25        water samples) 26 27
* results of the measurements, with the quantities and in the units in which they were 28        obtained (e.g., Sv/h, mSv, R/h, mrem/h, or count/min) 29 30
* dose or activity calculations, including occupancy factors and the assumptions made 31 32
* statement of conclusions from the survey and whether the findings comply with the 33        regulations 34 35
* recommendations regarding remedial actions and resurvey 36 37
* name of the persons doing surveys 38 39 Note: Regulatory Guide 4.16, Monitoring and Reporting Radioactive Materials in Liquid and 40 Gaseous Effluents from Nuclear Fuel Facilities, provides guidance on the development and 41 implementation of effluent monitoring programs described in license applications and for 42 monitoring and reporting effluent data.
43 44 45 46 E-4
 
1 Constraint on Releases of Airborne Radioactive Materials to the Environment 1 2
3 The regulations in 10 CFR 20.1302(b) state the following:
4 5      A licensee shall show compliance with the annual dose limit in 10 CFR 20.1301 by 6      (1) demonstrating by measurement or calculation that the total effective dose 7      equivalent to the individual likely to receive the highest dose from the licensed 8      operation does not exceed the annual dose limit; or (2) demonstrating that (i) the 9      annual average concentrations of radioactive material released in gaseous and 10      liquid effluents at the boundary of the unrestricted area do not exceed the values 11      specified in table 2 of Appendix B to 10 CFR Part 20 [Standards for Protection 12      Against Radiation] and (ii) if an individual were continuously present in an 13      unrestricted area, the dose from external sources would not exceed 0.002 rem 14      (0.02 mSv) in an hour and 0.05 rem (0.5 mSv) in a year.
15 16 In addition, 10 CFR 20.1101(d) states the following:
17 18      To implement the ALARA requirements of 10 CFR 20.1101(b), and notwithstanding 19      the requirements in 10 CFR 20.1301 of this part, a constraint on air emissions of 20      radioactive material to the environment, excluding radon-222 and its daughters, 21      shall be established by licensees other than those subject to 10 CFR 50.34(a),
22      such that the individual member of the public likely to receive the highest dose will 23      not be expected to receive a total effective dose equivalent in excess of 10 mrem 24      (0.1 mSv) per year from these emissions. If a licensee subject to this requirement 25      exceeds this dose constraint, the licensee shall report the exceedance as provided 26      in 10 CFR 20.2203 [Reports of exposures, radiation levels, and concentrations of 27      radioactive material exceeding the constraints or limits] and promptly take 28      appropriate corrective action to ensure against recurrence.
1  Further information regarding guidance on compliance with the Constraint Rule appears in Regulatory Guide 4.20, Constraint on Releases of Airborne Radioactive Materials to the Environment for Licensees Other Than Power Reactors.
E-5
 
1                                            APPENDIX F 2                                SURVEYS AND MONITORING 3
4 This appendix provides applicants and licensees with information for meeting the requirements 5 in Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for Protection 6 against Radiation, Subpart F, Surveys and Monitoring.
7 8 Ambient Radiation Level Surveys 9
10
* Dose-rate surveys and monitoring, at a minimum, should be conducted in locations 11          where individual adults, including workers, are exposed to radiation levels that might 12          result in radiation doses in excess of 10 percent of the occupational dose limits in 13          10 CFR 20.1201, Occupational dose limits for adults (see also 10 CFR 20.1502, 14          Conditions requiring individual monitoring of external and internal occupational dose).
15 16
* Dose-rate surveys, at a minimum, should be conducted in locations where members of 17          the public could receive a total effective dose equivalent of 1 millisievert (mSv) 18          (100 millirem (mrem)) in a year, or the dose in any unrestricted area from external 19          sources could exceed 0.02 mSv (2 mrem) in any 1 hour (see 10 CFR 20.1301(a)(2)).
20 21
* Dose-rate surveys should be done in a manner and frequency that is representative of 22          the use of radioactive materials. At a minimum, these surveys should be done daily in 23          areas of radioactive material use, where exposures to workers could reasonably occur, 24          (e.g., generator storage or elution and dose preparation stations). Other areas where 25          radiological conditions are not expected to change appreciably from day to day should 26          be surveyed weekly (e.g., radioactive waste storage areas).
27 28 Contamination Surveys 29 30 The U.S. Nuclear Regulatory Commission (NRC) regulations in 10 CFR 20.1501(a)(1) require, 31 in part, that licensees make, or cause to be made, surveys that may be necessary to comply 32 with the regulations in 10 CFR Part 20. The regulations in 10 CFR 20.2103, Records of 33 surveys, require that each licensee maintain records showing the results of surveys and 34 calibrations called for by 10 CFR 20.1501, General, and 10 CFR 20.1906(b) (see 35 10 CFR 20.2103(a)). In accordance with the regulations in 10 CFR 70.25(g), licensees shall 36 keep records of information important to the decommissioning of a facility. The regulations in 37 10 CFR 70.25(g)(1) require that licensees keep records of spills or other unusual occurrences 38 involving the spread of contamination in and around the facility, equipment, or site.
39 40 Licensees contamination surveys should be sufficient to identify areas of contamination that 41 might result in unacceptable levels of exposure to workers or to the public. Combined 42 removable and fixed contamination should be surveyed using appropriate radiation detection 43 equipment. Removable contamination can be detected and measured through wipe tests, which 44 should be analyzed using an appropriate counting instrument. Fixed contamination may be 45 measured directly at the surface of the contamination with the appropriate instrument detector 46 held at close proximity to the surface without direct contact. Table C-1 in appendix C to this 47 document gives examples of appropriate instruments to be used in contamination surveys and 48 monitoring.
49 50 F-1
 
1 Contamination surveys shall be conducted as follows:
2
* to evaluate radioactive contamination that could be present on surfaces of floors, walls, 3        laboratory furniture, or equipment 4
5
* after any spill or contamination event 6
7
* to evaluate contamination of users and the immediate work area at the end of each day 8        when licensed material is used 9
10
* in unrestricted areas at frequencies consistent with the types and quantities of materials 11        in use 12 13
* in areas adjacent to restricted areas and in all areas through which licensed materials 14        are transferred and temporarily stored before shipment 15 16 Survey Frequency 17 18 The regulation at 10 CFR 20.1501 requires that licensees make, or cause to be made, radiation 19 surveys that may be necessary for the licensee to comply with the NRC regulations. The 20 regulation at 10 CFR 20.1302, Compliance with dose limits for individual members of the 21 public, states (a) The licensee shall make or cause to be made, as appropriate, surveys of 22 radiation levels in unrestricted and controlled areas and radioactive materials in effluents 23 released to unrestricted and controlled areas to demonstrate compliance with the dose limits for 24 individual members of the public in &sect; 20.1301.
25 In general, surveys should be performed before the facility begins operation to establish a 26 baseline of background radiation levels and radioactivity from natural sources. These baseline 27 surveys should be performed under various conditions expected during routine plant operation.
28 Surveys should be conducted during the test operation of any new process or protective 29 equipment, during significant changes in input materials or workload, and during routine plant 30 operations with all potentially involved persons present and carrying out their functions.
31 Regulatory Guide 8.24, Health Physics Surveys During Enriched Uranium-235 Processing and 32 Fuel Fabrication, provides guidance on the types and frequencies of surveys that the NRC staff 33 considers acceptable for the protection of workers in facilities licensed by the NRC to process 34 special nuclear materials. Appendix B to Regulatory Guide 8.24 presents minimum acceptable 35 frequencies for meeting the requirements in 10 CFR 20.1501.
36 37 Personnel should survey for contamination in locations where radioactive materials are used or 38 handled, including areas where radioactive materials are eluted, prepared, assayed, dispensed, 39 or packaged for transport. These surveys should be done at a frequency appropriate to the 40 types and quantities of radioactive materials in use. The frequency of routine surveys should be 41 commensurate with the specific operations and nature of the work being conducted; the 42 quantities of material being processed; and the specific protective facilities, equipment, and 43 procedures used to protect workers from external radiation and the intake of radioactive 44 materials.
45 46 47 48 49 F-2
 
1 Contamination in Unrestricted Areas 2
3 Unrestricted areas inside a facility should be surveyed and smeared periodically for surface 4 contamination to ensure that contamination is adequately confined in the controlled areas.
5 Licensees should use uniform methods for surveying and collecting and analyzing smear 6 samples. When contamination is found, licensees should immediately decontaminate the 7 unrestricted areas to background levels.
8 9 A standardized method for smear testing of a relatively uniform area should be used over 10 extended periods of time to aid in comparing contamination at different times and places and to 11 evaluate trends. A smear taken from an area of about 100 square centimeters (cm2) is 12 acceptable to indicate levels of removable contamination.
13 When it is not possible to return an area to background levels, the licensee must ensure that the 14 amounts do not exceed the contamination levels listed in table F-1 1. Note: The contamination 15 levels in table F-1 are not to be used for releasing facilities for unrestricted use or termination of 16 the license pursuant to 10 CFR Part 20, Subpart E, Radiological Criteria for License 17 Termination. In particular, the contamination levels for most alpha emitters in table F-1 exceed 18 the occupational dose levels in 10 CFR Part 20, Subpart E. Table F-1 levels can continue to be 19 used for release of equipment and material from licensed material facilities during operational 20 activities before license termination (see 63 FR 64132, November 18, 1998).
1  Table F-1 is incorporated into draft NUREG-2212 from Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material, issued April 1993 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML103620647). See also Appendix A, Acceptable Surface Contamination Levels, to Regulatory Guide 8.24 (ML110400305). Table F-1 contamination levels can be used to determine whether equipment and materials used in facilities licensed under 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, can be released during operational activities and before license termination (see 63 FR 64132, November 18, 1998).
F-3
 
Table F-1 Acceptable Surface Contamination Levels Nuclide1                    Average2,3,6                Maximum2,4,6              Removable2,5,6 U-nat, U-235, U-238,            83.3 Bq/100 cm2            250 Bq/100 cm2              16.7 Bq/100 cm2 and associated decay products                  (5,000 dpm/100 cm2)        (15,000 dpm/100 cm2)        (1,000 dpm/100 cm2)
Transuranics, Ra-226,            1.7 Bq/100 cm2              5.0 Bq/100 cm2              0.3 Bq/100 cm2 Ra-228, Th-230, Th-228, Pa-231,                  (100 dpm/100 cm2)          (300 dpm/100 cm2)            (20 dpm/100 cm2)
Ac-227, I-125, I-129 Th-nat, Th-232, Sr-90,          16.7 Bq/100 cm2            50.0 Bq/100 cm2              3.3 Bq/100 cm2 Ra-223, Ra-224, U-232, I-126, I-131,I-          (1,000 dpm/100 cm2)        (3,000 dpm/100 cm2)          (200 dpm/100 cm2) 133 U-nat, U-235, U-238,            83.3 Bq/100 cm2            250 Bq/100 cm2              16.7 Bq/100 cm2 and associated decay products                  (5,000 dpm/100 cm2)        (15,000 dpm/100 cm2)        (1,000 dpm/100 cm2)
Other alpha emitters7            8.33 Bq/100 cm2            25 Bq/100 cm2                1.67 Bq/100 cm2 (500 dpm/100 cm2)          (1,500 dpm /100 cm2)        (100 dpm/100 cm2)
Beta-gamma emitters              83.3 Bq/100 cm2            250 Bq/100 cm2              16.7 Bq/100 cm2 (nuclides with decay modes other than alpha          (5,000 dpm/100 cm2)        (15,000 dpm/100 cm2)        (1,000 dpm/100 cm2) emission or spontaneous fission) except Sr-90 and others noted above.
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 cm2.
5 The  amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that area with a 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.
6 The  average and maximum radiation levels associated with surface contamination resulting from beta-gamma emitters should not exceed 0.2 millirads per hour (mrad/h) at 1 centimeter and 1.0 mrad/h at 1 centimeter, respectively, measured through not more than 7 milligrams per cm2 of total absorber.
7 Surface  contamination levels were derived using one order of magnitude less than the values for beta-gamma emitters.
1 F-4
 
1 Table F-1 provides the maximum acceptable residual levels for contaminated equipment that is 2 to be released for unrestricted use. The NRC provides additional guidance on the release of 3 equipment into unrestricted use in NUREG-1575, Supplement 1, Multi-Agency Radiation 4 Survey and Assessment of Materials and Equipment Manual (MARSAME), Appendix E, 5 Disposition Criteria, issued January 2009. Table F-1 values are acceptable criteria to assess 6 contamination in facilities during operation. Regulatory Guide 8.24, Health Physics Surveys 7 During Enriched Uranium-235 Processing and Fuel Fabrication, and Regulatory Guide 8.30, 8 Health Physics Surveys in Uranium Recovery Facilities, provide additional guidance on 9 surveys and contamination control in unrestricted areas.
10 11 Survey Record Requirements 12 13 The NRCs recordkeeping requirements do not specify the information that must be included in 14 a survey record. However, the information to be recorded should be commensurate with the 15 specific operations and nature of the activities being conducted and must be in sufficient detail 16 to meet the requirements in 10 CFR 20.2103, Records of surveys. In general, each survey 17 record should include, at a minimum, the following:
18 19
* diagrams of the areas identifying specific locations surveyed 20 21
* ambient radiation levels with appropriate units 22 23
* contamination levels with appropriate units 24 25
* make and model number of instruments used 26 27
* background levels 28 29
* name of the person making the evaluation and recording the results and the date 30 31
* corrective actions taken for elevated levels identified and results of repeated survey 32 33 The regulations at 10 CFR 70.25(g) require that licensees keep records of surveys and 34 information important to the decommissioning of a facility. The regulations in 35 10 CFR 70.25(g)(1) require that licensees keep records of spills or other unusual occurrences 36 involving the spread of contamination in and around the facility, equipment, or site.
37 38 Reports Under 10 CFR 20.2203 39 40 The regulations at 10 CFR 20.2203(a) require that for reportable events, in addition to the 41 notification required by 10 CFR 20.2202, Notification of incidents, each licensee must submit a 42 written report within 30 days after learning of any of the following occurrences, as stated below 43 in 10 CFR 20.2203(a)(1)-(4):
44 45        (1) Any incident for which notification is required by &sect; 20.2202; or 46 47        (2) Doses in excess of any of the following:
48 49            (i) The occupational dose limits for adults in &sect; 20.1201; or 50            (ii) The occupational dose limits for a minor in &sect; 20.1207; or F-5
 
1            (iii) The limits for an embryo/fetus of a declared pregnant woman in 2                  &sect; 20.1208; or 3            (iv) The limits for an individual member of the public in &sect; 20.1301; or 4            (v) Any applicable limit in the license; or 5            (vi) The ALARA constraints for air emissions established under &sect; 20.1101(d);
6                  or 7
8        (3) Levels of radiation or concentrations of radioactive material in 9
10            (i) A restricted area in excess of any applicable limit in the license; or 11            (ii) An unrestricted area in excess of 10 times any applicable limit set forth in 12                  this part or in the license (whether or not involving exposure of any 13                  individual in excess of the limits in &sect; 20.1301); or 14 15        (4) For licensees subject to the provisions of EPAs [U.S. Environmental 16            Protection Agencys] generally applicable environmental radiation standards 17            in 40 CFR Part 190, levels of radiation or releases of radioactive material in 18            excess of those standards, or of license conditions related to those 19            standards.
20 21 The regulations at 10 CFR 20.2203(b)(1) require that a report describe the extent of exposure of 22 individuals to radiation and radioactive material and include the following, as appropriate:
23 24
* estimates of each individuals dose 25 26
* the levels of radiation and concentrations of radioactive material involved 27 28
* the cause of the elevated exposures, dose rates, or concentrations 29 30
* corrective steps taken or planned to ensure against a recurrence, including the schedule 31        for achieving conformance with applicable limits, ALARA [as-low-as-is-reasonably 32        achievable] constraints, generally applicable environmental standards, and associated 33        license conditions 34 35 The regulations in 10 CFR 20.2203(b)(2) require that the report include the name, social 36 security account number, and date of birth for each occupationally overexposed individual (or 37 declared pregnant woman).
38 39 Air Sampling 40 41 Air sampling can be used to do the following:
42 43
* determine whether the confinement of radioactive materials is effective 44 45
* measure airborne radioactive material concentrations in the workplace 46 47
* estimate worker intakes of radioactive material 48 49
* determine posting requirements 50 F-6
 
1
* determine what protective equipment and measures are appropriate 2
3
* warn of significantly elevated levels of airborne radioactive material 4
5 Regulatory Guide 8.25, Air Sampling in the Workplace, and NUREG-1400, Air Sampling in 6 the Workplace, contain further guidance on air sampling.
7 8 Air Effluent Monitoring 9
10 Airborne radioactive effluents should be monitored at the release points (e.g., stack) to provide 11 accurate measurements to estimate public exposure. Licensees should verify the performance 12 of effluent monitoring systems by regular calibration of equipment and checks of filtration to 13 ensure their reliability.
14 15 Regulatory Guide 4.20, Constraints on Release of Airborne Radioactive Materials to the 16 Environment for Licensees Other Than Power Reactors, issued December 1996, and 17 Regulatory Guide 4.20, provide guidance on methods acceptable (calculation or COMPLY 18 code) to the U.S. Nuclear Regulatory Commission (NRC) for compliance with the constraint on 19 air emissions to the environment.
20 21 Effluent monitoring systems should be designed in accordance with American National 22 Standards Institute/Health Physics Society (ANSI/HPS) 13.1-2011, Sampling and Monitoring 23 Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities, 24 and ANSI N42.18-2004, Specification and Performance of On-Site Instrumentation for 25 Continuously Monitoring Radioactivity in Effluents.
26 27 Sanitary Sewerage Release Monitoring 28 29 The licensee should evaluate the concentrations of radioactive material in water that is released 30 to the environment and to the sanitary sewer. The licensee must show that these releases meet 31 the limits in 10 CFR 20.1301, Dose limits for individual members of the public, and 32 10 CFR 20.2003, Disposal by release into sanitary sewage, respectively.
33 34 Special Monitoring 35 36 Because of uncertainty in the time of intakes and the absence of other data related to the 37 exposure (e.g., physical and chemical forms, exposure duration), correlating positive results to 38 actual intakes for routine measurements can sometimes be difficult. Abnormal and inadvertent 39 intakes from situations such as inadequate engineering controls, inadvertent ingestion, 40 contamination of a wound, or skin absorption, should be evaluated on a case-by-case basis.
41 When determining whether potential intakes should be evaluated, consider the following 42 circumstances:
43 44
* presence of unusually high levels of facial or nasal contamination 45 46
* operational events with a reasonable likelihood that a worker was exposed to unknown 47        quantities of airborne radioactive material (e.g., loss of system or container integrity) 48 49
* known or suspected incidents of a worker ingesting radioactive material 50 F-7
 
1
* incidents that result in contamination of wounds or other skin absorption 2
3 References 4
5 American National Standards Institute (ANSI)/Health Physics Society, Sampling and Monitoring 6 Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities, 7 ANSI/HPS N13.1-2011, March 30, 2011.
8 9 ANSI, Specification and Performance of On-Site Instrumentation for Continuously Monitoring 10 Radioactivity in Effluents, ANSI N42.18-2004, 2004.
11 12 Code of Federal Regulations, Chapter I, Title 10 (10 CFR), Energy, 20.1201, Occupational 13 dose limits for adults.
14 15 10 CFR 20.1301, Dose limits for individual members of the public.
16 17 10 CFR 20.1302, Compliance with dose limits for individual members of the public.
18 19 10 CFR 20.1501, General.
20 21 10 CFR 20.1502, Conditions requiring individual monitoring of external and internal 22 occupational dose.
23 24 10 CFR 20.1906, Procedures for receiving and opening packages.
25 26 10 CFR 20.2103, Records of surveys.
27 28 10 CFR 70.25, Financial assurance and recordkeeping for decommissioning.
29 30 U.S. Nuclear Regulatory Commission (NRC), Federal Register Notice, Supplemental 31 Information on the Implementation of the Final Rule on Radiological Criteria for License 32 Termination, 63 FR 64132, November 18, 1998.
33 34 NRC, Air Sampling in the Workplace, NUREG-1400.
35 36 NRC, Multi-Agency Radiation Survey and Assessment of Materials and Equipment Manual 37 (MARSAME), NUREG 1575, Supplement 1.
38 39 NRC, Constraints on Release of Airborne Radioactive Materials to the Environment for 40 Licensees Other Than Power Reactors, Regulatory Guide 4.20.
41 42 NRC, Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication, 43 Regulatory Guide 8.24.
44 45 NRC, Air Sampling in the Workplace, Regulatory Guide 8.25.
46 47 NRC, Health Physics Surveys in Uranium Recovery Facilities, Regulatory Guide 8.30.
F-8
 
1                                              APPENDIX G 2                                        SAFETY CULTURE 3
4 Individuals and organizations undertaking regulated activities are expected to establish and 5 maintain a positive safety culture commensurate with the safety and security significance of 6 their activities and the nature and complexity of their organizations and functions. This applies to 7 all licensees, certificate holders, permit holders, authorization holders, holders of quality 8 assurance program approvals, vendors and suppliers of safety-related components, and 9 applicants for a license, certificate, permit, authorization, or quality assurance program approval, 10 subject to U.S. Nuclear Regulatory Commission (NRC) authority.
11 12 Nuclear safety culture is defined in the NRCs safety culture policy statement (76 FR 34773; 13 June 14, 2011) as the core values and behaviors resulting from a collective commitment by 14 leaders and individuals to emphasize safety over competing goals to ensure protection of 15 people and the environment. Individuals and organizations performing regulated activities bear 16 the primary responsibility for safely handling and securing these materials. Experience has 17 shown that certain personal and organizational traits are present in a positive safety culture.
18 A trait, in this case, is a pattern of thinking, feeling, and behaving that emphasizes safety, 19 particularly in goal-conflict situations (e.g., production versus safety, schedule versus safety, 20 and cost of the effort versus safety).
21 22 Organizations should ensure that personnel in the safety and security sectors have an 23 appreciation for the importance of each, emphasizing the need for integration and balance to 24 achieve both safety and security in their activities. Safety and security activities are closely 25 intertwined. While many safety and security activities complement each other, there may be 26 instances in which safety and security interests create competing goals. It is important that 27 consideration of these activities be integrated so as not to diminish or adversely affect either; 28 thus, mechanisms should be established to identify and resolve these differences. A safety 29 culture that accomplishes this would include all nuclear safety and security issues associated 30 with NRC-regulated activities.
31 32 The NRC, as the regulatory agency with an independent oversight role, reviews the 33 performance of individuals and organizations to determine compliance with requirements and 34 commitments through its existing inspection and assessment processes. However, the NRCs 35 safety culture policy statement and traits are not incorporated into the regulations. Many of the 36 safety culture traits may be inherent to an organizations existing radiation safety practices and 37 programs. For instance, in the handling of special nuclear material, licensees should have a 38 policy for reporting safety concerns to ensure safety culture is maintained. When safety has 39 become embodied in their culture, the staff should report a safety concern without any fear of 40 retaliation.
41 42 The NRCs safety culture policy statement can be found at https://www.nrc.gov/about-43 nrc/safety-culture/sc-policy-statement.html. More information on NRC activities relating to safety 44 culture appears at https://www.nrc.gov/about-nrc/regulatory/enforcement/safety-culture.html.
45 G-1
 
1                                                  APPENDIX H 2    MAJOR U.S. DEPARTMENT OF TRANSPORTATION REGULATIONS 3
4 The following are the major areas in the U.S. Department of Transportation regulations that are 5 most relevant for the transportation of licensed material shipped as Type A 1 quantities:
6 7
* Carriage by Public HighwayGeneral Information and Regulations, Subpart A, Title 49 8          of the Code of Federal Regulations (49 CFR) 177.816, 49 CFR 177.817, 9          49 CFR 177.834(a), 49 CFR 177.842: Driver training, shipping paper, general 10          requirements (secured against movement), Class 7 (radioactive) material.
11 12
* Emergency Response Information, Subpart G, 49 CFR 172.600, 49 CFR 172.602, 13          49 CFR 172.604: Applicability and general requirements, emergency response 14          information, emergency response telephone number.
15 16
* Hazardous Materials Table, 49 CFR 172.101, Purpose and use of hazardous materials 17          table, Appendix A, List of Hazardous Substances and Reportable Quantities, Table 2:
18          Radionuclides.
19 20
* Package Labeling, 49 CFR 172.400, 49 CFR 172.401, 49 CFR 172.403, 21          49 CFR 172.406, 49 CFR 172.407, 49 CFR 172.436, 49 CFR 172.438, 49 CFR 172.440:
22          General labeling requirements, prohibited labeling, radioactive materials, placement of 23          labels, specifications for radioactive labels.
24 25
* Package Markings, 49 CFR 172.300, 49 CFR 172.301, 49 CFR 172.303, 26          49 CFR 172.304, 49 CFR 172.310, 49 CFR 172.324: General marking requirements for 27          non-bulk packaging, prohibited marking, marking requirements, radioactive material, 28          hazardous substances in non-bulk packaging.
29 30
* Placarding of Vehicles, 49 CFR 172.500, 49 CFR 172.502, 49 CFR 172.504, 31          49 CFR 172.506, 49 CFR 172.516, 49 CFR 172.519, 49 CFR 172.556: Applicability, 32          prohibited and permissive placarding, general placarding requirements, providing and 33          affixing placards: highway, visibility and display of placards, specifications for radioactive 34          placards.
35 36
* Radiation Protection Program for Shippers and Carriers, Subpart I, 49 CFR 172.801, 37          49 CFR 172.803, 49 CFR 172.805: Applicability of the radiation protection program, 38          radiation protection program, record keeping, and notifications.
39 40
* Shipping Papers, 49 CFR 172.200-204: General entries, description, additional 41          description requirements, shippers certification.
42 43
* Training, Subpart H, 49 CFR 172.702, 49 CFR 172.704: Applicability and responsibility 44          for training and testing, training requirements.
1 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, Packaging and Transportation of Radioactive Material, or may be determined by procedures described in Appendix A, Determination of A1 and A2, to 10 CFR Part 71.
H-1
 
1
* ShippersGeneral Requirements for Shipments and Packaging, Subpart I, 2  49 CFR 173.403, 49 CFR 173.410, 49 CFR 173.412, 49 CFR 173.415, 49 CFR 173.431, 3  49 CFR 173.433, 49 CFR 173.435, 49 CFR 173.441, 49 CFR 173.443, 49 CFR 173.448, 4  49 CFR 173.475, 49 CFR 173.476: Definitions, general design requirements, additional 5  design requirements for Type A packages, authorized Type A packages, activity limits 6  for Type A packages, requirements for determining A 1 and A 2 , table of A 1 and A 2 values 7  for radionuclides, radiation level limitations, contamination control, general transportation 8  requirements, quality control requirements before each shipment, approval of special 9  form radioactive materials.
H-2
 
1                                              APPENDIX I 2                            NUCLEAR CRITICALITY GLOSSARY 3
4 This glossary defines technical and industry terms that are used in Chapter 3 of this document.
5 6 Abnormal condition. An event or condition not intended as a desirable or regularly occurring 7 condition in the facility or process design but which is anticipated as a contingency in criticality 8 safety evaluations. A condition that is reached by exceeding the safety limit(s) of one or more 9 controlled parameters.
10 11 Administrative control. Human actions, whether required or prohibited, relied on to prevent or 12 mitigate a specific accident sequence or to maintain subcriticality, and established in formal 13 plant procedures.
14 15 Area(s) of applicability. The range of physical parameters (e.g., isotopic abundance, 16 moderation, neutron energy, absorbers) characterizing a fissile material system over which a 17 given calculational method has been validated; the range of such parameters covered by the 18 chosen benchmark experiments and for which a bias has been determined.
19 20 Benchmark. A critical experiment that is widely accepted and whose physical characteristics 21 and their uncertainties have been well characterized, so that it is suitable for validation.
22 23 Bias. The numerical difference between the calculated and experimental values of k eff for a set 24 of benchmark experiments covering a particular area of applicability (often expressed as a 25 function of system parameters).
26 27 Concurrent. In the context of double contingency, two changes in process conditions are 28 concurrent if the effect of the first change persists until the second change occurs. This does not 29 mean simultaneous (where both upsets occur at the same time) but rather that the system is 30 affected by both changes during some time interval.
31 32 Consequence. (1) Occurrence of an accidental criticality, (2) the energy released in an 33 accidental criticality, normally expressed in terms of the number of fissions or dose to workers.
34 35 Contingency. In the context of double contingency, a change in process conditions or loss of a 36 criticality control that could result in one or more parameters exceeding their safety limits.
37 38 Control. Structures, systems, components, devices, equipment, or human actions relied on to 39 prevent or mitigate an accident sequence, or to maintain subcriticality by limiting a parameter 40 within specified limits and established and identified as such in formal plant safety basis 41 documentation.
42 43 Control system. A set of controls that function together to prevent or mitigate one or more 44 accident sequences, or to maintain subcriticality by limiting one or more parameters within 45 specified limits.
46 47 Controlled parameter. A parameter of a system that is maintained within a specified range to 48 ensure subcriticality.
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1 Credible abnormal condition. As used in meeting Title 10 of the Code of Federal Regulations 2 (10 CFR) 70.61(d), one of the abnormal conditions that must be considered in the context of 3 demonstrating compliance with the double contingency principle. Conditions resulting from 4 credible single failures and related sequences of events up to those that must be considered in 5 meeting the double contingency principle (see NUREG-1520, Standard Review Plan for Fuel 6 Cycle Facilities License Applications, Appendix 5-A, for more information).
7 8 Critical. (1) Having an actual k eff value greater than 1; (2) having a calculated k eff value greater 9 than the upper subcritical limit.
10 11 Critical mass. 1 For this documents purpose, a critical mass of special nuclear material (SNM) 12 means uranium enriched in the isotope uranium (U)-235 in quantities exceeding 350 grams of 13 contained U-235; or U-233 in quantities exceeding 200 grams; or plutonium in quantities 14 exceeding 200 grams; or any combination of them in accordance with the following formula:
15 (grams contained U-235/350) + (grams U-233)/200) + (grams plutonium (Pu)/200) > 1.
16 17 Criticality control. A control used to ensure subcriticality.
18 19 Criticality safety evaluation. A structured analysis demonstrating criticality safety for a given 20 process, including a demonstration that processes will be subcritical under normal and credible 21 abnormal conditions, and the specification of controls and limits to achieve that goal (also often 22 referred to, for example, as a nuclear criticality safety evaluation, analysis, or assessment).
23 24 Degraded. A control, control system, or controlled parameter is considered to be degraded 25 when the parameter is maintained within its safety limit but has reduced reliability and 26 availability, such that the possibility of those limits being exceeded is no longer unlikely.
27 28 Double contingency principle. Process designs should incorporate sufficient factors of safety 29 to require at least two unlikely, independent, and concurrent changes in process conditions 30 before a criticality accident is possible.
31 32 Double contingency protection. The condition of requiring at least two unlikely, independent, 33 and concurrent changes in process conditions before a criticality accident is possible.
34 35 Engineered control. Structures, systems, components, devices, or equipment relied on to 36 prevent or mitigate a specific accident sequence or to maintain subcriticality and established in 37 formal design documents.
38 39 Event. (1) A change in process conditions that has the potential to adversely affect safety; 40 (2) one of several occurrences that constitute an accident sequence.
41 42 Favorable geometry. Characteristic of structures, systems, devices, or equipment such that 43 fissile material maintained within specified dimensions will be subcritical under the most reactive 44 credible conditions (defined for a given isotopic composition and physicochemical form).
45 1  As used in 10 CFR Part 70, a critical mass of SNM means SNM in a quantity exceeding 700 grams of contained 235U; 520 grams of 233U; 450 grams of plutonium; 1,500 grams of contained 235U, if no uranium enriched to more than 4 percent by weight of 235U is present; 450 grams of any combination thereof; or one-half such quantities if massive moderators or reflectors made of graphite, heavy water, or beryllium may be present.
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1 Independent. In the context of double contingency, two changes in process conditions are 2 independent if the occurrence of one does not cause, or affect the probability of occurrence of, 3 the other; if the probability that both occur is independent of the order in which they occur; and if 4 there are no identifiable common-mode failures that can lead to criticality.
5 6 Isolated. (1) The condition whereby the flow of matter and energy between a system and 7 surrounding systems can be neglected for the purpose of doing a safety analysis; (2) being 8 separated by a sufficient distance from other systems or materials such that their presence has 9 a negligible effect on the systems k eff .
10 11 Lost. A control, control system, or controlled parameter is considered to be lost when the 12 measures that maintain the parameters within their safety limits cease to function as designed, 13 or cannot be verified to function as designed, whether or not the affected parameters actually 14 exceed their safety limits.
15 16 Management measures. Measures employed by plant management to ensure continued safe 17 operation of nuclear processes and continued effective operation of the nuclear criticality safety 18 (NCS) program, such as operator training, NCS engineer qualification, audits and assessments, 19 and administrative and operating procedures.
20 21 Margin of safety. The difference between the actual value of a parameter and the value of the 22 parameter at which the system is expected to be critical (taking bias and bias uncertainty into 23 account).
24 25 Margin of subcriticality. (1) The difference between the actual value of k eff and the value at 26 which the system is expected to be critical (taking bias and bias uncertainty into account);
27 (2) the difference between the calculated value of k eff (including uncertainties) and the value at 28 which the system is expected to be critical (taking bias and bias uncertainty into account), plus 29 any margin in k eff resulting from conservative modeling of system parameters.
30 31 Minimum margin of subcriticality. Margin in k eff beyond the bias and uncertainty in the bias, 32 to allow for any unknown or difficult to quantify uncertainties in calculating k eff (frequently 33 referred to as the arbitrary margin or administrative margin).
34 35 Normal condition. A condition specifically anticipated or allowed for as part of the normal 36 operation of the facility. A condition in which all controlled parameters are within their safety 37 limits.
38 39 Operating limit. The value of a controlled parameter to which actual operations are restricted, 40 with sufficient margin to ensure that exceeding the safety limit is unlikely.
41 42 Optimum. The value of a parameter that produces the highest k eff .
43 44 Parameter. The measurable or observable characteristic of a system that affects the value of 45 k eff . The parameters normally are mass, geometry, density, enrichment/isotopics, reflection, 46 moderation, concentration, interaction, absorption, volume, heterogeneity, physicochemical 47 form, and process variables.
48 49 Process condition. In the context of double contingency, the set of all characteristics or 50 attributes of a process important to safety (a change in the value of a parameter, or loss or 51 degradation of a control affecting the ability to maintain a parameter, for example).
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1 Reactivity. Loosely used synonymously with k eff . Used most frequently in most reactive 2 credible, to mean the physical conditions that produce the highest credible value of k eff.
3 4 Research and development. (1) Theoretical analysis, exploration, or experimentation, or 5 (2) the extension of investigative findings and theories of a scientific or technical nature into 6 practical application for experimental and demonstration purposes, including the experimental 7 production and testing of models, devices, equipment, materials, and processes.
8 9 Safe mass. The quantity of fissile material that is safely subcritical under the most reactive 10 credible conditions (defined for a given isotopic composition and physicochemical form),
11 including allowance for over batching.
12 13 Safe process condition. The defined range of acceptable values of one or more controlled 14 parameters and other process conditions.
15 16 Safety control. The same as control.
17 18 Safety limit. The value of a controlled parameter established by a criticality safety evaluation to 19 which the process will be controlled. This can be equal to the subcritical limit but can include 20 additional margin due to uncertainty and variability in the process (also referred to as the 21 analytical limit).
22 23 Safety margin. The same as margin of safety.
24 25 Sealed source. Any SNM encased in a capsule designed to prevent SNM leakage or escape.
26 27 Special nuclear material. (1) Plutonium, U-233, uranium enriched in the isotope U-233 or in 28 the isotope U-235, and any other material which the Commission, pursuant to the provisions of 29 section 51 of the Atomic Energy Act of 1954, as amended, determines to be SNM but does not 30 include source material, or (2) any material artificially enriched by any of the foregoing but does 31 not include source material.
32 33 Subcritical. Demonstrated not to be critical; having a value of k eff not greater than the upper 34 subcritical limit.
35 36 Subcritical limit. (1) The bounding value of a controlled parameter that has been demonstrated 37 to maintain a system subcritical in plant criticality safety evaluations; (2) the upper subcritical 38 limit.
39 40 Subcritical margin. The same as margin of subcriticality.
41 42 System. Discrete part of a fissile material operation that can be separated from other systems 43 for the purpose of doing a safety analysis and that is the subject of a criticality safety evaluation.
44 45 Unlikely. In the context of double contingency, (1) expected to occur rarely, if at all, during the 46 lifetime of a facility, or (2) judged to have a probability of occurrence less than 10-2 per year per 47 event.
48 49 Upper subcritical limit. The maximum value of k eff that is considered to be subcritical with an 50 acceptable degree of confidence (taking bias and bias uncertainty into account and including a 51 minimum margin of subcriticality).
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NRC FORM 335                                                                    U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (12-2010)                                                                                                              (Assigned by NRG, Add Vol., Supp., Rev.,
NRCMD 3.7                                                                                                                and Addendum Numbers, if any.)
BIBLIOGRAPHIC DATA SHEET (See instructions on the reverse)                                              NUREG-2212
: 2. TITLE AND SUBTITLE                                                                                                      3. DATE REPORT PUBLISHED Standard Review Plan for Applications for 10 CFR Part 70 Licenses for                                                        MONTH                  YEAR Possession and Use of Special Nuclear Materials of Critical Mass but not Subject                                        December                  2022 to the Requirements in 10 CFR Part 70, Subpart H
: 4. FIN OR GRANT NUMBER
: 5. AUTHOR(S)                                                                                                        6. TYPE OF REPORT Suzanne Ani, Marylin Diaz-Maldonado, James Downs, Timothy Harris, Technical Kenneth Kline, Jeremy Munson, Tyrone Naquin, Michael Norris,
: 7. PERIOD COVERED (Inclusive Dates)
Osiris Siurano-P&#xe9;rez, Jean Trefethen 8.PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if Contractor, provide name and mailing address.)
Division of Fuel Management (DFM)
Office of Nuclear material Safety and Safeguards (NMSS)
U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
: 9. SPONSORING ORGANIZATION - NAME AND ADDRESS (lf NRC, type "Same as above", if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address.)
DFM/NMSS, DPCP/NSIR, DPR/NSIR U.S Nuclear Regulatory Commission Washington, DC 20555-0001
: 10. SUPPLEMENTARY NOTES O. Siurano-P&#xe9;rez, R. Jervey, and J. Tobin - Project Managers
: 11. ABSTRACT (200 words or less)
This Standard Review Plan contains information intended to provide program-specific guidance for Title 10 of the Code of Federal Regulations (10 CFR) Part 70, Domestic Licensing of Special Nuclear Material, applications for materials licenses of critical mass of special nuclear materials (SNM) that are not subject to the requirements in 10 CFR Part 70, Subpart H, Additional Requirements for Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material (also known as greater than critical mass applicants or licensees). Specifically, NUREG-2212 describes the types of information required under 10 CFR 70.22, Contents of applications, to apply for a new (or renew an existing) materials license for possession and use of special nuclear material in quantities exceeding the thresholds for critical mass quantities of SNM specified in 10 CFR 150.11, Critical mass, but that, due to the nature of their activities, are not subject to the requirements in 10 CFR Part 70, Subpart H.
: 12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)                          13. AVAILABILITY STATEMENT unlimited NUREG-2212; Standard Review Plan; Applications for Materials Licenses of Critical                                                14. SECURITY CLASSIFICATION Mass; Greater Than Critical Mass; Critical Mass; 10 CFR Subpart H.                                                                (This Page) unclassified (This Report) unclassified
: 15. NUMBER OF PAGES
: 16. PRICE NRC FORM 335 (12-2010)
 
NUREG-2212 Standard Review Plan for Applications for 10 CFR Part 70 Licenses for Possession and Use December 2022 of Special Nuclear Materials of Critical Mass but Not Subject to the Requirements in 10 CFR Part 70, Subpart H}}

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