ML24304A864

DG-1425 (RG 1.183 Rev 2) Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors ACRS Version

ML24304A864
Person / Time
Issue date:	11/12/2024
From:	David Garmon-Candelaria NRC/NRR/DRA/ARCB
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References
DG-1425 RG 1.183 Rev 2
Download: ML24304A864 (105)

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U.S. NUCLEAR REGULATORY COMMISSION DRAFT REGULATORY GUIDE DG-1425 Proposed Revision 2 to Regulatory Guide 1.183 Issue Date: _______ 2024 Technical Office/Division: NRR/DRA and NRR/DSS Pre-Decisional/Public version for November and December 2024 meetings with the Advisory Committee on Reactor Safeguards This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff review or approval and does not represent an NRC final staff position. Public comments are being solicited on this DG and its associated regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal rulemaking Web site, http://www.regulations.gov, by searching for draft regulatory guide DG-1425. Alternatively, comments may be submitted to Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff. Comments must be submitted by the date indicated in the Federal Register notice.

Electronic copies of this DG, previous versions of DGs, and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at https://nrc.gov/reading-rm/doc-collections/reg-guides/. The DG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML24185A179. The regulatory analysis is associated with a rulemaking and may be found in ADAMS under Accession No. ML24239A776.

Pre-Decisional/Public version for November and December 2024 meetings with the Advisory Committee on Reactor Safeguards ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN-BASIS ACCIDENTS AT NUCLEAR POWER REACTORS A. INTRODUCTION Purpose This regulatory guide (RG) describes a method that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable in complying with regulations for design-basis accident (DBA) radiological dose consequence analysis using an alternative source term (AST). This guidance for light-water reactor (LWR) designs includes the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. This guide establishes an AST based in part on SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, issued April 2023 (Ref. 1), and identifies significant attributes of other accident source terms that may be acceptable. Analyses in SAND2023-01313 considered core-average burnup values spanning from those within the applicability of previous revisions of this guide up to higher burnup and enrichment values being considered for commercial application. However, the use of these source terms is not endorsed for mixed-oxide fuels or long-term accident tolerant fuel (ATF) concepts for which available data is limited. This guide also identifies acceptable radiological analysis assumptions for use in conjunction with the AST. Unusual site characteristics, plant design features, or other factors may require different assumptions, which the staff will consider on a case-by-case basis.

Applicability This RG applies to applicants for and holders of licenses and other approvals for nuclear power reactors under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 2), and 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 3), and to holders of renewed licenses under 10 CFR Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants (Ref. 4). In addition, although the guidance primarily reflects reviews of license amendment requests for light-water nuclear power plants licensed under 10 CFR Part 50, this RG could provide useful information to new reactor applicants and licensees under 10 CFR Part 50 or 10 CFR Part 52.

DG-1425, Page 2 Applicable Regulations 10 CFR Part 50 provides regulations for licensing production and utilization facilities.

o 10 CFR 50.67, Accident source term, allows applicable licensees to voluntarily revise the accident source term used in design-basis radiological consequence analyses. A licensee that seeks to revise its current accident source term must apply for a license amendment under 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit. The regulation in 10 CFR 50.67 also allows the NRC to issue these amendments if the applicants analysis demonstrates with reasonable assurance that certain dose reference values are met. In accordance with 10 CFR 50.67, a holder of an operating license issued before January 10, 1997, can voluntarily revise the accident source term used in design-basis radiological consequence analyses. Applicants and licensees under 10 CFR Part 52 are not required to comply with 10 CFR 50.67; however, they may find that this guidance is useful in evaluations of fission product releases and the radiological consequences of LWR DBAs.

Although the source term information is specific to LWR designs, this guide could provide useful information on radiological consequence analysis for non-LWR designs.

o General Design Criterion (GDC) 19, Control room, in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 requires that a control room be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents (LOCAs). It also states that the necessary design, fabrication, construction, testing, and performance criteria for structures, systems, and components (SSCs) important to safety are provided to permit access and occupancy of the control room under accident conditions.

o 10 CFR 50.34, Contents of applications; technical information, requires, in part, that each applicant for a construction permit or operating license must provide an analysis and evaluation of the design and performance of SSCs of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. This regulation also requires applicants to provide a description and safety assessment of the site on which the facility is to be located, with appropriate attention to features that affect facility design.

10 CFR Part 52 governs the issuance of early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities.

o 10 CFR 52.17, Contents of applications; technical information, requires an early site permit applicant to provide a site safety analysis report that contains an analysis and evaluation of the major SSCs of the facility that bear significantly on the acceptability of the site and to evaluate the offsite radiological consequences from accidents.

o 10 CFR 52.47, Contents of applications; technical information, 10 CFR 52.79, Contents of applications; technical information in final safety analysis report, 10 CFR 52.137, Contents of applications; technical information, and 10 CFR 52.157, Contents of applications; technical information in final safety analysis report, require applicants for standard design certification, combined license, standard design approval, and manufacturing license, respectively, to provide a final safety analysis report (FSAR) that, in part, presents a safety analysis of SSCs and to evaluate the offsite radiological consequences from accidents.

DG-1425, Page 3 Related Guidance NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP) (Ref. 5), provides guidance to the NRC staff for the review of safety analysis reports submitted as part of license applications for nuclear power plants.

o SRP, Introduction, section entitled, Scope of Review of License Applications (Initial Applications and Amendments, provides guidance to the NRC staff for establishing the scope and depth of reviews, including use of risk insights, acceptance criteria, and review guidelines.

o SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, provides guidance to the NRC staff for reviewing radiological consequence analyses for LWRs using ASTs.

o SRP Section 15.0.3, Design Basis Accident Radiological Consequences of Analyses for Advanced Light Water Reactors, provides guidance to the NRC staff for reviewing radiological consequence analyses for new LWR applications, including advanced evolutionary and passive LWRs.

Purpose of Regulatory Guides The NRC issues RGs to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs are acceptable if supported by a basis for the issuance or continuance of a permit or license by the Commission.

Paperwork Reduction Act This RG provides voluntary guidance for implementing the mandatory information collections in 10 CFR Parts 50 and 52 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 3150-0011 and 3150-0151, 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 email to Infocollects.Resource@nrc.gov, and to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0011 and 3150-0151), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW, 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.

DG-1425, Page 4 TABLE OF CONTENTS A. INTRODUCTION................................................................................................................................... 1 B. DISCUSSION.......................................................................................................................................... 6 C. STAFF REGULATORY GUIDANCE.................................................................................................. 11

1.

Implementation of Accident Source Term........................................................................ 11 1.1 Generic Considerations..................................................................................................... 11 1.2 Scope of Implementation.................................................................................................. 18 1.3 Scope of Required Analyses............................................................................................. 19 1.4 Risk Implications.............................................................................................................. 22 1.5 Submittal Requirements and Information......................................................................... 23 1.6 Final Safety Analysis Report Requirements..................................................................... 24

2.

Attributes of an Acceptable Accident Source Term......................................................... 24

3.

Accident Source Term...................................................................................................... 25 3.1 Fission Product Inventory................................................................................................. 25 3.2 Release Fractions.............................................................................................................. 26 3.3 Timing of Release Phases................................................................................................. 32 3.4 Radionuclide Composition................................................................................................ 32 3.5 Chemical Form.................................................................................................................. 33 3.6 Fuel Damage in Non-LOCA DBAs.................................................................................. 33 3.7 Assessment of Radiological Consequences due to Fuel Fragmentation Relocation and Dispersal for Analysis of a 10 CFR 50.46 Large-Break LOCA......................................................... 33

4.

Dose Calculation Methodology........................................................................................ 36 4.1 Offsite Dose Consequences.............................................................................................. 36 4.2 Control Room Dose Consequences................................................................................... 37 4.3 Other Dose Consequences................................................................................................ 39 4.4 Acceptance Criteria........................................................................................................... 39

5.

Analysis Assumptions and Methodology.......................................................................... 41 5.1 General Considerations..................................................................................................... 41 5.2 Accident-Specific Assumptions........................................................................................ 43 5.3 Atmospheric Dispersion Modeling and Meteorology Assumptions................................. 43 D. IMPLEMENTATION............................................................................................................................ 46 REFERENCES........................................................................................................................................... 47 APPENDIX A: Assumptions for Evaluating the Radiological Consequences of Light-Water Reactor Maximum Hypothetical Loss-of-Coolant Accidents...................................................... A-1 APPENDIX B: Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident.......................................................................................................................... B-1 APPENDIX C: Assumptions for Evaluating the Radiological Consequences of a Boiling-Water Reactor Rod Drop Accident......................................................................................................... C-1 APPENDIX D: Assumptions for Evaluating the Radiological Consequences of a Boiling-Water Reactor Main Steamline Break Accident..................................................................................... D-1 APPENDIX E: Assumptions for Evaluating the Radiological Consequences of a Pressurized-Water Reactor Steam Generator Tube Rupture Accident.......................................................... E-1 APPENDIX F: Assumptions for Evaluating the Radiological Consequences of a Pressurized-Water Reactor Main Steamline Break Accident......................................................................... F-1 APPENDIX G: Assumptions for Evaluating the Radiological Consequences of a Pressurized-Water Reactor Locked Rotor Accident...................................................................................... G-1 APPENDIX H: Assumptions for Evaluating the Radiological Consequences of a Pressurized-Water Reactor Control Rod Ejection Accident......................................................................... H-1

DG-1425, Page 5 APPENDIX I: Analytical Technique for Calculating Fuel-Design or Plant-Specific Steady-State Fission Product Release Fractions for Non-Loss-of-Coolant Accident Events................I-1 APPENDIX J: Acronyms......................................................................................................................... J-1 APPENDIX K: Knowledge Management Information............................................................................ K-1

DG-1425, Page 6 B. DISCUSSION Reason for Revision1 This revision (Revision 2) updates several areas of guidance and supports the Increased Enrichment rulemaking2 to facilitate the use of LWR fuel containing uranium enriched to greater than 5.0 and less than 20.0 weight percent uranium (U)-235, as stated in 10 CFR Part 50.KM-01 This guidance supports effective and efficient licensing actions for LWRs, including increased enrichment applications, by providing revised methods in three main areas:

(1) accounting for recent research and modeling of source terms (2) adopting a graded, risk-informed, and performance-based control room acceptance criteria framework (3) clarifying and, where appropriate, adding flexibility in the evaluation of radiological consequences associated with DBAs Specific to item 1, the NRC applied research results to update the LWR source terms for fuel enrichments up to 8 weight percent U-235 for pressurized-water reactors (PWRs) and up to 10 weight percent U-235 for boiling water reactors (BWRs). The updated source terms reflect the current understanding of severe accidents and fission product behavior since the publication of RG 1.183, Revision 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, in October 2023 (Ref. 6), and are appropriate for LWR fuels with burnups up to 80 GWd/MTU. Additionally, the staff applied research and analyses to develop mechanistic transport models to enable BWR applicants and licensees to credit the removal of fission products released during an accident through suppression pool scrubbing. Lastly, the staff provides guidance for evaluating the consequences of fuel fragmentation, relocation, and dispersal (FFRD).

Specific to item 2, this new guidance facilitates development of a licensee-specific design criterion based on a proposed range of acceptable values that are founded on regulatory precedents and recommendations from national and international radiation protection organizations.

Specific to item 3, updates to the RG include (1) further guidance on the DBA maximum hypothetical accident (MHA) LOCA,3 (2) a discussion of the use of source term in other regulatory areas (e.g., normal operating source terms used to evaluate design features for effluent processing and source terms based on 0.25-1 fuel defects used to determine the adequacy of shielding and ventilation design features), (3) clarifications of certain basic regulatory requirements such as diversity, redundancy, defense in depth, and margin to safety, (4) clarifications of the use of sensitivity analyses, (5) inclusion of language to allow for facility-specific control room occupancy factors, (6) high-level guidance on the use 1

Throughout this guide, the staff uses KM endnotes to point the reader to corresponding background information in Appendix K, Knowledge Management Information, to this RG.

2 Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, Docket ID NRC-2020-0034, RIN 3150-AK79. More information is available at https://www.regulations.gov/docket/NRC-2020-0034.

3 As described in a footnote to 10 CFR 50.67, the MHA (also referred to as the maximum credible accident) is that accident whose consequences, as measured by the radiation exposure of the surrounding public, would not be exceeded by any other accident whose occurrence during the lifetime of the facility would appear to be credible. The MHA LOCA, as used in this guide, refers to a loss of core cooling resulting in substantial meltdown of the core with subsequent release into containment of appreciable quantities of fission products. These evaluations assume containment integrity with offsite hazards evaluated based on design basis containment leakage.

DG-1425, Page 7 of best-estimate-plus-uncertainty analyses, and (7) inclusion of appendix K to provide information on changes in this revision.

Revision 1 of RG 1.183 addressed issues identified since the guide was originally issued in 2000.

These include (1) using the term MHA LOCA to clarify the accident that the staff finds acceptable to use to meet the description in the applicable regulations in section A above, with a clear delineation between source term assumptions and plant response,4 (2) adding transient release fractions from empirical data from in-pile, prompt power pulse test programs and analyses from several international publications of fuel rod performance under prompt power excursion conditions, (3) revising steady-state release fractions for accidents other than the LOCA based on a revision to the American National Standards Institute/American Nuclear Society Standard 5.4, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel, issued May 2011 (Ref. 8), (4) adding information to acknowledge that the RG may provide useful information for satisfying the radiological dose analysis requirements in 10 CFR Part 50 and 10 CFR Part 52 for new LWR applicants, including advanced evolutionary and passive LWR design and siting, (5) providing additional guidance for modeling BWR main steam isolation valve (MSIV) leakage, (6) adding guidance for ATF, high-burnup fuel, and increased enrichment source term analyses, (7) revising transport and decontamination models for the fuel handling DBA, (8) adding guidance for crediting holdup and retention of MSIV leakage within the main steamlines and condenser for BWRs, and (9) providing additional guidance on meteorological assumptions.

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Background===

An accident source term is intended to represent a major accident involving significant core damage not exceeded by that from any other credible accident. NRC staff experience in reviewing license applications has indicated the need to consider other accident sequences of lesser consequence but higher probability of occurrence. Facility-analyzed DBAs are not intended to be actual event sequences; rather, they are intended to be surrogates to enable deterministic evaluation of the response of engineered safety features (ESFs). These accident analyses are intentionally conservative to compensate for known uncertainties in accident progression, fission product transport, and atmospheric dispersion.

Probabilistic risk assessments (PRAs) and severe accident analyses provide useful insights into system performance and suggest changes in how the desired defense in depth is achieved. The 1985 Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants (Severe Reactor Accident Policy Statement) (50 FR 32138; August 8, 1985) describes the policy related to accidents more severe than the DBAs (Ref. 9). The Severe Reactor Accident Policy Statement recognizes the usefulness of PRAs in identifying severe accident vulnerabilities and providing additional insights to ensure that nuclear power plants do not pose undue risk to public health and safety. The Commissions Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities policy statement (60 FR 42622; August 16, 1995) calls for the use of PRA technology in all regulatory matters in a manner that complements the agencys deterministic approach and supports the traditional defense-in-depth philosophy, which continues to be an effective way to account for uncertainties in equipment and human performance (Ref. 10). Furthermore, the PRA policy statement encourages the use of PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) in regulatory matters, where practical within the bounds of the state of the art, to reduce unnecessary conservatisms in current regulatory requirements, RGs, license commitments, and staff practices. Staff Requirements 4

The MHA should be modeled with the deterministic substantial fuel melt source term being injected or overlaid into the radiological consequence analysis, notwithstanding the operation of safety-related equipment designed to preclude significant fuel failure. The purpose of this approach would be to test the adequacy of the containment and other safety-related systems. Safety-related systems may be credited as described in Regulatory Position 5.1.2, as this designation ensures reliability in performing their safety function.

DG-1425, Page 8 Memorandum (SRM)-SECY-98-144, Staff RequirementsSECY-98-144White Paper on Risk-Informed and Performance-Based Regulations, dated March 1, 1999, defines the terms and Commission expectations for risk-informed and performance-based regulation (Ref. 11).

In 1995, the NRC published NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants (Ref. 12). NUREG-1465 presents a representative accident source term for a BWR and for a PWR, which provides a more realistic portrayal of the amount of fission products present in containment following an accident than previously assumed in accident analyses. These source terms are characterized by the composition and magnitude of the radioactive material, the chemical and physical properties of the material, and the timing of the release to the containment. Because the source term was provided in terms of times and rates of appearance of radioactive fission products into containment as opposed to an instantaneous release of a fraction of radionuclide classes of concern, as had been done up to that point, the NUREG-1465 source term could be applied to the design of new and advanced LWRs.

The NRC staff considered the applicability of the revised source terms to operating reactors and determined that the current analytical approach, based on the source term in Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, issued March 1962 (Ref. 13), would continue to be adequate to protect public health and safety. Operating reactors licensed under that approach would not be required to reanalyze accidents using the revised source terms. However, the NRC staff determined that some operating reactor licensees might ask to use an AST in analyses to support cost -beneficial licensing actions.

The NRC staff therefore initiated several actions to provide a regulatory basis for operating reactor licensees to use an AST5 in design-basis analyses. These initiatives resulted in the development and issuance of the final rule for 10 CFR 50.67 (64 FR 71990; December 23, 1999) and the subsequent issuance of RG 1.183, Revision 0, in July 2000, as implementing guidance for the rule (Ref. 14 and 15).

In 2022, the Commission approved the staffs plan to initiate rulemaking to amend requirements for the use of LWR fuel containing uranium enriched to greater than 5.0 weight percent U-235 in SRM-SECY-21-0109, Rulemaking Plan on Use of Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, dated March 16, 2022 (Ref. 16). One of the issues addressed by the Increased Enrichment rulemaking is the control room design criteria that is used to evaluate the design of certain structures, systems and components important to fission product mitigation during design basis accidents. The approach to amending the control room design criteria and developing this guidance uses terms and concepts described in item 8, Risk-Informed, Performance-Based Approach, of SRM-SECY-98-144, which states that a risk-informed, performance-based approach to regulatory decision-making combines the risk-informed and performance-based elements, and goes on to state the following:

[the approach] applies these concepts to NRC rulemaking, licensing, inspection, assessment, enforcement, and other decision-making. A risk-informed, performance -based regulation is an approach in which risk insights, engineering analysis and judgment including the principle of defense-in-depth and the incorporation of safety margins, and performance history are used, to (1) focus attention on the most important activities; (2) establish objective criteria for evaluating performance; (3) develop measurable or calculable parameters for monitoring system and licensee performance; (4) provide flexibility to determine how to meet the established performance criteria in a way that will encourage and reward improved outcomes; and (5) focus on the results as the primary basis for regulatory decision-making.

5 The NUREG-1465 source terms have often been referred to as the revised source terms. In recognition that additional source terms may be identified in the future, 10 CFR 50.67 addresses alternative source terms. This RG endorses a source term derived from SAND2023-01313 and provides guidance on the acceptable attributes of other ASTs.

DG-1425, Page 9 Using these concepts, the staff developed a graded, risk-informed, and performance-based framework for the control room design criteria. NRC memorandum, Method for a Graded Risk-Informed Performance-Based Control Room Design Criteria Framework, issued July 2024 (Ref. 17), provides the methodology and technical basis for the graded, risk-informed, and performance-based framework described in this RG.

Several RGs and SRP chapters describe the NRCs traditional methods for calculating the radiological consequences of DBAs. The staff developed that guidance to be consistent with the TID-14844 source term and the whole-body and thyroid dose guidelines stated in 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance (Ref. 18). Many of those analysis assumptions and methods in that guidance are inconsistent with the AST and with the total effective dose equivalent (TEDE) criteria in 10 CFR 50.34, 10 CFR Part 52, and 10 CFR 50.67. This guide provides assumptions and methods that are acceptable to the NRC staff for performing design-basis radiological analyses using an AST.

RG 1.89, Revision 2, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, issued April 2023 (Ref. 19), offers guidance for equipment environmental qualification (EQ) using an AST. Reactor licensees licensed under 10 CFR Part 50 or applicants for licenses under 10 CFR Part 50 or 10 CFR Part 52 should use the applicable guidance in RG 1.89, Revision 2.

The NRC anticipates that, over time, licensees will adopt this revision of RG 1.183 because it provides modern consequence analysis approaches with greater modeling and operational flexibilities.

This was the NRCs experience with the introduction of the AST through RG 1.183, Revision 0, because the AST provided greater flexibilities when compared to legacy accident analysis guidance. However, each revision of this guide provides methods acceptable to the NRC for compliance with the applicable regulations as specified in the guidance. Once previous revisions of this guide become unnecessary because, for example, licensees voluntarily implement the guidance of more recent revisions, the staff may withdraw the earlier revisions, as justified.6 Since the regulatory positions in each revision of this guide were evaluated for applicability corresponding to each revision, combining regulatory positions and assumptions from multiple revisions of this guide, while potentially acceptable, will need appropriate technical justification and may require additional NRC staff review before receiving approval. For example, if a PWR licensee wants to increase the rod average burnup limit to 80 gigawatt-days-per metric ton uranium (GWd/MTU) and use fuel having enrichments of 8 weight-percent U-235 (technological advances covered by revision 2 of this guide) using guidance in both Revisions 0 and 1 of RG 1.183, then the licensees application will need to include appropriate technical bases for the applicability of that guidance to the licensees situation. These technical bases should include a justification of why continued use of the source terms in those revisions, or other source terms used in the application, are appropriate, including how the proposed source terms satisfy the attributes of an acceptable accident source term, as provided in Regulatory Position 2 of this guide.

6 Withdrawal of an RG, or a revision of an RG, means that the guidance no longer provides useful information or has been superseded by other guidance, technological innovations, congressional actions, or other events. The withdrawal of an RG, or a revision of an RG, requires the staff to consider backfitting in its decision-making; however, it does not alter any existing NRC licensing approvals or the acceptability of licensee commitments to them. However, withdrawn RGs, or revisions of RGs, should not be used in future requests or applications for NRC licensing actions without proper justification.

DG-1425, Page 10 Consideration of International Standards The International Atomic Energy Agency (IAEA) works with member states and other partners to promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety Requirements and Safety Guides for protecting people and the environment from harmful effects of ionizing radiation. This system of safety fundamentals, safety requirements, safety guides, and other relevant reports reflects an international perspective on what constitutes a high level of safety. To inform its development of this RG, the NRC considered IAEA Safety Requirements and Safety Guides pursuant to the Commissions International Policy Statement (Ref. 20) and Management Directive and Handbook 6.6, Regulatory Guides (Ref. 21).

The following IAEA Safety Requirements and Guides were considered in the development of this RG:

IAEA Specific Safety Guide (SSG), No. SSG-2, Revision 1, Deterministic Safety Analysis for Nuclear Power Plants, issued 2019 (Ref. 22). The NRC considers this RG to provide traditional deterministic radiological consequence analysis concepts and methodologies for DBAs similar to those conservative and combined best-estimate/conservative approaches described in SSG-2, Revision 1.

IAEA Safety Standard General Safety Requirements Part 7, Preparedness and Response for a Nuclear or Radiological Emergency, issued 2015 (Ref. 23). The NRC considers this RG to be consistent with the level of safety provided in Part 7.

DG-1425, Page 11 C. STAFF REGULATORY GUIDANCE This section and the appendices to this RG contain regulatory positions that establish a method acceptable to the NRC staff for complying with regulations for DBA dose consequence analysis using an AST.

This RG applies to MHA-LOCA models for applicants and licensees using zirconium-alloy cladded uranium dioxide (UO2) fuel rod designs within the following scope of applicability:

a.

The guidance applies to reactor core burnups up to a maximum rod-average of 80 GWd/MTU.

b.

The guidance applies to fuel enrichments up to 8 weight percent U-235 for PWRs and up to 10 weight percent for BWRs.

c.

The guidance applies to currently approved (as of the issuance date of this RG revision) fuel and cladding materials (e.g., coated zirconium alloy claddings and doped fuels). The guidance also applies to fuels with iron-chromium-aluminum (FeCrAl) alloy cladding (Ref. 24) and chromium-coated cladding (Ref. 25).

This RG applies to non-LOCA models for applicants and licensees within the following scope of applicability:

a.

The guidance applies to reactor core burnups up to a maximum rod-average burnup of 80 GWd/MTU.

b.

The guidance applies to fuel enrichments up to 8 weight percent U-235 for PWRs and up to 10 weight percent for BWRs.

c.

Regarding cladding material and fuel types, the guidance applies to currently approved (as of the issuance of this RG revision) zirconium-alloy cladded UO2 fuel rod designs at power levels below the burnup-dependent power envelopes depicted in figure 1 of this guide.

For fuel rod designs outside of the scope of applicability, appendix I provides an acceptable technique for calculating maximum steady-state non-LOCA release fractions.

1.

Implementation of Accident Source Term 1.1 Generic Considerations As used in this guide, the AST is an accident source term derived principally from SAND2023-01313; it differs from the SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, issued January 2011 (Ref. 26), TID-14844, and NUREG-1465 source terms used in the original and revised design and licensing of operating reactor facilities. ASTs may also be used by applicants for new LWRs, including advanced evolutionary and passive LWRs, under 10 CFR Part 50 and 10 CFR Part 52, and for existing operating reactor licensees under 10 CFR 50.34 and 10 CFR 50.67. This guide identifies an AST that is acceptable to the NRC staff and describes significant characteristics of other source terms that may be found acceptable. While the staff recognizes several potential uses of an AST, it is not possible to foresee all possible uses. Licensees may pursue technically justifiable uses of the ASTs in the most flexible manner in license amendment requests so long as these uses are compatible with maintaining a clear, logical, and consistent design basis and continue to comply with NRC regulations. These license amendment requests should demonstrate that the facility, as modified, will continue to provide sufficient safety margins, with adequate defense-in-

DG-1425, Page 12 depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameter inputs.

1.1.1 Evaluation of Defense in Depth and Safety Margins One aspect of the engineering evaluation is to show that the proposed change does not compromise the fundamental safety principles that are the basis of plant design and operation (i.e., activities such as maintenance, testing, inspection, and qualification).KM-02 During the design process, plant response and associated safety margins are evaluated using assumptions of physical properties and operating characteristics. National standards, the defense-in-depth philosophy, and the general design criteria in Appendix A to 10 CFR Part 50 are additional engineering considerations that influence plant design and operation.

RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Ref. 27), provides important discussions of the fundamental principles and elements of evaluating defense in depth and safety margin. For consistency between regulatory documents, these principles and elements are also covered in this RG in the context of license amendment requests pertinent to DBA radiological consequence analyses.

A licensees proposed change might affect safety margins and defense in depth incorporated into the current plant design and operation; therefore, the licensee should reevaluate the safety margins and layers of defense to support the proposed change. As part of this evaluation, the licensee should determine the impact of the proposed licensing-basis change on the functional capability, reliability, and availability of affected equipment. The plants licensing basis is the reference point for judging whether a proposed change unacceptably affects safety margins or defense in depth. Regulatory Guidance Positions 1.1.1.1 and 1.1.1.2 address the assessment of whether the proposed change maintains sufficient safety margins and remains consistent with the defense-in-depth philosophy.

1.1.1.1 Safety Margins This RGs methodology of radiological consequence analyses uses a system of coupled, first-order differential equations to describe the behavior of the source term traveling from the containment and other plant systems, to the environment and eventually to a dose receptor. The methodology bounds the complexities of reality for an engineered system. The models can be simplistic or very complex, ranging from a few dozen to several hundred input parameters. These input parameters not only model the design of the engineered system itself but other aspects as well, such as meteorological conditions, human performance, and health effects of exposure to ionizing radiation, various physical processes to mitigate the source term, and representative testing data. As such, the uncertainties in these models can be quite large, and results are typically presented as point estimates (i.e., without uncertainty characterizations).

A significant consideration in these radiological consequence models is the inclusion of safety factors placed on input parameters to achieve a conservative model and account for uncertainty. Safety factors conservatively bias the results to bound inherent uncertainties that exist in real-world engineering applications. These uncertainties can stem from various sources, such as variations in environmental conditions, operator actions, unexpected system failures, or unforeseen events. This conservatism ensures that the system not only meets its intended functions, but also maintains a high level of reliability to withstand conditions that may exceed original design expectations.

Licensees should evaluate their proposed uses of this guide, and the associated proposed facility modifications and changes to procedures, to determine whether the proposed changes are consistent with the principle that sufficient safety margins are maintained, including a margin to account for analysis

DG-1425, Page 13 uncertainties. The safety margins are products of specific values and limits in the technical specifications (TS) (which cannot be changed without NRC approval) and other values, such as assumed accident or transient initial conditions or assumed safety system response times. For example, changes, or the net effect of multiple changes, that result in a reduction in safety margins may require prior NRC approval. If the initial AST implementation, consistent with the guidance in this RG, is approved by the staff and becomes part of the facility design basis, licensees may use 10 CFR 50.59, Changes, tests and experiments, and its supporting guidance to assess facility modifications and changes to procedures that are described in the updated FSAR.

When the computed radiological consequence results are close to the established acceptance criteria, a margin analysis can be helpful in assessing the level of conservatism in the model. For instance, licensees can systematically evaluate how the variations or sensitivities in select input parameters and associated safety factors can lead to results that fall either above or below the established acceptance criteria. Additionally, licensees can assess the analytical margin included in the model by increasing the model fidelity. This is done by more realistically assessing simplifying modeling assumptions of highly complex engineered systems. Either approach is a valuable exercise as it aids in quantifying the models robustness and sensitivity to key variables, while also helping to identify the amount of margin inherent in the model.

In a margin analysis, it is acceptable for results to occasionally exceed the analysis of record results, provided they are within a certain acceptable range when balanced by conservatisms provided by other parameter and modeling assumptions. This is because real-world systems often encounter uncertainties, changing conditions, or unexpected events that may lead to degraded performance. To demonstrate that recalculations are not necessary, the margin analysis should be accompanied with sufficient explanation and justification to support the analysis of record by explaining how the model is robust and contains sufficient conservatisms. A margin analysis helps ensure that the analysis of record is robust enough to withstand uncertainties or unexpected variations, while avoiding unnecessary overdesign or detailed reanalysis. Likewise, by understanding and quantifying the sensitivities affecting acceptable margins, decision-makers gain valuable insights into the systems behavior and can make informed choices. If a margin analysis is performed to help demonstrate compliance, a discussion of the assessed parameters, their applied safety factor with justification, and impact on the results should be submitted for review by the staff.

1.1.1.2 Defense in Depth Defense in depth is an element of the NRCs safety philosophy that employs successive measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused event occurs at a nuclear facility. The defense-in-depth philosophy has traditionally been applied in plant design and operation to provide multiple means to accomplish safety functions and prevent the release of radioactive material. It has been and continues to be an effective way to account for uncertainties in equipment and human performance and to account for the potential for unknown and unforeseen failure mechanisms or phenomena that, because they are unknown or unforeseen, are not reflected in either the PRA or traditional engineering analyses. SRM-SECY-98-144 provides additional information on defense in depth as an element of the NRCs safety philosophy.

When the NRC issues an operating or combined license, the staff ensures that the facility as described in the FSAR meets the Commissions expectations for defense in depth. However, when a licensee adopts an AST, using this RG, it may propose changes to the facility (e.g., design, operating procedures, dose acceptance criteria for the control room) that may adversely impact defense in depth.

Therefore, an applicant who adopts this RG should evaluate whether the facilitys defense in depth continues to be adequate with the proposed changes.

DG-1425, Page 14 Revision 3 of RG 1.174 lists seven considerations in evaluating proposed changes that may impact defense in depth:

1.

Preserve a reasonable balance among the layers of defense.

2.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

3.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

4.

Preserve adequate defense against potential common-cause failure.

5.

Maintain multiple fission product barriers.

6.

Preserve sufficient defense against human errors.

7.

Continue to meet the intent of the plants design criteria.

The applicant should examine the applicable items from the above list and explain how, if the proposed changes are approved, the facilitys defense in depth would continue to be adequate. For example, an applicant who wishes to use control room dose acceptance criteria above 10-rem-TEDE should demonstrate how changes made to the facility in response to regulatory initiatives such as 10 CFR 50.160, Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities, and Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities10 CFR 50.54(f), dated November 23, 1988, ensure the robustness of the containment for station blackout sequences for those plant designs with containments determined to possess risk outliers associated with extended loss of emergency alternating current (AC) or loss of heat sink scenarios (Ref. 28). Likewise, control room habitability and occupancy during severe accidents, such as extended loss of emergency AC power, should be examined within the Commissions comprehensive radiation protection and emergency planning framework to ensure adequate protection of workers during actions taken to maintain the reactor in a safe condition.

1.1.2 Integrity of Facility Design Basis The DBA source term used for dose consequence analyses is a fundamental assumption and the basis for much of the facility design. Additionally, many aspects of an operating reactor facility are derived from the radiological design analyses that incorporated the TID-14844 accident source term.

Although a complete reassessment of all facility radiological analyses would be desirable, the NRC staff determined that recalculation of all design analyses for operating reactors would generally not be necessary. Regulatory Position 1.3 of this guide contains guidance on which analyses should be updated as part of the AST implementation submittal and which may need to be updated in the future as additional modifications are made.

This approach for operating reactors creates two tiers of analysesone based on the previous TID-14844 source term, and one based on an AST. The radiological acceptance criteria would also differ, as some analyses are based on whole-body and thyroid criteria, and some are based on TEDE criteria.

Full implementation of an AST revises the plant licensing basis to specify the AST in place of the previous TID-14844 accident source term and establishes the TEDE dose as the new acceptance criteria.

Selective implementation of an AST also revises the plant licensing basis and may establish the TEDE dose as the new acceptance criteria. Selective implementation differs from full implementation only in the

DG-1425, Page 15 scope of the change. In either case, the facility design bases should clearly indicate that the source term assumptions and radiological criteria in the affected analyses have been superseded and that future revisions of these analyses, if any, will use the updated approved assumptions and criteria.

Radiological analyses generally should be based on assumptions and inputs that are consistent with corresponding data used in other design-basis safety analyses unless the use of these data would result in nonconservative results or otherwise conflict with regulatory guidance.

1.1.3 Source Term ApplicationsKM-03 Use of radiological source terms in DBA radiological consequence analyses is deeply embedded in the regulatory policy and practices of the NRC, even as the licensing approaches have evolved. For LWRs, the source term refers to the magnitude and mix of the radionuclides released from damaged fuel (i.e., a substantially melted core), expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form and the timing of their release. It is based on the concept of defense in depth in which power plant design, operation, siting, and emergency planning make up independent layers of nuclear safety. This approach encourages nuclear plant designers to incorporate several lines of defense to maintain the effectiveness of physical barriers between radiation sources and materials from workers, members of the public, and the environment. It centers on the concept of DBAs, assessment of which aims to determine the effectiveness of each line of defense. The DBAs establish and confirm the design basis of the nuclear facility, including its safety-related SSCs and items important to safety, ensuring that the plant design meets the safety and numerical radiological criteria stated in regulations and subsequent guidance. From this foundation, specific safety requirements have evolved through a number of criteria, procedures, and evaluations, as reflected in regulations, RGs, standard review plans, TS, license conditions, and various regulatory technical information documents.

Compliance with the NRCs requirements ensures that radioactive material is used safely and securely. For nuclear power plants, the NRC requires applicants and licensees to demonstrate through analysis that the reactors ESFs will prevent or mitigate a release of radioactive material from the reactor for a range of accidents. Some of these accident analyses assume that the fission product barriers provided by the fuel cladding and reactor coolant system boundaries are breached to evaluate the adequacy of containment structures and systems designed to mitigate the release of radioactive material. The NRC further requires applicants and licensees to demonstrate that, if there were a release from the reactor fuel to the environment, the resulting doses to members of the public and workers would be below the NRCs regulatory criteria. For this demonstration, licensees have historically used radiological source terms provided in NRC regulatory guidance. For advanced reactor designs, license applicants are developing design-specific source terms. The NRCs Nuclear Power Reactor Source Term web page (https://www.nrc.gov/reactors/new-reactors/advanced/references/nuclear-power-reactor-source-term.html) provides information to aid in developing advanced reactor source terms, and Regulatory Position 2 of this RG includes information on the attributes of an acceptable accident source term.

Additionally, for nonpower reactors, information on radiation source terms is provided in documents such as NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, issued February 1996 (Ref. 29), and Safety Reports Series No. 53, Derivation of the Source Term and Analysis of the Radiological Consequences of Research Reactor Accidents, issued 2008 (Ref.

30); and on the NRCs Non-Power Facilities web page (https://www.nrc.gov/reactors/non-power.html).

DG-1425, Page 16 1.1.3.1 Various Radiological Source Terms Used by the NRC The SRP uses various source terms for a variety of purposes, including the following:

a.

The normal operational source term, based on operational reactor experience as described in American National Standards Institute/American Nuclear Society N18.1, is addressed in SRP Section 11.1, Coolant Source Terms, for reactor coolant (primary and secondary) and reactor steam design details and in SRP Section 11.2, Liquid Waste Management System, and SRP Section 11.3, Gaseous Waste Management System, for system design features used to process and treat liquid and gaseous effluents before they are released or recycled (Ref. 31).

b.

The anticipated operational occurrence source term, based on the TS or the design-basis source term, whichever is more limiting, is used to determine the effects of events like primary to secondary leaks and BWR steam leaks. SRP Section 11.1 addresses this for reactor coolant (primary and secondary) and reactor steam design details.

c.

Design-basis source terms, based on 0.25-1 percent fuel defects, as described in SRP Section 12.2, Radiation Sources, are used to determine the adequacy of shielding and ventilation design features.

d.

Post-accident shielding source term (for vital area access, including work in the area), as described in SRP Section 12.2, is based on guidance from NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980 (Ref. 32), item II.B.2; RG 1.89; or this guide.

e.

The EQ source term may or may not be more limiting than the stated accident source term.

Further guidance in developing source terms used to assess dose and dose rates to equipment is addressed in SRP Section 3.11, Environmental Qualification of Mechanical and Electrical Equipment; SRP Section 12.2; and RG 1.89.

f.

Accident source terms for non-MHA-LOCA accidents, based on design-basis events, are addressed in SRP Chapter 15, Transient and Accident Analysis. These analyses are generally used to develop TS and limiting conditions for operation, consistent with 10 CFR 50.36, Technical specifications.

g.

The MHA-LOCA source term, used to demonstrate compliance with reactor site criteria and safety analysis requirements (10 CFR Part 100 and 10 CFR 50.34), is required for the purposes of licensing nuclear power plants to represent radionuclide releases to reactor containments associated with a substantial meltdown of the reactor core. The consequences of these radionuclide releases are evaluated, assuming that the containment remains intact and leaks occur at design-basis leak rates, as described in the TS. Radionuclides that leak from the containment are termed the radiological release to the environment. The magnitude of the radiological release to the environment can be estimated from the containment leak rate and the radionuclide inventory suspended in the containment atmosphere as a function of time. The radionuclide inventory suspended in the containment atmosphere depends on the amount released to the containment, as well as the effectiveness of natural and engineered processes that lead to radionuclide deposition and retention within containment. The postulated radionuclide release to the containment is termed the in-containment source term.

DG-1425, Page 17 1.1.3.2 Emergency Preparedness ApplicationsKM-04 The regulations in 10 CFR 50.47, Emergency plans, include the requirements for emergency preparedness at nuclear power plants. Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, to 10 CFR Part 50 states additional requirements. These regulations require the plume exposure pathway emergency planning zone (EPZ) to consist of an area of about 10 miles (16 kilometers) in radius for water-cooled reactors with an authorized power level greater than 250 megawatts thermal (MWt). The EPZ size may be determined on a case-by-case basis for water-cooled reactors with an authorized power level less than 250 MWt. NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants, issued December 1978 (Ref. 33), includes the planning basis for many of these requirements.7 This joint effort by the U.S. Environmental Protection Agency (EPA) and the NRC considered the principal characteristics (such as nuclides released and distances) likely to be involved in a spectrum of design-basis and severe (core melt) accidents. Specifically, the technical basis for the 10-mile plume exposure EPZ was based on evaluation of the offsite consequences of the DBA, less severe beyond-DBAs, and less probable but more severe beyond-DBAs for large LWRs. The estimated radiological consequences of these accidents were then compared to the guidance in EPA-520/1-75-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, issued September 1975 (Ref. 35). The DBAs and less severe accidents were those in which the containment is expected to be intact. In more severe accidents, the containment is expected to either fail or be bypassed. Each of these three accident classes were assessed against the three dose-based figures-of-merit. However, no single accident scenario was used as the basis of the required emergency preparedness. The objective of the planning is to provide public protection encompassing a spectrum of events, with a sufficient basis for extension of response efforts for unanticipated events. The NRC and EPA issued these requirements after a long period of involvement with many stakeholders, including the Federal Emergency Management Agency, other Federal agencies, local and State governments (and in some cases foreign governments), private citizens, utilities, and industry groups.

Light-water small modular reactors may choose to implement a performance-based emergency preparedness framework under 10 CFR 50.160, which includes the determination of a facility-specific plume exposure pathway EPZ in accordance with the criteria in 10 CFR 50.33(g)(2)(i). Appendix A to RG 1.242, Performance-Based Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities (Ref. 36), provides a general methodology for establishing plume exposure pathway EPZ size. In RG 1.242, the guidance on event selection for the EPZ sizing analysis states that for LWR power reactors, the events should include the design-basis events, DBAs, and beyond-design-basis events evaluated in SRP Chapter 15 and SRP Chapter 19, Severe Accidents. Light-water small modular reactor applicants may use the guidance in this guide (i.e.,

RG 1.183) that applies to their design to calculate the radiological consequences of DBAs as part of the plume exposure pathway EPZ sizing analysis.

Although the NRC based the AST in this guide on a limited spectrum of severe accidents, the characteristics are tailored specifically for use in DBA analysis. The AST is not representative of the spectrum of events that make up the planning basis of emergency preparedness. Therefore, the AST is insufficient by itself as a basis for determining the plume exposure pathway EPZ for applicants and licensees under 10 CFR 50.160, for requesting a case-by-case review under 10 CFR 50.33(g)(1), or for requesting, through an exemption, an EPZ less than the 10 miles required by 10 CFR 50.47 and Appendix E to 10 CFR Part 50.

7 NUREG-0654/FEMA-REP-1, Revision 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, issued November 1980 (Ref. 34), also addresses this planning basis.

DG-1425, Page 18 This guidance does not, however, preclude the appropriate use of insights based on the AST in establishing emergency response procedures, such as those associated with emergency dose projections, protective measures, and severe accident management guides.

1.1.4 Applicability to New Light-Water Reactor Applications, Including Advanced Evolutionary and Passive Designs The NRC originally created RG 1.183 for use by existing nuclear power reactors to satisfy regulations under 10 CFR 50.34 and 10 CFR 50.67. Revision 1 of RG 1.183 extended the applicability of the RG for use by new LWR power reactors, including advanced evolutionary and passive LWR designs, in satisfying the radiological dose analysis requirements in 10 CFR Part 50 and 10 CFR Part 52 for safety and siting analyses. New LWR applicants and licensees may use the guidance in this RG that is applicable to their design to meet the accident radiological consequence analysis requirements in 10 CFR Part 50 or 10 CFR Part 52 for permits, licenses, approvals, or certifications. To review these applications, the staff will use the methodology and assumptions stated in this RG that apply to the design.

1.2 Scope of Implementation The AST described in this guide is characterized by radionuclide composition and magnitude, the chemical and physical form of the radionuclides, and the timing of their release. The accident source term is a fundamental assumption and the basis of much of the facility design.

For operating reactors to which 10 CFR 50.67 applies, a complete implementation of an AST would upgrade all existing radiological analyses and would consider the impact of all five characteristics of a source term as defined in 10 CFR 50.2, Definitions. However, the NRC staff has determined that there could be implementations for which this level of reanalysis may not be necessary. For holders of operating licenses, as defined in the applicability section of 10 CFR 50.67, two categories of AST implementation are defined: full and selective. Regulatory Positions 1.2.1 and 1.2.2 describe these categories.

For the radiological consequence analyses for new LWR permit, license, approval, or certification applications (e.g., those under 10 CFR 50.34(a)(1) or 10 CFR Part 52), the accident source term should consider all characteristics of a source term as defined in 10 CFR 50.2 and detailed in Regulatory Position 2 of this guide. Full and selective implementations, as used in the regulatory positions that follow, do not apply to new reactor applicants.

1.2.1 Full Implementation Full implementation is a modification of the facility design basis that addresses all characteristics of the AST: specifically, the composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release. Full implementation revises the plant licensing basis to specify the AST in place of the previous accident source term and establishes the TEDE as the new acceptance criterion. This applies not only to the analyses performed in the application (which may include only a subset of the plant analyses), but also to all future design-basis analyses. A full implementation is considered when, at a minimum, the application includes a reanalyzed MHA LOCA using the guidance in appendix A to this guide. Regulatory Position 1.3 in this guide provides additional guidance on the analysis. Since the AST and the TEDE criteria will become part of the facility design basis, new applications of the AST will not require prior NRC approval unless the new application involves a change to a TS, or unless the licensees 10 CFR 50.59 evaluation concludes that prior NRC approval is required. However, a change from an approved AST to a different AST that is not approved for use at that facility would require a license amendment under 10 CFR 50.67.

DG-1425, Page 19 1.2.2 Selective Implementation Selective implementation is a modification of the facility design basis that (1) is based on one or more of the characteristics of the AST or (2) entails reevaluation of a limited subset of the design-basis radiological analyses. The NRC staff will allow licensees flexibility in adopting technically justified selective implementations, if they maintain a clear, logical, and consistent design basis. An example of an application of selective implementation would be one in which a licensee desires to use the release timing insights of the AST to increase the required closure time for a containment isolation valve by a small amount. Another example would be a request to remove the charcoal filter media from the spent fuel building ventilation exhaust. In the latter example, the licensee may need to reanalyze only DBAs that credited the iodine removal performed by the charcoal media. Regulatory Position 1.3 of this guide provides additional analysis guidance. NRC approval for the AST (and the TEDE dose criterion) will be limited to the particular selective implementation proposed by the licensee. The licensee would be able to make subsequent modifications to the facility and changes to procedures based on the selected AST characteristics incorporated into the design basis under the provisions of 10 CFR 50.59. However, the use of other AST characteristics or the use of TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics would require prior staff approval under 10 CFR 50.67. As an example, a licensee with an implementation involving only timing, such as relaxed closure time on isolation valves, could not use 10 CFR 50.59 as a mechanism to implement a modification involving a reanalysis of the MHA LOCA. However, the licensee could extend use of the timing characteristic to adjust the closure time on isolation valves not included in the original approval.

1.3 Scope of Required Analyses 1.3.1 Design-Basis Radiological Analyses For several regulatory requirements, compliance is demonstrated, in part, by the evaluation of the radiological consequences of DBAs. The current licensing basis may include, but is not limited to, the following:

a.

EQ of equipment (10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants);

b.

control room habitability (GDC 19 of Appendix A to 10 CFR Part 50);

c.

emergency response facility habitability (paragraph IV.E.8 of Appendix E to 10 CFR Part 50);

d.

AST (10 CFR 50.67);

e.

environmental reports (10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions) (Ref. 37);

f.

power reactor siting (10 CFR 100.11 for applications before January 10, 1997, and 10 CFR 100.21, Non-seismic siting criteria, which references criteria in 10 CFR 50.34(a)(1),

for subsequent applications);8 and

g.

power reactor applications for early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses (10 CFR Part 52), construction permits (10 CFR 50.34(a)), or operating licenses (10 CFR 50.34(b)).

8 For licensees that have implemented an AST, the dose criteria of 10 CFR 50.67 supersede those of 10 CFR 100.11.

DG-1425, Page 20 There may be other areas in which the TS bases and various licensee commitments refer to evaluations that use an AST. A plants licensing bases may also include, but are not limited to, the following sections of NUREG-0737:

a.

post-accident access shielding (II.B.2),

b.

post-accident sampling capability (II.B.3),

c.

accident monitoring instrumentation (II.F.1),

d.

leakage control (III.D.1.1),

e.

emergency response facilities (III.A.1.2) and

f.

control room habitability (III.D.3.4).

For applications under 10 CFR Part 52 (e.g., 10 CFR 52.47(a)(8)), 10 CFR 50.34(f) requires that each applicant for a design certification, design approval, combined license, or manufacturing license shall demonstrate compliance with the technically relevant portions of the requirements related to the accident at the Three Mile Island nuclear reactor in 10 CFR 50.34(f)(1)-(3), except for 10 CFR 50.34(f)(1)(xii), 10 CFR 50.34(f)(2)(ix), and 10 CFR 50.34(f)(3)(v). These requirements include the NUREG-0737 sections listed above.

1.3.2 Reanalysis Guidance Any full or selective implementation of an AST and any associated facility modification should be supported by an evaluation of all significant radiological and non-radiological impacts of the proposed actions. This evaluation should consider the impact of the proposed changes on the facilitys compliance with the regulations and commitments listed above, as well as any other facility-specific requirements.

These impacts may be caused by (1) the associated facility modifications or (2) the differences in the AST characteristics. The scope and extent of the reevaluation will necessarily depend on the specific facility modification proposed9 and on whether a full or selective implementation is being pursued. Licensees do not need to perform a complete recalculation of all facility radiological analyses, but licensees should evaluate all impacts of the proposed changes and update the affected analyses and design bases appropriately. The NRC considers an analysis to be affected if the proposed modification changes one or more assumptions or inputs used in the analysis, in such a way that the results, or the conclusions drawn from the results, are no longer valid. A licensee may use NRC-approved generic analyses, such as those performed by owner groups or vendor topical reports, provided that the licensee justifies the applicability of the generic conclusions to the specific facility and implementation. Sensitivity analyses, discussed below, may also be an option. If affected design-basis analyses are to be recalculated, the licensee should assess and update, as appropriate, all affected assumptions and inputs and address all selected characteristics of the AST and the TEDE criteria. A license amendment request should describe the licensees reanalysis effort and provide statements on the acceptability of the proposed implementation, including modifications, with respect to each of the applicable analysis requirements and commitments identified in Regulatory Position 1.3.1 of this guide.

The NRC staff has evaluated the impact of the AST on three representative operating reactors (Ref. 38). This evaluation determined that radiological analysis results based on the TID-14844 source term assumptions and the whole-body and thyroid methodology generally bound the results from analyses based on the AST and TEDE methodology. Licensees may use the applicable conclusions of this evaluation in addressing the impact of the AST on design-basis radiological analyses. However, this does 9

For example, a proposed modification to change the timing of a containment isolation valve from 2.5 seconds to 5.0 seconds might be acceptable without any dose calculations. However, a proposed modification that would delay containment spray actuation could involve recalculation of MHA LOCA doses, reassessment of the containment pressure and temperature transient, recalculation of sump pH, reassessment of the emergency diesel generator loading sequence, integrated doses to equipment in the containment, and more.

DG-1425, Page 21 not exempt licensees from evaluating the remaining radiological and non-radiological impacts of the AST implementation and the impacts of the associated plant modifications. For example, a selective implementation based on the timing insights of the AST may change the required isolation time for the containment purge dampers from 2.5 seconds to 5.0 seconds. This application may be acceptable without dose calculations. However, the licensee may need to evaluate the ability of the damper to close against increased containment pressure or the ability of ductwork downstream of the dampers to withstand increased stresses.

An application for full implementation is considered when, at a minimum, the application includes a reanalyzed MHA LOCA using the guidance in appendix A to this guide. The licensee should update other design-basis analyses in accordance with the guidance in this section.

A selective implementation of an AST and of any associated facility modification based on the AST should evaluate all the radiological and non-radiological impacts of the proposed actions as they apply to the particular implementation. The licensee should update design-basis analyses in accordance with the guidance in this section. There is no requirement that an MHA-LOCA analysis be performed.

The analyses performed should address all impacts of the proposed modification, the selected characteristics of the AST, and if dose calculations are performed, the TEDE criteria. For selective implementations based on the timing characteristic of the AST (e.g., a change in the closure timing of a containment isolation valve), reanalysis of radiological calculations may not be necessary if the modified elapsed time remains a fraction (e.g., 0.25) of the time between accident initiation and the onset of the gap release phase. Longer time delays may be considered on an individual basis. For longer time delays, it may be necessary to evaluate the radiological consequences and other impacts of the delay, such as blockage by debris in sump water. If affected design-basis analyses are to be recalculated, all affected assumptions and inputs should be assessed and updated as appropriate. Selected characteristics of the AST and the TEDE criteria should be addressed.

1.3.3 Use of Sensitivity or Scoping Analyses It may be possible to demonstrate by sensitivity or scoping evaluations that existing analyses have sufficient margin and need not be recalculated. As used in this guide, a sensitivity analysis is an evaluation that considers how the overall results vary as an input parameter (in this case, an AST characteristic) is varied, for a given set of assumptions. A scoping analysis is a brief evaluation that uses conservative, simple methods to show that the results of the analysis bound those obtainable from a more complete treatment. Sensitivity analyses are particularly applicable to suites of calculations that address diverse components or plant areas but that are otherwise largely based on generic assumptions and inputs.

Such cases might include post-accident vital area access dose calculations, shielding calculations, and equipment EQ (integrated dose). It may be possible to identify a bounding case, reanalyze that case, and use the results to draw conclusions about the remaining analyses. It may also be possible to show that, for some analyses, the whole-body and thyroid doses determined with the previous source term would bound the TEDE obtained using the AST. Where present, arbitrary designer margins may be adequate to bound any impact of the AST and the TEDE criteria. If sensitivity or scoping analyses are used, the license amendment request should discuss the analyses performed and the conclusions drawn. Scoping or sensitivity analyses should not constitute a significant part of the evaluations for the design-basis exclusion area boundary (EAB), low-population zone (LPZ), or control room dose unless a clear and defensible basis exists for their doing so. Sensitivity analyses should avoid including well-defined parameters, such as atmospheric dispersion factors, that are based on site-specific data. Sensitivity studies used for the purpose of identifying a bounding design-basis case should not vary parameters that are not a part of the licensing basis.

DG-1425, Page 22 1.3.4 Updating Analyses Following Implementation Full implementation of the AST replaces the previous accident source term with the approved AST and the TEDE criteria for all design-basis radiological analyses. The implementation may have been supported in part by sensitivity or scoping analyses that concluded that many of the design-basis radiological analyses would remain bounding for the AST and the TEDE criteria and would not require updating. After the implementation is complete, there may be a subsequent need (e.g., a planned facility modification) to revise these analyses or to perform new analyses. For these recalculations, all affected analyses should address all characteristics of the AST and the TEDE criteria incorporated into the design basis. Reevaluation using the previously approved source term may not be appropriate. Since the AST and the TEDE criteria are part of the approved design basis for the facility, use of the AST and TEDE criteria in new applications at the facility does not constitute a change in analysis methodology that would require NRC approval.10 This guidance also applies to selective implementations, to the extent that the affected analyses are within the scope of the approved implementation as described in the facility design basis. In these cases, the updated analyses should consider the characteristics of the AST and the TEDE criteria identified in the facility design basis. Use of other characteristics of the AST or TEDE criteria that are not part of the approved design basis and changes to previously approved AST characteristics require prior NRC staff approval under 10 CFR 50.67.

1.3.5 Equipment Environmental Qualification A proposed plant modification associated with the AST implementation may affect current EQ analyses. To address these impacts, the licensee should update EQ analyses that have assumptions or inputs affected by the plant modification. New facilities proposing to implement an AST should use the guidance in RG 1.89, Revision 2.

1.4 Risk Implications This guide provides regulatory assumptions that licensees should use in calculating the radiological consequences of DBAs. These assumptions have no direct influence on the probability of the design-basis initiator. These analysis assumptions cannot increase the core damage frequency (CDF) or the large early release frequency. However, facility modifications made possible by the AST could have an impact on risk. If the proposed implementation of the AST involves changes to the facility design that would invalidate assumptions made in the facilitys PRA, the licensee should evaluate the impact on the existing PRAs.

Evaluations should consider the risk impact of proposed implementations that seek to remove or adjust performance requirements of existing engineered safeguards equipment based on the reduced postulated doses. The NRC staff may request risk information if there is reason to question adequate protection of public health and safety.

10 In performing screenings and evaluations pursuant to 10 CFR 50.59, it may be necessary to compare dose results (figure of merit) expressed in terms of whole-body and thyroid with results expressed in terms of TEDE. Either figure of merit represents different systems of dosimetry. There is no methodology that converts the figures of merit between systems.

Therefore, to calculate the desired figure of merit, the appropriate dosimetry methodology (i.e., dose conversion factors) must be applied.

DG-1425, Page 23 The licensee may elect to use risk insights in support of proposed changes to the design basis that are not addressed in currently approved NRC staff positions. Guidance on appropriate use of risk insights appears in RG 1.174, Revision 3.

1.5 Submittal Requirements and Information According to 10 CFR 50.90, an application for a license amendment must fully describe the changes desired and should follow, as far as applicable, the form prescribed for original applications.

RG 1.70, Revision 3, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition), issued November 1978 (Ref. 39), provides additional guidance. For license amendment requests to revise the accident source term used in design-basis radiological consequence analyses, 10 CFR 50.67 states that the NRC may issue the amendment only if the applicants analysis demonstrates with reasonable assurance that the identified dose criteria are met. The licensees analyses, which will become part of the design and licensing basis of the facility, are key aspects of the submittal information.

This information submitted in the amendment request, along with the current plant design as documented in the FSAR, staff safety evaluation reports, regulatory guidance, other licensee commitments, and staff experience gained in approving similar requests for other plants also provide important information. The methods and analyses described in this guidance are acceptable to the NRC staff for meeting the requirements in 10 CFR 50.67. However, if licensees propose to use alternatives different from this guidance, licensees should provide sufficient justification to demonstrate that the requirements for the pertinent regulations are satisfied. Importantly, as noted above and stated in 10 CFR 50.67, the amendment may be issued only if the applicants analysis demonstrates with reasonable assurance that the identified dose criteria are met, and therefore, licensees should ensure that they present adequate information, such as analysis assumptions, inputs, and methods, in the submittal to support the staffs assessment.11 The amendment request should describe the licensees analyses of the radiological and non-radiological impacts of the proposed modification in sufficient detail to support review by the NRC staff.

Consistent with 10 CFR 50.90, the licensee must, as far as applicable, follow the form prescribed for original applications. Typically, original applications include FSAR pages and TSs. Licensees should submit affected FSAR pages and TSs annotated with changes that reflect the revised analyses.

Additionally, the NRC staff recommends that licensees submit the actual calculation documentation. In lieu of submitting marked-up FSAR pages, licensees should include a detailed list, preferably in tabular format, of all changes being proposed between the current facility licensing basis and the requested license amendment, as well as appropriate justification for each change.

If the licensee has used a currently approved version of an NRC-sponsored computer code, the NRC staffs review and confirmatory analyses will often be more efficient if the licensee identifies the code used and submits the inputs used in the calculations performed using that code. In many cases, this information will help support efficiency in the NRC staffs review and confirmatory analyses. This recommendation does not constitute a requirement that the licensee use NRC-sponsored computer codes.

Applications for licenses, certifications, and approvals under 10 CFR Part 52 have requirements similar to those stated above for license amendment submittals. RG 1.206, Revision 1, Applications for Nuclear Power Plants, issued October 2018 (Ref. 40), provides additional guidance on combined license applications.

11 The analyses required by 10 CFR 50.67 are important to the design basis of the facility, and 10 CFR 50.34 requires design basis safety analyses and evaluations; they are a significant input to the evaluations required by 10 CFR 50.92, Issuance of amendment, or 10 CFR 50.59.

DG-1425, Page 24 1.6 Final Safety Analysis Report Requirements The regulations in 10 CFR 50.71, Maintenance of records, making of reports, include the requirements for updating the facilitys FSAR. Specifically, 10 CFR 50.71(e) requires that the FSAR be updated to include all changes made in the facility or in procedures described in the FSAR, as well as all safety analyses and evaluations performed by the licensee in support of approved requests for license amendments or in support of conclusions that changes did not require a license amendment in accordance with 10 CFR 50.59. The analyses required by 10 CFR 50.67 are subject to this requirement. The licensee should update the affected radiological analysis descriptions in the FSAR to reflect the design-basis changes to the methodology and input. The analysis descriptions should contain sufficient detail to identify the methodologies used, significant assumptions and inputs, and numerical results. RG 1.70 provides additional guidance. The licensee should remove the descriptions of superseded analyses from the FSAR in the interest of maintaining a clear design basis.

2.

Attributes of an Acceptable Accident Source Term The NRC did not provide an acceptable AST in 10 CFR 50.67. Regulatory Position 3 of this guide identifies an AST that is acceptable to the NRC staff for use in new power reactor applications and operating power reactors. The NRC, its contractors, various national laboratories, peer reviewers, and others have expended substantial effort in performing severe accident research and in developing the source terms in SAND2023-01313. However, future research may identify opportunities for changes in these source terms. The NRC staff will consider applications for an AST different from that identified in this guide. All proposed ASTs should be of at least the same quality as the source terms in SAND2023-01313. An acceptable AST has the following attributes:

a.

The AST is based on major accidents hypothesized for the purposes of design analyses, or on consideration of possible accidental events that could result in hazards not exceeded by those from other accidents considered credible. The AST addresses events that involve a substantial meltdown of the core with the subsequent release of appreciable quantities of fission products.

b.

The AST is expressed in terms of the times and rates of appearance of radioactive fission products released into containment, the types and quantities of the radioactive species released, and the chemical forms of iodine released.

c.

The AST is not based on a single accident scenario but instead represents a spectrum of credible severe accident events. Risk insights may be used, not to select a single risk-significant accident, but rather to establish the range of events to be considered. However, risk insights alone are not an acceptable basis for excluding a particular event. Relevant insights from applicable severe accident research on the phenomenology of fission product release and transport behavior may be considered.

d.

The AST has a defensible technical basis supported by sufficient experimental and empirical data, is verified and validated, and is documented in a scrutable form that facilitates public review and discourse.

e.

The AST is peer reviewed by appropriately qualified subject -matter experts. The peer review comments, and their resolution should be part of the documentation supporting the AST.

DG-1425, Page 25 Regulatory Position 3 of this guide also identifies steady-state fission product release fractions residing in the fuel rod void volume (plenum and pellet-to-cladding gap) that are acceptable to the NRC staff for use in new power reactor applications and operating power reactors. As an alternative to these bounding release fractions, appendix I to this guide provides an acceptable analytical procedure for calculating plant-specific or fuel-rod-design-specific fission product release fractions. The NRC internal memorandum, Technical Basis for Draft RG 1.183 Revision 1 (2021) Non-LOCA Fission Product Release Fractions, dated July 28, 2021 (Ref. 41), provides an example calculation illustrating the application of this analytical procedure.

3.

Accident Source Term12 This regulatory position provides an AST that is acceptable to the NRC staff. It offers guidance on the fission product inventory, release fractions, timing of the release phases, radionuclide composition, chemical form, and fuel damage for LOCA and non-LOCA DBAs. The data in Regulatory Positions 3.1 through 3.5 are fundamental to the definition of an AST. Once approved, the AST assumptions or parameters specified in these positions become part of the facilitys design basis. The NRC will evaluate deviations from this guidance against Regulatory Position 2. After the NRC staff has approved an implementation of an AST, subsequent changes to the AST will also require NRC staff review under 10 CFR 50.67.

3.1 Fission Product Inventory The inventory of fission products in the reactor core and available for release to the containment should be based on the maximum full-power operation of the core with, as a minimum, currently licensed values for fuel enrichment and fuel burnup, and an assumed core power equal to the currently licensed rated thermal power times the approved core power measurement uncertainty factor (e.g., 1.02). These parameters should be examined to maximize fission product inventory. For non-LOCA DBAs, the NRC staff will consider on a case-by-case basis the use of more explicit methods to calculate the fission product inventory and reactor coolant system activity, such as using the burnup, power history, and peaking factor from the accident analyses (e.g., chapter 14 or 15) in the updated FSAR, for each rod predicted to fail. The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach maximum values.13 The core inventory should be determined using an appropriate computer code for calculating isotope generation and depletion. The core inventory factors (in curies per megawatt thermal) provided in TID-14844 and used in some analysis computer codes were derived for low-burnup, low-enrichment fuel and should not be used with higher burnup or higher enrichment fuels. The code should model the fuel geometries, material composition, and burnup, and the cross-section libraries used should be applicable to the projected fuel burnup.

For the MHA LOCA, all fuel assemblies in the core are assumed to be affected, and the analysis should use the core -average inventory. For non-LOCA DBA events that do not involve the entire core, the fission product inventory of each damaged fuel rod is determined by dividing the total core inventory by the number of fuel rods in the core. To account for differences in power level across the core, the analysis should apply the radial peaking factors (for PWRs, these are contained in the facilitys core operating limits report or TSs) in determining the inventory of the damaged rods.

The licensee should not adjust the fission product inventory for events postulated to occur during power operations at less than full-rated power or those postulated to occur at the beginning of core life.

12 The data in this regulatory position do not apply to cores containing mixed-oxide fuel or to long-term ATF concepts.

13 Note that for some radionuclides, such as cesium-137, maximum values will not be reached before fuel offload. Thus, the maximum inventory at the end of life should be used.

DG-1425, Page 26 For events postulated to occur while the facility is shut down (e.g., a fuel handling accident), the licensee may model radioactive decay from the time of shutdown.

3.2 Release Fractions For the MHA LOCA, table 114 (for BWRs) and table 2 (for PWRs) in this RG list the core inventory release fractions, by radionuclide group, for the gap release and early in-vessel damage phases.

These fractions are applied to the maximum core inventory described in Regulatory Position 3.1.

The limitations on the use of tables 1 and 2 in this RG are based on several reference documents.

First, tables 1 and 2 are based, in part, on accident source terms from SAND2023-01313.KM-04 These source terms were derived by examining a set of accident sequences for current LWR designs for licensees using zirconium-alloy cladded UO2 fuel rod designs with reactor core burnups up to a maximum rod-average of 80 GWd/MTU (and fuel enrichments up to 8 weight percent U-235 for PWRs and up to 10 weight percent for BWRs), for currently approved (as of the issuance date of this RG revision) fuel and cladding materials. They reflect the current understanding of severe accidents and fission product behavior. Second, the NRC internal memorandum, Applicability of Source Term for Accident Tolerant Fuel, High Burn Up and Extended Enrichment, dated May 13, 2020 (Ref. 42), and insights from a literature review on ATFs reported in NUREG/CR-7282, Review of Accident Tolerant Fuel Concepts with Implications to Severe Accident Progression and Radiological Releases, issued July 2021 (Ref. 43),

extend the applicability of these source terms to doped fuels. In addition, the chromium-coated cladding (Ref. 25) and FeCrAl cladding (Ref. 24) source term analyses also support the applicability of tables 1 and 2 to these two near-term cladding deigns. Lastly, the NRC internal memorandum, Letter Report on Evaluation of the Impact of Fuel Fragmentation, Relocation, and Dispersal for the Radiological Design Basis Accidents in Regulatory Guide 1.183, dated July 20, 2021 (Ref. 44), assesses the impact of FFRD behavior on the accident source terms from SAND2011-0128. The conclusions of the July 20, 2021, memorandum also extend to the SAND2023-01313 source term. Based on information in that memorandum, for the purposes of assessing the radiological consequences of the MHA LOCA, the impact of FFRD does not need to be considered for the range of applicability of burnups and enrichments in SAND2023-01313.

3.2.1 Pathway-Specific Release Fractions Using Mechanistic Transport ModelingKM-05 For the BWR AST, table 1 shows separate pathway-specific release fractions for the containment leakage pathway, the main steamline leakage pathway through the MSIVs, and the liquid leakage pathway from the suppression pool water volume. The purpose of breaking the BWR AST into three pathways is to consider the retention of fission products in the suppression pool in the determination of release fractions. The containment source term in table 1.1 represents the cumulative containment inventory fraction excluding the suppression pool inventory in each phase of the MHA LOCA. The suppression pool source term in table 1.2 represents the cumulative suppression pool inventory fraction in each phase of the MHA LOCA. In contrast to these, the steamline source term in table 1.3 represents the time-averaged, airborne core inventory fraction present in the steamline between the first steam relief valve and the first MSIV over all steamlines. Appendix A contains additional information. (Ref. 45) 14 In this guide, table 1 refers collectively to tables 1.1, 1.2, and 1.3.

DG-1425, Page 27 Table 1.1. BWR Core Inventory Fraction Released into Containment Atmosphere Group Gap Release Phase Early In-Vessel Phase Noble Gases 1.6x10-2 9.5x10-1 Halogens 1.3x10-6 6.0x10-2 Alkali Metals 1.2x10-6 6.0x10-3 Tellurium Metals 0.0 3.8x10-2 Barium, Strontium 0.0 3.0x10-4 Noble Metals 0.0 7.4x10-6 Cerium Group 0.0 0.0 Lanthanides 0.0 0.0 Molybdenum 0.0 1.0x10-4 Table 1.2. BWR Core Inventory Fraction Retained in the Suppression Pool Group Gap Release Phase Early In-Vessel Phase Noble Gases 0.0 0.0 Halogens 5.0x10-3 6.5x10-1 Alkali Metals 5.0x10-3 3.1x10-1 Tellurium Metals 3.0x10-3 5.2x10-1 Barium, Strontium 6.0x10-4 4.7x10-3 Noble Metals 0.0 6.0x10-3 Cerium Group 0.0 0.0 Lanthanides 0.0 0.0 Molybdenum 1.9x10-5 1.2x10-1

DG-1425, Page 28 Table 1.3. BWR Core Inventory Time-Averaged Fraction in Steamline15 Group Gap Release Phase16 Early In-Vessel Phase16 Noble Gases 2.9x10-5 1.1x10-3 Halogens 5.6x10-6 5.1x10-5 Alkali Metals 5.1x10-6 1.3x10-5 Tellurium Metals 3.2x10-6 2.7x10-5 Barium, Strontium 6.1x10-7 2.4x10-7 Noble Metals 0.0 2.4x10-7 Cerium Group 0.0 0.0 Lanthanides 0.0 0.0 Molybdenum 3.3x10-9 3.0x10-6 Table 2. PWR Core Inventory Fraction Released into Containment Atmosphere Group Gap Release Phase Early In-Vessel Phase Noble Gases 2.6x10-2 9.3x10-1 Halogens 7.0x10-3 5.8x10-1 Alkali Metals 3.0x10-3 5.0x10-1 Tellurium Metals 6.0x10-3 5.5x10-1 Barium, Strontium 1.0x10-3 2.0x10-3 Noble Metals 0.0 8.0x10-3 Cerium Group 0.0 0.0 Lanthanides 0.0 0.0 Molybdenum 2.0x10-5 1.5x10-1 For non-LOCA DBAs, table 3 (for BWRs) and table 4 (for PWRs) list the maximum steady-state fission product release fractions residing in the fuel rod void volume (plenum and pellet-to-cladding gap),

by radionuclide group, available for release upon cladding breach. The licensing bases of some facilities may include non-LOCA events that assume the release of the gap activity from the entire core. For events involving the entire core, the core-average gap fractions of tables 1 and 2 may be used, and the radial peaking factor may be omitted.

15 This represents the core inventory released to the portion of the steamline that lies between the first SRV and the MSIV. It is the time-averaged airborne fraction summed over all steamlines.

16 The release fractions in this table already take fission product deposition into account, so credit should not be taken for deposition inboard of the first MSIV during the gap and early in-vessel phases.

DG-1425, Page 29 The applicability of the steady-state fission product release fractions in tables 3 and 4 is limited to currently approved (as of the issuance of Revision 2 of this RG) full-length UO2 fuel rod designs operating up to a maximum rod-average burnup of 80 GWd/MTU at power levels below the burnup-dependent power envelopes depicted in figures 1 and 2. In figures 1 and 2, the bounding rod-average power refers to the rod-average linear heat generation rate of the peak rod. Licensees should make adjustments to account for power uncertainties and plant maneuvering when comparing operating power histories to figures 1 and 2. If it can be demonstrated that local power level, rate of fission gas release, and cumulative fission gas release remain less than those of the limiting co-resident UO2 fuel rod, then the steady-state fission product release fractions in tables 3 and 4 apply to fuel rod designs containing integral burnable absorbers (e.g., gadolinia). One acceptable means of demonstrating this is by using an NRC-approved fuel performance code that has fission gas release models that are applicable to the integral burnable absorber fuel designs. If BWR part-length fuel rods are treated as full-length fuel rods with respect to overall quantity of fission products, then table 3 steady-state fission product release fractions apply to these part-length fuel rod designs. Applicability to future fuel rod designs, including chromium-coated zirconium (Zr) cladding, non-Zr claddings, doped UO2 fuel, high-density fuel, and mixed-oxide fuel, will be evaluated on a case-by-case basis. Appendix I provides an acceptable analytical technique for calculating plant-specific or fuel-rod-design-specific fission product release fractions.

Table 3. BWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap Group Fraction I-131 0.04 I-132 0.04 Kr-85 0.40 Other Noble Gases 0.04 Other Halogens 0.03 Alkali Metals 0.20 Table 4. PWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap Group 14 x 14 Fraction 17 x 17 Fraction I-131 0.09 0.03 I-132 0.10 0.03 Kr-85 0.52 0.50 Other Noble Gases 0.08 0.02 Other Halogens 0.05 0.02 Alkali Metals 0.26 0.25

DG-1425, Page 30 Figure 1. Maximum allowable power operating envelope for PWR steady-state release fractions Figure 2. Maximum allowable power operating envelope for BWR steady-state release fractions

DG-1425, Page 31 For non-LOCA DBAs involving a rapid increase in fuel rod power, such as the BWR control rod drop accident and PWR control rod ejection accident, additional fission product releases may occur because of pellet fracturing and grain boundary separation. This transient fission gas release (TFGR) increases the amount of activity available for release into the reactor coolant system for fuel rods that experience cladding breach. The empirical database suggests that TFGR is sensitive to both local fuel burnup and peak radial average fuel enthalpy rise. As a result, separate low-burnup and high-burnup TFGR correlations for stable, long-lived radionuclides (e.g., krypton (Kr)-85 and cesium-137) are provided, as follows:

pellet burnup < 50 GWd/MTU, TFGR (long-lived isotopes) = maximum [ (0.26

• H) - 13) / 100, 0 ],

(Equation 1) pellet burnup > 50 GWd/MTU, TFGR (long-lived isotopes) = maximum [ (0.26

• H) - 5) / 100, 0 ],

(Equation 2)

Where:

TFGR = transient fission gas release fraction, and H = increase in radial average fuel enthalpy, calories per gram.

An investigation into the effect of differences in diffusion coefficients and radioactive decay on fission product transient release concluded that different radionuclides require adjustments to the above empirically based correlations (Ref. 46). For stable, long-lived noble gases (e.g., Kr-85) and alkali metals (e.g., cesium-137), the transient fission product release is equivalent to the above burnup-dependent correlations. For volatile, short-lived radioactive isotopes such as halogens (e.g., iodine (I)-131) and xenon (Xe) and Kr noble gases except Kr-85 (e.g., Xe-133, Kr-85m), the transient fission product release correlations should be multiplied by a factor of 0.333. The low-burnup and high-burnup TFGR correlations for volatile, short-lived radioisotopes are as follows:

pellet burnup < 50 GWd/MTU, TFGR (short-lived isotopes) = 0.333

• maximum [ (0.26

• H) - 13) / 100, 0 ], (Equation 3) pellet burnup > 50 GWd/MTU, TFGR (short-lived isotopes) = 0.333

• maximum [ (0.26

• H) - 5) / 100, 0 ], (Equation 4)

Where:

TFGR = transient fission gas release fraction, and H = increase in radial average fuel enthalpy, calories per gram.

For the remaining high-temperature non-LOCA DBAs that predict fuel rod cladding failure, such as the PWR reactor coolant pump locked rotor and main steamline break, additional fission product releases may occur because of fuel pellet fragmentation (e.g., fracturing of high-burnup rim region) due to loss of pellet-to-cladding mechanical constraint or impact loads. TFGR has been experimentally observed under a variety of accident conditions. At the time of issuance of Revision 2 of this RG, no consensus exists on the mechanism or the computation of TFGR for these events; therefore, an acceptable method to address TFGR for non-LOCA DBAs other than reactivity-initiated accidents would be to prevent balloon and burst failures though design and analysis. Though not fully applicable to non-LOCA and non-reactivity-initiated DBAs, NRC Research Information Letter 2021-13, Interpretation of Research on Fuel Fragmentation Relocation, and Dispersal at High Burnup, issued December 2021 (Ref. 47),

provides data that can be used to provide a bounding estimate of TFGR for high-temperature DBAs.

DG-1425, Page 32 The total fraction of fission products available for release equals the steady-state fission product release fractions in tables 3 and 4 plus any TFGR prompted by the accident conditions. TFGR may be calculated separately for each axial node based on local accident conditions (e.g., fuel enthalpy rise) and then combined to yield the total TFGR for a particular damaged fuel rod. An NRC internal memorandum (Ref. 41) documents the technical bases of the steady-state fission product release fractions and TFGR correlations.

The non-LOCA fission product release fractions and TFGR correlations do not include the additional contribution associated with fuel melting. The event-specific appendices to this RG provide guidance for adjusting these gap inventories for fuel rods that are predicted to experience limited fuel centerline melting.

3.3 Timing of Release Phases Table 5 provides the onset and end time of each sequential release phase for LOCA DBAs. The specified onset is the time following the initiation of the accident (i.e., time = 0). The early in-vessel release phase immediately follows the gap release phase. The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase.17 For non-LOCA DBAs in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage.

The applicability of table 5 is consistent with the applicability of tables 1 and 2.

Table 5. MHA-LOCA Release Phases Phase PWRs BWRs Onset End Time Onset End Time Gap Release 0.5 minutes 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 2 minutes 0.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Early In-Vessel 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 5.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 0.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 7.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> For facilities licensed with a leak-before-break methodology, the licensee may assume the onset of the gap release phase to be 10 minutes. The licensee may propose an alternative time for the onset of the gap release phase based on facility-specific calculations using suitable analysis codes or based on an accepted topical report shown to apply to the specific facility. In the absence of approved alternatives, the licensee should use the gap release phase onsets in table 5.

3.4 Radionuclide Composition Table 6 lists the elements in each radionuclide group that should be considered in design-basis analyses.

17 This statement excludes the effects of radioactive decay in the core inventory on the linear release modeled. In lieu of treating the release in a linear ramp manner, the activity for each phase can be modeled as being released instantaneously at the start of that release phase (i.e., in step increases).

DG-1425, Page 33 Table 6. Radionuclide Groups Group Elements Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se Barium, Strontium Ba, Sr Noble Metals Ru, Rh, Pd, Co Lanthanides La, Nd, Eu, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np, Zr Molybdenum Mo, Tc, Nb 3.5 Chemical Form Of the radioiodine released from the reactor coolant system to the containment in a postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the gap and the fuel pellets. Except for elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. The transport of these iodine species following release from the fuel may affect these assumed fractions. The accident-specific appendices to this RG contain additional details.

3.6 Fuel Damage in Non-LOCA DBAs The amount of fuel damage caused by non-LOCA DBAs should be analyzed to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel cladding is breached. Cladding failure mechanisms include high-temperature failure modes (e.g., critical heat flux, local oxidation, and ballooning) and pellet-to-cladding mechanical interaction.

Appendix B to this guide addresses the modeling of the amount of fuel damage caused by a fuel handling accident.

3.7 Assessment of Radiological Consequences due to Fuel Fragmentation Relocation and Dispersal for Analysis of a 10 CFR 50.46 Large-Break LOCA Recent experimental findings indicate that under certain transient conditions, portions of a reactor containing high-burnup fuel operating at sufficiently high power can fragment and escape from fuel cladding that has burstKM-06 (Ref. 47). The escaped fuel fragments could subsequently be distributed throughout a reactor coolant system. FFRD is an important concern because of the potential dose impacts on members of the public and workers and because of the potential challenges to the coolability of the reactor core.

The staffs understanding of FFRD phenomena has continued to advance. In 2012, the NRC issued NUREG-2121, Fuel Fragmentation, Relocation, and Dispersal during the Loss-of-Coolant Accident (Ref. 48), which captured the results of over 90 LOCA tests performed in eight different

DG-1425, Page 34 programs over 35 years. The NRC concluded from this review that FFRD is an important consideration during 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, LOCA analyses for a reactor using higher burnup or increased enrichment fuels and that additional research into this phenomenon was required. In 2015, the NRC published SECY-15-0148, Evaluation of Fuel Fragmentation, Relocation and Dispersal under Loss-of-Coolant Accident (LOCA)

Conditions Relative to the Draft Final Rule on Emergency Core Cooling System Performance during a LOCA (50.46c) (Ref. 49). SECY-15-0148 concluded that immediate regulatory action was not needed to address FFRD phenomena at that time based on existing fuel design limits and assumptions on how high-burnup fuel would be used.

To assess the radiological consequences of 10 CFR 50.46 LOCA analyses that predict FFRD, a new acceptance criterion has been established in table 7. This acceptance criterion is consistent with the criteria applied to other non-MHA-LOCA DBAs, such that the radiological consequences associated with an FFRD event would be like those expected from other accidents.

In 2021, the NRC assessed FFRD impacts on the MHA-LOCA source term (also referred to as the in-containment source term) (Ref. 50). The FFRD-induced source term comes from the fission product gases generated within the reactor fuel. At a microscopic level, gas bubbles can form within grains of fuel pellets and at grain boundaries. The pressure of this gas increases with fuel burnup. The higher pressure in the grain boundaries drives fission gases to the gap/plena of fuel rods (i.e., the small space between the outer surface of a fuel pellet and the inner surface of the cladding). Under accident conditions, the average fuel temperature rises, which increases the gas pressure in the fuel, grain boundaries, and fuel-clad gap. Some species that are solid at operating temperatures may vaporize during accident conditions, further increasing pressure both within the intergranular gas bubbles and within the fuel-cladding gap. Large pressure differences between the fuel rod and the coolant can lead to clad ballooning, which removes the mechanical restraint provided by the cladding on the fuel pellets. This loss of restraint results in the formation of stresses in the fuel and can lead to fuel fragmentation. Under accident conditions, pressures can burst the fuel clad. This results in a sudden reduction of the gas pressure in the fuel-clad gap to that of the surroundings, resulting in a sudden large pressure differential between the gases in the grain boundaries and the surrounding gas. These changes in mechanical forces can cause the pellets to fragment and release into the reactor coolant system, creating a radiological source term above the normal operational source term, as described in Regulatory Position 1.1.4.

Without a best-estimate FFRD-induced source term, licensees would use a fraction of the applicable MHA-LOCA release fractions presented in Regulatory Position 3.2. The release fraction would be determined based on the total mass of FFRD predicted. This is appropriate since the scenarios considered in the development of the MHA-LOCA source term used to demonstrate compliance with requirements in 10 CFR 50.34(a)(1)(ii)(D), 10 CFR 50.67, and 10 CFR 100.11 exclude the effects of emergency core cooling. This MHA-LOCA source term is the result of a postulated substantial meltdown of the core. As such, the MHA-LOCA source term involves far greater radiological releases from the fuel than from FFRD. However, best-estimate FFRD-induced source terms may be considered on a case-by-case basis.

The release timing for an FFRD-induced source term is assumed to be instantaneous. Appendix A provides the assumptions acceptable to the staff for evaluating the radiological consequences of MHA-LOCAs and can be used for a 10 CFR 50.46 LOCA analysis that predicts FFRD.

This regulatory guide does not provide guidance on demonstrating compliance with the 10 CFR 50.46 requirements or how to estimate the total mass of fuel released because of FFRD.

DG-1425, Page 35 Table 7a Accident Dose Criteria for EAB, LPZ, and Control Room Locations Accident or Case EAB and LPZ Dose Criteria (TEDE)

Control Room Dose Criteriab (TEDE)

Analysis Release Durationc MHA LOCA 0.25 sievert (Sv)

(25 rem)

See table 8d 30 days for containment, emergency core cooling systems (ECCS), and MSIV (BWR) leakage 10 CFR 50.46 LOCA with FFRD 0.063 Sv (6.3 rem) 0.10 Sv (10.0 rem) 30 days for containment, ECCS, MSIV (BWR) leakage BWR Main Steamline Break Instantaneous puff Fuel Damage or Pre-Accident Spike 0.25 Sv (25 rem) 0.10 Sv (10.0 rem)

Equilibrium Iodine Activity 0.025 Sv (2.5 rem) 0.05 Sv (5.0 rem)

BWR Rod Drop Accident 0.063 Sv (6.3 rem) 0.10 Sv (10.0 rem) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> PWR Steam Generator Tube Rupture Affected steam generator: time to isolatee Unaffected steam generator(s): until shutdown cooling is in operation and releases from the steam generator have been terminated Fuel Damage or Pre-Accident Spike 0.25 Sv (25 rem) 0.10 Sv (10.0 rem)

Concurrent Iodine Spike 0.025 Sv (2.5 rem) 0.05 Sv (5.0 rem)

PWR Main Steamline Break Until shutdown cooling is in operation and releases from the steam generators have been terminated Fuel Damage or Pre-Accident Spike 0.25 Sv (25 rem) 0.10 Sv (10.0 rem)

Concurrent Iodine Spike 0.025 Sv (2.5 rem) 0.05 Sv (5.0 rem)

PWR Locked Rotor Accident 0.025 Sv (2.5 rem) 0.05 Sv (5.0 rem)

Until shutdown cooling is in operation and releases from the steam generators have been terminated PWR Control Rod Ejection Accident 0.063 Sv (6.3 rem) 0.10 Sv (10.0 rem)

Containment pathway: 30 days Secondary system: until shutdown cooling is in operation and releases from the steam generators have been terminated Fuel Handling Accident 0.063 Sv (6.3 rem) 0.10 Sv (10 rem) 30 days a

For PWRs with steam generator alternative repair criteria, different dose criteria may apply to steam generator tube rupture and main steamline break analyses.

b The control room exposure period is 30 days for all accidents.

c The column labeled Analysis Release Duration summarizes the assumed radioactivity release durations identified in the individual appendices to this guide. These appendices contain complete descriptions of the release pathways and durations.

d Table 8 presents a graded, risk-informed, and performance-based framework for the control room dose criteria. The framework is applicable to the MHA LOCA.

e Tube rupture in the affected steam generator may result in the need to control the steam generator water level using steam dumps. These releases may extend the duration of the release from the affected steam generator beyond the initial isolation.

DG-1425, Page 36

4.

Dose Calculation Methodology The NRC staff has determined (e.g., in Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants: Final Rule (61 FR 65157; December 11, 1996)) that there is an implied synergy between the ASTs and TEDE criteria and between the TID-14844 source terms and the whole-body and thyroid dose criteria (Ref. 51). The TEDE criteria should not be used with results calculated from TID-14844. The guidance in this Regulatory Position applies to all dose calculations performed with an AST pursuant to 10 CFR 50.67 and 10 CFR Part 52. The regulatory position also provides guidance for determining control room and offsite doses and the control room and offsite dose acceptance criteria. Certain selective implementations may not require dose calculations, as described in Regulatory Position 1.3 of this guide.

4.1 Offsite Dose Consequences The licensee should use the following assumptions in determining the TEDE for persons located at or beyond the EAB:

a.

The dose calculations should determine the TEDE. TEDE is the sum of the effective dose equivalent (for external exposures) (EDEX) and the committed effective dose equivalent (for internal exposures) (CEDE). The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides that have significant dose consequences and significant released radioactivity.18

b.

The exposure-to-CEDE factors for inhalation of radioactive material should be derived from the data in International Commission on Radiological Protection (ICRP) Publication 30, Limits for Intakes of Radionuclides by Workers, issued in 1979 (Ref. 52). Table 2.1 of Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, issued in 1988 (Ref. 53), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed Effective yield doses that correspond to the CEDE.

c.

Table III.1 of Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water, and Soil, issued in 1993 (Ref. 54), provides external effective dose equivalent (EDE) conversion factors acceptable to the NRC staff. The factors in the column headed Effective yield doses that correspond to the EDE.

d.

No correction should be made for depletion of the effluent plume by deposition on the ground.

e.

The TEDE should be determined for an individual at the most limiting EAB location. The maximum EAB TEDE for any 2-hour period following the start of the radioactivity release should be determined and used in determining compliance with the dose criteria in 10 CFR 50.6719 and 10 CFR Part 52. The maximum 2-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a sliding sum over the increments for successive 2-hour periods. The maximum TEDE obtained is taken as the result of the analysis. The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see analysis release duration in table 7). The analysis should assume that the 18 The prior practice of basing inhalation exposure on only radioiodine and not including radioiodine in external exposure calculations is not consistent with the definition of TEDE and the characteristics of the revised source term.

19 For the EAB TEDE, the maximum 2-hour value is the basis for screening and evaluation under 10 CFR 50.59. Changes to doses outside of the 2-hour window are considered only in the context of their impact on the maximum 2-hour EAB TEDE.

DG-1425, Page 37 most limiting 2-hour EAB /Q value occurs simultaneously with the limiting release to the environment (see also Regulatory Position 5.3 of this guide). In calculations of the maximum EAB TEDE for an individual, the maximum 2-hour EAB /Q value and a breathing rate of this individual of 3.5x10-4 cubic meters per second (m3/s) should be used for the entire duration of the release to the environment to ensure that the limiting case is identified.

If multiple release paths are analyzed separately, additional processing is needed to identify the maximum 2-hour TEDE that is the sum of all paths, since the maximum periods may not be the same for each path. In these cases, it will be necessary to assess each release using the maximum 2-hour EAB /Q value, sum the doses for each pathway for each time increment, and then identify the maximum 2-hour EAB TEDE. As a conservative alternative, the maximum 2-hour TEDE for each path could be summed to determine the value for the accident.

f.

The TEDE should be determined for the most limiting receptor at the outer boundary of the LPZ for the duration of the accident. This value should be used in determining compliance with the dose criteria in 10 CFR 50.67 and 10 CFR Part 52.

For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be assumed to be 3.5x10-4 m3/s.

From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate should be assumed to be 1.8x10-4 m3/s. After that and until the end of the accident, the rate should be assumed to be 2.3x10-4 m3/s.

4.2 Control Room Dose Consequences The following guidance should be used in determining the TEDE for persons located in the control room.

4.2.1 Sources of Radiation The TEDE analysis should consider all sources of radiation that will cause exposure of control room personnel. The applicable sources will vary from facility to facility but typically will include the following:

a.

contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility,

b.

contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope,

c.

radiation shine from the external radioactive plume released from the facility,

d.

radiation shine from radioactive material in the reactor containment, and

e.

radiation shine from radioactive material in systems and components inside or external to the control room envelope (e.g., radioactive material buildup in recirculation filters).

4.2.2 Material Releases and Radiation Levels The radioactive material releases and radiation levels used in the control room dose analysis should be determined using the same source term, in-plant transport, and release assumptions used for

DG-1425, Page 38 determining the EAB and the LPZ TEDE values, unless these assumptions would produce nonconservative results for the control room.

4.2.3 Transport Models The models used for the transport of radioactive material into and through the control room20 and the shielding models used to determine radiation dose rates from external sources should be structured to provide suitably conservative estimates of the exposure of control room personnel.

4.2.4 Engineered Safety Features The licensee may assume credit for ESFs that mitigate airborne radioactive material within the control room. Such features may include control room isolation or pressurization or intake or recirculation filtration. Guidance appears in SRP Section 6.5.1, ESF Atmosphere Cleanup Systems, and RG 1.52, Revision 4, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, issued September 2012 (Ref. 57). The control room design is often optimized for the MHA LOCA, and the protection afforded for other accident sequences may not be as advantageous. In most designs, control room isolation is actuated by ESF signals or radiation monitors. In some cases, the ESF signal is effective only for selected accidents, placing reliance on the radiation monitors for the remaining accidents. Several aspects of radiation monitors can delay control room isolation, including the delay for activity to build up to concentrations equivalent to the alarm setpoint and the effects of different radionuclide accident isotopic mixes on monitor response.

4.2.5 Personal Protective Equipment The licensee should generally not take credit for the use of personal protective equipment or prophylactic drugs such as potassium iodide. The NRC may consider deviations on a case-by-case basis.

4.2.6 Dose Receptor Occupancy Factor21 Occupancy factors are used to estimate the amount of time a receptor, or individual, is present at a particular location to estimate the amount of time a receptor was exposed to a radioactive source (e.g., plume). An occupancy factor is typically expressed in terms of fraction per day. With detailed information on a receptors habits, an analyst can determine appropriate locations and occupancy factors required for the analysis.KM-07 Without facility-specific receptor information on occupational habits within the control room during an event, these analyses assume that the dose receptor is the hypothetical maximum exposed individual. The default occupancy factors assume that the receptor is present in the control room 100 percent of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60 percent of the time between 1 and 20 The iodine protection factor methodology of Reference 30 may not be adequately conservative for all DBAs and control room arrangements because it models a steady-state control room condition. Since many analysis parameters change over the duration of the event, the iodine protection factor methodology should be used only with caution. The NRC computer codes HABIT (Ref. 53) and RADTRAD (Ref. 54) incorporate suitable methodologies.

21 Considering the requirements of 10 CFR Part 26, Fitness for Duty Programs, and greater focus on the integral role humans play in the safe operation of a nuclear power plant, this regulatory position was updated to credit facility-specific emergency plan staffing operations as the basis for the occupancy factor.

DG-1425, Page 39 4 days, and 40 percent of the time from 4 days to 30 days.22 For the duration of the event, the licensee should assume the breathing rate of this individual to be 3.5x10-4 m3/s (Ref. 60).

Licensees develop and maintain staffing plans to execute their emergency plans. These plans include facility-specific considerations that impact the licensees ability to staff emergency planning functions. Licensees can credit facility-specific staffing plans to define facility-specific occupancy factors that differ from the default values provided earlier in this section. This provision acknowledges the unique requirements of each facility and fosters alignment with other regulatory requirements that apply during accident scenarios. Facility-specific occupancy factors should align with the licensing basis and appropriately account for uncertainties (e.g., shift changes) with the intent of maintaining operational flexibility during an event.

4.2.7 Dose Conversion Factor The licensee should calculate control room doses using the dose conversion factors identified in Regulatory Position 4.1 for use in offsite dose analyses. The calculation should consider all radionuclides, including progeny from the decay of parent radionuclides that have significant dose consequences, and the released radioactivity. The EDE from photons may be corrected for the difference between finite cloud geometry in the control room and the semi-infinite cloud assumption used in calculating the dose conversion factors. The following expression may be used to correct the semi-infinite cloud dose, EDEX, to a finite cloud dose, EDEXfinite, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 58):

1173 338

.0 V

EDEX EDEX finite

=

(Equation 5)

The expression may be already incorporated into certain radiological assessment codes and would not need to be separately added to the dose results of finite volumes such as control room doses when using those codes.

4.3 Other Dose Consequences The licensee should use the guidance in Regulatory Positions 4.1 and 4.2, as applicable, to reassess the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737.

The licensee should update design envelope source terms from NUREG-0737 for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE. Integrated radiation exposure of plant equipment should be determined using the guidance in RG 1.89, Revision 2.

4.4 Acceptance Criteria The accident dose radiological criteria for the EAB, the outer boundary of the LPZ, and the control room are in 10 CFR 50.34, 10 CFR Part 52, 10 CFR 50.67, and GDC 19 in Appendix A to 10 CFR Part 50. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation. For events with a higher probability of 22 These occupancy factors are already included in the determination of the /Q values using the Murphy and Campe methodology described in Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19, issued August 1974 (Ref. 56), and should not be credited twice. The ARCON96 code (Ref. 57) does not incorporate these occupancy factors into the determination of the /Q values. Therefore, dose calculations using ARCON96

/Q values should include the occupancy factors.

DG-1425, Page 40 occurrence, postulated EAB and LPZ doses should not exceed the criteria in table 7. The accident dose for the EAB should not exceed the acceptance criteria for any 2-hour period following the onset of the fission product release. The accident dose for the LPZ should not exceed the acceptance criteria during the entire period of the passage of the fission product release. To support the Increased Enrichment Rulemaking, the staff developed a graded, risk-informed, and performance-based framework for the control room for the MHA-LOCA.KM-08 Table 8 presents this framework (Ref. 17 gives the background of this framework methodology). The framework leverages facility-specific risk insights based on PRA acceptability.

In accordance with 10 CFR 50.67 and GDC 19, as proposed for amendment in the Increased Enrichment Rulemaking, licensees may use a 10-rem-TEDE control room design criterion in assessing an AST, without further justification. The licensee may establish a control room design criterion value higher than 10-rem-TEDE but not greater than 25-rem-TEDE, provided that the licensee demonstrates that the specified value is commensurate with the risk of the plant. Acceptability of the PRA models used to demonstrate that the specified criterion is commensurate with the risk of the plant is determined for the following aspects: scope, level of detail, conformance with PRA technical elements (i.e., technical robustness), and plant representation and PRA configuration control. The PRA models should be consistent with the philosophy in RG 1.174 and the technical adequacy expectations for the model in RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities (Ref. 61).

For instance, the use of overall CDF results from an NRC-approved license amendment request that incorporates the risk-informed completion time program into the facilitys technical specifications (i.e.,

Technical Specifications Task Force Traveler (TSTF)-505, Provide Risk-Informed Extended Completion TimesRITSTF Initiative 4b) would be acceptable (Ref. 62). The baseline PRA model should estimate the overall CDF for all significant sources of risk both internal and external to the plant (e.g., internal, flood, fires, seismic, high winds, and others).

For licensees with an acceptable PRA model, licensees should follow each step, listed below:

1. Select the corresponding facility-specific control room design criteria in table 8 based on the overall CDF.

2. Present a summary of the baseline PRA model (e.g., original submittal and staff safety evaluation) and any adjustments made to the model since the initial approval.

3. Reassess23 the facility-specific control room design criterion, as necessary, for subsequent license amendment requests that affect the MHA LOCA.

The acceptance criteria for the various NUREG-0737 items generally reference GDC 19 or specify criteria derived from GDC 19. These criteria are generally specified in terms of whole-body dose or its equivalent to any body organ. For facilities applying for, or having received, approval to use an AST, licensees should update the applicable criteria for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii).

For new reactor applicants, the technical support center (TSC) habitability acceptance criterion is based on the requirement of paragraph IV.E.8 of Appendix E to 10 CFR Part 50 to provide an onsite TSC from which effective direction can be given and effective control can be exercised during an emergency.

The radiation protection design of the TSC is acceptable if the total calculated radiological consequences 23 This assessment is considered a snapshot in time and not a living assessment in that the applicable control room design criterion should be recomputed only when the licensee pursues changes that impact DBA analyses through license amendment requests. Licensees do not need to maintain a living assessment of some applicable, up-to-date control room design criterion under this framework.

DG-1425, Page 41 for the postulated fission product release fall within the exposure acceptance criteria specified for the control room (5-rem-TEDE for the duration of the accident).

Table 8. Guidelines for Control Room Location Based on a Graded, Risk-Informed, and Performance-Based Framework Overall CDF Graded Control Room Design Criteria (rem-TEDE)

CDF 1.E-5 25 1E-5 < CDF 5E-5 20 5E-5 < CDF 1E-4 15 CDF > 1E-4; or licensee not adopting the graded framework to determine acceptance criteria 10

5.

Analysis Assumptions and Methodology 5.1 General Considerations 5.1.1 Analysis Quality The analyses discussed in this guide are reanalyses of the design-basis safety analyses required by 10 CFR 50.67 or evaluations required by 10 CFR 50.34, 10 CFR Part 52, and GDC 19. These analyses are a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR 50.59 and 10 CFR Part 52.

The licensee should prepare, review, and maintain these analyses in accordance with quality assurance programs that comply with Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50.

These design-basis analyses were structured to provide a conservative set of assumptions to test the performance of one or more aspects of the facility design. Many physical processes and phenomena are represented by conservative bounding assumptions rather than being modeled directly. The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion. Licensees should exercise caution in proposing deviations based on data from specific accident sequences, since the DBAs were never intended to represent any specific accident sequence; the proposed deviation may not be conservative for other accident sequences.

5.1.2 Credit for Engineered Safeguard Features The licensee may take credit for accident mitigation features that are classified as safety related, are required to be operable by TSs, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. However, the licensee should not take credit for ESFs that would affect the generation of the source term described in tables 1 and 2. Additionally, the licensee should assume the single active component failure that results in the most limiting radiological consequences. Assumptions about the occurrence and timing of a loss of offsite power should be selected with the objective of

DG-1425, Page 42 maximizing the postulated radiological consequences. The licensee should consider design-basis delays in the actuation of these features, especially for features that rely on manual intervention.

5.1.3 Assignment of Numerical Input Values The licensee should select the numerical values to be used as inputs to the dose analyses with the objective of determining a conservative postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but nonconservative in another portion of the same analysis.

For example, an assumption of minimum containment system spray flow is usually conservative for estimating iodine scrubbing but, in many cases, may be nonconservative when determining sump pH.

Sensitivity analyses may be needed to determine the appropriate value to use. As a conservative alternative, the limiting value applicable to each portion of the analysis may be used in the evaluation of that portion. A single value may not be applicable for a given parameter for the duration of the event, particularly for parameters affected by changes in density. For a parameter addressed by TSs, the value used in the analysis should be that identified in the TSs.24 If a range of values or a tolerance band is specified, the value that would result in a conservative postulated dose should be used. If the parameter is based on the results of less frequent surveillance testing (e.g., steam generator nondestructive testing), the degradation that may occur between periodic tests should be considered in establishing the analysis value.

Best-estimate-plus-uncertainty methods provide more realistic representations of uncertainty when compared to point estimate deterministic methods. This additional realism in the representation of uncertainty can enhance the understanding of plant-specific analytical margins. Methods for quantification of uncertainty in engineering models are well established. For example, one common method for uncertainty analysis uses Monte Carlo statistical techniques, which incorporate a random sampling of representative distributions of various model parameter inputs. Important considerations in the application of uncertainty methods include the selection of parameters that are appropriate to the model and distributions used in the analysis. When using best-estimate-plus-uncertainty techniques in the context of this guide, the proposed distributions may be based on measurements or on justifiable engineering judgment when data are limited. Descriptive statistics from the final distribution results should be provided with the submittals applying these techniques, and values from the 95th percentile should be used when demonstrating compliance. Sampling from within the parameter distributions must encompass the entire range of values to ensure a comprehensive representation of potential values.

Extrapolation from the data will not be accepted, nor will triangular distributions. For such an analysis, a sufficient number of iterations must be performed to ensure a high level of confidence in the accuracy and reliability of the obtained results. A 95/95 confidence level (i.e., the uncertainty analysis demonstrates with 95 percent confidence that 95 percent of the results will be below the applicable acceptance criteria) is considered acceptable for comparing best-estimate predictions. Best-estimate-plus-uncertainty approaches that apply statistical sampling techniques and parameter distributions will be evaluated on a case-by-case basis. Licensees should use a best-estimate-plus-uncertainty approach when results produce a considerable margin to the regulatory acceptance criteria.

5.1.4 Applicability of the Prior and the Proposed Licensing Basis The NRC staff considers the implementation of an AST to be a significant change to the design basis of the facility that is voluntarily initiated by the licensee. The characteristics of the ASTs and the revised dose calculation methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facilitys design-basis analyses. Licensees should consider and address 24 Note that for some parameters, the technical specification value may be adjusted, for analysis purposes, by factors provided in other regulatory guidance. For example, ESF filter efficiencies are based on the guidance in RG 1.52, rather than the surveillance test criteria in the technical specifications. Generally, these adjustments address possible changes in the parameter between scheduled surveillance tests.

DG-1425, Page 43 new or unreviewed issues created by a particular site-specific implementation of the AST where the implementation conflicts with the facilitys licensing basis. However, prior design bases that are unrelated to the use of the AST, or are unaffected by the AST, may continue as part of the facilitys design basis.

Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.

5.2 Accident-Specific Assumptions The appendices to this RG provide accident -specific assumptions that are acceptable to the staff for performing site-specific analyses as required by 10 CFR 50.34, 10 CFR Part 52, 10 CFR 50.67, and GDC 19. Licensees should review their licensing-basis documents for guidance on the analysis of radiological DBAs other than those provided in this guide. The DBAs addressed in these attachments were selected from accidents that may involve damage to irradiated fuel. This guide does not address all DBAs with radiological consequences. The inclusion or exclusion of a particular DBA in this guide does not mean that an analysis of that DBA is required or not required. Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST and changes to the facility or to the radiological analyses.

The NRC staff has determined that the analysis assumptions in the appendices to this guide provide an integrated approach to performing the individual analyses, and licensees should address each assumption or propose acceptable alternatives. Such alternatives may be justifiable based on plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing-basis consideration. The assumptions in the appendices are consistent with the AST identified in Regulatory Position 3. Although applicants are free to propose alternatives to these assumptions for consideration by the NRC staff, the use of staff positions inconsistent with these assumptions is beyond the scope of this guidance.

5.3 Atmospheric Dispersion Modeling and Meteorology Assumptions Atmospheric dispersion factors (/Q values) for the EAB, the LPZ, the control room, and, as applicable, the onsite emergency response facility (i.e., the TSC)25 that the NRC approved during initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified in this guide, provided that such values remain relevant to the particular accident, the release characteristics that affect plume rise, the locations from where radiological material is released, and the receptor locations. Licensees should ensure that any previously approved values remain accurate and do not include any misapplication of a methodology or calculational errors in the identified values.

RG 1.145, Revision 1, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, issued November 1982 (Ref. 63), and RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, issued June 2003 (Ref. 64), document methods for determining /Q values.

RG 1.145 and RG 1.194 should be used if the FSAR /Q values are to be revised, or if values are to be determined for new release points, receptor distances, or release characteristics that affect plume rise. In addition to calculating control room /Q values, the modeling methodology outlined in RG 1.194 may be modified to estimate offsite /Q values at offsite boundaries out to distances of 1,200 m if using the procedures consistent with RG 1.249, Use of ARCON Methodology for Calculation of Accident-Related Offsite Atmospheric Dispersion Factors (Ref. 65). EAB /Q values are determined for 25 The radiological habitability analysis for an onsite TSC is performed to support the emergency preparedness review of emergency facilities. Reevaluation of TSC habitability as part of a license amendment request may be needed if a radiological analysis of the TSC is included in that plants current licensing basis. For an onsite TSC, the atmospheric dispersion modeling is handled similarly to that for the control room.

DG-1425, Page 44 the limiting 2-hour period within a 30-day period following the start of the radioactivity release. Control room /Q values are generally determined for initial averaging periods of 0-2 hours and 2-8 hours, and LPZ /Q values are generally determined for an initial averaging period of 0-8 hours. Control room and LPZ /Q values are also generally determined for averaging periods of 8-24 hours, 24-96 hours, and 96-720 hours.

The source term defined in TID-14844 assumes that the entire source term is instantaneously released into the containment atmosphere. Therefore, the maximum release rate coincides with the most conservative 0-2-hour /Q value. In contrast, the AST is assumed to develop over specified time intervals, with the maximum release rate occurring sometime after accident initiation. To ensure a conservative dose analysis, the period of the most adverse release of radioactive materials to the environment, with respect to doses, should be assumed to occur coincident with the period of most unfavorable atmospheric dispersion. One acceptable methodology for calculating the control room and LPZ /Q values is as follows. If the 0-2-hour /Q value is calculated, this value should be used coincident with the maximum 2-hour release to the environment. If the maximum 2-hour release occurs at the beginning of the period of releases to the environment, the 2-8-hour /Q value should be used for the remaining 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the first 8-hour time period. If the maximum 2-hour release occurs sometime after the beginning of the releases, the 2-8-hour /Q value should be used before and after the maximum 2-hour release for a combined total of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The 8-24, 24-96, and 96-720-hour /Q values should similarly be used for the remainder of the release duration. Figure 3 provides examples of aligning /Q values with the maximum 2-hour release.

DG-1425, Page 45 Figure 3. Example alignments of /Q values with the maximum 2-hour release period

DG-1425, Page 46 D. IMPLEMENTATION Licensees generally are not required to comply with the guidance in this regulatory guide. If the NRC proposes to use this regulatory guide in an action that would constitute backfitting, as that term is defined in 10 CFR 50.109, Backfitting, and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests (Ref. 66); affect the issue finality of an approval issued under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants; or constitute forward fitting, as that term is defined in Management Directive 8.4, then the NRC staff will apply the applicable policy in Management Directive 8.4 to justify the action.

If a licensee believes that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may inform the NRC staff in accordance with Management Directive 8.4.

DG-1425, Page 47 REFERENCES26 These references indicate the versions of the documents available at the time of issuance of this regulatory guide (RG). Licensees or applicants using this RG should check all referenced documents to verify that no change has occurred since the issuance of the RG.

1.

Sandia National Laboratories, SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, Albuquerque, New Mexico, April 2023 (ML23097A087).

2.

U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy.

3.

CFR, Licenses, Certifications, and Approvals of Nuclear Power Plants, Part 52, Chapter I, Title 10, Energy.

4.

CFR, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, Part 54, Chapter I, Title 10, Energy.

5.

U.S. Nuclear Regulatory Commission (NRC), NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Washington, DC.

6.

NRC, RG 1.183, Revision 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Washington, DC, October 2023 (ML23082A305).

7.

Deleted.

8.

American National Standards Institute (ANSI)/American Nuclear Society (ANS),

ANSI/ANS-5.4, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel, La Grange Park, Illinois, May 2011 (not publicly available in ADAMS).

9.

NRC, Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants, Federal Register, Vol. 50, pp. 32138 (50 FR 32138), Washington, DC, August 8, 1985.

10.

NRC, Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Final Policy Statement, Federal Register, Vol. 60, pp. 42622-42629 (60 FR 42622), Washington, DC, August 16, 1995.

11.

NRC, SECY-98-144, White Paper on Risk-Informed and Performance-Based Regulations, March 1, 1999 (ML003753601).

12.

NRC, NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, Washington, DC, February 1995 (ML041040063).

26 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

DG-1425, Page 48

13.

U.S. Atomic Energy Commission (now U.S. NRC), TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites, Washington, DC, March 1962 (ML021720780).

14.

NRC, Use of Alternative Source Terms at Operating Reactors, Federal Register, Vol. 64, No.

246, pp. 71990-72002 (64 FR 71990), Washington, DC, December 23, 1999.

15.

NRC, RG 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Washington, DC, July 2000 (ML003716792).

16.

Staff Requirements - SRM-SECY-21-0109, Rulemaking Plan on Use of Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, March 16, 2022 (ML22075A103).

17.

Dickson, E., NRC, internal memorandum to K. Hsueh, Method for a Graded Risk-Informed Performance-Based Control Room Design Criteria Framework, Washington, DC, July 2024 (ML24212A254).

18.

CFR, Reactor Site Criteria, Part 100, Chapter I, Title 10, Energy.

19.

NRC, RG 1.89, Revision 2, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Washington, DC, April 2023 (ML22272A602).

20.

NRC, Nuclear Regulatory Commission International Policy Statement, Federal Register, Vol. 79, No. 132, pp. 39415-39418 (79 FR 39415), Washington, DC, July 10, 2014.

21.

NRC, Management Directive 6.6, Regulatory Guides, Washington, DC.

22.

International Atomic Energy Agency (IAEA), Specific Safety Guide (SSG), No. SSG-2 (Rev. 1),

Deterministic Safety Analysis for Nuclear Power Plants, IAEA Safety Standards, Vienna, Austria, 2019.

23.

IAEA, General Safety Requirements Part 7, Preparedness and Response for a Nuclear or Radiological Emergency, IAEA Safety Standards, Vienna, Austria, 2015.

24.

Sandia National Laboratories, SAND2024-10670, Iron-Chromium-Aluminum Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis - High Burnup Fuel Source Term Accident Sequence Analysis Supplement, Albuquerque, New Mexico, June 2024 (ML24229A069).

25.

Sandia National Laboratories, SAND2024-10673, Cr-coated Accident Tolerant Fuel Concept Source Term Accident Sequence Analysis -High Burnup Fuel Source Term Accident Sequence Analysis Supplement, Albuquerque, New Mexico, June 2024 (ML24229A063).

26.

Sandia National Laboratories, SAND2011-0128, Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup or MOX Fuel, Albuquerque, New Mexico, January 2011 (ML20093F003).

DG-1425, Page 49

27.

NRC, RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Washington, DC, January 2018 (ML17317A256).

28.

NRC, Generic Letter 88-20, Individual Plant Examinations for Severe Accident Vulnerabilities-10 CFR 50.54(f), November 23, 1988 (ML031200499).

29.

NRC, NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Washington, DC, February 1996 (https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1537/).

30.

IAEA, Safety Reports Series No. 53, Derivation of the Source Term and Analysis of the Radiological Consequences of Research Reactor Accidents, IAEA Safety Standards, Vienna, Austria, 2008.

31.

ANSI/ANS, ANSI/ANS-18.1, American National Standard Radioactive Source Term for Normal Operation of Light Water Reactors, La Grange Park, Illinois, 1999 (not publicly available in ADAMS).

32.

NRC, NUREG-0737, Clarification of TMI Action Plan Requirements, Washington, DC, November 1980 (ML102560051).

33.

NRC, NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants, Washington, DC, December 1978 (ML051390356).

34.

NRC, NUREG-0654/FEMA-REP-1, Revision 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Washington, DC, November 1980 (ML040420012).

35.

U.S. Environmental Protection Agency (EPA), EPA-520/1-75-001 Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, Washington, DC, September 1975.

36.

NRC, RG 1.242, Performance-Based Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities, Washington, DC, November 2023 (ML23226A036).

37.

CFR, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, Part 51, Chapter I, Title 10, Energy.

38.

NRC, SECY-98-154, Results of the Revised (NUREG-1465) Source Term Rebaselining for Operating Reactors, Washington, DC, June 30, 1998 (ML992880064).

39.

NRC, RG 1.70, Revision 3, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition), Washington, DC, November 1978 (ML011340122).

40.

NRC, RG 1.206, Revision 1, Applications for Nuclear Power Plants, Washington, DC, October 2018 (ML18131A181).

DG-1425, Page 50

41.

Clifford, P., NRC, internal memorandum to R. Lukes, Technical Basis for Draft RG 1.183 Revision 1 (2021) Non-LOCA Fission Product Release Fractions, Washington, DC, July 28, 2021 (ML21209A524).

42.

Case, M., NRC, internal memorandum to J. Donoghue and M. Franovich, Applicability of Source Term for Accident Tolerant Fuel, High Burn Up and Extended Enrichment, Washington, DC, May 13, 2020 (ML20126G376).

43.

NRC, NUREG/CR-7282, Review of Accident Tolerant Fuel Concepts with Implications to Severe Accident Progression and Radiological Releases, Washington, DC, July 2021 (ML21210A321).

44.

Esmaili, H., NRC, internal memorandum to K. Hsueh, Letter Report on Evaluation of the Impact of Fuel Fragmentation, Relocation, and Dispersal for the Radiological Design Basis Accidents in Regulatory Guide 1.183, Washington, DC, July 20, 2021 (ML21197A067).

45.

Sandia National Laboratories, SAND2024-10674, Multi-region Tabular Source Terms for BWR Containment Design Leakage Assessments, Albuquerque, New Mexico, June 2024 (ML24229A044).

46.

Pacific Northwest National Laboratory, Report 18212, Revision 1, Update of Gap Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard, Richland, Washington, June 2011 (ML112070118).

47.

NRC, Research Information Letter 2021-13, Interpretation of Research on Fuel Fragmentation Relocation, and Dispersal at High Burnup, Washington, DC, December 2021 (ML21313A145).

48.

NRC, NUREG-2121, Fuel Fragmentation, Relocation and Dispersal during the Loss-of-Coolant Accident, Washington, DC, March 2012 (ML12090A018).

49.

NRC, SECY-15-0148, Evaluation of Fuel Fragmentation, Relocation and Dispersal Under Loss-of-Coolant Accident (LOCA) Conditions Relative to the Draft Final Rule on Emergency Core Cooling System Performance during a LOCA (50.46c), Washington, DC, November 30, 2015 (ML15230A200).

50.

Salay, M., Corson, J., Campbell, S., NRC, Office of Research Informal Assistance Request, FFRD Impact on the Containment Source Term, June 12, 2021 (ML21197A069).

51.

NRC, Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants, Federal Register, Vol. 61, No. 239, pp. 65157-65177 (61 FR 65157), Washington, DC December 11, 1996.

52.

International Commission on Radiological Protection (ICRP), ICRP Publication 30, Limits for Intakes of Radionuclides by Workers, Pergamon Press, London, 1979.2 2

Copies of International Commission on Radiological Protection (ICRP) documents may be obtained through the ICRP website: http://www.icrp.org/; 280 Slater Street, Ottawa, Ontario K1P 5S9, Canada; tel: +1(613) 947-9750, fax: +1(613) 944-1920.

DG-1425, Page 51

53.

EPA, EPA-520/1-88-020, Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Washington, DC, 1988.3

54.

EPA, EPA-402-R-93-081, Federal Guidance Report 12, External Exposure to Radionuclides in Air, Water, and Soil, Washington, DC, 1993.

55.

NRC, NUREG/CR-6210, Supplement 1, Computer Codes for Evaluation of Control Room Habitability (HABIT V1.1), Washington, DC, October 1998 (not publicly available in ADAMS).

56.

NRC, NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation, Washington, DC, April 1998 (ML15092A284);

NUREG/CR-7220, SNAP/RADTRAD 4.0: Description of Models and Methods, Washington, DC, June 2016 (ML16160A019).

57.

NRC, RG 1.52, Revision 4, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, Washington, DC, September 2012 (ML12159A013).

58.

Murphy, K.G., and K.W. Campe, Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19, Proceedings of 13th AEC Air Cleaning Conference, U.S. Atomic Energy Commission (now U.S. NRC), Washington, DC, August 1974 (ML19116A064).

59.

NRC, NUREG/CR-6331, Revision 1, Atmospheric Relative Concentrations in Building Wakes, Washington, DC, May 1997 (ML17213A190).

60.

ICRP, ICRP Publication 2, Report of Committee II on Permissible Dose for Internal Radiation, Pergamon Press, London, 1959.

61.

NRC, RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Washington, DC, December 2020 (ML20238B871).

62.

NRC, Technical Specifications Task Force (TSTF) Traveler-505, TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, July 2, 2018 (ML18183A493).

63.

NRC, RG 1.145, Revision 1, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Washington, DC, November 1982 (ML003740205).

64.

NRC, RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, Washington, DC, June 2003 (ML031530505).

3 Copies of EPA Library Services documents may be obtained through the agencys website: https://www.epa.gov/libraries.

DG-1425, Page 52

65.

NRC, RG 1.249, Use of ARCON Methodology for Calculation of Accident-Related Offsite Atmospheric Dispersion Factors, Washington, DC, August 2023 (ML22024A241).

66.

NRC, Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Washington, DC.

DG-1425, Appendix A, Page A-1 APPENDIX A ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF LIGHT-WATER REACTOR MAXIMUM HYPOTHETICAL LOSS-OF-COOLANT ACCIDENTS The assumptions in this appendix are acceptable to the staff of the U.S. NRC for evaluating the radiological consequences of maximum hypothetical accident (MHA) loss-of-coolant accidents (LOCAs) at light-water reactors. These assumptions supplement the guidance in the main body of this guide.

Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. A-1), defines LOCAs as those postulated accidents that result from a loss-of-coolant inventory at rates that exceed the capability of the reactor coolant makeup system. Leaks up to a double-ended rupture of the largest pipe of the reactor coolant system (RCS) are included. Separate mechanistic analyses are performed using a spectrum of break sizes to evaluate fuel and emergency core cooling system performance for conformance with 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. The NRCs regulatory framework ensures that licensees design their emergency core cooling systems (ECCSs) to meet acceptance criteria that limit peak cladding temperature, maximum cladding oxidation, and maximum hydrogen generation and that reactor cores remain in coolable geometries such that long-term cooling can be provided following a LOCA. Therefore, the ECCS functions to limit reactor core damage during a LOCA such that resultant radiological consequences are very low, especially when compared to acceptance criteria that apply to the MHA LOCA.

The MHA LOCA, like all design-basis accidents (DBAs), is a conservative surrogate accident that is intended to challenge aspects of the facility design. Specifically, the MHA LOCA is the DBA that is used to evaluate the siting of new reactors and the design of systems, structures, and components (SSCs) that function to limit fission product releases to the environment (e.g., containments). Since light-water reactors, with their ECCSs, are designed such that LOCAs are not expected to result in significant radiological releases to the environment, the NRC regulatory framework requires that MHA-LOCA analyses assume a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission products. This approach ensures that the designs of SSCs that function to limit fission product releases are appropriately analyzed.

The MHA-LOCA analysis should calculate the limiting dose consequences to the public offsite and the workers in the control room, assuming a deterministic substantial core damage source term, discussed below, released into an intact containment.

A-1.

Source Term Regulatory Position 3 of this guide provides acceptable assumptions about core inventory and the release of radionuclides from the fuel.

A-1.1 If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical form of radioiodine released to the containment should be assumed to be 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine species, including those from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created

DG-1425, Appendix A, Page A-2 during the MHA LOCA event (e.g., radiolysis products). Except for elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

A-2.

Transport in Primary Containment Acceptable assumptions related to the transport, reduction, and release of radioactive material in and from the primary containment in a pressurized-water reactor (PWR) or the drywell in a boiling-water reactor (BWR) are as follows:

A-2.1 The radioactivity released from the fuel should be assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment (in a PWR) or the drywell (in a BWR) as it is released. This distribution should be adjusted if there are internal compartments that have limited ventilation exchange. The suppression pool free air volume may be included, provided there is a mechanism to ensure mixing between the drywell and the wetwell. The release into the containment or drywell should be assumed to terminate at the end of the early in-vessel release phase.

A-2.2 Reduction in airborne radioactivity in the containment due to natural deposition within the containment may be credited. Section 6.5.2, Containment Spray as a Fission Product Cleanup System, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP) (Ref. A-2), describes an acceptable model for removal of iodine and aerosols. The NRC staff does not accept the practice of deterministically assuming that a 50 percent plateout of iodine is released from the fuel, or the aerosol reductions (decontamination) calculated in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, issued July 1996 (Ref. A-3), because the deterministic 50 percent plateout value and the aerosol reductions calculated in NUREG/CR-6189 are inconsistent with the characteristics of the revised source terms. However, the methods used in NUREG/CR-6189 may be credited on a case-by-case basis if they are adjusted to incorporate the revised MHA-LOCA source term in this revision of Regulatory Guide (RG) 1.183. When these adjusted NUREG/CR-6189 methods are used, the DBA analyses should use the 10th-percentile values unless otherwise justified. Some licensees may consider specific containment design features to evaluate aerosol fission product removal. The amount of removal will be evaluated on a case-by-case basis. Reduction in airborne aerosol radioactivity in the containment by both sprays and gravitational settling should be evaluated on a case-by-case basis.

A-2.3 Reduction in airborne radioactivity in the containment by containment spray systems that have been designed and are maintained in accordance with section 6.5.2 of the SRP (Ref. A-2) may be credited. Section 6.5.2 of the SRP and NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays, issued June 1993 (Ref. A-4), describe acceptable models for the removal of iodine and aerosols (DBA analyses should use the 10th-percentile values).

The evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray droplets. In addition, since spray droplets are assumed to be ineffective once they impact a structure, the obstructions in drywells and containments (particularly in BWR Mark I and Mark II drywells) should be considered in the determination of decontamination factors (DFs) and removal coefficients credited for the drywell or containment.

The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed region volume per hour, unless other rates are justified. On a case-by-case basis, the licensee may consider containment mixing rates determined by the cooldown rate in the sprayed region and the buoyancy-driven flow that results. The containment

DG-1425, Appendix A, Page A-3 building atmosphere may be considered a single well-mixed volume if the spray covers at least 90 percent of the containment building space and an engineered-safety-feature (ESF) ventilation system is available for adequate mixing of the unsprayed compartments.

As provided in section 6.5.2 of the SRP, the maximum DF for elemental iodine is based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached and that the elemental iodine removal is limited to 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />s-1. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays. The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled Total in tables 1 and 2 of this guide, multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., the SRP methodology treats aerosols as particulate).

A-2.4 Reduction in airborne radioactivity in the containment by in-containment recirculation filter systems may be credited if these systems meet the guidance of RG 1.52, Revision 4, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, issued September 2012 (Ref. A-5). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.

A-2.5 Historically, reduction in airborne radioactivity in the containment by suppression pool scrubbing has not been credited in licensing actions for operating BWRs; however, Regulatory Position 3.2 now provides release fractions for pathway-specific release fractions using mechanistic transport modeling that the staff considers acceptable. For suppression pool solutions having a pH less than 7, elemental iodine vapor should be conservatively assumed to evolve into the containment atmosphere.

A-2.6 Reduction in airborne radioactivity in the containment by retention in ice condensers, or other ESFs not addressed above, should be evaluated on a case-by-case basis. See SRP Section 6.5.4, Ice Condenser as a Fission Product Cleanup System (Ref. A-2).

A-2.7 The evaluation should assume that the primary containment (including the wetwell for Mark I and II containment designs) will leak at the peak pressure technical specification (TS) leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50 percent of the TS leak rate. For BWRs, leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by plant configuration and analyses, to a value not less than 50 percent of the TS leak rate. Leakage from sub-atmospheric containments is assumed to terminate when the containment is brought to and maintained at a sub-atmospheric condition as defined by the TS.

The licensees evaluation of the post-accident containment pressure response may credit safety-related systems to further reduce containment pressure, and thereby the associated containment leak rate, in the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and for the duration of the accident. Regulatory Position 5.1.2 addresses credit for ESFs.

A-2.8 If the primary containment is routinely purged during power operations, the licensee should analyze releases via the purge system before containment isolation and should sum the resulting doses with the postulated doses from other release paths. The purge release evaluation should assume that 100 percent of the radionuclide inventory in the RCS liquid is released to the containment at the initiation of the MHA LOCA. This inventory should be based on the TS RCS

DG-1425, Appendix A, Page A-4 equilibrium activity. Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the licensee should consider release fractions associated with the gap release and early in-vessel release phases as applicable.

A-3.

Dual Containments For facilities with dual containment systems, the acceptable assumptions related to the transport, reduction, and release of radioactive material in and from the secondary containment or enclosure buildings are as follows:

A-3.1 Leakage from the primary containment should be considered to be collected, processed by ESF filters, if any, and released to the environment via the secondary containment exhaust system during periods in which the secondary containment has a negative pressure as defined in the TS.

Credit for an elevated release should be assumed only if the point of physical release is more than 2.5 times the height of any adjacent structure.

A-3.2 Leakage from the primary containment is assumed to be released directly to the environment as a ground-level release during any period in which the secondary containment does not have a negative pressure as defined in the TS.

A-3.3 The effect of high windspeeds on the ability of the secondary containment to maintain a negative pressure should be evaluated on a case-by-case basis. The windspeed to be assumed is the 1-hour average value that is exceeded only 5 percent of the total number of hours in the dataset. Ambient temperatures used in these assessments should be the 1-hour average value that is exceeded in either 5 percent or 95 percent of the total numbers of hours in the dataset, whichever is conservative for the intended use (e.g., if high temperatures are limiting, use those exceeded only 5 percent of the time) (Ref. A-6).

A-3.4 Credit for dilution in the secondary containment may be allowed when adequate means to cause mixing can be demonstrated. Otherwise, the leakage from the primary containment should be assumed to be transported directly to exhaust systems without mixing. Credit for mixing, if found to be appropriate, should generally be limited to 50 percent. This evaluation should consider the magnitude of the containment leakage in relation to contiguous building volume or exhaust rate, the location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede streamflow between the release and the exhaust.

A-3.5 Primary containment leakage that bypasses the secondary containment should be evaluated at the bypass leak rate incorporated in the TS. If the bypass leakage is through water (e.g., via a filled piping run that is maintained full), credit for retention of iodine and aerosols may be considered on a case-by-case basis. Similarly, deposition of aerosols and elemental halogens in gas-filled lines may be considered on a case-by-case basis.

A-3.6 Reduction in the amount of radioactive material released from the secondary containment because of ESF filter systems may be taken into account, provided that these systems meet the guidance of RG 1.52 (Ref. A-5).

A-4.

Assumptions on Engineered-Safety-Feature System Leakage ESF systems that recirculate sump water outside of the primary containment are assumed to leak during their intended operation. This release source includes leakage through valve packing glands, pump

DG-1425, Appendix A, Page A-5 shaft seals, flanged connections, and other similar components. This release source may also include leakage through valves isolating interfacing systems (Ref. A-7). The licensee should analyze the radiological consequences from the postulated leakage and combine them with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the MHA LOCA. The following assumptions are acceptable for evaluating the consequences of leakage from ESF components outside the primary containment for BWRs and PWRs:

A-4.1 With the exception of noble gases, all fission products released from the fuel to the containment (as defined in the applicable tables 1 through 2 of this guide) should be assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. For BWRs, table 1.2 provides a mechanistic model of the core inventory fraction retained in the suppression pool based on analyses from SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, issued April 2023 (Ref. A-8). Table 1.2 is a suitably conservative mechanistic model for the transport of airborne activities in the suppression pool. For PWRs, in lieu of the deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameter values that make spray and deposition models conservative in estimating containment airborne leakage are nonconservative in estimating the buildup of sump activity.

A-4.2 The leakage should be taken as 2 times1 the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the TS, or licensee commitments to item III.D.1.1 of NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980 (Ref. A-9), would require declaring such systems inoperable. Design leakage from any systems not included in the TS that transport primary coolant sources outside of containment should be added to the total leakage. The applicant should justify the design leakage used. The leakage should be assumed to start at the earliest time when the recirculation flow occurs in these systems, and to end at the latest time when the releases from these systems are terminated. It should account for the ESF leakage at accident conditions. Design leakage through valves isolating ESF recirculation systems from tanks vented to the atmosphere (e.g., the pump miniflow return to the refueling water storage tank in the emergency core cooling system) should also be considered.

A-4.3 With the exception of iodine, all radioactive materials in the recirculating liquid should be assumed to be retained in the liquid phase.

A-4.4 If the temperature of the leakage exceeds 212 degrees Fahrenheit (°F), the fraction of total iodine (i.e., aerosol, elemental, and organic) in the liquid that becomes airborne should be assumed to equal the fraction of the leakage that flashes to vapor. This flash fraction (FF) should be determined using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment, using the following formula:

1 The multiplier of 2 is used to account for increased leakage in these systems over the duration of the accident and between surveillances or leakage checks.

DG-1425, Appendix A, Page A-6 fg f

f h

h h

FF 2

1

=

Where:

hf1 is the enthalpy of liquid at system design temperature and pressure, hf2 is the enthalpy of liquid at saturation conditions (14.7 pounds per square inch absolute, 212°F), and hfg is the heat of vaporization at 212°F.

A-4.5 If the temperature of the leakage is less than 212°F or the calculated FF is less than 10 percent, the amount of iodine that becomes airborne should be assumed to be 10 percent of the total iodine activity in the leaked fluid, unless a smaller amount can be substantiated. The justification of such values should consider the sump pH history; changes to the leakage pH caused by pooling on concrete surfaces, leaching through piping insulation, evaporation to dryness, and mixing with other liquids in drainage sumps; area ventilation rates and temperatures; and subsequent re-evolution of iodine.

A-4.6 The radioiodine that is postulated to be available for release to the environment is assumed to be 97 percent elemental and 3 percent organic.2 Reduction in release activity by dilution or hold-up within buildings, or by ESF ventilation filtration systems, may be credited where applicable.

Filter systems used in these applications should be evaluated using the guidance of RG 1.52 (Ref. A-5).

A-5.

Main Steam Isolation Valve Leakage in Boiling-Water Reactors For BWRs, the main steam isolation valves (MSIVs) have design leakage that may result in a radioactivity release. The licensee should analyze and combine the radiological consequences from postulated MSIV leakage with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the MHA LOCA.

Two methods are presented below to compute aerosol deposition within main steamlines. Each method computes similar removal coefficients that are suitable for radiological consequence calculations.

These methods are not valid if credit is also taken for aerosol removal from drywell sprays, or for other containment aerosol removal processes, when modeling the MSIV leakage release pathway without accounting for the change in particle size distribution due to these containment removal processes. The two MSIV leakage models are the following:

a.

reevaluated Accident Evaluation Branch (AEB)-98-03, Assessment of Radiological Consequences for the Perry Pilot Plant Application Using the Revised (NUREG-1465)

Source Term, dated December 9, 1998 (Ref. A-11), with multi-group, and

b.

numerical integration.

2 The 97 percent elemental, 3 percent organic speciation is a conservative deterministic assumption based on the hypothesis that most of the iodine released to the environment will be in elemental form, with a small percentage converted to organic, as supported in section 3.5 of NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, issued February 1995 (Ref. A-10).

DG-1425, Appendix A, Page A-7 The assumptions in the following subsections are acceptable for evaluating the consequences of MSIV leakage.

A-5.1 The source of the MSIV leakage should be assumed to be the BWR core inventory time-averaged fraction in the steamline, as presented in table 1.3 (see Ref. A-12). This represents the overall, time-averaged fraction of the fission product inventory in the portion of the steamline from just downstream of the first steam relief valve to the first MSIV in all steamlines. To ascertain the fission product concentration, C, for the region-specific BWR source term analyses, the licensee should divide the fission product inventory by the plant-specific volume of this portion of the steamline multiplied by the number of steamlines.

Table 1.3 is a suitably conservative mechanistic model for the transport of airborne radionuclides in the steamlines. When using this approach, no credit should be taken for aerosol deposition upstream of the inboard MSIV before the end of the early in-vessel release phase. Up to the end of the early in-vessel phase (i.e., the gap and early in-vessel phases), the concentration, C, of aerosols should be modeled as a constant source. After the early in-vessel release phase, credit for aerosol deposition may be taken, as aerosols and steam condense from contact with cooling surfaces. Modeling aerosol deposition may take on the form of an exponentially declining input rate as = (0), where is the rate constant having units of inverse time (t-1). Reference A-13 provides an acceptable long-term removal rate constant of 0.12 hr-1 (per hour) in the cases where the main steam line is cooled.

For new BWR designs or license amendments that propose changes from a referenced design control document, other models of MSIV source concentration will be considered on a case-by-case basis. In general, the concepts used in developing the guidance for BWR Mark I, II, and III plants may be followed as applicable to designs under consideration.

A-5.2 The chemical form of radioiodine released to the drywell should be assumed to be 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide. Apart from elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

A-5.3 All the MSIVs should be assumed to leak at the maximum leak rate above which the TS would require declaring the MSIVs inoperable. The leakage should be assumed to continue for the duration of the accident, as specified in table 7 of this guide, and should be assigned to steamlines so that the accident dose is maximized. Postulated leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by site-specific analyses.

A-5.4 A reduction in MSIV releases caused by holdup and deposition in the main steam piping and main condenser, including the treatment of air ejector effluent by offgas systems, may be credited if the components and piping systems used in the release path can perform their safety function during and following a safe shutdown earthquake and are powered by emergency power sources.

These reductions are allowed for steam system piping segments that are enclosed by physical barriers, such as closed valves. The piping segments and physical barriers should be designed, constructed, and maintained to seismic Category I guidelines as specified in RG 1.29, Seismic Design Classification for Nuclear Power Plants (Ref. A-14). Alternatively, operating license holders may evaluate and demonstrate the piping segments and barriers to be rugged as described in Regulatory Position A-5.5. The amount of reduction allowed will be evaluated on a case-by-case basis and is to be justified based on the alternative drain pathways established by operating procedures and the potential leakage pathways to the environment.

DG-1425, Appendix A, Page A-8 On March 3, 1999, the NRC staff issued a safety evaluation (Ref. A-15) of the GE topical report NEDC-31858P, Revision 2, [Boiling Water Reactor Owners Group] BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, issued September 1993 (Ref. A-16). In its safety evaluation, the staff found the BWROG report to be an acceptable method for direct reference in individual submittals on MSIV leakage, subject to the conditions and limitations described in the safety evaluation. For the purposes of DBA radiological consequence analyses, based on the information in the BWROG report, proposed MSIV leakage limits in excess of 200 standard cubic feet per hour (scfh) per steamline and in excess of 400 scfh for total MSIV leakage will be considered on a case-by-case basis with sufficient justification. The single valve limitation is based on the consideration that leakage in excess of 200 scfh may indicate a substantial valve defect. The total MSIV leakage limitation of 400 scfh is based on considerations of the relationship of MSIV leakage rate to the allowable containment leakage rate (La), as well as providing defense in depth related to the single valve limitation.

Consistent with the BWROG report, the following information related to the reliability of the pathway to the main condenser should be provided when an alternative drain pathway is credited:

a.

the alternative drain pathway and the basis for its functional reliability, commensurate with its intended safety-related function;

b.

the maintenance and testing program for the active components (such as valves) in the alternative drain pathway, and a confirmation that the valves that are required to open the alternative drain pathway are included in the inservice testing program;

c.

how the alternative drain pathway addresses the single failure of active components to verify its availability to convey MSIV leakage to the condenser;

d.

a secondary pathway to the condenser; and

e.

emergency operating procedures that may be required to identify necessary operator actions to mitigate MSIV leakage consequences utilizing the alternative drain pathway if a highly reliable power source is available or to identify necessary operator actions to mitigate MSIV leakage consequences using the alternative drain pathway if a highly reliable power source is unavailable.

A-5.5 Licensees that have already evaluated the seismic ruggedness of the steamlines, alternate drain paths, and the main condenser, and who have obtained prior staff approval, may credit the piping addressed in that approval. Licensees that have not previously applied for such approval may do so in accordance with the guidance in Reference A-14 or using the revised seismic analysis method described below in Revised Seismic Analysis of the Alternative Drain Pathway for establishing a seismically rugged alternative drain pathway to the condenser.

Licensees choosing either of these methods should define the alternative drain pathway to the main condenser and provide the basis for its reliability. The basis for reliability should include the qualification and redundancy of valves that must change position to establish the pathway, operator training, and procedures governing establishment of the alternative drain pathway as described in Regulatory Position A-5.4.

Revision 1 to this guide provided an updated seismic analysis method. This seismic analysis method, described below in Seismic Analysis of the Alternative Drain Pathway, provides a

DG-1425, Appendix A, Page A-9 method to demonstrate that the defined drain path to the condenser is seismically rugged. The NRC staffs determination of reasonable assurance that SSCs in the pathway will not undergo gross seismic failure is based upon the staffs consideration of engineering information, operating experience, and probabilistic insights related to seismic events, including the following:

margin in material strength provided by use of an appropriate code of record for design of the pathway(s) to the condenser, and by construction to augmented quality standards in the areas of material certification, testing, and nondestructive examination; additional margin in material strength provided by the high-pressure and high-temperature design of the SSCs in the drain pathway to accommodate seismic loads after an accident considered for the analysis in this RG; low failure probability (5 percent or less) for the SSCs in the drain pathway at safe-shutdown earthquake seismic accelerations for the majority of operating plants, especially BWRs (ranging from 0.12g to 0.25g peak ground acceleration), based on a conservative median fragility value to represent these SSCs; available information from post-earthquake walkdowns performed for the nuclear power plants at Kashiwazaki-Kariwa in Japan and at North Anna Power Station in the United States, and consideration of the walkdowns performed by licensees to identify weaknesses in SSCs when exposed to seismic events (including beyond-design-basis seismic events) as part of the NRCs post-Fukushima actions resulting from Near-Term Task Force Recommendation 2.3: Seismic (Ref. A-17); and aging management programs to address material degradation due to aging in SSCs in the alternative pathway are addressed, for licensees that currently have extended operating licenses or will apply for such licenses in the future.

The application of these design and construction quality standards provides reasonable confidence regarding the robustness of the SSCs in the alternative drain pathway. Additionally, probabilistic information developed through earthquake operating experience and analysis of walkdown results confirms a low likelihood of failure at design-basis seismic accelerations.

Consideration of dynamic loading conditions or complete walkdowns of the credited pathway provides greater assurance of ruggedness for sites with higher seismic hazards.

Seismic Analysis of the Alternative Drain Pathway All licensees choosing this alternative should describe the code of record used for the main steamlines and the extent of quality assurance measures applied to the design, materials, and fabrication of the steamlines and attached piping. The description should include the alternate pathway identified to the main condenser and the basis for the reliability of this pathway. The licensee should also provide the following information for the site, as applicable for the site seismic hazard tier:

Note: The method for seismic analysis of the alternative drain pathway presented below is intended to be used with the aerosol deposition models in Regulatory Position A-5.6.

DG-1425, Appendix A, Page A-10 (1)

If the piping and valves in the alternate pathway have been subjected to dynamic seismic analysis for the as-built configuration to a code of record (e.g., American Society of Mechanical Engineers (ASME) B31.1, Power Piping (Ref. A-18)), and the magnitude of the seismic response spectrum for the analysis equals or exceeds the licensees safe-shutdown earthquake, then a description of the dynamic analysis provides sufficient justification.

(2)

If the SSCs in the alternate pathway have not been subjected to dynamic seismic analysis to a code of record (e.g., ASME B31.1), and if the peak spectral acceleration of the ground motion response spectrum based on the licensees most recent site-specific probabilistic seismic hazard is at most 0.4g, then the justification should include the following:

a discussion of seismic capacity and margin present in the relevant SSCs, including the condenser, based on their design code(s) of record, insights from plant-specific seismic assessments performed as part of the Individual Plant Examination for External Events for relevant SSCs, a walkdown of a sample of the relevant SSCs, including the condenser, performed by knowledgeable licensee staff members, to verify that they have been constructed as designed, and confirmatory calculations for a sample of piping supports, to verify that they provide acceptable flexibility at terminal ends of piping and major branch connections.

The extent of the selected samples should be justified based on the plant-specific seismic hazard and quality assurance practices applied to design and fabrication. Details of the walkdown(s), including the qualifications of the licensee staff members performing them, should be retained in archival documentation.

(3)

If the SSCs in the alternate pathway have not been subjected to dynamic seismic analysis to a code of record (e.g., ASME B31.1), and if the peak spectral acceleration of the ground motion response spectrum based on the licensees most recent site-specific probabilistic seismic hazard is greater than 0.4g, then the justification should include the following:

a discussion of seismic capacity and margin present in the relevant SSCs, including the condenser, based on their design code(s) of record, insights from the Individual Plant Examination for External Events as described above, walkdown(s) of the SSCs in the alternate pathway, including the condenser, performed by knowledgeable licensee staff members, to ensure that items adversely affecting the seismic capacity of relevant SSCs (e.g., loose or missing anchorages and degraded pipe supports) are identified and corrected, and

DG-1425, Appendix A, Page A-11 confirmatory calculations for a sample of piping supports, to verify that they provide acceptable flexibility at terminal ends of piping and major branch connections.

The extent of the selected samples should be justified based on the plant-specific seismic hazard and quality assurance practices applied to design and fabrication. Details of the walkdown(s), including the qualifications of the licensee staff members performing them, should be retained in archival documentation.

A-5.6 For BWRs with Mark I, II, or III containment designs, aerosol deposition in horizontal volumes that meet Regulatory Position A-5.4 or A-5.5 may be credited as described below.3 The NRC staff will consider aerosol deposition models for BWR designs other than those with Mark I, II, or III containment designs on a case-by-case basis.

A-5.6.1 Reevaluated AEB-98-03 with the multi-group method: At the beginning of the accident, aerosol deposition removal coefficients for the main steamline piping between the MSIVs and downstream of the MSIVs may apply an updated AEB 98-03 use of the Stokes settling velocity physics parameters with the multi-group method. After the early in-vessel release phase, credit for aerosol deposition upstream of the inboard MSIV may also be credited as steam condenses from contact with cooling surfaces as described in Regulatory Position A-5.1. The method below computes both total effective aerosol removal efficiencies (TEAREs) (i.e., filter efficiencies) and equivalent removal coefficients (hr-1).

When evaluating the Stokes settling velocity, use the aerodynamic mass median diameter (AMMD),, based on a distribution directly measured from experiments to evaluate the settling velocity where the specific aerosol parameter distributions of shape factor, density, and volume-equivalent diameter do not need to be defined. Therefore, the Stokes settling velocity can be rewritten in terms of the aerodynamic diameter,, as follows:

=

()

(Equation A-1)

Where:

= aerosol unit density = 1.0 g/cm3,

= aerosol aerodynamic diameter,

= gravitational acceleration,

() = Cunningham slip factor as a function of, and

= viscosity.

The document State-of-the-Art Report on Nuclear Aerosols, issued in 2009 (Ref. A-19),

provides a summary of experimental observations from integral experiments involving irradiated fuel to infer characteristics of aerosols under light-water reactor severe accident conditions. The State-of-the-Art Report recommends the use of a log-normal distribution for aerosols in the RCS (AMMD 1.0 microns (m) with a geometric standard deviation,, of 2.0), and provides PHÉBUS-Fission Product aerosol measurements in containment (AMMD of 3.0 m and of 2.0). Considering the MHA-LOCA modeling approach, which considers no pipe break and where the deposition properties after reflood are based on the characteristics of the RCS and containment aerosol (i.e., the approach considers the effects of an active emergency core cooling 3

The credit described in this regulatory position will supersede the aerosol settling estimates previously given in the NRC staff document AEB 98-03 (Ref. A-11) when Revisions 1 and 2 of RG 1.183 are used.

DG-1425, Appendix A, Page A-12 system), the methods in Regulatory Positions A-5.6.1 and A-5.6.2 should assume a log-normal aerosol diameter distribution with an AMMD of 2.0 m and of 2.0. Assume as fixed values a

() of 1 and a viscosity of 1.93x10-5 Pascal-second. Licensees should perform at least 10,000 trials to develop a settling velocity distribution dataset. Note that while the NRC memorandum of Reference A-20 addresses the methods discussed in Regulatory Positions A-5.6.1 and A-5.6.2, it does not establish regulatory positions. For example, that memorandum does not endorse input parameters such as the AMMD, assumed in example calculations, and statements on the validity of the existing 20-group method.

The multi-group method should include the following assumptions and steps to estimate removal coefficients:

a.

Discretize the settling velocity dataset into at least 2,000 equal-width groups. Assign a relative probability to each group by dividing the number of data points within each group by the sample size (e.g., 10,000 trials) to determine the group probabilities.

Identify the midpoint of each group to represent the settling velocity for that group.

b.

Compute each groups aerosol filter efficiency using the following method. By rearranging equations 2, 3 and 4 from Reference A-11, the filter efficiency,, is computed by using the group settling velocity, the settling area, the volumetric flow rate, and the volume of the well-mixed region being modeled as follows:

= 1

= 1

= 1

(Equation A-2)

Where:

= removal, or filter, efficiency,

= settling velocity (ft/hr),

A = settling area (ft2),

= outgoing concentration of nuclides in the pipe segment volume,

= initial concentration of nuclides in the pipe segment volume, Q = volumetric flow rate into pipe segment volume (ft3/hr), and

= equivalent removal coefficient (hr-1).

Account for the effect of the changing settling velocity distribution in the downstream volumes by adjusting the downstream volume efficiencies by multiplying them by the prior volume aerosol filter removal efficiency.

c.

Compute the TEAREs (i.e., filter efficiencies) and equivalent removal coefficients, (hr-1), for a credited volume by the following method. Compute the probability-weighted aerosol filter efficiency by multiplying the aerosol filter efficiency by the group probability from step 1. Then sum all the probability-weighted aerosol removal efficiencies to obtain the TEARE. By solving for in equation A-2 the removal coefficients are computed to yield:

DG-1425, Appendix A, Page A-13

=

(Equation A-3)

Where:

filt = TEAREs, Q = volumetric flow rate into credited volume, and V = well-mixed pipe free volume.

A-5.6.2 Numerical Integration: At the beginning of the accident, aerosol deposition removal coefficients for the main steamline piping between the MSIVs and downstream of the MSIVs may apply the methods use of the Stokes settling velocity and physics parameters of Regulatory Position A-5.6.1 with the numerical integration method. After the early in-vessel release phase, credit for aerosol deposition upstream of the inboard MSIV may also be credited, as steam condenses from contact with cooling surfaces as described in Regulatory Position A-5.1. The equation for a normalized number distribution (()) of particles of aerodynamic diameter ()

is given (Ref. A-21) as follows:

() =

(Equation A-4)

Where:

= geometric standard deviation, and

= geometric mean (which, for a log-normal distribution, is the same as the median diameter).

According to the Hatch-Choate equations, the AMMD is related to the median diameter (), in meters, as follows:

= [3 ln

]

(Equation A-5)

Discretize the range of particle diameters, da, from 1x10-9 m to 1x10-3 m into 150 groups. For each group, apply equation A-4 to compute the normalized number distribution, (()). Then, for each discretized group, (1) compute its settling velocity by applying equation A-1, (2) use equation 3 from Reference A-20 to compute inboard and outboard concentrations of particles leaving the volumes, (3) sum up the inboard and outboard concentrations using an appropriate numerical integration technique (such as the trapezoidal method), and (4) use equation A-2 to then compute the filter efficiencies. Finally, use equation A-3 to convert filter efficiencies,,

into removal coefficients (in units of hr-1).

A-5.6.3 Aerosol deposition removal coefficients for the condenser using a multi-group method and numerical integration are acceptable and will be evaluated on a case-by-case basis.

A-5.7 Reduction of the amount of released elemental iodine by plateout deposition on steam system piping may be credited, but the amount of reduction in concentration allowed will be evaluated on a case-by-case basis. The model should assume well-mixed volumes. Reference A-22 provides guidance on an acceptable model.

DG-1425, Appendix A, Page A-14 A-5.8 Reduction of the amount of released organic iodine (e.g., using the Brockman-Bixler model in RADTRAD (Ref. A-23)) should not be credited.

A-5.9 In the absence of collection and treatment of releases by ESFs such as the MSIV leakage control system, or as described in Regulatory Position A-5.4, then the MSIV leakage should be assumed to be released to the environment as an unprocessed, ground-level release.

A-5.10 Hold-up and dilution of MSIV leakage releases into the turbine building should not be assumed.

A-6.

Containment Purging The licensee should analyze the radiological consequences from post-LOCA primary containment purging as a combustible gas or pressure control measure. If the installed containment purging capabilities are maintained for the purposes of severe accident management and are not credited in any design-basis analysis, radiological consequences need not be evaluated. If primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with the consequences postulated for other fission product release paths to determine the total calculated radiological consequences of the LOCA. The licensee may consider the reduction in the amount of radioactive material released through ESF filter systems using the guidance in RG 1.52 (Ref. A-5).

DG-1425, Appendix A, Page A-15 APPENDIX A REFERENCES1 A-1.

U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy.

A-2.

U.S. Nuclear Regulatory Commission (NRC), NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Washington, DC.

A-3.

NRC, NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, Washington, DC, July 1996 (ML100130305).

A-4.

NRC, NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays, Washington, DC, June 1993 (ML063480542).

A-5.

NRC, Regulatory Guide (RG) 1.52, Revision 4, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, Washington, DC, September 2012 (ML12159A013).

A-6.

NRC, Information Notice 88-76, Recent Discovery of a Phenomenon Not Previously Considered in the Design of Secondary Containment Pressure Control, Washington, DC, September 19, 1988 (ML031150101).

A-7.

NRC, Information Notice 91-56, Potential Radioactive Leakage to Tank Vented to Atmosphere, Washington, DC, September 19, 1991 (ML031190264).

A-8.

Sandia National Laboratories, SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, Albuquerque, New Mexico, April 2023 (ML23097A087).

A-9.

NRC, NUREG-0737, Clarification of TMI Action Plan Requirements, Washington, DC, November 1980 (ML102560051).

A-10. NRC, NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, Washington, DC, February 1995 (ML041040063).

A-11. NRC, Accident Evaluation Branch (AEB)-98-03, Assessment of Radiological Consequences for the Perry Pilot Plant Application Using the Revised (NUREG-1465) Source Term, Washington, DC, December 9, 1998 (ML011230531).

1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

DG-1425, Appendix A, Page A-16 A-12. Sandia National Laboratories, SAND2024-10674, Multi-region Tabular Source Terms for BWR Containment Design Leakage Assessments, Albuquerque, New Mexico, June 2024 (ML24229A044).

A-13. Yuan, Z., Campbell, S., Salay, M., Esmaili, H., NRC, RES/FSCB 2024-01, Analysis of the Source Term in the Main Steamlines of BWRs after Early In-Vessel Phase, Washington, DC, July 2024 (ML24222A207).

A-14. NRC, RG 1.29, Seismic Design Classification for Nuclear Power Plants, Washington, DC.

A-15. NRC, letter to T.A. Green, Boiling Water Reactor Owners Group (BWROG) Projects, Safety Evaluation of GE Topical Report, NEDC-31858P, Revision 2, BWROG Report for Increasing MSIV Leakage Limits and Elimination of Leakage Control Systems, September 1993, Washington, DC, March 3, 1999 (ML010640286).

A-16. Boiling Water Reactor Owners Group (BWROG), NEDC-31858P, BWROG Report for Increasing MSIV Leakage Rate Limits and Elimination of Leakage Control Systems, Birmingham, Alabama, September 1993 (Not publicly available).

A-17. Leeds, E.J. and Johnson, M. R., NRC, letter to all power reactor licensees and holders of construction permits in active or deferred status, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, Washington, DC, March 12, 2012 (ML12053A340).

A-18. American Society of Mechanical Engineers (ASME), ASME B31.1, Power Piping, New York, NY, 2001.

A-19. Allelein, H.-J.A., NEA/CSNI/R(2009)5, State-of-the-Art Report on Nuclear Aerosols, Nuclear Energy Agency, Committee on the Safety of Nuclear Installations, Paris, 2009.

A-20. Dickson, E., NRC, internal memorandum to K. Hsueh, Technical Basis for Draft RG 1.183 Revision 1 (2021) Re-evaluated AEB-98-03 Settling Velocity Method, the Multi-Group Method, and the Numerical Integration Method, Washington, DC, July 29, 2021 (ML21141A006).

A-21. Williams, M.M.R., Aerosol Science Theory and Practice, Pergamon Press, New York, New York, 1991.

A-22. J.E. Cline and Associates, Inc., MSIV Leakage Iodine Transport Analysis, letter report, Rockville, Maryland, March 26, 1991 (ML003683718).

A-23. NRC, NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal and Dose Estimation, Washington, DC, April 1998 (ML15092A284); NUREG/CR-7220, SNAP/RADTRAD 4.0: Description of Models and Methods, Washington, DC, June 2016 (ML16160A019).

DG-1425, Appendix B, Page B-1 APPENDIX B ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT This appendix provides assumptions acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of a fuel handling accident at a light-water reactor. These assumptions supplement the guidance in the main body of this guide.

B-1.

Source Term Regulatory Position 3 of this guide provides acceptable assumptions regarding core inventory and the release of radionuclides from the fuel. The following assumptions also apply:

B-1.1 The number of fuel rods assumed to be damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples); the height of the drop; and the compression, torsion, and shear stresses on the irradiated fuel rods. The analysis should also consider damage to adjacent fuel assemblies, if applicable (e.g., for events over the reactor vessel).

B-1.2 The fission product release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.

B-1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool should be assumed to be 95 percent cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. This regulatory position and those in B-2 use, in part, the transport models discussed in References B-1 and B-2.

All the gap activity in the damaged rods is assumed to be released over two phases:

Phase 1the instantaneous release from the rising bubbles (from start of accident to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). Elemental iodine and organic iodine are conservatively assumed to be in vapor form. Elemental iodine is subsequently decontaminated by passage through the overlying pool of water into the building atmosphere.

Phase 2the protracted release due to re-evolution as elemental iodine (starts at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and ends at 30 days). CsI is conservatively assumed to completely dissociate into the pool water. Because of the low pH of the pool water, CsI (as well as Phase 1 absorbed elemental iodine within the pool) slowly re-evolves as elemental iodine into the building atmosphere.

B-1.4 The radioactive material available for release is assumed to be from the assemblies with the peak inventory. The fission product inventory for the peak assembly represents an upper limit value.

The inventory should be calculated assuming the maximum achievable operational power history and burnup. These parameters should be examined to maximize fission product inventory. This inventory calculation should include appropriate assembly peaking factors.

DG-1425, Appendix B, Page B-2 B-2.

Phase 1 ReleaseInitial Gaseous Release and Water Depth The elemental iodine decontamination factor (DF) is a function of bubble size and rise time through the water column, both of which are functions of fuel pin pressure. If the water depth is between 19 and 23 feet, the DF for elemental iodine can be computed based on a best-estimate rod pin pressure for the limiting fuel rods in the reactor core at the most limiting time in life. The time period between reactor shutdown and the movement of fuel may be used to compute radioactive decay and reduced decay power. The internal gas temperature, and thus the pin pressure, may be determined using the limiting pool water temperature near the fuel rods, and basing these values on a full-core offload.

For water depths between 19 and 23 feet, the elemental iodine DF based on pin pressure is computed using the following equations:

= 81.046.

(Equation B-1)

= 9.2261, (Equation B-2)

= 0.0002 + 1.0009, (Equation B-3) where t is the bubble rise time in seconds, computed as a function of pin pressure, x, in pounds per square inch gauge (psig), and d is the bubble diameter in centimeters, computed as a function of pin pressure, x (psig).

If the depth of water is not between 19 and 23 feet, the DF will need to be determined on a case-by-case basis. The DF for organic iodine is assumed to be 1.

B-3.

Phase 2 ReleaseRe-evolution Release The re-evolution calculation results in a simple exact transient solution. It has the flexibility to account for the effect of potential filtration and other removal mechanisms. The following site-specific and general parameters are needed:

a.

Vpool = total pool free volume,

b.

Spool = total pool surface area,

c.

Qrecirc = volumetric flow of recirculation system (to evaluate effects of filtration),

d.

NI-131gap = fuel pin radioactive iodine in gap (moles),

e.

NI-127gap = fuel pin nonradioactive iodine in gap (moles),

f.

KL = mass transfer coefficient, 3.66x10-6 m/s (Ref. B-1), and

g.

pH = bounding design acidity value of the pool.

Note: For this approach, Vpool, Spool, KL, and Qrecirc must use consistent units. (To calculate concentrations in moles per liter (M), Vpool must be converted to liters.)

Calculation Sequence:

1.

Calculate amount of iodine (radioactive and nonradioactive) in the fuel pin gap using tables 3 and 4 from the main body of this guide.

2.

Calculate volatile iodine fraction in pool.

DG-1425, Appendix B, Page B-3

3.

Calculate removal coefficients.

4.

Evaluate release as either (a) an overall release (neglecting time), or (b) a time-dependent release.

Step 1Calculate amount of iodine in the fuel pin gap using tables 3 and 4 in the main body of this guide.

Both the radioactive and the nonradioactive iodine (e.g., I-131 and I-127) in the pool affect the radioactive iodine evolution. The calculations operate on moles, so iodine isotope quantities must be converted to moles.

In the example equations that follow, I-131 is used to represent the radioactive iodine and I-total is used to represent the stable iodine; however, all radioactive and nonradioactive iodine should be considered. For a given mass of iodine, the number of moles of iodine can be calculated from the mass, m, in grams (g) and its atomic weight, M, as follows:

-=

-()

(Equation B-4)

-=

-()

(Equation B-5)

Alternatively, for radioactive materials, the number of moles can be calculated from the activity in becquerels (Bq):

-=

(Equation B-6)

Activities in curies (Ci) must be converted to becquerels (1 Ci = 3.7x1010 Bq).

The radioactive iodine concentration can be found using radiological decay formulas that account for time before fuel movement. If this is done, the activity of the other iodine isotopes at the time before fuel movement should be added to the I-131 activity.

Step 2Calculate volatile iodine fraction in pool.

The next step is to determine the fraction of iodine atoms in the pool that are in I2 (volatile) form:

Calculate the total concentrations in the pool:

Ct = total I concentration (M) (moles I atoms / L) = ( NI-total_gap + NI-131gap) / Vpool.

(Equation B-7)

Note: Vpool must be converted to liters for concentrations to be calculated in moles/liter.

Calculate the H+ concentration:

Ch = [H+] = 10-pH.

(Equation B-8)

DG-1425, Appendix B, Page B-4 Calculate the [I2]/[I-]2 concentration ratio, Ri (Ref. B-2):1 Ri = [I2]/[I-]2 = Ch2 / (6.05x10-14 + 1.47x10-9 Ch).

(Equation B-9)

Calculate the fraction of I atoms in I2 form:

o First evaluate Bm (negative B for the quadratic equation below):

Bm = 4 Ct + 1 / Ri.

(Equation B-10) o Then evaluate the volatile fraction, Xe (fraction of I atoms in I2 form):

Xe = (Bm - 16

) / (4 Ct).

(Equation B-11)

Step 3Calculate applicable removal coefficients.

The evolution removal coefficient, e, is calculated using the mass transfer coefficient, the pool surface-to-volume ratio, and the fraction of I that is in I2 form:

e = KL Xe Spool / Vpool (Equation B-12)

The removal rate is reduced to account for the fraction of iodine that is volatile and thus available to evolve to the gas space. This evolution rate applies to both nonradioactive and radioactive iodine.

Step 4Evaluate release as an overall release.

The removal coefficient is used to model the time-dependent concentration of radionuclides released from the pool as follows:

Qe = e Vpool (Equation B-13)

a.

If recirculation filtration is credited, f is used.

b.

Alternatively, one can model a loop and filter instead of using f.

B-4.

Noble Gases and Particulates The retention of noble gases in the water in the fuel pool or reactor cavity is negligible (i.e., the DF is 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., the DF is infinite).

1 The combined speciation rate is from NUREG/CR-5950, Iodine Evolution and pH Control, issued December 1992 (Ref. B-3).

DG-1425, Appendix B, Page B-5 B-5.

Fuel Handling Accidents within the Fuel Building For fuel handling accidents postulated to occur within the fuel building, the following assumptions are acceptable to the NRC staff:

B-5.1 The radioactive material that escapes from the fuel pool to the fuel building is assumed to be released to the environment over a 2-hour time period for the initial fuel gap gas release, which accounts for time-independent releases from the fuel pool. The release rate is generally assumed to be a linear or exponential function over this time period. Time-dependent releases from the fuel pool due to the re-evolution of iodine are to be considered releases directly from the pool to the environment outside the fuel building.

B-5.2 Engineered-safety-feature (ESF) filtration systems may reduce the amount of radioactive material released from the fuel pool. Such a reduction may be considered using the guidance in Regulatory Guide (RG) 1.52, Revision 4, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, issued September 2012 (Ref. B-4). The radioactivity release analyses should determine and account for delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system.2 B-5.3 The radioactivity release from the fuel pool should be assumed to be drawn into the ESF filtration system without mixing or dilution in the fuel building. If mixing can be demonstrated, credit for mixing and dilution may be considered on a case-by-case basis. This evaluation should consider the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede streamflow between the surface of the pool and the exhaust plenums.

B-6.

Fuel Handling Accidents within the Containment For fuel handling accidents postulated to occur within the containment, the following assumptions are acceptable to the NRC staff:

B-6.1 If the containment is isolated3 during fuel handling operations, no radiological consequences need to be analyzed.

B-6.2 If the containment is open during fuel handling operations but designed to automatically isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed for the isolated pathway.

2 These analyses should consider the time needed for the radioactivity concentration to reach levels corresponding to the monitor setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as applicable.

3 Containment isolation does not imply containment integrity as defined by technical specifications for non-shutdown modes.

The term isolation is used here collectively to encompass both containment integrity and containment closure, which is typically in place during shutdown periods. For isolation to be credited in the analysis, the technical specifications should address the appropriate form of isolation.

DG-1425, Appendix B, Page B-6 B-6.3 If the containment is open during fuel handling operations (e.g., a personnel air lock or equipment hatch is open),4 the radioactive material that escapes from the reactor cavity pool to the containment is assumed to be released to the environment over a 2-hour period for the initial fuel gap gas release, which accounts for time-independent releases from the reactor cavity. The release rate is generally assumed to be a linear or exponential function over this period.

Time-dependent releases from the reactor cavity pool due to the re-evolution of iodine are to be considered releases directly from the pool to the environment outside the containment.

B-6.4 A reduction in the amount of radioactive material released from the containment by ESF filtration systems may be considered using the guidance of RG 1.52 (Ref. B-4). The radioactivity release analyses should determine and account for delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system.5 B-6.5 Credit for dilution or mixing of the activity released from the reactor cavity by natural or forced convection inside the containment may be considered on a case-by-case basis. Such credit is generally limited to 50 percent of the containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede streamflow between the surface of the reactor cavity and the exhaust plenums.

4 Technical specifications that allow such operations usually include administrative controls to close the airlock, hatch, or open penetrations within 30 minutes. Such administrative controls generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur. Radiological analyses generally should not credit this manual isolation.

5 These analyses should consider the time needed for the radioactivity concentration to reach levels corresponding to the monitor setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as applicable.

DG-1425, Appendix B, Page B-7 APPENDIX B REFERENCES1 B-1.

U.S. Nuclear Regulatory Commission (NRC), Re-evaluation of the Fission Product Release and Transport for the Design-Basis Accident Fuel Handling Accident (ML19248C647), Enclosure 4 to internal memorandum from M.J. Case to M.X. Franovich, Closeout to Research Assistance Request for Independent Review of Regulatory and Technical Basis for Revising the Design-Basis Accident Fuel Handling Accident, Washington, DC, November 23, 2019 (ML19114A117 (package)).

B-2.

Dickson, E., NRC, internal memorandum to K. Hsueh, Example Calculation of Re-evaluated Fuel Handling Accident Fission Product Transport Model for Draft RG 1.183 Revision 1 (2021),

Washington, DC, August 30, 2021 (ML21190A040).

B-3.

NRC, NUREG/CR-5950, Iodine Evolution and pH Control, Washington, DC, December 1992 (ML063460464).

B-4.

NRC, Regulatory Guide 1.52, Revision 4, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, Washington, DC, September 2012 (ML12159A013).

1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

DG-1425, Appendix C, Page C-1 APPENDIX C ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BOILING-WATER REACTOR ROD DROP ACCIDENT This appendix provides assumptions acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of a rod drop accident at a boiling-water reactor. These assumptions supplement the guidance in the main body of this guide.

C-1.

Regulatory Position 3 of this guide provides assumptions acceptable to the NRC staff regarding core inventory. The fission product release from the breached fuel to the coolant is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. In addition to the combined fission product inventory (steady-state gap plus transient release), the release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting, and on the assumption that 100 percent of the noble gases and 50 percent of the iodines contained in that fraction are released to the reactor coolant.1 C-2.

If no or minimal fuel breach2 is postulated for the limiting event, the released activity should be the maximum coolant activity allowed by the TSs (typically a pre-accident spike of 4.0 microcuries per gram dose equivalent (DE) iodine (I)-131).

C-3.

The assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material from the fuel and the reactor coolant are as follows:

C-3.1 The activity released from the fuel from the gap and/or from fuel pellets is assumed to be instantaneously mixed in the reactor coolant within the pressure vessel.

C-3.2 Credit should not be assumed for partitioning in the pressure vessel or for removal by the steam separators.

C-3.3 Of the activity released from the reactor coolant within the pressure vessel, 100 percent of the noble gases, 10 percent of the iodine, and 1 percent of the remaining radionuclides are assumed to reach the turbine and condensers.

C-3.4 Of the activity that reaches the turbine and condenser, 100 percent of the noble gases, 10 percent of the iodine, and 1 percent of the particulate radionuclides are assumed available for release to the environment. The turbine and condensers leak to the environment as a ground-level release at a rate of 1 percent per day3 for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leakage is assumed to terminate. No credit should be assumed for dilution or hold-up within the turbine building.

Radioactive decay during hold-up in the turbine and condenser may be assumed.

1 Calculated values of the combined release (gap activity plus fuel melt) are limited to a total of 1.0.

2 Minimal fuel breach is defined for use in this appendix as an amount of damage that will yield reactor coolant system activity concentration levels less than the maximum technical specification limits. The activity assumed in the analysis should be based on the activity associated with the projected fuel breach or the maximum technical specification values, whichever maximizes the radiological consequences. In determining the DE I-131, only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

3 If there are forced flowpaths from the turbine or condenser, such as unisolated motor vacuum pumps or unprocessed air ejectors, the leak rate should be assumed to be the flow rate associated with the most limiting of these paths. Credit for collection and processing of releases, such as by offgas or standby gas treatment, will be considered on a case-by-case basis.

DG-1425, Appendix C, Page C-2 C-3.5 In lieu of the transport assumptions in Regulatory Positions C-3.2 through C-3.4 above, a more mechanistic analysis may be used on a case-by-case basis. Such analyses account for the quantity of contaminated steam carried from the pressure vessel to the turbine and condensers, based on a review of the minimum transport time from the pressure vessel to the first main steam isolation valve and the closure time for this valve.

C-3.6 The iodine species released from the reactor coolant within the pressure vessel should be assumed to be 95 percent cesium iodide as an aerosol, 4.85 percent elemental iodine, and 0.15 percent organic iodide. The release from the turbine and condenser should be assumed to be 97 percent elemental and 3 percent organic.

DG-1425, Appendix D, Page D-1 APPENDIX D ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BOILING-WATER REACTOR MAIN STEAMLINE BREAK ACCIDENT This appendix provides assumptions acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of a main steamline break accident at a boiling-water reactor. These assumptions supplement the guidance in the main body of this guide.

Source Term D-1.

Regulatory Position 3 of this guide provides assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel.

D-2.

If no or minimal fuel breach1 is postulated for the limiting event, the released activity should be the maximum coolant activity allowed by technical specifications (TS). The iodine concentration in the primary coolant is assumed to correspond to the following two cases in the standard TS for the nuclear steam supply system vendor:

D-2.1 The concentration that is the maximum value permitted (typically 4.0 microcuries per gram (Ci/g) dose equivalent (DE) iodine (I)-131) and corresponds to the conditions of an assumed pre-accident spike.

D-2.2 The concentration that is the maximum equilibrium value permitted for continued full-power operation (typically 0.2 Ci/g DE I-131).

D-3.

The activity released from the fuel should be assumed to mix instantaneously and homogeneously in the reactor coolant. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. Noble gases should be assumed to enter the steam phase instantaneously.

Transport D-4.

Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows:

D-4.1 The main steamline isolation valves should be assumed to close in the maximum time allowed by TS.

D-4.2 The total mass of coolant released should be assumed to be the amount in the steamline and connecting lines at the time of the break, plus the amount that passes through the valves before closure.

1 Minimal fuel breach is defined for use in this appendix as an amount of damage that will yield reactor coolant system activity concentration levels less than the maximum technical specification limits. The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining DE I-131, only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

DG-1425, Appendix D, Page D-2 D-4.3 All radioactivity in the released coolant should be assumed to be released to the environment instantaneously as a ground-level release. No credit should be assumed for plateout, hold-up, or dilution within facility buildings.

D-4.4 The iodine species released from the main steamline should be assumed to be 95 percent cesium iodide as an aerosol, 4.85 percent elemental iodine, and 0.15 percent organic iodide.

DG-1425, Appendix E, Page E-3 APPENDIX E ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED-WATER REACTOR STEAM GENERATOR TUBE RUPTURE ACCIDENT This appendix provides assumptions acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of a steam generator tube rupture (SGTR) accident at a pressurized-water reactor. These assumptions supplement the guidance in the main body of this guide.

Source Term E-1.

Regulatory Position 3 of this guide provides assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel.

E-2.

If no or minimal fuel breach1 is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by technical specifications (TS). Two cases of iodine spiking should be assumed:

E-2.1 A reactor transient has occurred before the postulated SGTR and has raised the primary coolant iodine concentration to the maximum value permitted at full-power operations by the TS (typically 60 microcuries per gram (Ci/g) dose equivalent (DE) iodine (I)-131). This is the pre-accident iodine spike case.

E-2.2 The primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value specified in the TS (typically 1.0 Ci/g DE I-131). This is the concurrent iodine spike case. A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel pins assumed to have defects.

E-3.

The activity released from the fuel, if any, should be assumed to be released instantaneously and homogeneously through the primary coolant. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.

E-4.

The specific activity in the steam generator liquid at the onset of the SGTR is at the maximum value permitted by secondary activity TS (typically 0.1 Ci/g).

E-5.

Iodine releases from the steam generators to the environment should be assumed to be 97 percent elemental iodine and 3 percent organic iodide.

1 Minimal fuel breach is defined for use in this appendix as an amount of damage that will yield reactor coolant system activity concentration levels less than the maximum technical specification limits. The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining DE I-131, only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

DG-1425, Appendix E, Page E-4 Transport E-6.

Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows:

E-6.1 The primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate limiting condition for operation specified in the TS. The primary-to-secondary leak rate at later stages of the transient may be reduced if justified by plant-specific design and engineering analyses. The leakage should be apportioned between affected and unaffected steam generators in a manner that maximizes the calculated dose.

E-6.2 The density used in converting volumetric leak rates (e.g., in gallons per minute) to mass leak rates (e.g., in pounds mass per hour) should be consistent with the basis of surveillance tests used to show compliance with leak rates in the TS. These tests are typically based on cool liquids.

Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gram per cubic centimeter (62.4 pounds mass per cubic foot).

E-6.3 The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature in the bulk of the primary system is less than 100 degrees Celsius (212 degrees Fahrenheit). The release of radioactivity from the unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated. The release of radioactivity from the affected steam generator should be assumed to continue until shutdown cooling is in operation and releases from the steam generator have been terminated, or the steam generator is isolated from the environment so that no release is possible, whichever occurs first.

E-6.4 All noble gas radionuclides released from the primary system should be assumed to be released to the environment without reduction or mitigation.

E-6.5 The transport model described in this section should be used for iodine and particulate releases from the steam generators. Figure E-1 illustrates this model, which is summarized as follows:

DG-1425, Appendix E, Page E-5 Figure E-1. Transport model E-6.5.1 A portion of the primary-to-secondary leakage will flash to vapor, based on the thermodynamic conditions in the reactor and secondary coolant (bulk water in figure E-1).

For the unaffected steam generators used for plant cooldown, the primary-to-secondary leakage, discussed in Regulatory Position E-6.1, can be assumed to mix with the secondary water without flashing during periods of total tube submergence.

E-6.5.2 The leakage in the affected steam generator that immediately flashes to vapor will rise through the secondary water of the steam generator and enter the steam space. Credit may be taken for scrubbing in the generator, using the models in NUREG-0409, Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident, issued January 1978 (Ref. E-1), during periods of total submergence of the tubes.

E-6.5.3 The leakage in the affected steam generator that does not immediately flash is assumed to mix with the secondary water.

E-6.5.4 The radioactivity in the secondary water is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient.2 A partition coefficient of 100 may be assumed for iodine. The retention of noniodine particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators.

E-6.6 During periods of steam generator dryout, all primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation.

Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. E-2). If the tubes are uncovered, a portion of the primary-to-secondary leakage will flash and atomize, based on the thermodynamic conditions in the reactor and secondary coolant, and will be released to the environment with no mitigation. The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction) needs to be considered. The impact of restoration strategies described in emergency operating procedures for steam generator water levels should be evaluated.

2 In this appendix, the partition coefficient is defined as follows:

PC mass of I per unit mass of liquid mass of I per unit mass of gas

=

2 2

DG-1425, Appendix E, Page E-6 APPENDIX E REFERENCES1 E-1.

U.S. Nuclear Regulatory Commission (NRC), NUREG-0409, Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident, Washington, DC, January 1978 (ML19269F014).

E-2.

NRC, Information Notice 88-31, Steam Generator Tube Rupture Analysis Deficiency, Washington, DC, May 25, 1988 (ML031150151).

1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

DG-1425, Appendix F, Page F-1 APPENDIX F ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED-WATER REACTOR MAIN STEAMLINE BREAK ACCIDENT This appendix provides assumptions acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of a main steamline break (MSLB) accident at a pressurized-water reactor. These assumptions supplement the guidance in the main body of this guide.

Source Term F-1.

Regulatory Position 3 of this regulatory guide provides assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel.

F-2.

If no or minimal fuel breach1 is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by the technical specifications (TS). Two cases of iodine spiking should be assumed:

F-2.1 A reactor transient has occurred before the postulated MSLB and has raised the primary coolant iodine concentration to the maximum value permitted by the TS (typically 60 microcuries per gram (Ci/g) dose equivalent (DE) iodine (I)-131). This is the pre-accident iodine spike case.

F-2.2 The primary system transient associated with the MSLB causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value specified in the TS (typically 1.0 Ci/g DE I-131). This is the concurrent iodine spike case. A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel gap assumed to have defects.

F-3.

The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. The fuel damage estimate should assume that the highest-worth control rod is stuck at its fully withdrawn position.

F-4.

The specific activity in the steam generator liquid at the onset of the MSLB should be assumed to be at the maximum value permitted by secondary activity TS (typically 0.1 Ci/g DE I-131).

1 Minimal fuel breach is defined for use in this appendix as an amount of damage that will yield reactor coolant system activity concentration levels less than the maximum technical specification limits. The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining DE I-131, only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

DG-1425, Appendix F, Page F-2 F-5.

Iodine releases from the steam generators to the environment should be assumed to be 97 percent elemental iodine and 3 percent organic iodide. These fractions apply both to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

Transport F-6.

Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows:

F-6.1 The secondary water in the faulted2 steam generator is assumed to rapidly blow down to the environment. The duration of the blowdown is obtained from thermal-hydraulic analysis codes.

The activity in the faulted steam generator secondary water is assumed to be released to the environment without mitigation.

F-6.2 For facilities that have not implemented alternative repair criteria, the primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate limiting condition for operation specified in the TS. For facilities with traditional steam generator specifications (both per generator and for the total of all generators), the leakage should be apportioned between faulted and unaffected steam generators in a manner that maximizes the calculated dose. For example, for a four-loop facility with a limiting condition for operation of 1.9x103 liters per day (500 gallons per day) for any one generator, not to exceed 3.8 liters per minute (1 gallon per minute) from all generators, it would be appropriate to assign 1.9x103 liters per day (500 gallons per day) to the faulted generator and 1.2x103 liters per day (313 gallons per day) to each of the unaffected generators.

For facilities that have implemented alternative repair criteria, the primary-to-secondary leak rate in the faulted steam generator should be assumed to be the maximum accident-induced leakage derived from the repair criteria and burst correlations. For the unaffected steam generators, the leak rate limiting condition for operation specified in the TS is equally apportioned between the unaffected steam generators.

F-6.3 The density used in converting volumetric leak rates (e.g., in gallons per minute) to mass leak rates (e.g., in pounds mass per hour) should be consistent with the basis of the parameter being converted. The leak rate correlations for alternative repair criteria are generally based on the collection of cooled liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate TS are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gram per cubic centimeter (62.4 pounds mass per cubic foot).

F-6.4 The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature in the bulk of the primary system is less than 100 degrees Celsius (212 degrees Fahrenheit). The primary-to-secondary leak rate at later stages of the transient may be reduced if justified by plant-specific design and engineering analyses. The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

2 In this appendix, faulted refers to the state of the steam generator in which the secondary side has been depressurized by an MSLB, in such a way that protective system response (main steamline isolation, reactor trip, safety injection, etc.) has occurred.

DG-1425, Appendix F, Page F-3 F-6.5 All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.

F-6.6 The transport model described in this section should be used for iodine and particulate releases from the steam generators.

F-6.6.1 For the unaffected steam generators used for plant cooldown, the primary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergence.

F-6.6.2 The radioactivity in the secondary water of the unaffected generators is assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. A partition coefficient of 100 may be assumed for iodine. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators.

F-6.6.3 The primary-to-secondary leakage to the faulted steam generator is assumed to flash to vapor and be released to the environment with no mitigation.

Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. F-1). If the tubes are uncovered, a portion of the primary-to-secondary leakage will flash and atomize, based on the thermodynamic conditions in the reactor and secondary coolant, and will be released to the environment with no mitigation. The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction) needs to be considered. The impact of restoration strategies described in emergency operating procedures for steam generator water levels should be evaluated.

DG-1425, Appendix F, Page F-4 APPENDIX F REFERENCES1 F-1 U.S. Nuclear Regulatory Commission, Information Notice 88-31, Steam Generator Tube Rupture Analysis Deficiency, Washington, DC, May 25, 1988 (ML031150151).

1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

DG-1425, Appendix G, Page G-1 APPENDIX G ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED-WATER REACTOR LOCKED ROTOR ACCIDENT This appendix provides assumptions acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of a locked rotor accident at a pressurized-water reactor. These assumptions supplement the guidance in the main body of this guide.

Source Term G-1.

Regulatory Position 3 of this regulatory guide provides assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel.

G-2.

If no fuel damage is postulated for the limiting event, a radiological analysis is not required, as the consequences of this event are bounded by the consequences projected for the main steamline break outside containment.

G-3.

The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.

G-4.

The chemical form of radioiodine released from the fuel should be assumed to be 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97 percent elemental iodine and 3 percent organic iodide. These fractions apply both to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.

Transport G-5.

Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows:

G-5.1 The primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate limiting condition for operation specified in the technical specifications (TS). The primary-to-secondary leak rate at later stages of the transient may be reduced if justified by plant-specific design and engineering analyses. The leakage should be apportioned between the steam generators in a manner that maximizes the calculated dose.

G-5.2 The density used in converting volumetric leak rates (e.g., in gallons per minute) to mass leak rates (e.g., in pounds mass per hour) should be consistent with the basis of surveillance tests used to show compliance with leak rate TS. These tests are typically based on cool liquids. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gram per cubic centimeter (62.4 pounds mass per cubic foot).

G-5.3 The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature in the bulk of the primary system of the leakage is less than 100 degrees Celsius (212 degrees Fahrenheit). The

DG-1425, Appendix G, Page G-2 release of radioactivity should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

G-5.4 All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.

G-5.5 The transport model described in Regulatory Positions E-6.5 and E-6.6 of appendix E to this guide should be used for iodine and particulates.

DG-1425, Appendix H, Page H-1 APPENDIX H ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED-WATER REACTOR CONTROL ROD EJECTION ACCIDENT This appendix provides assumptions acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of a control rod ejection accident at a pressurized-water reactor. These assumptions supplement the guidance in the main body of this guide.

Two release paths are considered: (1) release via containment leakage and (2) release via the secondary plant. Each release path is evaluated independently, as if it were the only pathway available. The consequences of this event are acceptable if the dose from each path considered separately is less than the acceptance criterion in table 7 in this guide.

Source Term H-1.

Regulatory Position 3 of this guide provides assumptions acceptable to the NRC staff regarding core inventory. The fission product release from the breached fuel to the coolant is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. In addition to the combined fission product inventory (steady-state gap plus transient release), the release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting, and on the assumption that 100 percent of the noble gases and 50 percent of the iodines contained in that fraction are released to the reactor coolant.1 H-2.

If no fuel breach is postulated for the limiting event, a radiological analysis is not required, as the consequences of this event are bounded by the consequences projected for the maximum hypothetical loss-of-coolant accident, main steamline break, and steam generator tube rupture.

H-3.

In the case of the first release path, 100 percent of the activity released from the fuel should be assumed to be released instantaneously and homogeneously through the containment atmosphere.

In the case of the second release path, 100 percent of the activity released from the fuel should be assumed to be completely dissolved in the primary coolant and available for release to the secondary system.

H-4.

The chemical form of radioiodine released to the containment atmosphere should be assumed to be 95 percent cesium iodide, 4.85 percent elemental iodine, and 0.15 percent organic iodide. If containment sprays do not actuate or are terminated before sump water is accumulated, or if the containment sump pH is not controlled at values of 7 or greater, the iodine species should be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids created during the control rod ejection accident event (e.g., pyrolysis and radiolysis products). With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.

H-5.

Iodine releases from the steam generators to the environment should be assumed to be 97 percent elemental iodine and 3 percent organic iodide.

1 Calculated values of the combined release (gap activity plus fuel melt) are limited to a total of 1.0.

DG-1425, Appendix H, Page H-2 Transport from Containment H-6.

Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material in and from the containment are as follows:

H-6.1 A reduction in the amount of radioactive material available for leakage from the containment that is due to natural deposition, containment sprays, recirculating filter systems, dual containments, or other engineered safety features may be considered. Refer to appendix A to this guide for guidance on acceptable methods and assumptions for evaluating these mechanisms.

H-6.2 The containment should be assumed to leak at the leak rate incorporated in the technical specifications (TS) at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at 50 percent of this leak rate for the remaining duration of the accident. Peak accident pressure is the maximum pressure defined in the TS for containment leak testing. Leakage from sub-atmospheric containments is assumed to be terminated when the containment is brought to a sub-atmospheric condition, as defined in TS.

Transport from Secondary System H-7.

Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material in and from the secondary system are as follows:

H-7.1 A leak rate equivalent to the primary-to-secondary leak rate limiting condition for operation specified in the TS should be assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated. The primary-to-secondary leak rate at later stages of the transient may be reduced if justified by plant-specific design and engineering analyses.

H-7.2 The density used in converting volumetric leak rates (e.g., in gallons per minute) to mass leak rates (e.g., in pounds mass per hour) should be consistent with the basis of surveillance tests used to show compliance with leak rate TS. These tests are typically based on cooled liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gram per cubic centimeter (62.4 pounds mass per cubic foot).

H-7.3 All noble gas radionuclides released to the secondary system are assumed to be released to the environment without reduction or mitigation.

H-7.4 The transport model described in Regulatory Positions E-6.5 and E-6.6 of appendix E to this guide should be used for iodine and particulates.

DG-1425, Appendix I, Page I-1 APPENDIX I ANALYTICAL TECHNIQUE FOR CALCULATING FUEL-DESIGN OR PLANT-SPECIFIC STEADY-STATE FISSION PRODUCT RELEASE FRACTIONS FOR NON-LOSS-OF-COOLANT ACCIDENT EVENTS This appendix provides an acceptable analytical technique for calculating steady-state fission product release fractions residing in the fuel rod void volume (plenum and pellet-to-cladding gap), based on either specific fuel rod designs or more realistic fuel rod power histories. This analytical procedure was used, along with bounding fuel rod power histories, to calculate the release fractions listed in tables 3 and 4 of Regulatory Position 3.2 in the main body of this guide. Lower release fractions are achievable using less aggressive rod power histories or less limiting fuel rod designs (e.g., 17x17 versus 14x14 fuel rod designs). The analytical technique outlined in this section is one acceptable means of calculating maximum steady-state release fractions.

Steady-state gap inventories represent radioactive fission products generated during normal steady-state operation that have diffused within the fuel pellet, have been released into the fuel rod void space (i.e., rod plenum and pellet-to-cladding gap), and are available for release upon fuel rod cladding failure. Given the continued accumulation of long-lived radioactive isotopes and the inevitable decay of short-lived radioactive isotopes, the most limiting time in life (i.e., maximum gap fraction) for a particular radioactive isotope varies with fuel rod exposure and power history. The analytical technique described in this appendix specifies the use of fuel rod power profiles based on core operating limits or limiting fuel rod power histories. In addition, this analytical technique produces a composite worst time-in-life (i.e., maximum gap fraction for each radioactive isotope). Therefore, the steady-state fission product gap inventories calculated using this analytical approach will be significantly larger than realistic fuel rod or core-average source terms. One acceptable means of capturing more realism in the calculation of the steady-state release fractions would be to calculate burnup-dependent release fractions for each radionuclide. The resulting burnup-dependent release fractions could then be used to calculate radiological consequences at different times in life.

The U.S. Nuclear Regulatory Commission (NRC) codeveloped the Fuel Analysis under Steady-State and Transients (FAST) (formerly FRAPCON and FRAPTRAN) fuel rod thermal-mechanical fuel performance code to perform independent audit calculations for licensing activities. While calibrated and validated against a large empirical database, FAST and its predecessors are not NRC-preapproved codes and may not be used to calculate plant-specific, fuel-specific, or cycle-specific gap inventories that are in accordance with the acceptable analytical procedure below without further justification.

The analytical technique used to calculate steady-state gap inventories should have the following attributes:

I-1.

For stable, long-lived radioactive isotopes, such as krypton (Kr)-85, an NRC-approved fuel rod thermal-mechanical performance code with established modeling uncertainties should be used to predict the integral fission gas release (FGR). The code should include the effects of thermal conductivity degradation with burnup and should have been verified against measured fuel temperatures and stable FGR data up to the licensed burnup of the particular fuel rod design.

I-1.1 Long-lived radioactive isotopes will continue to accumulate throughout exposure, with insignificant decay because of their long half-lives. For this reason, maximum gap inventories for long-lived isotopes are likely to occur near or at the end of life of the fuel assembly.

DG-1425, Appendix I, Page I-2 I-1.2 Cesium is expected to behave differently from noble gases once it reaches the grain boundaries.

At this point, it may react with other constituents in the fuel to form less volatile compounds that may then accumulate on the grain boundaries as solids or liquids. Cesium released from the fuel may also react with the zirconium in the cladding to form more stable (i.e., nongaseous) compounds. These effects tend to decrease the inventory of gaseous cesium available for release in the event of a cladding breach. To account for these effects, the following relationship is recommended:

(Gap Inventory)Cs-134, Cs-137 = (Release Fraction)Kr-85 * (0.5),

where (Gap Inventory)Cs-134, Cs-137 is the amount of gaseous cesium available for release, and (Release Fraction)Kr-85 is calculated using an approved fuel performance code.

I-2.

For volatile, short-lived radioactive isotopes, such as iodine (I) (i.e., I-131, I-132, I-133, and I-135) and xenon (Xe) and Kr noble gases (except for Kr-85) (i.e., Xe-133, Xe-135, Kr-85m, Kr-87, and Kr-88), the release-to-birth (R/B) fraction should be predicted using either an NRC-approved release model or the NRC-endorsed release model from the American National Standards Institute (ANSI)/American Nuclear Society (ANS) standard ANSI/ANS-5.4, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel, issued May 2011 (Ref. I-1). The prediction should use fuel parameters at several depletion time steps from an NRC-approved fuel rod thermal-mechanical performance code. The fuel parameters necessary for use in the NRC-endorsed ANSI/ANS-5.4 model calculations of the R/B fraction are local fuel temperature, fission rate, and axial node/pellet burnup. Consistent with Regulatory Position I-1, the code should include the effects of thermal conductivity degradation with burnup and should have been verified against measured fuel temperatures and stable FGR data up to the licensed burnup of the particular fuel rod design.

Because of their relatively short half-lives, the amount of activity associated with volatile radioactive isotopes depends on their rate of production (i.e., fission rate and cumulative yield),

rate of release, and rate of decay. Maximum R/B ratios for short-lived isotopes are likely to occur at approximately the maximum exposure at the highest power level (i.e., the inflection point in the power operating envelope).

I-2.1 NUREG/CR-7003, Background and Derivation of ANS-5.4 Standard Fission Product Release Model, issued January 2010 (Ref. I-2), provides guidance on using the NRC-endorsed ANSI/ANS-5.4 release model to calculate short-lived R/B fractions.

I-2.1.1 For nuclides with half-lives less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, no gap inventories are provided. Because of their rapid decay (relative to the time for diffusion and transport), the gap fractions for these nuclides will be bounded by the calculated gap fractions for longer lived nuclides under the headings Other Noble Gases and Other Halogens.

DG-1425, Appendix I, Page I-3 I-2.1.2 For nuclides with half-lives less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, an approved fuel performance code is applied to predict the R/B fraction using equation 12 in NUREG/CR-7003 and its definitions of terms, as follows:

Where:

R is the release rate (atoms per cubic centimeter per second (atoms/(cm3-s))),

B is the production rate (atoms/(cm3-s)),

S is surface area (cm2),

V is volume (cm3),

accounts for precursor enhancement effects and is defined below in Table I-1, D is the diffusion coefficient (cm2/s), and is the half-life (s-1).

I-2.1.3 For nuclides with half-lives greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the R/B fraction is predicted by multiplying the fractal scaling factor (Fnuclide) by the predicted Kr-85m R/B fraction using equation 13 of NUREG/CR-7003, as follows:

The R/B fraction for the isotope I-132 should be calculated using this equation even though its half-life is less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (2.28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />), because its precursor of tellurium (Te)-132 has a half-life of 3.2 days, which controls the release of I-132.

I-2.1.4 Table I-1 lists the fractal scaling factors for each nuclide, calculated using the following equation from NUREG/CR-7003:

nuclide m

i nuclide m

i m

i D

V S

B R

=

m Kr i

m Kr i

nuclide nuclide i

D V

S F

B R

85 85

=

25

.0 85 85

=

m Kr nuclide m

Kr nuclide nuclide F

DG-1425, Appendix I, Page I-4 Table I-1. Fractal Scaling Factors for Short-Lived Nuclides NUCLIDE NUREG/CR-7003, TABLE 1 FRACTAL SCALING FACTOR Half-Life Decay Constants (1/sec)

Alpha Xe-133 5.243 days 1.53x10-6 1.25 2.276 Xe-135 9.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 2.12x10-5 1.85 1.301 Xe-135m 15.3 minutes 7.55x10-4 23.50 1.005 Xe-137 3.82 minutes 3.02x10-3 1.07 0.328 Xe-138 14.1 minutes 8.19x10-4 1.00 0.447 Xe-139 39.7 seconds 1.75x10-2 1.00 0.208 Kr-85m 4.48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 4.30x10-5 1.31 1.000 Kr-87 1.27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> 1.52x10-4 1.25 0.721 Kr-88 2.84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> 6.78x10-5 1.03 0.840 Kr-89 3.15 minutes 3.35x10-3 1.21 0.330 Kr-90 32.3 seconds 2.15x10-2 1.11 0.203 I-131 8.04 days 9.98x10-7 1.00 2.395 I-132 2.28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> 8.44x10-5 137*

2.702 I-133 20.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 9.26x10-6 1.21 1.439 I-134 52.6 minutes 2.20x10-4 4.40 0.900 I-135 6.57 hours6.597222e-4 days <br />0.0158 hours <br />9.424603e-5 weeks <br />2.16885e-5 months <br /> 2.93 x10-5 1.00 1.029

• The I-132 alpha term accounts for the significant contribution from precursor Te-132.

I-3.

Release fractions should be calculated using an NRC-approved fuel rod thermal-mechanical code, which provides a methodology for calculating high-confidence stable fission gas release and fuel temperatures.

I-3.1 For short-lived isotopes, the 2011 release model standard ANSI/ANS-5.4 recommends multiplying the best-estimate predictions by a factor of 5.0 to obtain upper tolerance release

DG-1425, Appendix I, Page I-5 fractions. When using an NRC-approved fuel rod thermal-mechanical code, which may produce different values for R/B of Kr-85m, an alternative factor may be derived following the same procedure, as described in sections 4.4.1 and 4.4.2 of NUREG/CR-7003, with nominal predicted fuel rod temperatures. For long-lived isotopes, established model uncertainties associated with the NRC-approved fuel rod thermal-mechanical code should be applied, either deterministically or sampled within a statistical application methodology, to obtain high-confidence upper tolerance release fractions.

I-4.

Nominal fuel design specifications (excluding tolerances) may be used.

I-5.

Actual in-reactor fuel rod power histories may diverge from reload core depletion calculations because of unplanned shutdowns or power maneuvering. Therefore, the rod power history or histories used to predict gap inventories should bound anticipated operation. Rod power histories from the fuel rod design analysis, based on thermal-mechanical operating limits from the core operating limits report or on radial falloff curves, should be used. The fuel rod power history used to calculate gap inventories should be verifiable.

I-5.1 The calculation supporting the bounding gap inventories in tables 3 and 4 in the main body of this guide used a segmented power history for both the boiling-water reactor and pressurized-water reactor limiting designs. Seven power histories were considered, each running at 90 percent of the bounding rod-average power, except that they ran at the linear heat generation rate limit for approximately 9 to 10 gigawatt-days per metric ton of uranium burnup (rod-average) at seven burnup intervals. Given that no single fuel rod will dominate the bounding power envelope, a segmented power history approach is an acceptable alternative to assigning fuel rod power at the maximum, burnup-dependent power level over the fuel rod lifetime.

I-6.

Higher local power density (Fq) promotes more local FGR. Higher rod-average power (Fr), along with a flatter axial power distribution (Fz), promotes more FGR along the fuel stack. Sensitivity cases should be evaluated to ensure that the limiting fuel rod power history is captured.

I-7.

Each fuel rod design (e.g., UO2, UO2Gd2O3, part-length, full-length) should be evaluated.

I-8.

The minimum acceptable number of radial and axial nodes as defined in ANSI/ANS-5.4 should be used, along with the methodology of summing the release for these nodes, to determine the overall release from the fuel pellets to the fuel void volume.

DG-1425, Appendix I, Page I-6 APPENDIX I REFERENCES1 I-1 American National Standards Institute (ANSI)/American Nuclear Society (ANS),

ANSI/ANS-5.4, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel, La Grange Park, Illinois, May 2011 (not publicly available in ADAMS).

I-2 U.S. Nuclear Regulatory Commission, NUREG/CR-7003, Background and Derivation of ANS-5.4 Standard Fission Product Release Model, Washington, DC, January 2010 (ML100130186).

1 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public website at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. For problems with ADAMS, contact the Public Document Room staff at 301-415-4737 or (800) 397-4209, or email pdr.resource@nrc.gov. The NRC Public Document Room (PDR), where you may also examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.

DG-1425, Appendix J, Page J-1 APPENDIX J ABBREVIATIONS ADAMS Agencywide Documents Access and Management System AEB Accident Evaluation Branch AMMD aerodynamic mass median diameter ANS American Nuclear Society ANSI American National Standards Institute ASME American Society of Mechanical Engineers AST alternative source term ATF accident tolerant fuel Bq becquerel(s)

BWR boiling-water reactor BWROG Boiling Water Reactor Owners Group CDF core damage frequency CEDE committed effective dose equivalent CFR Code of Federal Regulations Ci curie(s) cm centimeter(s)

CsI cesium iodide DBA design-basis accident DE dose equivalent DF decontamination factor EAB exclusion area boundary ECCS emergency core cooling system EDE effective dose equivalent EDEX effective dose equivalent from external sources EPA U.S. Environmental Protection Agency EPZ emergency planning zone EQ environmental qualification ESF engineered safety feature F

Fahrenheit FAST Fuel Analysis under Steady-State and Transients FeCrAl iron-chromium-aluminum

DG-1425, Appendix J, Page J-2 FF flash fraction FFRD fuel fragmentation relocation and dispersal FGR fission gas release FOIA Freedom of Information Act FSAR final safety analysis report g

gram(s) g/cm3 gram(s) per cubic centimeter GDC general design criterion/criteria GWd/MTU gigawatt-day(s) per metric ton uranium I

iodine IAEA International Atomic Energy Agency ICRP International Commission on Radiological Protection Kr krypton LOCA loss-of-coolant accident LPZ low-population zone LWR light-water reactor Ci/g microcurie(s) per gram m

micron(s) m3/s cubic meter(s) per second MHA maximum hypothetical accident MSIV main steam isolation valve MSLB main steamline break MWt megawatt(s) thermal NRC U.S. Nuclear Regulatory Commission OMB Office of Management and Budget PRA probabilistic risk assessment psig pounds per square inch gauge PWR pressurized-water reactor R/B release-to-birth RADTRAD RADionuclide Transport, Removal, and Dose Estimation RCS reactor coolant system RG regulatory guide s

second(s) scfh standard cubic foot/feet per hour

DG-1425, Appendix J, Page J-3 SGTR steam generator tube rupture SRM staff requirements memorandum SRP Standard Review Plan (NUREG-0800)

SSC structure, system, or component SSG Specific Safety Guide Sv sievert(s)

Te tellurium TEARE total effective aerosol removal efficiency TEDE total effective dose equivalent TFGR transient fission gas release TID technical information document TMI Three Mile Island TS technical specification(s)

TSC technical support center UO2 uranium dioxide Xe xenon Zr zirconium

DG-1425, Appendix K, Page K-1 APPENDIX K KNOWLEDGE MANAGEMENT INFORMATION During the development of Revision 1 and Revision 2 of Regulatory Guide (RG) 1.183, the staff interacted with several stakeholders who expressed a desire for background information on this RG. As a result, the staff involved in the preparation of Revision 2 developed this appendix to explain the basis for several of the more significant changes to the guide. Licensees should not use the information in this appendix as guidance for licensing actions. Instead, the explanations should be used for informational purposes only and should be read within the context of the reference that appears in the body of this guide. Each item below corresponds to a change made in the body of this guide.

KM-01 (p. 6)

In response to industry interest in light-water reactor (LWR) fuels enriched to between 5.0 to 10.0 weight percent uranium (U)-235, the staff submitted a rulemaking plan in SECY-21-0109, Rulemaking Plan on Use of Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, dated December 20, 2021 (Agencywide Documents Access and Management System Accession No. ML21232A237). The plan asked for Commission approval to begin rulemaking to amend U.S. Nuclear Regulatory Commission (NRC) requirements to facilitate the use of LWR fuel containing uranium enriched to greater than 5.0 weight percent U-235. In SECY-21-0109, the staff recommended rulemaking, in part, to reduce the number of exemption requests associated with implementing accident tolerant fuels, thus ensuring regulatory efficiency and stability. Rulemaking on this topic would allow the staff to thoroughly review the potential regulatory implications of fuels enriched to greater than 5.0 weight percent U-235 and identify and assess the potential costs and benefits of changing regulatory requirements that impact their use.

Rulemaking would also provide options for a generic resolution of these issues and invite stakeholder participation in decisions affecting this regulatory area, rather than deciding issues on a case-by-case basis as in the current regulatory framework.

The Commission approved the staffs plan to initiate rulemaking to amend requirements for the use of LWR fuel containing uranium enriched to greater than 5.0 weight percent U-235 in Staff Requirements Memorandum (SRM)-SECY-21-0109, Staff RequirementsSECY-21-0109Rulemaking Plan on Use of Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, dated March 16, 2022 (ML22075A103). The Commission stated that the provisions of the rule should apply only to high-assay, low-enriched uranium (HALEU) fuel, both for nonproliferation and safeguards reasons, and that the staffs analysis should focus on the range of enrichment most likely to be contemplated in future applications.

In addition, the Commission directed the following:

Fuel fragmentation, relocation, and dispersal (FFRD) issues relevant to fuels of higher enrichment and burnup levels should be appropriately addressed and analyzed in the regulatory basis for this rulemaking.

DG-1425, Appendix K, Page K-2 The staff should take a risk-informed approach when developing this rule and the associated regulatory basis and guidance.

The staff should work expeditiously with stakeholders to identify and develop necessary regulatory guidance and technical bases to support effective and efficient licensing of increased enrichment applications.

KM-02 (p. 12)

Regulatory Position 1.1.1 has been updated to provide additional guidance on the evaluation of defense in depth and safety margin. Important elements from the Standard Review Plan and RG 1.174 have been incorporated, as appropriate.

KM-03 (p. 15)

The updated Regulatory Position 1.1.4 expands the discussion of regulatory source terms based on recommendations from the Advisory Committee on Reactor Safeguards (ML23256A179) and resolutions to Differing Professional Opinions (DPO) 2020-002 and DPO 2021-001 (ML21067A645 and ML23263A639, respectively). Updated information includes the agencys use of source terms and their various applications within the agencys regulatory framework.

KM-04 (pp. 17, 26)

The technical basis for the 10-mile plume exposure emergency planning zone was based on evaluation of the offsite consequences of the design-basis accident (DBA), less severe beyond-DBAs, and less probable but more severe beyond-DBAs. The estimated radiological consequences of these accidents were then compared to the U.S. Environmental Protection Agency guidance in EPA-520/1-75-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, issued September 1975. The DBAs and less severe accidents are those in which the containment is expected to be intact. More severe accidents are those in which the containment is expected to either fail or be bypassed. Each of these three accident classes are assessed against three dose-based figures-of-merit criteria.

KM-05 (p. 26)

In support of the Increased Enrichment Rulemaking, the staff focused on three primary areas that could prepare the agency for licensing increased enrichment:

(1) amending the control room design criteria, (2) updating regulatory source terms, and (3) updating guidance transport models and other assumptions.

With respect to area 3, the analyses in SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, issued April 2023 (ML23097A087),

were assessed. These assessments focused on the best-estimate transport and spatial distribution of radionuclide inventories within various locations of the facility that could be applied to inform the development of mechanistic transport models.

Results of these assessments were the development of mechanistic transport models for each of the three boiling-water reactor release pathways. Application of these models allows for the direct credit of suppression pool scrubbing without the need for additional analysis or calculations. (Revisions 0 and 1 of RG 1.183 do not directly allow for credit of suppression pool scrubbing.)

The models are consistent with Commission policy to update regulatory guidance as stated in SRM-SECY-99-144.

DG-1425, Appendix K, Page K-3 KM-06 (p. 33)

In support of the Increased Enrichment Rulemaking, the Commission approved the staffs plan to initiate rulemaking to amend requirements for the use of LWR fuel containing uranium enriched to greater than 5.0 weight percent U-235 in SRM-SECY-21-0109. The Commission directed that FFRD issues relevant to fuels of higher enrichment and burnup levels should be appropriately addressed and analyzed in the regulatory basis for this rulemaking.

The staff focused on the impact of loss-of-coolant accident (LOCA) analyses under Title 10 of the Code of Federal Regulations (10 CFR) 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. These analyses predict FFRD on the maximum hypothetical accident (MHA)-LOCA source term. In 2021, the NRC assessed FFRD impacts on the MHA-LOCA source term (ML21197A069).

Insights from the assessment resulted in the updated regulatory positions and guidance with the development of a new dose acceptance criteria for a 10 CFR 50.46 LOCA analysis that predicts FFRD.

KM-07 (p. 38)

In support of the Increased Enrichment Rulemaking, the staff focused on three primary areas that could prepare the agency for licensing increased enrichment:

(1) amending the control room design criteria, (2) updating regulatory source terms, and (3) updating guidance transport models and other assumptions.

With respect to area 3, modeling assumptions were assessed to understand whether new or updated information warranted a revision.

Occupancy factors are used to estimate the amount of time an individual is present at a particular location, typically expressed as a fraction per day. RG 1.183, Regulatory Position 4.2.6, assumes a time-stepped occupancy factor of 100 percent of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60 percent of the time between 1 and 4 days, and 40 percent of the time from 4 days to 30 days (about 317 hours0.00367 days <br />0.0881 hours <br />5.241402e-4 weeks <br />1.206185e-4 months <br /> total, or 13.2 days). The origin of this assumption is a report by Murphy and Campe, Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19, published in Proceedings of the 13th AEC Air Cleaning Conference, August 1974. This report is based on a summary of lessons learned from a review of over 50 control room designs that describe typical time-stepped occupancy factors.

In reconsidering accident modeling assumptions, the staff determined that additional flexibility could be included in its guidance as it pertains to the occupancy factors that are assumed in analyses that are within the scope of this guide. Specifically, licensee programs that are implemented to meet the requirements in 10 CFR Part 26, Fitness-for-Duty Programs, can be used to develop assumptions that more accurately represent worker occupancy during analyzed accidents. The regulation in 10 CFR 26.205(d) requires licensees to control the work hours of individuals. Specifically, 10 CFR 26.205(d)(1) requires that any individuals work hours do not exceed the following limits: 16 work hours

DG-1425, Appendix K, Page K-4 in any 24-hour period, 26 work hours in any 48-hour period, and 72 work hours in any 7-day period. Under emergency conditions, 10 CFR 26.207(d) provides licensees an exemption from the requirements of 10 CFR 26.205(c) and (d) during declared emergencies, as defined in the licensees emergency plan.

Therefore, this regulatory position was updated to credit facility-specific emergency plan staffing operations as the basis for the occupancy factor. This update is consistent with SRM-SECY-99-144, White Paper on Risk-Informed and Performance-Based Regulations, dated March 1, 1999 ((ML003753601), which describes the policy for a performance-based approach to give the licensee more flexibility in demonstrating compliance.

KM-08 (p. 40)

In support of the Increased Enrichment Rulemaking, the staff focused on three primary areas that could prepare the agency for licensing increased enrichment:

(1) amending the control room design criteria, (2) updating regulatory source terms, and (3) updating guidance transport models and other assumptions.

With respect to area 2, the staff developed a graded, performance-based, and risk-informed framework to support the amended control room design criteria.

When developing the framework, the staff considered several key Commission-directed probabilistic risk assessment (PRA) policies, which advocate certain changes to the development and implementation of its regulations using risk-informed and, ultimately, performance-based approaches. First, the staff reviewed the Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants (50 FR 32138; August 8, 1985), which describes the policy related to accidents more severe than the DBAs. This policy statement recognizes the usefulness of PRAs in identifying severe accident vulnerabilities and in providing additional insights to ensure that nuclear power plants do not pose an undue risk to public health and safety. Next, the staff reviewed the Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities policy statement (60 FR 42622; August 16, 1995), which formalized the Commissions commitment to risk-informed regulation through the expanded use of PRA. This policy statement states, in part, the following:

The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data, and in a manner that complements the NRCs deterministic approach and supports the NRCs traditional defense-in-depth philosophy.

Then, the staff reviewed SRM-SECY-98-144. This SRM defines the terms and Commission expectations for risk-informed and performance-based regulation.

The methodology described in this paper uses the terms and concepts of SRM-SECY-98-144, Item 8, Risk-Informed, Performance-Based Approach, which reads as follows:

DG-1425, Appendix K, Page K-5 A risk-informed, performance-based approach to regulatory decision-making combines the risk-informed and performance-based elements discussed in Items 5 and 7, above, and applies these concepts to NRC rulemaking, licensing, inspection, assessment, enforcement, and other decision-making.

Stated succinctly, a risk-informed, performance-based regulation is an approach in which risk insights, engineering analysis and judgment including the principle of defense-in-depth and the incorporation of safety margins, and performance history are used, to (1) focus attention on the most important activities, (2) establish objective criteria for evaluating performance, (3) develop measurable or calculable parameters for monitoring system and licensee performance, (4) provide flexibility to determine how to meet the established performance criteria in a way that will encourage and reward improved outcomes, and (5) focus on the results as the primary basis for regulatory decision-making.

The purpose of this framework is to enable a performance-based evaluation using traditional deterministic radiological consequence analyses methods within defined risk-informed boundaries. These boundaries are defined by acceptable radiation exposure guidelines for radiation workers during accident and emergency conditions and acceptable contemporary nuclear facility risk profiles using modern PRA methods. The intent of such a framework is to provide flexibility when determining how to meet an established acceptance criteria in a way that encourages and rewards improved outcomes. In practice, this method produces a framework that justifies a higher control room design criterion with a lower plant-specific risk metric.