NUREG-2246, DFC, Fuel Qualification for Advanced Reactors

NUREG-2246 DFC, Fuel Qualification for Advanced Reactors

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Issue date:	06/30/2021
From:	Paul Clifford, Timothy Drzewiecki, Jordan Hoellman, Jeffrey Schmidt, Vanwert C NRC/NRR/DANU/UARP
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NUREG-2246 DFC
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NUREG-2246 Fuel Qualification for Advanced Reactors Draft Report for Comment Office of Nuclear Reactor Regulation

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NUREG-2246 Office of Nuclear Reactor Regulation Fuel Qualification for Advanced Reactors Draft Report for Comment Manuscript Completed: June 2021 Date Published: June 2021 Prepared by:

T. Drzewiecki J. Schmidt C. Van Wert P. Clifford Jordan Hoellman, NRC Project Manager

COMMENTS ON DRAFT REPORT Any interested party may submit comments on this report for consideration by the NRC staff.

Comments may be accompanied by additional relevant information or supporting data. Please specify the report number NUREG-2246 in your comments, and send them by the end of the comment period specified in the Federal Register notice announcing the availability of this report.

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For any questions about the material in this report, please contact: Timothy Drzewiecki, Reactor Systems Engineer, 301-415-5184 or by e-mail at Timothy.Drzewiecki@nrc.gov or Jordan Hoellman, Project Manager, 301-415-5481 or by email at Jordan.Hoellman2@nrc.gov.

Please be aware that any comments that you submit to the NRC will be considered a public record and entered into the Agencywide Documents Access and Management System (ADAMS). Do not provide information you would not want to be publicly available.

iii ABSTRACT 1

Proposed advanced reactor designs use fuel designs and operating environments (e.g., neutron 2

energy spectra, fuel temperatures, neighboring materials) that differ from the large experience 3

base available for traditional light-water reactor fuel. The purpose of this report is to identify 4

criteria that will be useful for advanced reactor designs through an assessment framework that 5

would support regulatory findings associated with nuclear fuel qualification. The report begins by 6

examining the regulatory basis and related guidance applicable to fuel qualification, noting that 7

the role of nuclear fuel in the protection against the release of radioactivity for a nuclear facility 8

depends heavily on the reactor design. The report considers the use of accelerated fuel 9

qualification techniques and lead test specimen programs that may shorten the timeline for 10 qualifying fuel for use in a nuclear reactor at the desired parameters (e.g., burnup). The 11 assessment framework particularly emphasizes the identification of key fuel manufacturing 12 parameters, the specification of a fuel performance envelope to inform testing requirements, the 13 use of evaluation models in the fuel qualification process, and the assessment of the 14 experimental data used to develop and validate evaluation models and empirical safety criteria.

15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38

v TABLE OF CONTENTS ABSTRACT.................................................................................................................... iii LIST OF FIGURES........................................................................................................ vii LIST OF TABLES........................................................................................................... ix ABBREVIATIONS AND ACRONYMS............................................................................ xi 1 INTRODUCTION.....................................................................................................1-1 1.1 Purpose........................................................................................................................1-1 1.2 Safety Case..................................................................................................................1-1 1.3 Scope...........................................................................................................................1-2 2 BACKGROUND......................................................................................................2-1 2.1 Regulatory Basis..........................................................................................................2-1 2.2 Related Guidance.........................................................................................................2-2 2.2.1 NUREG-0800, Section 4.2..............................................................................2-3 2.2.2 ATF-ISG-2020-01...........................................................................................2-3 2.2.3 Regulatory Guide 1.233..................................................................................2-4 2.2.4 Guidance in Development...............................................................................2-4 2.3 Accelerated Fuel Qualification......................................................................................2-5 2.4 Lead Test Specimens...................................................................................................2-5 2.5 Assessment Framework...............................................................................................2-6 3 FUEL QUALIFICATION ASSESSMENT FRAMEWORK........................................3-1 3.1 G1Fuel Manufacturing Specification..........................................................................3-1 3.1.1 G1.1Dimensions.........................................................................................3-2 3.1.2 G1.2Constituents........................................................................................3-2 3.1.3 G1.3End State Attributes.............................................................................3-2 3.2 G2Safety Criteria......................................................................................................3-3 3.2.1 G2.1Design Limits during Normal Operation and Anticipated Operational Occurrences....................................................................3-3 3.2.2 G2.2Radionuclide Release Limits...............................................................3-5 3.2.3 G2.3Safe Shutdown....................................................................................3-8 3.3 Assessment Framework for Evaluation Models.......................................................... 3-10 3.3.1 EM G1Evaluation Model Capabilities........................................................ 3-11 3.3.2 EM G2Evaluation Model Assessment....................................................... 3-12 3.4 Assessment Framework for Experimental Data.......................................................... 3-15 3.4.1 ED G1Independence of Validation Data.................................................... 3-16 3.4.2 ED G2Test Envelope................................................................................ 3-16 3.4.3 ED G3Data Measurement......................................................................... 3-17 3.4.4 ED G4Test Conditions.............................................................................. 3-18 4

SUMMARY

AND CONCLUSIONS..........................................................................4-1 5 REFERENCES........................................................................................................5-1 APPENDIX A LIST OF ALL GOALS............................................................................ A-1

vii LIST OF FIGURES Figure 2-1 AFQ Process Workflow (Terrani, et al., 2020)...................................................2-5 Figure 3-1 Decomposition of the Main Goal.......................................................................3-1 Figure 3-2 Decomposition of G1, Fuel Manufacturing Specification.................................3-2 Figure 3-3 Decomposition of G2, Safety Criteria..............................................................3-3 Figure 3-4 Decomposition of G2.1, Design Limits During Normal Operation and AOOs...............................................................................................................3-4 Figure 3-5 Decomposition of G2.2, Radionuclide Release Limits.....................................3-5 Figure 3-6 Decomposition of G2.2.2, Criteria for Barrier Degradation..............................3-6 Figure 3-7 Decomposition of G2.2.3, Conservative Modeling of Radionuclide Retention and Release....................................................................................3-7 Figure 3-8 Decomposition of G2.3, Safe Shutdown.........................................................3-8 Figure 3-9 Decomposition of G2.3.1, Maintaining Coolable Geometry.............................3-8 Figure 3-10 Decomposition of G2.3.2, Control Element Insertion.................................... 3-10 Figure 3-11 Decomposition of the Main Goal for Evaluation Model Assessment................ 3-11 Figure 3-12 Decomposition of EM G1, Evaluation Model Capabilities.............................. 3-11 Figure 3-13 Decomposition of EM G2, Evaluation Model Assessment............................. 3-13 Figure 3-14 Decomposition of EM G2.2, Demonstrated Prediction Ability Over Test Envelope....................................................................................................... 3-14 Figure 3-15 Decomposition of the Main Goal for Data Assessment................................... 3-15 Figure 3-16 Decomposition of ED G3, Data Measurement.............................................. 3-17 Figure 3-17 Decomposition of ED G4, Test Specimens................................................... 3-18

ix LIST OF TABLES Table A-1 List of Goals in Fuel Qualification Assessment Framework.............................. A-1 Table A-2 List of Goals in Evaluation Model Assessment Framework.............................. A-2 Table A-3 List of Goals in Experimental Data Assessment Framework............................ A-2

xi ABBREVIATIONS AND ACRONYMS AFQ accelerated fuel qualification AOO anticipated operational occurrence ARDC advanced reactor design criterion ED experimental data EM evaluation model FAST fission accelerated steady-state test FQAF fuel qualification assessment framework G

goal GDC general design criterion/criteria GESTAR General Electric standard application for reactor fuel LWR light-water reactor OBE operating basis earthquake PCMI pellet-clad mechanical interaction PCMM predictive capability maturity model PIRT phenomena identification and ranking table SAFDL specified acceptable fuel design limit SARRDL specified acceptable radionuclide release design limit SSC structure, system, and component SSE safe shutdown earthquake TRISO tristructural-isotropic U-10Zr uranium alloy with 10 weight percent zirconium U-Pu-10Zr uranium-plutonium alloy with 10 weight percent zirconium UO2 uranium dioxide

1-1 1 INTRODUCTION 1

1.1 Purpose 2

The objective of nuclear fuel qualification is to demonstrat[e] that a fuel product fabricated in 3

accordance with a specification behaves as assumed or described in the applicable licensing 4

safety case, and with the reliability necessary for economic operation of the reactor plant 5

(Crawford, et al., 2007). Proposed advanced reactor designs have fuel designs and operating 6

environments (e.g., neutron energy spectra, fuel temperatures, neighboring materials) that differ 7

from the large experience base available for traditional light-water reactor (LWR) fuel. Nuclear 8

fuel affects many aspects of the overall design of a nuclear power plant, and qualification of 9

nuclear fuel has traditionally involved long development times. The purpose of this report is to 10 provide a fuel qualification assessment framework for use with advanced reactor designs that 11 satisfies regulatory requirements. Specifically, the framework provides criteria derived from 12 regulatory requirements that, when satisfied, would support regulatory findings necessary for 13 licensing. The framework follows a top-down approach in which a set of base goals1 support 14 high-level regulatory requirements.2 This report provides the bases for the identified base 15 goals and clarifying examples for the types of information that an applicant would need to 16 provide in order for the NRC to determine that these goals are satisfied and regulatory 17 requirements are met. Appendix A lists all goals within the framework.

18 19 This framework relies on regulatory requirements that are applicable to applications for design 20 certifications, combined licenses, manufacturing licenses, or standard design approvals. While 21 the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.43(e) are not 22 applicable to applications for a construction permit, the remaining requirements, identified in 23 Section 2.1, are generically applicable to power reactor applications. Accordingly, the framework 24 provides applicants with criteria for satisfying regulatory requirements for applications for a 25 design certification, combined license, manufacturing license, standard approval, and for the 26 development of a fuel qualification plan to support a construction permit application.

27 28 1.2 Safety Case 29 The role of nuclear fuel in the protection against the release of radioactivity can vary depending 30 on the reactor design3. For example, facilities that use traditional oxide fuels with metal cladding 31 are designed with robust barriers (e.g., containment buildings) to prevent the release of 32 radioactive material under postulated accident conditions, whereas a facility that uses 33 tristructural-isotropic (TRISO) fuel may credit a series of barriers (including barriers within the 34 fuel itself) to prevent the release of radioactive material (i.e., a functional containment (NRC, 35 2018a)). Thus, in the nuclear fuel qualification process, it is essential to specify the fission 36 product retention functions of the nuclear fuel (this is addressed under Goal (G) 2, Safety 37 Criteria, in Section 3.2 of this report).

38 1 A base goal is a goal that is not decomposed any further but is supported by evidence.

2 High-level in this context refers to its position in the framework. Regulatory requirements are located near the top of the framework and lower-level goals are provided that, if satisfied, provide bases for satisfying the regulatory requirements.

3 Fuel qualification literature often use the term safety case. This term is undefined but generally refers to the safety functions that the fuel is relied upon to perform. Principally among these safety-functions is the protection against the release of radionuclides.

1-2 1.3 Scope 1

Nuclear fuel affects many aspects of nuclear safety, including neutronic performance 2

(e.g., reactivity feedback), thermal-fluid performance (e.g., margin to critical heat flux limits), fuel 3

mechanical performance, reactor core seismic behavior, fuel transportation, and storage. The 4

scope of this report focuses on the identification and understanding of fuel life-limiting failure 5

and degradation mechanisms due to irradiation during reactor operation. The assessment 6

criteria in Section 3 of this report draw on regulatory experience gained from licensing solid fuel 7

reactor designs (particularly LWR designs), results from advanced reactor fuel testing 8

performed to-date, and accelerated fuel qualification (AFQ) considerations. An attempt has 9

been made to develop generically applicable criteria. However, some criteria may not apply to 10 liquid fuel forms (e.g., molten salt reactor fuel), and these fuel forms may require additional or 11 alternate criteria (see Section 2.2.4 for guidance on molten salt reactor fuel).

12

2-1 2 BACKGROUND 1

2.1 Regulatory Basis 2

Nuclear fuel qualification to support reactor licensing involves the development of a basis to 3

support findings associated with regulatory requirements that apply to the nuclear facility. This 4

section discusses these requirements and their relationship to this report. Note that satisfying 5

the fuel qualification framework criteria only partially addresses the requirements associated 6

with the nuclear facility. This is because the fuel qualification framework provides a means to 7

identify the safety criteria for the fuel and it is the safety criteria for the fuel that establish the 8

performance criteria for some structures, systems, and components (SSCs) of the facility.

9 Therefore, addressing the criteria in the fuel qualification framework provides the information 10 necessary to meet the regulations, but does not in and of itself satisfy regulatory requirements.

11 The requirements are fully addressed through the description and analysis of these SSCs in an 12 application.

13 14 The relevant regulatory requirements are as follows:

15 16 10 CFR 50.43(e)(1)(i) requires demonstration of the performance of each safety feature 17 of the design through either analysis, appropriate test programs, experience, or a 18 combination thereof. The assessment framework developed in Section 3 of this report 19 (1) provides a means to identify the safety features of the fuel necessary to comply with 20 regulatory requirements (see Goal (G) 2, Safety Criteria, in Section 3.2), and (2) 21 clarifies the types of evidence (e.g., analysis, testing, experience) typically expected to 22 demonstrate these safety features. In accordance with the scope of this report, the 23 safety features assessed in Section 3 are associated with the identification and 24 understanding of fuel life-limiting failure and degradation mechanisms that are due to 25 irradiation during reactor operation.

26 27 The regulation in 10 CFR 50.43(e)(1)(iii) requires that sufficient data exist on the safety 28 features of the design to assess the analytical tools used for safety analyses over a 29 sufficient range of normal operating conditions, transient conditions, and specified 30 accident sequences, including equilibrium core conditions. This range appears in G2.1.1, 31 Definition of Fuel Performance Envelope, which is discussed in Section 3.2.1.1 of this 32 report. Additionally, the evaluation model assessment framework in Section 3.3 provides 33 criteria for assessing analytical tools, and the experimental data assessment framework 34 in Section 3.4 provides criteria for data adequacy.

35 36 General Design Criterion (GDC) 2 and Advanced Reactor Design Criterion4 (ARDC) 2, 37 Design bases for protection against natural phenomena, of Appendix A, General 38 Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic licensing of 39 production and utilization facilities, requires that SSCs important to safety be designed 40 to withstand the effects of natural phenomena such as earthquakes, tornadoes, 41 hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety 42 functions. Appendix S to 10 CFR 50, Earthquake engineering criteria for nuclear power 43 plants, implements GDC 2 as it pertains to seismic events and defines specific 44 4 Regulatory Guide 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors, (NRC, 2018b) provides guidance on how the GDC in Appendix A to 10 CFR Part 50 may be adapted for non-LWR designs.

2-2 earthquake criteria for nuclear power plants. This appendix established definitions for 1

safe shutdown earthquake (SSE), operating basis earthquake (OBE), and safety 2

requirements for relevant SSCs. These SSCs are necessary to assure the integrity of 3

the reactor coolant pressure boundary, the capability to shut down the reactor and 4

maintain it in a safe-shutdown condition, or the capability to prevent or mitigate the 5

consequences of accidents that could result in potential offsite exposures. The safety 6

functions generally associated with nuclear fuel include control of reactivity, cooling of 7

radioactive material, and confinement of radioactive material5. The requirements related 8

to natural phenomena can be partially addressed by satisfying G2.3, Safe Shutdown, 9

which is discussed in Section 3.2.3.

10 11 GDC 10 and ARDC 10, Reactor Design, require that specified acceptable fuel design 12 limits (SAFDLs) or specified acceptable radionuclide release design limits (SARRDLs) 13 not be exceeded during any condition of normal operation, including the effects of 14 anticipated operational occurrences (AOOs). This requirement can be partially 15 addressed by satisfying G2.1, Design Limits during Normal Operation and AOOs, 16 which is discussed in Section 3.2.1.

17 18 GDC 27 and ARDC 26, Combined Reactivity Control Systems Capability, require, in 19 part, the ability to achieve and maintain safe shutdown under postulated accident 20 conditions and assurance that the capability to cool the core is maintained. This 21 requirement can be partially addressed by satisfying G2.3, Safe Shutdown, which is 22 discussed in Section 3.2.3.

23 24 GDC 35 and ARDC 35, Emergency Core Cooling, require an emergency core cooling 25 system that provides sufficient cooling under postulated accident conditions; they also 26 require that fuel and clad damage that could interfere with continued effective core 27 cooling is prevented. This requirement can be partially addressed by satisfying G2.3, 28 Safe Shutdown, which is discussed in Section 3.2.3.

29 30 The regulations in 10 CFR 50.34(a)(1)((ii)(D), 10 CFR 52.47(a)(2)(iv), and 31 10 CFR 52.79(a)(1)(vi) require an evaluation of a postulated fission product release. This 32 requirement can be partially addressed by satisfying G2.2, Radionuclide Release 33 Limits, which is discussed in Section 3.2.2.

34 The fuel qualification assessment framework in Section 3 of this report provides guidance to 35 facilitate an efficient and transparent licensing review in the area of fuel qualification. The 36 guidance provided in this report is not a substitute for the Commissions regulations, and 37 compliance with the guidance is not required.

38 39 2.2 Related Guidance 40 Several guidance documents are available or are in development that address nuclear fuel 41 qualification. This section discusses these guidance documents and their relationship to this 42 report.

43 5 These fundamental safety functions are identified in the IAEA safety glossary (IAEA, 2018) and are also incorporated into NRC regulations. Reactivity control is specified by GDC 27 and ARDC 26; heat removal is specified by GDC/ARDC 10, GDC 27, ARDC 26, and GDC/ARDC 35; radionuclide retention is specified by GDC/ARDC 10 and is associated with the requirements under 10 CFR 50.34(a)(1)((ii)(D), 10 CFR 52.47(a)(2)(iv),

and 10 CFR 52.79(a)(1)(vi).

2-3 2.2.1 NUREG-0800, Section 4.2 1

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear 2

Power Plants: LWR Edition, Section 4.2, Revision 3, Fuel System Design, issued March 2007 3

(NRC, 2007), lists acceptance criteria that staff considers in a licensing review for a LWR fuel 4

system. Section 3.2 of this report captures the objectives of the fuel system safety review as 5

follows:

6 7

Assurance that the fuel system is not damaged as a result of normal operation and 8

AOOs can be demonstrated, in part, by meeting G2.1, Design Limits during Normal 9

Operation and AOOs, which is discussed in Section 3.2.1.

10 11 Assurance that fuel system damage is never so severe as to prevent control element 12 insertion when required can be demonstrated, in part, by meeting G2.3, Safe 13 Shutdown, which is discussed in Section 3.2.3. Section 3.2.3.2 discusses the specific 14 item of control element insertion.

15 16 Assurance that the number of fuel rod failures is not underestimated for postulated 17 accidents can be demonstrated, in part, by meeting G2.2, Radionuclide Release Limits, 18 which is discussed in Section 3.2.2.

19 20 Assurance that coolability is always maintained can be demonstrated, in part, by 21 meeting G2.3, Safe Shutdown, which is discussed in Section 3.2.3. Section 3.2.3.1 22 discusses the specific item of maintaining a coolable geometry.

23 NUREG-0800, Section 4.2, provides guidance regarding traditional LWR fuel and the licensing 24 bases for traditional LWR power plants. Specifically, NUREG-0800, Section 4.2, evaluates fuel 25 system designs for known fuel failure mechanisms from traditional LWR fuel (i.e., uranium 26 dioxide (UO2) fuel with zirconium-alloy cladding), identifies specific testing for addressing key 27 LWR fuel phenomena, and includes empirical acceptance criteria based on testing of LWR fuel 28 samples. As such, the specific acceptance criteria provided in NUREG-0800, Section 4.2, may 29 not apply or may not suffice to address advanced reactor technologies that use different fuel 30 forms, or address situations in which the fuel plays different roles in the protection against the 31 release of radionuclides. However, this report incorporates lessons learned from the 32 development of the acceptance criteria in NUREG-0800, Section 4.2, as follows:

33 34 The significant effect of fuel manufacturing parameters on fuel performance is addressed 35 through G1, Fuel Manufacturing Specification, which is discussed in Section 3.1.

36 37 Limitations on test facilities and the risks of obtaining irradiated fuel data are discussed 38 in the experimental data assessment framework in Section 3.4 and are also mentioned 39 in Section 3.2.2.3.1.

40 2.2.2 ATF-ISG-2020-01 41 ATF-ISG-2020-01, Supplemental Guidance Regarding the Chromium-Coated Zirconium Alloy 42 Fuel Cladding Accident Tolerant Fuel Concept, issued January 2020 (NRC, 2020a), provides 43 supplementary guidance to NUREG-0800, Section 4.2. The guidance was developed using a 44

2-4 phenomena identification and ranking table (PIRT) process6 and is specific to applications 1

involving fuel products with chromium-coated zirconium alloy cladding. Like the guidance in 2

NUREG-0800, Section 4.2, the specific phenomena identified in ATF-ISG-2020-01 may not 3

apply to advanced reactor technologies. However, the PIRT process may be used to identify 4

failure mechanisms and necessary features of an evaluation model, as discussed in the 5

evaluation model assessment framework in Section 3.3 of this report.

6 7

2.2.3 Regulatory Guide 1.233 8

Regulatory Guide 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and 9

Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for 10 Licenses, Certifications, and Approvals for Non-Light-Water Reactors, issued June 2020 (NRC, 11 2020b), provides guidance for a modern, risk-informed approach to licensing reviews. This 12 approach emphasizes assessing facility risk by quantifying event frequencies and the 13 associated radiological consequences. The consequence evaluation aspect of the risk 14 assessment is addressed, in part, by G2.2, Radionuclide Release Limits, which is discussed in 15 Section 3.2.2.

16 17 Additionally, Regulatory Guide 1.233 discusses fundamental safety functions. Fuel qualification 18 partially addresses the fundamental safety functions of control of reactivity, cooling of 19 radioactive material, and confinement of radioactive material by incorporating the role of the 20 fuel in these safety functions in G2, Safety Criteria, which is discussed in Section 3.2 of this 21 report, as follows:

22 23 Confinement of radioactive material is partially addressed by G2.1, Design Limits during 24 Normal Operation and AOOs, and G2.2, Radionuclide Release Limits.

25 26 Control of reactivity and cooling or radioactive material are partially addressed by G2.3, 27 Safe Shutdown.

28 29 2.2.4 Guidance in Development 30 The U.S. Nuclear Regulatory Commission (NRC) staff is currently developing guidance in 31 additional areas related to fuel qualification. As discussed in Section 1.3, the safety case for 32 reactors that use nonsolid fuel forms may require additional or alternative criteria to those in this 33 report. To that end, the NRC is supporting the development of a proposed methodology for 34 molten salt reactor fuel salt qualification (ORNL, 2018) (ORNL, 2020).

35 36 Additionally, G2 addresses the role of the fuel in the protection against the release of 37 radioactivity, as discussed in Section 1.2. G2 is supported by source term considerations, as 38 detailed in G2.2.1, Radionuclide Retention Requirements, and G2.2.3, Conservative Modeling 39 of Radionuclide Retention and Release. Furthermore, G2.1, Design Limits during Normal 40 Operation and AOOs, discusses SARRDLs, which involve the use of a source term. The NRC 41 is supporting the development of source term guidance for non-LWRs which may affect this 42 aspect of fuel qualification (SAND, 2020) (INL, 2020). 7 43 6 See Regulatory Guide 1.203, Transient and Accident Analysis Methodologies, for more information on the PIRT process (NRC, 2005).

7 The guidance developed on source term does not alter the fuel qualification framework. Both the guidance on source term and the fuel qualification framework accommodate a graded approach to source term where simplified, conservative models can be used to reduce the data requirements.

2-5 1

2.3 Accelerated Fuel Qualification 2

AFQ involves, in part, the use of advanced modeling and simulation to inform constituent and 3

system selection and to enable integral fuel performance analyses (Terrani, et al., 2020). The 4

AFQ process, shown in Figure 2-1, supports the identification of important parameters and 5

phenomena for targeted characterization through separate-effects tests.

6 7

8 9

Figure 2-1 AFQ Process Workflow (Terrani, et al., 2020) 10 11 Advanced separate-effects testing techniques, such as fission accelerated steady-state testing 12 (FAST) (Beausoleil II, Povirk, & Curnutt, 2020) and MiniFuel (Petrie, Burns, Raftery, Nelson, &

13 Terrani, 2019), can reduce the time needed to achieve a given burnup and provide basic data 14 on material behavior and property evolution under irradiation conditions. The information 15 obtained through these analyses and separate-effects tests could help justify the adequacy of 16 the evaluation model as part of Evaluation Model (EM) G1, Evaluation Model Capabilities, 17 which is discussed in Section 3.3.1. Additionally, validated physics-based models may support 18 some extrapolation of evaluation models beyond the limits of available integral test data, as 19 noted under EM G.2.2.4, Restricted Domain, in Section 3.3.2.2.4. Ultimately, the AFQ process 20 relies on integral irradiation test data to validate engineering scale fuel performance codes and 21 to confirm the performance and safety of the fuel system under prototypic conditions.

22 Accordingly, the integral test data produced as part of the AFQ process appear to be consistent 23 with the considerations in the experimental data assessment framework discussed in 24 Section 3.4.

25 26 2.4 Lead Test Specimens 27 Much of the data necessary to qualify fuel for use come from irradiated test specimens. Lead 28 test specimens have been successfully used in operating reactors to obtain data at the needed 29 exposures and are discussed in NUREG-0800, Section 4.2, as well as in Section 3.4.2 of this 30 report. Section 3.4.2 of this report further discusses the potential for use of lead test specimens 31 beyond what has been traditionally used for LWRs that can be useful for advanced reactor 32 technologies.

33 34

2-6 2.5 Assessment Framework 1

The top-down development of an assessment framework is not a novel approach in the 2

regulatory process. Similar assessment frameworks have been developed in the code scaling, 3

applicability, and uncertainty evaluation methodology (NRC, 1989), the evaluation model 4

development and assessment process (NRC, 2005), and the objectives hierarchy discussed in 5

NUREG/BR-0303, Guidance for Performance-Based Regulation, issued December 2002 6

(NRC, 2002). Another top-down assessment framework, developed for thermal margin 7

evaluations for LWRs, was based on many years of safety reviews (NRC, 2019). Assessment 8

frameworks have facilitated safety reviews and have been shown to increase transparency 9

about information needs, to promote efficiency by focusing attention on areas of recognized 10 importance, and to clarify the logic behind decisions.

11 12

3-1 3 FUEL QUALIFICATION ASSESSMENT FRAMEWORK 1

This section on the fuel qualification assessment framework (FQAF) systematically identifies 2

fuel safety criteria. The comprehensive list of safety criteria, called a fuel assessment 3

framework, is informed by existing regulatory requirements, regulatory guidance, and staff 4

experience with safety reviews for nuclear fuel in both LWRs and non-LWRs. The fuel 5

assessment framework is developed using a top-down approach that starts with the high-level 6

goal (G) that the fuel be qualified for use and then decomposes this goal into subgoals. Meeting 7

the subgoals indicates that the higher-level goal is met. Each subgoal can either be further 8

decomposed into other subgoals, or if no further decomposition is deemed necessary, the 9

subgoal may be considered a base goal and evidence must be provided to demonstrate that the 10 base goal is satisfied. In this report, base goals are identified by the use of grey boxes.

11 12 Consistent with the purpose of fuel qualification (see Section 1.1) and with a regulatory focus on 13 safety, this report uses the following definition for fuel qualification:

14 15 Fuel is qualified for use if reasonable assurance exists that the fuel, fabricated 16 in accordance with its specification, will perform as described in the safety 17 analysis.

18 19 This statement is captured figuratively in Figure 3-1, which decomposes fuel qualification into 20 two supporting goals. These goals are further decomposed into lower level supporting goals, 21 until criteria are obtained which can be directly verified by evidence. The subsections that follow 22 describe the process, criteria, and associated evidence necessary to demonstrate fuel 23 qualification.

24 25 26 Figure 3-1 Decomposition of the Main Goal 27 28 3.1 G1Fuel Manufacturing Specification 29 Fuel performance during normal operation and accident conditions can be highly sensitive to the 30 fuel fabrication process. For example, failure criteria during reactivity-initiated accidents for 31 LWRs with zirconium-based cladding depend upon the heat treatment of the cladding because 32 of its impact on microstructure (NRC, 2020c). Similarly, key manufacturing parameters have 33 been identified for TRISO fuel that must be controlled to ensure satisfactory performance (EPRI, 34

3-2 2020). Staff recognizes that manufacturing processes for a nuclear fuel product may evolve 1

over the product life cycle; therefore, a complete manufacturing specification is not expected as 2

part of the licensing documentation. However, the licensing documentation should include 3

sufficient information to ensure the control of key parameters affecting fuel performance during 4

the manufacturing process. The goal G1 is decomposed as shown in Figure 3-2 to identify the 5

specific types of information to be included in licensing documentation.

6 7

8 9

Figure 3-2 Decomposition of G1, F uel Manufacturing Specification 10 11 3.1.1 G1.1Dimensions 12 Key dimensions and tolerances for fuel components that affect performance should be 13 specified. Consistent with the scope of this report, as discussed in Section 1.3, these 14 dimensions and tolerances should be specific to components that affect fuel life-limiting failure 15 and degradation mechanisms that are due to irradiation during reactor operation (e.g., fuel pellet 16 and cladding dimensions, key assembly dimensions). It is recognized that some of dimensions 17 can be controlled by an approved change process (e.g., General Electric Standard Application 18 for Reactor Fuel (GESTAR)).

19 20 3.1.2 G1.2Constituents 21 Key constituents of fuel components (e.g., uranium dioxide (UO2) fuel, uranium-plutonium-22 zirconium fuel alloys with specified concentrations (U-Pu-10Zr), cladding material) should be 23 specified, along with allowances for impurities.

24 25 3.1.3 G1.3End State Attributes 26 End state attributes for the materials within fuel components (e.g., microstructure) should be 27 specified or otherwise justified. The information necessary to capture the desired end state of 28 the material may take several forms. For example, specific manufacturing processes 29 (e.g., cold-working, heat treatments, acid pickling, deposition techniques) that are essential to 30 create the microstructure may be indicated in lieu of end state attributes. In some cases, it may 31

3-3 be preferable to use performance-based end state attributes that can be supported through 1

periodic testing and reporting (NRC, 2016). Additionally, it may be possible to demonstrate 2

insensitivity to manufacturing processes so that end state attributes need not be specified in 3

licensing documentation. Licensing documentation should provide sufficient justification for 4

cases where a specific material is insensitive to manufacturing processes.

5 6

3.2 G2Safety Criteria 7

An evaluation of the safety case involves an assessment against safety criteria, which are 8

generally associated with protection against the release of radioactive material but also address 9

the fundamental safety functions of heat removal and reactivity control. In general, many safety 10 criteria for nuclear fuel depend on the events to which the fuel is subjected. Specifically, nuclear 11 fuel is expected to retain its integrity under conditions of normal operation, including the effects 12 of AOOs, but some degree of fuel failure can be accommodated for low-frequency design-basis 13 accident conditions (i.e., those not expected to occur during the life of the plant). The goal G2 is 14 decomposed as shown in Figure 3-3 to address the varying types of safety criteria for the range 15 of events for which nuclear fuel should be qualified.

16 17 18 Figure 3-3 Decomposition of G2, Safety Criteria 19 20 3.2.1 G2.1Design Limits during Normal Operation and Anticipated Operational 21 Occurrences 22 Fuel integrity is expected to remain intact under conditions of normal operation, including the 23 effects of AOOs. Alternatively, some designs may use SARRDLs, which allow a small degree of 24 radionuclide release from the fuel (NRC, 2018b). Multiple degradation mechanisms and failure 25 modes may exist; limits need to be established to protect against all of them. At the highest 26 level, the assessment of a fuel against design limits for normal operation and AOOs requires 27 knowledge of the conditions that the fuel is exposed to (i.e., the performance envelope) and a 28 method to assess the fuel performance under those conditions (i.e., an evaluation model).

29 These supporting goals, shown in Figure 3-4, are discussed below.

30 31

3-4 1

Figure 3-4 Decomposition of G2.1, Design Limits During Normal Operation and AOOs 2

3 3.2.1.1 G2.1.1Definition of Fuel Performance Envelope 4

The fuel performance envelope specifies the environmental conditions and radiation exposure 5

under which the fuel is required to perform. This performance envelope informs the safety 6

analysis and technical specifications for the design (i.e., limiting conditions for operation). It is 7

noted that irradiation-induced growth and fission product swelling of fuel components are often 8

life-limiting phenomena for the fuel design. The envelope may be specified by fuel designers 9

and may constrain the design of the reactor and associated systems. Alternatively, a reactor 10 design may be proposed that imposes constraints on fuel performance. In support of G2.1, the 11 goal G2.1.1 can be met by specifying the conditions (e.g., temperatures, pressures, power),

12 exposure, and transient conditions that the fuel is expected to encounter during normal 13 operation, including AOOs. Additionally, G2.1.1 supports G2.2, which addresses the fuel 14 contribution to the source term during design-basis accidents, as discussed in Section 3.2.2.1.

15 Accordingly, this goal can be fully satisfied by specifying the conditions the fuel is expected to 16 encounter during normal operation, AOOs, and design-basis accidents.

17 18 3.2.1.2 G2.1.2Evaluation Model 19 This goalthat evaluation models are available to assess fuel performance against design 20 limits to protect against fuel failure and degradation mechanismsrequires the specification of 21 means of evaluating fuel for performance, failure, and degradation. The assessment of 22 evaluation models supports several goals and is further decomposed. Therefore, Section 3.3 23 provides a separate assessment framework for evaluation models. G2.1.2 is satisfied by 24 meeting the supporting goals in that framework for fuel performance during normal operation 25 and AOOs.

26 27

3-5 3.2.2 G2.2Radionuclide Release Limits 1

Radiological consequences under postulated accident conditions are an essential consideration 2

in nuclear power plant licensing. Under postulated accident conditions, some fuel failure is 3

possible, which contributes to the accident source term. As radionuclide inventory originates 4

from the nuclear fuel, fuel qualification should include characterizing the behavior of the fuel 5

under accident conditions, so that its contribution to the accident source term can be determined 6

in a suitably conservative manner. Accordingly, the goal G2.2the ability to demonstrate 7

margin to radionuclide release limits under accident conditions, in relation to fuel qualification 8

is supported by three goals related to the fuel contribution to the accident source term. These 9

goals, shown in Figure 3-5 (along with G2.1.1, which also supports G2.2), are discussed further 10 below.

11 12 13 Figure 3-5 Decomposition of G2.2, Radionuclide Release Limits 14 15 3.2.2.1 G2.1.1Definition of Fuel Performance Envelope 16 Section 3.2.1.1 already discussed G2.1.1. In support of G2.2, this goal can be satisfied by 17 specifying the design-basis accident conditions to which the fuel is subjected. Design-basis 18 accident conditions depend on reactor design; however, as discussed in Section 3.2.1.1, 19 conditions to which the fuel is subjected during design-basis accidents may be specified 20 independent of the reactor design, leading to constraints on the design of the reactor and 21 associated systems. The types of design-basis accident conditions that should be considered 22 include transient overpower events (e.g., reactivity-initiated accidents), transient undercooling 23 events (e.g., loss-of-coolant accidents), and externally applied loads (e.g., fuel handling, 24 transportation, seismic activity, and major piping failures).

25 26 3.2.2.2 G2.2.1Radionuclide Retention Requirements 27 The role of nuclear fuel in the safety case can vary between reactor designs and fuel types. For 28 example, traditional LWR fuel that uses UO2 pellets with zircalloy cladding is not credited to 29 retain cladding integrity under large-break loss-of-coolant accidents8, while advanced reactor 30 designs may credit retention of radionuclides within the fuel under accident conditions.

31 Additionally, plant site characteristics such as proximity to population and weather patterns may 32 further influence radionuclide retention requirements (even for the same reactor and fuel 33 8 NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, (NRC, 1995) states that, Assuming that the coolant loss cannot be accommodated by the reactor coolant makeup systems or the emergency core cooling systems, fuel cladding failure would occur with the release of radioactivity located in the gap between the fuel pellet and the fuel cladding.

3-6 design). To satisfy G2.2.1, the degree of radionuclide retention within the fuel system should be 1

specified.

2 3

3.2.2.3 G2.2.2Criteria for Barrier Degradation 4

Radionuclide barrier (e.g. fuel cladding) failure and degradation mechanisms under accident 5

conditions (e.g., pellet-clad mechanical interaction (PCMI) and high enthalpy failure, 6

temperature-induced reactions and phase transformations) must be understood when the 7

design credits retention of barrier integrity (e.g., during reactivity-initiated accidents in LWRs, or 8

considering the potential for fission product attack of the silicon carbide layer in TRISO fuel at 9

high temperatures). As such, the goal G2.2.2 is decomposed into two supporting goals, shown 10 in Figure 3-6.

11 12 13 Figure 3-6 Decomposition of G2.2.2, Criteria for Barrier Degradation 14 15 3.2.2.3.1 G2.2.2(a)Conservative Criteria 16 Criteria used to determine barrier degradation should be suitably conservative. These criteria 17 are expected to be established based on transient testing and irradiated fuel samples, as 18 discussed under G2.2.2(b). Ideally, to establish a statistical confidence level (e.g., 95/95),

19 criteria would be established through a regression analysis using experimental data, then 20 validated by assessment against a separate and independent set of data (see Section 3.4.1 21 (Experimental Data (ED) G1) for a discussion on data independence). However, this ideal 22 scenario may not be realized due to challenges associated with obtaining irradiated fuel 23 samples and conducting transient testing for design-basis accident conditions. The amount of 24 experimental data supporting the criteria should be proportional to the degree of understanding 25 of key degradation and performance phenomena (NRC, 2020c). If the data collected are not 26 sufficient to support statistical modeling, a conservative or bounding approach may be required.

27 28 3.2.2.3.2 G2.2.2(b)Experimental Data 29 This goal is satisfied through an evaluation against the experimental data assessment 30 framework in Section 3.4.

31

3-7 1

3.2.2.4 G2.2.3Conservative Modeling of Radionuclide Retention and Release 2

Consistent with the requirements specified as part of G2.2.1 and discussed in Section 3.2.2.2, 3

radionuclide retention and release behavior of the fuel under accident conditions should be 4

modeled conservatively. This goal is related to the barrier degradation criteria specified in 5

G2.2.2 and discussed in Section 3.2.2.3, but it differs in its focus on radionuclide retention within 6

the fuel matrix (e.g., UO2 pellet or uranium alloy with 10 percent zirconium (U-10Zr) fuel ingot) or 7

fuel particle (e.g., fuel compact for a TRISO-based fuel). This goal is decomposed into two 8

supporting goals, as shown in Figure 3-7.

9 10 11 12 Figure 3-7 Decomposition of G2.2.3, Conservative Modeling of Radionuclide Retention 13 and Release 14 15 3.2.2.4.1 G2.2.3(a)Conservative Transport Model 16 The model of radionuclide transport within the fuel matrix should be conservative. As in the case 17 of barrier degradation criteria, discussed in Section 3.2.2.3.1, challenges associated with 18 obtaining and testing irradiated fuel samples may make it difficult to obtain sufficient data to 19 characterize the transport model in a statistical manner; therefore, conservative or bounding 20 estimates may be required. Additionally, previous source term models for LWRs have generally 21 included some degree of expert judgment. A clarifying example of how to develop a suitably 22 conservative radionuclide transport model is available in regulatory guidance on accident source 23 terms (NRC, 2000).

24 25 3.2.2.4.2 G2.2.3(b)Experimental Data 26 This goal is satisfied through an evaluation against the experimental data assessment 27 framework in Section 3.4.

28 29

3-8 3.2.3 G2.3Safe Shutdown 1

Safe shutdown of a nuclear plant refers to a state in which the reactor is subcritical, decay heat 2

is being removed, and radionuclide inventory is contained. The international atomic energy 3

agency (IAEA) refers to this as a safe state (IAEA, 2018). The ability to achieve safe shutdown 4

in any scenario needs to be assured. Therefore, criteria need to be established to ensure that a 5

coolable geometry is maintained in all scenarios and that fuel system damage is never so 6

severe as to prevent control element (e.g., control rod) insertion when required. These 7

supporting goals, captured in Figure 3-8, are discussed below.

8 9

10 Figure 3-8 Decomposition of G2.3, Safe Shutdown 11 12 3.2.3.1 G2.3.1Maintaining Coolable Geometry 13 The maintenance of a coolable geometry is identified as a supporting goal in achieving and 14 maintaining safe shutdown. It is further decomposed into the supporting goals shown in 15 Figure 3-9, which are discussed below.

16 17 18 Figure 3-9 Decomposition of G2.3.1, Maintaining Coolable Geometry 19 20 21

3-9 3.2.3.1.1 G2.3.1(a)Identification of Phenomena 1

Phenomena that could cause the loss of coolable geometry should be specified. Existing NRC 2

regulations and guidance applicable to design basis accidents specify some acceptance criteria 3

for these events that are intended to prevent such phenomena from significantly altering core 4

geometry under postulated accident conditions. Examples of phenomena that could cause the 5

loss of coolable geometry include: (1) fuel melt, (2) fuel swelling and fuel pellet and cladding 6

fragmentation and dispersal during transient overpower events, and (3) loss of cladding ductility 7

or long-term cladding phase stability during loss-of-coolant accidents.

8 9

3.2.3.1.2 G2.3.1(b)Evaluation Models 10 Several evaluation models may be needed to demonstrate that coolable geometry is 11 maintained. These models typically involve the use of conservative criteria and the evidence 12 needed to meet this goal depends on the associated phenomena. For example, a 13 conservatively chosen criterion such as the onset of fuel melting should not require a detailed 14 evaluation model supported by integral testing, but an empirically based criterion such as 15 energy deposition for fuel dispersal or peak cladding temperature for cladding embrittlement 16 requires the demonstration of an appropriate margin against experimental data. Historical 17 examples of acceptable empirical criteria include those developed for transient overpower 18 (NRC, 2020c) and loss-of-coolant accidents (Hache & Chung, 2000). In addition to these 19 empirical models for demonstrating a coolable geometry, analytical models have been used to 20 demonstrate that coolable geometry is maintained for internal and external events (Framatome, 21 2018).

22 23 The evaluations performed to demonstrate coolable geometry vary in terms of complexity, form 24 simple conservative criteria to detailed dynamic response models. The most general case that 25 applies to all these situations is the generic evaluation model assessment discussed in 26 Section 3.3. Accordingly, this goal is satisfied through a comparative assessment against the 27 evaluation model assessment framework in Section 3.3. The application of the evaluation model 28 assessment framework should follow a graded approach in accordance with the level of 29 understanding of the physical phenomena and conservatism in the criteria.

30 31 3.2.3.2 G2.3.2Control Element Insertion 32 Control element insertion is identified as a supporting goal in achieving and maintaining safe 33 shutdown. It is further decomposed into the supporting goals shown in Figure 3-10, which are 34 discussed below.

35 36

3-10 1

Figure 3-10 Decomposition of G2.3.2, Control Element Insertion 2

3 3.2.3.2.1 G2.3.2(a)Identification of Criteria 4

Criteria should be specified to ensure that the control element insertion path is not obstructed 5

during normal operation or accident conditions. These criteria should consider loads from both 6

internal and external (e.g., seismic) events. An example of such a criterion for traditional LWRs 7

is the stress limit imposed on the control rod guide tubes to inhibit distortion of the insertion 8

path.

9 10 3.2.3.2.2 G2.3.2(b)Evaluation Model 11 The evaluation performed to demonstrate that control element insertion can be assured has 12 typically involved a stress analysis to ensure that the control element insertion path is not 13 deformed as a result of internal and external events. This is typically done using a separate 14 evaluation model. Accordingly, this goal is satisfied through a comparative assessment against 15 the evaluation model assessment framework in Section 3.3.

16 17 3.3 Assessment Framework for Evaluation Models 18 The term evaluation model here is used in the generic sense. Typically, an evaluation model is 19 an analytical tool, a computer code, or a combination of such tools. However, the use of a 20 sophisticated tool such as a computer code may not be necessary to evaluate fuel performance.

21 For example, a simple mathematical expression or set of data can serve as an evaluation 22 model, if sufficient evidence exists to support its use.

23 24 The evaluation model assessment framework developed here is designed to be generically 25 applicable. In particular, it supports G2.1.2, which addresses the evaluation of design limits 26 under conditions of normal operation and AOOs, G2.3.1(b), which addresses maintaining 27 coolable geometry, and G2.3.2(b), which addresses control element insertion. The evaluation 28 model assessment framework presented here overlaps conceptually with the goals previously 29 established for criteria for barrier degradation (Section 3.2.2.3) and radionuclide retention and 30 release (Section 3.2.2.4). The latter two goals, however, have historically involved empirical 31 evaluation models based on destructive testing using irradiated nuclear fuel under accident 32 conditions. Accordingly, goals for barrier degradation and radionuclide retention and release are 33 provided separately from the evaluation model assessment framework of this section.

34

3-11 1

The top-level goal of an acceptable evaluation model is supported by the goals of (1) adequate 2

modeling capabilities and (2) assessment against experimental data. These supporting goals 3

are shown in Figure 3-11 and discussed below.

4 5

6 Figure 3-11 Decomposition of the Main Goal for Evaluation Model Assessment 7

8 3.3.1 EM G1Evaluation Model Capabilities 9

The evaluation model capabilities goal is decomposed into three supporting goals as shown in 10 Figure 3-12. This decomposition is informed by the predictive capability maturity model (PCMM) 11 framework, which identifies representation and geometric fidelity and physics and material 12 model fidelity as assessment elements (SAND, 2007). The evaluation model assessment 13 framework also considers other elements of the PCMM framework. Specifically, EM G2 14 addresses model validation and uncertainty quantification and sensitivity analysis; see 15 Section 3.3.2. The remaining elements of the PCMM framework, code verification and 16 solution verification, are expected to be addressed as part of a quality assurance program for 17 the design, analysis, and fabrication of a nuclear power facility. The goals supporting EM G1, 18 shown in Figure 3-12, are discussed below.

19 20 21 Figure 3-12 Decomposition of EM G1, Evaluation Model Capabilities 22 23

3-12 3.3.1.1 EM G1.1Geometry Modeling 1

The evaluation model should be capable of modeling the geometry of the fuel system. Table 3 2

of the PCMM provides guidance on the levels of maturity needed to assess the geometry, 3

including consideration of peer review (SAND, 2007). It is recognized that some fuel designs 4

may require simplifying assumptions to address difficulties in geometric modeling. For example, 5

TRISO-based particulate fuel involves coupled phenomena occurring at different geometric 6

scales (e.g., micro-scale within the TRISO particle, meso-scale within the fuel compact, and 7

macro-scale within the reactor core). Geometric modeling for such particulate fuel could involve 8

simplifications and assumptions that a less heterogeneous fuel design may not require.

9 Additionally, the evaluation model should be able to capture geometric changes due to 10 irradiation and exposure to the in-reactor environment (e.g., fuel swelling, cladding creep, oxide 11 layer growth). Irrespective of imposed simplifications, the geometric modeling scheme should be 12 appropriately justified, and the integrated evaluation model should be validated through the 13 assessment process under EM G2.

14 15 3.3.1.2 EM G1.2Material Modeling 16 The evaluation model should be capable of modeling material properties of the fuel system and 17 its surrounding environment. This includes changes in material properties due to irradiation and 18 exposure to the in-reactor environment (e.g., thermal conductivity degradation in nuclear fuel, 19 changes to melting temperature, eutectic formation, changes to Youngs modulus). Table 3 of 20 the PCMM provides guidance on the levels of maturity needed to assess the material modeling, 21 including considerations for model calibration against test data and peer review (SAND, 2007).

22 The material modeling scheme should be justified, and the integrated evaluation model should 23 be validated through the assessment process under EM G2.

24 25 3.3.1.3 EM G1.3Physics Modeling 26 The evaluation model should be capable of modeling the physical processes that affect fuel 27 performance. This goal requires knowledge of failure mechanisms, including changes due to 28 irradiation and exposure to the in-reactor environment for the specified fuel, as well as fuel 29 contribution to the SARRDL, if applicable. The evaluation model is expected to include sufficient 30 physics modeling to address known degradation mechanisms (e.g., cladding oxidation and 31 hydrogen pickup, fuel rod internal pressure, cladding strain, fuel assembly growth and wear, 32 stress and fatigue for fuel components). Table 3 of the PCMM provides guidance on the levels 33 of maturity needed to assess the physics modeling, including considerations for model 34 calibration against test data and peer review (SAND, 2007). The physics models incorporated 35 into the evaluation model should be justified, and the integrated evaluation model should be 36 validated through the assessment process under EM G2. Means of justification include the use 37 of an expert panel to develop a PIRT (PNNL, 2019) and internal review based on past 38 experience, legacy data (ANL, 2018), or separate-effects testing (Beausoleil II, Povirk, &

39 Curnutt, 2020) (Petrie, Burns, Raftery, Nelson, & Terrani, 2019).

40 41 3.3.2 EM G2Evaluation Model Assessment 42 Evaluation model assessment is an essential process that provides confidence in the 43 application of the evaluation model. To ensure that evaluation model predictions are suitably 44 conservative, they should be assessed against appropriate experimental data. For statistically 45 based modeling approaches, any bias or uncertainty in the evaluation model prediction should 46 be adequately quantified, so that design and safety analyses can account for such bias or 47

3-13 uncertainty. For conservative modeling approaches, the evaluation model should suitably bound 1

the experimental data. The assessment process comprises two supporting goals, shown in 2

Figure 3-13, which are discussed below.

3 4

5 Figure 3-13 Decomposition of EM G2, Evaluation Model Assessment 6

7 3.3.2.1 EM G2.1Experimental Data 8

This goal is satisfied through an evaluation against the experimental data assessment 9

framework in Section 3.4.

10 11 12 3.3.2.2 EM G2.2Demonstrated Prediction Ability over Test Envelope 13 EM G2.2 involves the comparison of evaluation model predictions against experimental data, 14 which should establish uncertainties and biases and identify limitations in the applicability of the 15 evaluation model. EM G2.2 is satisfied by meeting the four supporting goals shown in 16 Figure 3-14, which are discussed below.

17 18

3-14 1

Figure 3-14 Decomposition of EM G2.2, Demonstrated Prediction Ability Over Test 2

Envelope 3

3.3.2.2.1 EM G2.2.1Quantification of Error 4

Uncertainties and biases for figures of merit need to be sufficiently understood to establish 5

confidence in the evaluation model. It is expected that, to determine uncertainties and biases, 6

the predictions of the evaluation model for assessment cases will be compared against 7

assessment data, and the differences between measured and predicted values will be 8

quantified. If sufficient data exist, then statistical confidence levels can be placed on the 9

uncertainties of the evaluation model predictions. However, a more bounding or conservative 10 approach can also be taken (e.g., applying a bias or penalty to the model predictions, showing 11 that the model is inherently conservative). EM G2.2.1 can be satisfied by a statement on the 12 evaluation model biases and uncertainties, along with justification through a quantification of the 13 ratio of predicted to measured values for assessment cases.

14 15 3.3.2.2.2 EM G2.2.2Span of Validation Data 16 Assessment data should be distributed throughout the fuel performance envelope. The fuel 17 performance envelope, discussed in Sections 3.2.1.1 and 3.2.2.1, is used to specify the test 18 envelope; accordingly, assessment data should be available to assess the evaluation model 19 over the entire span of the performance envelope. However, it is recognized that certain regions 20 of the fuel performance envelope may not require data. For example, post-irradiation 21 examination of an integral test specimen may not be necessary for low-burnup fuel. In such 22 cases, it may suffice to provide justification that those regions do not require data (e.g., that 23 limiting phenomena are known not to be present below a specified burnup). EM G2.2.2 can be 24 satisfied by demonstrating that assessment data are available over the entire performance 25 envelope, and by justifying any gaps in assessment data.

26 27 3.3.2.2.3 EM G2.2.3Data Density 28 Assessment data should be appropriately distributed throughout the fuel performance envelope.

29 As discussed in Section 3.3.2.2.2, it may be acceptable to have regions in the performance 30 envelope where the evaluation model is not directly supported by assessment data from integral 31

3-15 experiments. However, in regions that do require assessment data, a sufficient number of data 1

points should be available for assessment of the evaluation model. It is reasonable to expect 2

data density to be greater near conditions of normal operation, as fuel designers may require 3

additional data to satisfy fuel reliability targets. However, any sparse data regions (i.e., regions 4

of low data density) in the fuel performance envelope need adequate justification. EM G.2.2.3 5

can be satisfied by justifying the data density throughout the fuel performance window.

6 7

3.3.2.2.4 EM G2.2.4Restricted Domain 8

Use of the evaluation model should be restricted to application domains for which the model has 9

been assessed. Application of an evaluation model outside of the supporting test envelope (see 10 Section 3.4.2) may be justified based on physical arguments (e.g., that the evaluation model 11 provides a simplified or bounding treatment of physical phenomena). Justification for 12 extrapolation of a model outside of the test envelope is strengthened by the use of 13 physics-based models, such as those discussed in Section 2.3, which are informed by 14 fundamental information about fuel evolution and behavior, as opposed to empirically derived 15 models (Terrani, et al., 2020). EM G2.2.4 can be satisfied by specifying the application domain 16 of the evaluation model as supported by the test envelope and by additional physical arguments 17 as necessary.

18 19 3.4 Assessment Framework for Experimental Data 20 The assessment of experimental data is the largest area of review for fuel qualification. The 21 assessment framework developed here supports all goals requiring evaluations against 22 assessment data. Because a fuel qualification program involves several types of experiments 23 (e.g., steady-state irradiation of integral test specimens, transient ramp testing, design-basis 24 accident testing), and because of transient test facility limitations and challenges associated 25 with irradiated fuel testing, it is recognized that the level of evidence expected to support a goal 26 can vary depending on the type of data collected. The assessment framework presented in this 27 section discusses this variance in the level of evidence as applicable. The main goal for 28 assessment data is decomposed, as shown in Figure 3-15, into four supporting goals, which are 29 discussed below.

30 31 32 Figure 3-15 Decomposition of the Main Goal for Data Assessment 33 34

3-16 3.4.1 ED G1Independence of Validation Data 1

Assessment data consist of experimentally measured values that are used to quantify the error 2

in the evaluation model. Ideally, assessment data should be independent from any data used in 3

the development (i.e., training) of the evaluation model. Although it may seem appropriate to 4

use training data, training data cannot provide an accurate assessment because the evaluation 5

model has already been tuned to those data. That is, quantifying the error of the training data 6

would only show how well the model can predict the data used to generate it, not how well the 7

model can predict data not used to generate it. Substantially more data points appear in the 8

application domain (an infinite number) than were used to generate the model, and these are 9

the points of most interest in future uses of the model; therefore, the focus should be on 10 estimating the error over those points, not on the points used to generate the model. Thus, 11 experimental data that were not used to train the model should be held in reserve and used to 12 validate the model. Maintaining validation data separate from the model development process 13 helps avoid a potential source of bias that could provide a distorted indication of the models 14 accuracy for future uses.

15 16 In some instances, however, the validation data and the training data are one and the same.

17 Methods exist in machine learning for determining whether the selection of the training data 18 affects the resulting uncertainty; such methods include random subsampling and k-fold 19 cross-validation. In each of these methods, the available data are randomly separated into 20 subsets of training and validation data. The training data are used to develop the coefficients of 21 the model, and the validation data are used to determine the overall uncertainty of the model.

22 The process is then repeated with different randomly selected training and validation data sets.

23 These methods can provide reasonable estimates of the impact of using the same data for 24 training and validation.

25 26 The discussion of data independence has so far considered scenarios where a sufficient 27 number of data points exist to train and validate a model using statistical approaches (i.e.,

28 model regression and the calculation of confidence intervals). It is recognized, however, that 29 only limited data may be available because of the challenges associated with obtaining 30 irradiated fuel samples. Experience from transient overpower testing has shown that it may be 31 acceptable to develop criteria without separating the data into training and validation sets (NRC, 32 2020c). Similarly, fission gas release and swelling models have been proposed based on a 33 limited amount of test data (Lee, Kim, & Jung, 2001). ED G1 can be satisfied by demonstrating 34 that the data used in the evaluation model assessment are sufficiently independent.

35 36 3.4.2 ED G2Test Envelope 37 Data should be collected over a test envelope that spans the performance envelope (see 38 Section 3.2.1.1). The performance envelope should address normal operation, AOOs, and 39 postulated accident conditions. The development of the test envelope should consider 40 (1) steady-state integral testing of the fuel system in a prototypical environment, (2) high-power 41 and undercooling tests to address AOO conditions and to assess design margins, (3) power 42 ramp testing to assess fuel performance during anticipated power changes, and 43 (4) design-basis accident tests to establish margin to fuel breach and contribution to the source 44 term under accident conditions. Typical design-basis accident scenarios of interest include 45 overpower events (e.g., reactivity-initiated accidents) and undercooling events (e.g.,

46 loss-of-coolant accidents).

47 48

3-17 Many of the data necessary for fuel qualification come from irradiated test specimens. However, 1

test specimens at the desired conditions may sometimes be unavailable. In such situations, it 2

may be possible to use lead test specimens to extend the burnup limits of a fuel type. In some 3

cases, direct examination of lead test specimens may provide a basis to support extending 4

applicability of an evaluation model to a new burnup range. In other cases, irradiated lead test 5

specimens may become the subject of subsequent tests under transient or accident conditions 6

to assess evaluation models applicable under such conditions.

7 8

Lead test specimen programs have traditionally allowed for the placement of a limited number of 9

test specimens in nonlimiting regions of the reactor core to maximize the safety margin.

10 However, an extended use of lead test specimens (e.g., relaxation of the number and/or 11 location of the test specimens) may be allowable if justified by a safety analysis that includes 12 margin to account for the uncertainty in the performance of fuel above its burnup limit. The use 13 of fuel above its qualified limit should be supported by sufficient monitoring to detect potential 14 failures. Methods are available, such as gas tagging (McCormick & Schenter, 1974) (Pollack, 15 Lewis, & Kelly, 2013), that can be used to identify the precise source of potential fuel failures.

16 Additionally, if lead test specimens are subjected to conditions beyond existing data ranges, a 17 licensing review may be necessary to ensure the appropriate level of safety before the extended 18 limits are applied to the fuel design. ED G2 can be satisfied by demonstrating that the test 19 envelope addresses the necessary performance envelope for the fuel design.

20 21 3.4.3 ED G3Data Measurement 22 An understanding of measurement accuracy is essential to establish confidence in the data 23 used to develop and assess evaluation models. This goal is decomposed, as shown in 24 Figure 3-16, into three supporting goals, which are discussed below.

25 26 27 Figure 3-16 Decomposition of ED G3, Data Measurement 28 29 3.4.3.1 ED G3.1Test Facility Quality Assurance 30 Experimental data should be collected under an appropriate quality assurance program.

31 Standards such as the American Society of Mechanical Engineers (ASME) Nuclear Quality 32 Assurance (NQA)-1 are available for test facility quality assurance. Provisions may also be 33 applied to existing data to make them compliant with quality assurance requirements (ANL, 34 2020). ED G3.1 can be satisfied by demonstrating that data were collected under an appropriate 35 quality assurance program or by otherwise justifying the use of existing data.

36

3-18 1

3.4.3.2 ED G3.2Measurement Techniques 2

Data should be collected using established or otherwise proven measurement techniques. The 3

use of novel or first-of-a-kind measurement techniques should be adequately justified. ED G3.2 4

can be satisfied by specifying the measurement techniques and justifying the use of any novel 5

or first-of-a-kind techniques.

6 7

3.4.3.3 ED G3.3Experimental Uncertainties 8

An error analysis should be performed to assess sources of bias and uncertainty in each 9

experiment. Measurement uncertainty should be quantified when possible, and its overall 10 impact on assessment data should be discussed. ED G3.3 can be satisfied by providing an 11 experimental error analysis.

12 13 3.4.4 ED G4Test Conditions 14 The test conditions should be representative of prototypical conditions. Test specimens used in 15 experiments should be representative of the proposed fuel design (i.e., the fuel design 16 submitted for safety review). This goal is decomposed, as shown in Figure 3-17, into two 17 supporting goals, which are discussed below.

18 19 20 Figure 3-17 Decomposition of ED G4, Test Specimens 21 22 3.4.4.1 ED G4.1Manufacturing of Test Specimens 23 Test specimens should be fabricated consistently with the manufacturing specification. (This 24 goal is associated closely with G1, Fuel Manufacturing Specification (Section 3.1), which 25 emphasized that fuel performance during normal operation and accident conditions can be 26 highly sensitive to the fuel fabrication process.) It may be possible to provide justification for any 27 acceptable differences in fabrication between the fuel and test specimens. Such justifications 28 will be addressed case by case. ED G4.1 can be satisfied by demonstrating that test specimens 29 are fabricated consistently with the fuel manufacturing specification.

30 31

3-19 3.4.4.2 ED G4.2Evaluation of Test Distortions 1

Test distortions should be evaluated. Test distortions arise from differences between the test 2

and the actual conditions under which the fuel is expected to perform (e.g., differences in test 3

dimensions, initial and boundary conditions). An example of an expected test distortion is the 4

geometry distortion typical of transient testing in a test reactor, as test reactors are generally too 5

small to accommodate full-size fuel designs. ED G4.2 can be satisfied by an analysis of test 6

distortions and justification for any identified distortions.

7 8

9

4-1 4

SUMMARY

AND CONCLUSIONS 1

Section 3 of this report presents a systematic evaluation and justification of the requirements for 2

qualifying nuclear fuel, and the table in Appendix A includes a concise list of the criteria 3

identified to support a determination that nuclear fuel is qualified for use. These criteria provide 4

a basis to support regulatory findings in the area of fuel qualification, as follows:

5 6

The regulation in 10 CFR 50.43(e)(1)(i), requiring that the performance of each safety 7

feature of the design has been demonstrated, is satisfied for the fuel by demonstrating 8

that the safety criteria (G2 of the FQAF, discussed in Section 3.2) can be satisfied, which 9

requires information to provide assurance that the fuel will perform as described in the 10 safety analysis.

11 12 The regulation in 10 CFR 50.43(e)(1)(iii) requires that sufficient data exist on the safety 13 features of the design to assess the analytical tools used for safety analyses over a 14 sufficient range of normal operating conditions, transient conditions, and specified 15 accident sequences, including equilibrium core conditions. This requirement can be 16 satisfied by (1) specifying the fuel performance envelope, which covers a sufficient range 17 of conditions (G2.1.1 of the FQAF), and (2) by demonstrating that assessed evaluation 18 models and empirical criteria are capable of evaluating the fuel performance over the 19 performance envelope (G2.1.2, G2.2.2, G2.2.3, G2.3.1, and G2.3.2(b) of the FQAF).

20 Sections 3.2.1.1, 3.2.2.1, 3.2.2.3, 3.2.2.4, 3.2.3.1, and 3.2.3.2.2 discuss these topics 21 further.

22 23 GDC 2 and ARDC 2 require that SSCs important to safety be designed to withstand the 24 effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, 25 tsunami, and seiches without loss of capability to perform their safety functions. G2.3 of 26 the FQAF (discussed in Section 3.2.3) partially addresses this requirement through 27 assurance of the ability to achieve and maintain safe shutdown.

28 29 GDC 10 and ARDC 10 require that SAFDLS or SARRDLs not be exceeded under any 30 conditions of normal operation, including the effects of AOOs. This requirement is 31 satisfied, in part, by demonstrating margin to design limits under conditions of normal 32 operation, including the effects of AOOs (G2.1 of the FQAF, discussed in Section 3.2.1).

33 34 GDC 27 and ARDC 26 require, in part, the ability to achieve and maintain safe shutdown 35 under postulated accident conditions. G2.3 of the FQAF (discussed in Section 3.2.3) 36 partially addresses this requirement through assurance of the ability to achieve and 37 maintain safe shutdown.

38 39 GDC 35 and ARDC 35 require an emergency core cooling system that provides 40 sufficient cooling under postulated accident conditions. They also require that fuel and 41 clad damage that could interfere with continued effective core cooling is prevented. G2.3 42 of the FQAF (discussed in Section 3.2.3) partially addresses these requirements through 43 assurance of the ability to achieve and maintain safe shutdown.

44 45 The regulations in 10 CFR 50.34(a)(1)((ii)(D), 10 CFR 52.47(a)(2)(iv), and 46 10 CFR 52.79(a)(1)(vi) require an evaluation of a postulated fission product release. This 47 requirement is partially addressed by demonstrating margin to radionuclide release limits 48 under accident conditions (G2.2 of the FQAF, discussed in Section 3.2.2).

49

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A-1 APPENDIX A LIST OF ALL GOALS Table A -1 List of Goals in Fuel Qualification Assessment Framework GOAL Fuel is qualified for use G1 Fuel is manufactured in accordance with a specification G1.1 Key dimensions and tolerances of fuel components are specified G1.2 Key constituents are specified with allowance for impurities G1.3 End state attributes for materials within fuel components are specified or otherwise justified G2 Margin to safety limits can be demonstrated G2.1 Margin to design limits can be demonstrated under conditions of normal operation and AOOs G2.1.1 Fuel performance envelope is defined G2.1.2 Evaluation model is available (see EM Assessment Framework)

G2.2 Margin to radionuclide release limits under accident conditions can be demonstrated G2.1.1 Fuel performance envelope is defined G2.2.1 Radionuclide retention requirements are specified G2.2.2 Criteria for barrier degradation and failure are suitably conservative (a)

Criteria are conservative (b)

Experimental data are appropriate (see ED Assessment Framework)

G2.2.3 Radionuclide retention and release from fuel matrix are modeled conservatively (a)

Model is conservative (b)

Experimental data are appropriate (see ED Assessment Framework)

G2.3 Ability to achieve and maintain safe shutdown is assured G2.3.1 Coolable geometry is ensured (a)

Criteria to ensure coolable geometry are specified (b)

Evaluation models are available (see EM Assessment Framework)

G2.3.2 Control element insertion can be demonstrated (a)

Criteria are provided to ensure that control element insertion path is not obstructed (b)

Evaluation model is available (see EM Assessment Framework)

A-2 Table A -2 List of Goals in Evaluation Model Assessment Framework GOAL Evaluation model is acceptable for use EM G1 Evaluation model contains the appropriate modeling capabilities EM G1.1 Evaluation model is capable of modeling the geometry of the fuel system EM G1.2 Evaluation model is capable of modeling the material properties of the fuel system EM G1.3 Evaluation model is capable of modeling the physics relevant to fuel performance EM G2 Evaluation model has been adequately assessed against experimental data EM G2.1 Data used for assessment are appropriate (see ED Assessment Framework)

EM G2.2 Evaluation model is demonstrably able to predict fuel failure and degradation mechanisms over the test envelope EM G2.2.1 Evaluation model error is quantified through assessment against experimental data EM G2.2.2 Evaluation model error is determined throughout the fuel performance envelope EM G2.2.3 Sparse data regions are justified EM G2.2.4 Evaluation model is restricted to use within its test envelope Table A -3 List of Goals in Experimental Data Assessment Framework GOAL Experimental data used for assessment are appropriate ED G1 Assessment data are independent of data used to develop/train the evaluation model ED G2 Data has been collected over a test envelope that covers the fuel performance envelope ED G3 Experimental data have been accurately measured ED G3.1 The test facility has an appropriate quality assurance program ED G3.2 Experimental data are collected using established measurement techniques ED G3.3 Experimental data account for sources of experimental uncertainty ED G4 Test specimens are representative of the fuel design ED G4.1 Test specimens are fabricated consistent with the fuel manufacturing specification ED G4.2 Distortions are justified and accounted for in the experimental data

NUREG-2246 Timothy J. Drzewiecki, Jeffrey S. Schmidt, Christopher Van Wert, and Paul Clifford Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Same as above Proposed advanced reactor designs use fuel designs and operating environments (e.g., neutron energy spectra, fuel temperatures, neighboring materials) that differ from the large experience base available for traditional light-water reactor fuel. The purpose of this report is to identify criteria that will be useful for advanced reactor designs through an assessment framework that would support regulatory findings associated with nuclear fuel qualification. The report begins by examining the regulatory basis and related guidance applicable to fuel qualification, noting that the role of nuclear fuel in the protection against the release of radioactivity for a nuclear facility depends heavily on the reactor design. The report considers the use of accelerated fuel qualification techniques and lead test specimen programs that may shorten the timeline for qualifying fuel for use in a nuclear reactor at the desired parameters (e.g., burnup). The assessment framework particularly emphasizes the identification of key fuel manufacturing parameters, the specification of a fuel performance envelope to inform testing requirements, the use of evaluation models in the fuel qualification process, and the assessment of the experimental data used to develop and validate evaluation models and empirical safety criteria.

Fuel qualification June 2021 Technical Fuel Qualification for Advanced Reactors

NUREG-2246 Fuel Qualification for Advanced Reactors June 2021