ML19242E192

Fuel Burnup and Enrichment Extension Preparation Strategy

ML19242E192
Person / Time
Issue date:	08/30/2019
From:	Office of Nuclear Reactor Regulation
To:
Proffitt J
References
Download: ML19242E192 (19)

Text

1 APPENDIX A: FUEL BURNUP AND ENRICHMENT EXTENSION 2 PREPARATION STRATEGY 3

4 Based on stakeholder interactions, the NRC staff is aware of industrys plans to request higher 5 fuel burnup limits along with the deployment of near-term ATF concepts. Additionally, the staff 6 expects that the extension of fuel burnup limits, and the economic drive to achieve those 7 burnups, will result in requests to increase fuel enrichments to greater than the current standard 8 of 5 weight percent U235. Therefore, the staff is proactively assessing the current knowledge 9 and experimental database associated with extending both burnup and enrichment for light 10 water reactor (LWR) fuels. This plan focuses on the strategy to prepare the NRC for review of 11 future licensing actions in which industry requests to go beyond current licensed limits with 12 burnups up to ~75 GWd/MTU rod-average and enrichments up to ~8 weight percent U235. Staff 13 will continue to engage with industry and the fuel vendors on these topics and adjust this 14 strategy as industry plans for higher burnup and increased enrichments evolve.

15 16 Overview of Preparatory Activities 17 18 As with other ATF activities related to advanced cladding and fuel materials, the staff has 19 grouped its burnup and enrichment preparatory activities into four tasks. The highlights of each 20 task are briefly described below; subsequent sections within this appendix describe these tasks 21 in greater detail.

22 23 Task 1: Regulatory Framework: In-Reactor Performance 24

• Participate in coordinated PIRT exercises on in-reactor performance of fuels with 25 increased enrichment under a wide array of conditions, performance -based metrics, and 26 analytical criteria to ensure acceptable performance.

27

• Perform a scoping study to (1) evaluate the applicability of existing regulations and 28 guidance for higher burnups and increased enrichment, (2) identify changes to, or the 29 need for, new regulations and guidance, and (3) identify any key policy issues.

30

• Identify consensus standards that need to be updated for higher burnups and increased 31 enrichment and participate in the update process where appropriate.

32

• Determine and clarify the regulatory criteria that need to be satisfied for higher burnup 33 fuels and fuels with increased enrichment and the regulatory options available to 34 applicants and vendors.

35

• If needed, resolve policy issues and initiate rulemaking and guidance development 36 activities.

37 1

38 Task 2: Regulatory Framework: Fuel Cycle, Transportation, and Storage 39 40

• 10 CFR Part 70, Domestic Licensing of Special Nuclear Materials is performance 41 based; therefore, the staff does not anticipate identification of gaps or deficiencies in 42 these regulations for the licensing of enrichment facilities to produce increased 43 enrichment material or fuel fabrication facilities to fabricate increased enrichment fuel.

44 The staff has previously licensed plants that produce uranium fuel enriched to the levels 45 addressed in this plan.

46

• 10 CFR Part 71, Packaging and Transportation of Radioactive Material; and 47 10 CFR Part 72, Licensing Requirements for the independent Storage of Spent Nuclear 48 Fuel and High-Level Radioactive Waste, and Reactor-Related Greater Than Class C 49 Waste, are largely performance based; therefore, the staff does not anticipate 50 identification of gaps or deficiencies in these regulations.

51

• Gaps in the review guidance may develop as the fuel cycle industry develops plans for 52 manufacturing, transporting, and storing higher burnup and increased enrichment fuel.

53 The NRC will monitor the fuel cycle industrys plans and identify and develop any 54 necessary regulatory guidance in a timely manner.

55

• The NRC is engaging industry to understand the details and timing of its plans to 56 produce uranium hexafluoride (UF6) that is enriched above the current limit (5 weight 57 percent U235) and fabricate increased enrichment LWR fuel.

58 59 Task 3: Probabilistic Risk-Assessment Activities 60 61 Like the impacts of ATF cladding and fuel matrix concepts, higher burnups and increased 62 enrichments manifest in a probabilistic risk assessment (PRA) via impacts on the plants 63 response to a postulated accident, in the form of changes to assumptions about sequence 64 timing, success criteria, and severe accident phenomenology. The PRA activities described in 65 the main body of this document (i.e., the activities originally crafted to address changes in plant 66 response to beyond-design-basis accidents associated with ATF) may adequately encompass 67 the PRA-related work needed to address the impacts of higher burnups and increased 68 enrichments. The specific timeframes and nature of the industry activities and associated NRC 69 deterministic technical basis development will dictate this. For instance, the pilot PRA model 70 work described in Section Error! Reference source not found. may be able to accommodate 71 the potential burnup and enrichment changes combined with the other cladding and fuel 72 response impacts associated with ATF. The degree of coverage provided by the pre-existing 73 planning will also depend on the degree to which burnup and enrichment changes impact other 74 agency uses of PRA information (such as in assessing environmental impacts associated with 75 postulated accidents). At this time, the staff is assessing whether higher burnup and increased 76 enrichments warrant any additional or different ATF-related PRA work, and the staff will adjust 77 its planning accordingly.

78 79 2

80 Task 4: Developing Independent Confirmatory Calculation Capabilities 81 82 Independent confirmatory calculations are one of the tools that the staff can use in its safety 83 review of topical reports and license amendment requests. Confirmatory calculations provide 84 the staff insight on the phenomenology and potential consequences of transient and accident 85 scenarios. In addition, sensitivity studies help to identify risk significant contributors to the 86 safety analyses and assist in focusing the staffs review.

87 88 The staffs approach to modifying and validating existing NRC codes and performing 89 confirmatory analysis for burnup and enrichment extension will be similar to the approach for 90 ATF described in Section Error! Reference source not found. in the ATF Project Plan. At this 91 time, the NRC plans to modify its codes that are developed to analyze fuel performance, 92 thermal hydraulics, neutronics, and severe accidents and source terms to support confirmatory 93 analysis of fuels with higher burnup and increased enrichment. See Section Error! Reference 94 source not found. and Appendix B of the ATF Project Plan for further details.

3

95 A.1 Task 1: Regulatory Framework: In-Reactor Performance 96 97 Higher fuel burnups and increased enrichments present new and unique technical issues that 98 current guidance, review plans, and regulatory criteria may not readily address. To prepare the 99 agency to conduct meaningful and timely licensing reviews of higher fuel burnup and/or 100 increased enrichment proposals, well-developed and vetted positions are needed on potential 101 policy issues that may arise during the review and licensing process. These positions must be 102 communicated to stakeholders clearly and early.

103 104 This task addresses the changes to the in-reactor regulatory framework that may be required to 105 support the implementation of higher fuel burnups and increased enrichments considering the 106 technical issues they present. Generally, the technical issues associated with higher fuel 107 burnups and increased enrichments respectively fall into two categories, fuel integrity (cladding 108 and/or fuel pellet) and nuclear criticality safety. ECCS performance embrittlement mechanisms 109 and fuel fragmentation, relocation, and dispersal (FFRD) are examples of fuel integrity technical 110 issues. Spent fuel pool criticality, and potential fast critical conditions during accident scenarios 111 are examples of the technical issues that fall under nuclear criticality safety. The regulatory 112 framework changes that may be necessary to address each technical issue are likely to be 113 different, and the staff anticipates that such changes will need to be made before either higher 114 fuel burnups or increased enrichments can be licensed for general use.

115 116 With regard to the regulations at Appendix A, General Design Criteria for Nuclear Power 117 Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, the 118 NRC staff has concluded that the general design criteria (GDC) discussed therein will not be 119 affected by higher burnups and increased enrichments. While higher burnups and increased 120 enrichments may impact the way compliance with regulatory requirements The degree to which 121 existing regulations and guidance need revision or new regulatory requirements and guidance 122 need to be established, depends on the level of departure from existing burnup and enrichment 123 limits. is demonstrated, the actual principal design and performance requirements provided by 124 the GDC remain applicable. Note that loading increased enrichment fuel designs in a specific 125 plant will ultimately need to meet relevant plant-specific criteria. This is especially important for 126 those reactors in the United States that were licensed before the issuance of the GDC (about 127 40 percent of the operating plants).

128 129 Beyond the GDC, higher burnups and the use of fuel with increased enrichment may affect the 130 regulations and guidance related to fuel design and performance and nuclear criticality safety 131 listed in Tables A.1 and A.2, below. The staff plans to map the technical issues and potential 132 failure issues to these requirements and guidance to determine the scope of changes that are 133 necessary.

4

134 135 Table A.1 Potentially Affected Regulations Regulation Affected by:

Title (10 CFR) Burnup Enrichment 20 Standards for Protection Against Radiation 50.34 Contents of Applications; Technical Information Acceptance Criteria for Emergency Core Cooling 50.46 Systems for Light-Water Nuclear Power Reactors 50.67 Accident Source Term 50.68 Criticality Accident Requirements 50, ECCS Evaluation Models Appendix K Environmental Protection Regulations for 51 Domestic Licensing and Related Regulatory Functions (specifically, Tables S-3 and S-4) 100 Reactor Site Criteria 136 137 5

138 Table A.2 Potentially Affected Guidance Guidance Affected by:

Title Document Burnup Enrichment Cladding Swelling and Rupture Models for NUREG-0630 LOCA Analysis Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:

NUREG-0800 LWR Edition (Section 4.2, Fuel System Design in particular for burnup)

Accident Source Terms for Light-Water NUREG-1465 Nuclear Power Plants Standard Review Plans for Environmental NUREG-1555 Reviews for Nuclear Power Plants:

Environmental Standard Review Plan Fuel Fragmentation, Relocation, and Dispersal NUREG-2121 During the Loss-of-Coolant Accident NUREG/CR-FRAPCON-3.5 7022 Vol. 1-2 NUREG/CR-FRAPTRAN 1.5 7023 Vol. 1-2 Material Property Correlations: Comparisons NUREG/CR-Between FRAPCON-3.5, FRAPTRAN 1.5, and 7024 MATPRO NUREG/CR- Cladding Behavior During Postulated Loss-of-7219 Coolant Accidents Alternative Radiological Source Terms for RG 1.183 Evaluating Design Basis Accidents at Nuclear Power Reactors Methods and Assumptions for Evaluating Radiological Consequences of Design Basis RG 1.195 Accidents at Light-Water Nuclear Power Reactors RG 1.203 Transient and Accident Analysis Methods Pressurized Water Reactor Control Rod DG 1327 Ejection and Boiling Water Reactor Control Rod Drop Accidents 139 140 A.1.1 Additional Considerations 141 142 Aspects of higher burnup and increased enrichment fuel designs or the implementation strategy 143 could expand the scope, level of complexity, and schedule of the staffs review. Specifically, an 144 increase in fuel burnup and U235 enrichment could impact the scope of the staffs environmental 145 review and have implications for the license renewal generic environmental impact statement 146 (GEIS) associated with a plants licensing basis.

147 6

148 Higher fuel burnups and increased enrichments may affect the NRCs generic environmental 149 findings as documented in the GEIS. Licensees seeking to adopt higher fuel burnups and 150 increased enrichments beyond the current licensed limits will need to submit a license 151 amendment request, and this submittal will need to provide sufficient information as to the 152 potential environmental impacts of the request to facilitate the staffs review. Such information 153 should consider justification for the continued applicability of the existing generic basis as 154 presented in the GEIS. The NRC staff will need to undertake a review of the justification, and 155 this could be a source of additional complexity. To minimize this additional complexity, the staff 156 may need to reassess the current basis for Table S-3 and Table S-4 of 10 CFR Part 51 as 157 documented in the GEIS, how it may be impacted by higher burnup fuels and increased 158 enrichment, and the changes that may be necessary to generate an updated technical basis 159 and GEIS. The necessity of this effort will become clearer as NRC continues engagement with 160 industry and the fuel vendors.

161 162 A.1.2 Lead Test Assemblies 163 164 Lead Test Assembly (LTA) programs provide pool-side post-irradiation examination data 165 collection, irradiated material for subsequent hot-cell examination and research, and 166 demonstration of in-reactor performance. This characterization of irradiated material properties 167 and performance is essential for qualifying analytical codes and methods and developing the 168 safety design bases for higher burnup fuels and fuels with increased enrichments.

169 170 The NRC has recently published a letter to the Nuclear Energy Institute (ADAMS Accession 171 No. ML18100A045) that documents the agencys position concerning criteria for the insertion of 172 LTAs under 10 CFR 50.59 without additional NRC review and approval. LTA programs for 173 higher burnup and increased enrichment may fall outside the guidance in the letter and require 174 LARs, depending on the scope of the LTA campaign and the licensing basis of the reactor.

175 176 A.1.3 Licensing Strategy 177 178 The staff expect industry to take an incremental approach in moving to higher burnup and 179 increased enrichment. Therefore, the NRC staff envisions near-term and longer-term strategies 180 for moving forward with the licensing of higher burnup fuels and fuels with increased 181 enrichments. In the near-term, licensees will need to request exemptions to existing regulations 182 on a licensee-specific basis for the use of these technologies and demonstrate compliance with 183 safety requirements along with the exemption criteria. Should widespread adoption of these 184 technologies become apparent, the NRC will utilize the longer-term strategy of rulemaking to 185 update existing regulations to facilitate a more predictable licensing process.

186 7

187 A.1.4 Deliverables 188 189 Table A.3 Anticipated In-Reactor Deliverables*

Title Due Date Map of technical issues and failure mechanisms to regulations, 6-9 months from completion and guidance documents. of the PIRT exercise 18-24 months from Develop or revise guidance to address any identified completion of the PIRT necessary changes.

exercise 24-36 months from Develop rulemaking to address any identified necessary identification of required changes.

change 190 191

• The technical lead is the NRR Division of Safety Systems, Nuclear Performance and Code Review Branch 8

192 A.2 Task 2: Regulatory Framework: Fuel Cycle, Transportation, and Storage 193 194 The regulatory activities on ATF of higher burnup/increased enrichment present different 195 challenges at the various stages of the front and back end of the fuel cycle. The NRC 196 recognizes that these challenges have different timelines, with increased enrichment being the 197 near-term technical issue that must be addressed for successful deployment of ATF.

198 199 For the front end of the fuel cycle, which includes fuel assembly fabrication and transportation of 200 feed material and fresh fuel assemblies, increased enrichment may present additional technical 201 and regulatory issues; however current guidance, review plans, and regulatory criteria are 202 adequate to address these issues. To prepare the agency to conduct near-term timely licensing 203 and certification reviews of increased enrichment levels for ATF, discussion of 204 licensing/certification strategies and approaches between applicants and NRC will be 205 undertaken to address any potential technical or policy issues that may arise. Any issues the 206 NRC identifies will be communicated to stakeholders promptly.

207 208 For the back end of the fuel cycle, which includes transportation and storage of spent fuel at 209 higher burnups/increased enrichments, the NRC will continue to monitor industrys initiatives 210 and licensing actions for reactor operation, and assess whether revisions to current guidance, 211 review plans and regulatory criteria may be warranted. The NRC recognizes that licensing and 212 certification actions related to the transportation and storage of such spent fuel will not occur in 213 the near term. The NRC will engage with industry as plans on the back end of the fuel cycle are 214 developed and will update this plan to reflect those actions. Therefore, the rest of the 215 discussion in the plan will focus on near-term issues related to increased enrichment of ATF.

216 217 This task contemplates the changes to the regulatory framework that may be required to 218 support the implementation of increased enrichment for ATF considering the technical or 219 regulatory issues they present. When considering the safe transportation of material for the 220 front end of the fuel cycle, the notable technical issue associated with increased enrichments 221 pertains to nuclear criticality safety for UF6 transportation and fresh fuel assemblies. Fuel 222 assemblies (both fresh and irradiated) that rely on the fuel assembly structural performance to 223 remain intact after evaluation of accident conditions and the criticality evaluation of a single UF6 224 package without using the exception in 10 CFR 71.55(g) are examples of the technical issues 225 that fall under fuel integrity and nuclear criticality safety, respectively. Benchmarking criticality 226 analyses for increased enrichment fuel and burnup credit analyses for spent fuel storage and 227 transport are also examples of the technical issues that fall under nuclear criticality safety. The 228 regulatory framework changes that may be necessary to address each technical issue are likely 229 to be different, however the staff does not anticipate that such changes will need to be made 230 before either higher fuel burnup or increased enrichment can be licensed/certified for general 231 use in reactor.

232 233 9

234 A.2.1 Regulatory Infrastructure Analysis 235 236 The regulatory requirements in 10 CFR Part 70, 10 CFR Part 71 and 10 CFR Part 72 govern the 237 use of radioactive material for fuel enrichment and fabrication facilities, transportation, and spent 238 fuel storage. For increased enrichment in UF6 feed material and fresh fuel assemblies, changes 239 to the regulations are not necessary to accommodate industry plans; however, licensing and 240 certification challenges may exist. The criticality regulations in 10 CFR 71.55(g) grant an 241 exception from the consideration of moderation intrusion for the transportation of UF6 enriched 242 to 5 weight percent or less. Transportation of UF6 enriched to greater than 5 weight percent will 243 require the design and certification of new packages, the modification of currently approved 244 packages, or an exemption from the regulations that require evaluation of a single package with 245 optimum moderation for enrichments greater than 5 weight percent.

246 247 Table A.4 identifies the current guidance documents for the review of fuel facility licensing, 248 transportation package certification, and spent fuel storage licensing and certification and 249 identifies whether the guidance document is affected by industry plans to use higher enriched, 250 higher burnup fuel.

251 252 Table A.4 NRC Fuel Cycle Review Guidance Review Guidance Title Affected By Document Burnup Enrichment Standard Review Plan for Transportation NUREG-16091 Packages for Radioactive Material Standard Review Plan for Transportation NUREG-16171 Packages for Spent Nuclear Fuel Standard Review Plan for Fuel Cycle NUREG-1520 Facilities License Applications Managing Aging Processes In Storage NUREG-2214 (MAPS) Report Standard Review Plan for Spent Fuel Dry NUREG-2215 Storage Systems and Facilities Dry Storage and Transportation of High NUREG-2224 Burnup Spent Nuclear Fuel Spent Fuel Storage https://www.nrc.gov/reading-rm/doc-collecti and Transportation ons/isg/spent-fuel.html3 Interim staff guidance2 253 1

Note that NUREG-1607 and NUREG-1617 are being combined into a single standard review plan, NUREG-2216, Standard Review Plan for Transportation Package Approval, which is scheduled to be completed in the summer of 2020.

2 After completion of NUREG-2215 and NUREG-2216, all existing Interim Staff Guidance documents issued by the Division of Spent Fuel Management will be retired.

3 In particular, SFST-ISG-8, Revision 3, Burnup Credit in the Criticality Safety Analysis of PWR Spent Fuel in Transport and Storage Casks, is affected by both higher burnup and increased enrichment.

10

254 These review guidance documents draw on industry experience in the fabrication, 255 transportation, and storage of Zrclad- UO2 fuel with up to 5 weight percent enrichment and 256 burnup up to approximately 62 GWd/MTU rod average. The NRC may need to supplement its 257 guidance to address safety related- issues associated with increased enrichments and higher 258 burnups. When staff believes that supplemental information or guidance would facilitate the 259 preparation and review of applications involving the enrichment, fabrication, transportation, and 260 storage of higher burnup and/or increased enrichment fuel, it will discuss this with stakeholders 261 and take action where practical.

262 263 A.2.2 Facility, Transportation, and Storage Reviews 264 265 The regulatory reviews to support the development and deployment of increased enrichment 266 fuel will occur in several fuel cycle areas over the near term to support production (enrichment 267 and fuel fabrication) and transportation of UF6 feed material and fresh fuel assemblies. The 268 sections below discuss these various reviews.

269 270 A.2.2.1 Uranium Enrichment and Fuel Fabrication Facility Reviews 271 272 The uranium enrichment facilities that produce enriched uranium as well as fabrication 273 operations that would produce conventional fuel (e.g., Zr-clad UO2) with increased enrichment 274 will conduct operations that are similar to currently licensed operations. These licensees will 275 have to submit amendments to produce or use uranium with increased enrichment. Fuel 276 fabrication operations that use new processes for producing a different type of fuel material 277 (e.g., uranium alloy or U3Si2) are expected to submit amendments to address both increased 278 enrichment as well as the new processes.

279 280 The staff is currently engaged with licensees of fuel cycle facilities to understand the status of 281 their plans and the anticipated timing of their license amendment submittals.

282 283 A.2.2.2 Unirradiated Fuel Transportation Package and Storage Cask Reviews 284 285 As industry prepares for the batch loading of higher burnup and increased enrichment fuel, the 286 staff expects to receive requests for the approval of transportation packages that allow large-287 scale (i.e., batch) shipment of uranium feed material (currently UF6) and unirradiated fuel 288 assemblies. The staff will review these requests against the requirements of 10 CFR Part 71 289 and will use NUREG-1609 and pertinent interim staff guidance for the safety reviews. The NRC 290 staff will support PIRT efforts that focus on the identification and evaluation of material 291 properties used in the safety analyses of transportation packages with higher burnup and 292 increased enrichment. These PIRT efforts are expected to help the staff develop additional 293 regulatory guidance for transportation of fuel with increased enrichment, if required.

294 295 The staff is currently engaged with fuel cycle facility certificate holders to understand the status 296 of their plans and the anticipated timing of their certificate amendment submittals.

297 11

298 A.2.2.3 Irradiated Fuel Transportation Package and Spent Fuel Storage Reviews 299 300 The back end of the fuel cyclespent fuel storage and transportpresents some challenges 301 that are similar to the front end. For example, benchmarking criticality safety is still an issue for 302 the back end for enrichments between 5 and 8 weight percent, but additional challenges may 303 exist depending on the licensing/certification strategy. Other areas where challenges may exist 304 include, performance of the cladding material during vacuum drying, aging while in dry cask 305 storage, fatigue data for transportation, and benchmarking the isotopic depletion analyses for 306 use in the shielding analyses for higher burnup fuels and for use in burnup credit criticality 307 analyses.

308 309 The staff is currently engaged with fuel cycle facility certificate holders to understand the status 310 of their plans and the anticipated timing of their certificate amendment submittals.

311 312 A.2.2.4 Potential Challenges 313 314 NRC staff has identified technical challenges for transportation of unirradiated fuel and spent 315 fuel with higher burnup and increased enrichment.

316 317 A.2.2.4.1 Challenges for Transportation of Unirradiated Fuel 318 319 In addition to challenges for approval of transport of UF6 at increased enrichments (greater than 320 5 weight percent), it should be noted that American National Standards Institute (ANSI) N14.1, 321 Nuclear Materials Uranium Hexafluoride - Packagings For Transport, only applies to 322 enrichments up to 5 weight percent 235U for the 30B and 30C cylinders. DOT regulations in 49 323 CFR 173.420 state that UF6 packagings (whether fissile, fissile excepted, or non-fissile) must be 324 designed, fabricated, inspected, tested and marked in accordance with American National 325 Standard N14.1 that was in effect at the time the packaging was manufactured. In addition to 326 an NRC approval for shipment in a packaging using a 30B or 30C cylinder, a special permit 327 from DOT will be needed.

328 329 Benchmarking criticality analyses for fissile material enriched to greater than 5 weight percent 330 235U presents a challenge due to the limited number of critical experiments in that range.

331 Applicants for package approval could overcome this challenge by performing:

332 333

• new critical experiments to validate criticality calculations for 5-8 wt% enriched uranium, 334

• relying on sensitivity/uncertainty analysis methods to develop new critical experiments, 335

• relying on sensitivity/uncertainty analysis methods to determine that existing 336 experiments are applicable to 5-8 wt% enriched uranium, or 337

• reduce the allowable maximum k-effective to account for uncertainties in criticality code 338 performance due to lack of applicable critical experiments for benchmarking.

339 340 12

341 A.2.2.4.2 Challenges for Transportation and Storage of Spent Fuel 342 343 In addition to the benchmarking challenge listed above, other challenges exist for the storage 344 and transportation of spent fuel. Evaluation of material performance during vacuum drying, 345 aging while in storage, and cladding material properties are needed to evaluate structural 346 performance during normal storage, transport, and accident conditions.

347 348 Aging effects during long-term, dry cask storage include evaluation of impacts of potential 349 operable age-related phenomena on cladding performance. Those mechanisms described in 350 NUREG-2214 that may be affected by higher burnup and increased enrichment include creep, 351 hydrogen absorption, oxidation, delayed hydride cracking, and irradiation hardening. In 352 addition, the impacts of both potential higher end-of-life rod internal pressures on the credibility 353 of age-related phenomena, and the increased pellet swelling on the mechanical performance of 354 the cladding should be evaluated. There is also a need for experimental confirmation to 355 determine whether unknown age-related phenomena impact the spent fuel during storage and 356 transport after storage.

357 358 In addition to cladding material properties discussed above for unirradiated fuel, fatigue 359 performance data will be needed to evaluate vibration normally incident to transport as required 360 in 10 CFR 71.71(c)(5).

361 362 A transportation package or storage cask that is evaluated containing spent fuel will have the 363 same benchmarking concerns listed above for unirradiated material. If a package/cask is 364 evaluated for burnup credit, instead of fresh fuel, the isotopic depletion analyses will need to be 365 validated for the increased enrichment and burnup levels. In addition to validating the criticality 366 analysis, the accuracy of depletion calculations to calculate the source term for the shielding 367 analyses should be evaluated for burnups greater than 62 GWd/MTU rod average.

368 369 However, these challenges would not preclude an effective and efficient staff review.

370 371 A.2.2.5 Anticipated Regulatory Actions 372 373 Near -term regulatory actions consist of the reviews of fuel cycle facilities license amendments.

374 There are other regulatory actions needed to support increased enrichment and higher burnup; 375 however, only one fuel cycle facility have shared plans to submit a license amendment. Other 376 expected regulatory actions will be identified in future revisions of the plan after industry plans 377 become clearer.

378 379 13

380 A.3 Task 3: Probabilistic Risk Assessment Activities 381 382 The NRC uses PRAs to estimate risk: to investigate what can go wrong, how likely it is, and 383 what the consequences could be. The results of PRAs provide the NRC with insights into the 384 strengths and weaknesses of the design and operation of a nuclear power plant. PRAs cover a 385 wide range of NRC regulatory activities, including many risk-informed licensing and oversight 386 activities (e.g., risk-informed technical specification initiatives, the significance determination 387 process portion of the Reactor Oversight Process). These activities make use of both 388 plant-specific licensee PRA models and plant-specific NRC PRA models. The NRC uses the 389 former models predominantly for licensing and operational activities and the latter models 390 predominantly for oversight activities. A key tenet of risk-informed decision-making is that these 391 models reflect the as-designed, as-operated plant. For this reason, these models should be 392 updated to reflect significant plant modifications. The introduction into the reactor core of fuels 393 intended for higher burnups and fuels with increased enrichments may affect these models, 394 particularly once the reactor core composition significantly influences the plants response to a 395 postulated accident (e.g., higher initial decay heat from increased unrainum-235 enrichment).

396 397 Developing capabilities to support risk-informed regulatory activities following the 398 implementation of higher fuel burnups and increased enrichments could require significant NRC 399 resource. Information about the industrys intended approach is needed to create a meaningful 400 plan. Early NRC interactions with the industry and vendors regarding higher burnup and 401 increased enrichment activities, such as fuel technology update meetings and early 402 preapplication meetings, will be used to encourage an approach that is consistent with 403 regulatory requirements and staff guidance. Just as with the ATF project plan, this project plan 404 recognizes that the staffs PRA-related preparatory work involves two separate, but closely 405 related, aspects:

406 407 (1) The staff needs to prepare for, and review, PRA-related information submitted as part of 408 the licensing process for the batch loading of fuels with increased enrichments and 409 higher burnups as well as the incorporation of these technologies into the licensing 410 basis.

411 412 (2) The staff needs to develop PRA-related capabilities to do the following effectively:

413 414

• Review risk-informed licensing applications and ensure that applicants are using 415 acceptable PRA models once higher fuel burnups and increased enrichments are 416 implemented.

417

• Perform risk-informed oversight evaluations (e.g., significance determination 418 process) once higher fuel burnups and increased enrichments are implemented.

419 420 Item 1 is highly dependent on the approach taken by each vendor or licensee, or both, in its 421 licensing application, while item 2 is somewhat independent of the licensing approach.

422 Therefore, this project plan currently focuses more attention on item 2.

423 14

424 In the near-term, increases in fuel burnup and enrichment limits are expected to be only 425 marginally greater than current limits, and this may have only a limited or no impact on PRA 426 modeling. However, in the long term, increases in fuel burnup and enrichment limits are 427 expected to be appreciably greater than current limits, and this may have a more significant 428 impact on PRA modeling.

429 430 PRA activities for higher burnups and increased enrichments will be analogous to the activities 431 for ATF described in Section Error! Reference source not found. of this document. In 432 particular, NRC staff must ensure that licensees PRAs continue to use acceptable models and 433 assumptions as part of the implementation of higher burnup fuels and fuels with increased 434 enrichments and update the NRCs models (as necessary) to reflect any plant modifications 435 made to accommodate these new technologies. Also analogous to the activities for ATF, it is 436 envisioned that much of the analytical investigation needed to assess PRA-related impacts and 437 support PRA-related changes in the agencys SPAR models due to higher burnups and 438 increased enrichments can use the independent confirmatory calculational capabilities currently 439 being developed by the NRC. These capabilities are discussed in Section A.4 of this project 440 plan. See Section Error! Reference source not found. of this document for further information 441 on the analogous PRA activities NRC will take in response to higher burnups and increased 442 enrichments.

443 444 Engagement on PRA-related topics both among the NRC staff and with external stakeholders is 445 important at all stages. Effective interaction will foster a common understanding of the 446 acceptability of PRA methods used to model plant modifications and the impact that will 447 ultimately be realized when these modifications are integrated into PRAs and risk-informed 448 processes. Effective interaction can also ensure that information required to develop PRA 449 modeling assumptions related to plant modifications is properly coordinated with the 450 deterministic review. In this case, relevance of PRAs has been identified early in the process, 451 and time is available to address the PRA-related needs in a thoughtful and symbiotic manner.

452 453 For the purpose of identifying PRA-related milestones, the following key assumptions are 454 necessary:

455 456

• The timing of PRA-related efforts will be cross-coordinated with those of the previously 457 identified partner areas (e.g., severe accident analysis) to allow the leveraging of 458 deterministic work to make the PRA-related efforts efficient.

459

• Near-term TR/LAR reviews will start in 2020, with long-term licensing reviews occurring 460 no earlier than 2023.

461

• This plan does not account for rulemaking initiatives that might be requested to facilitate 462 rapid adoption of increased enrichments (e.g., modifications to 10 CFR 50.68, Criticality 463 Accident Requirements).

464 465 The PRA-related milestones for higher burnups and increased enrichment activities are listed 466 below in Table A.6. It should be noted that it may be feasible to merge the work outlined in 15

467 Table A.6 with the existing ATF PRA-related milestones found in Table 9.1, depending on the 468 nature and timing of the higher burnup and increased enrichment activities relative to that of the 469 ATF activities.

470 471 Table A.6 PRA Activities for Higher Burnups and Increased EnrichmentsMilestones Milestone Input Needed Lead Needed By Time/

Duration Participate in internal and external discussions and knowledge development related 1 to higher burnups and N/A Ongoing N/A increased enrichments (e.g., internal working group meetings, public meetings)

Complete licensing reviews, including potential TRs or More information regarding industry guidance, related to the 2 the specific licensing TBD TBD risk-informed aspects of approach licensing higher burnup fuels and increased enrichments Complete a SPAR pilot of a Deterministic knowledge 1 year before the BWR and PWR subject plant for base being developed first long-term core higher burnups and increased under other tasks load4 of higher 3 enrichments to assess 6 months (e.g., independent burnup fuels and CDF/LERF impacts, gain risk confirmatory code fuels with increased insights, and identify potential analysis) enrichment improvements to guidance Update guidance (as necessary) to support licensing and oversight functions for Completion of the items Before the core 4 plants making modifications (if 1 year above load any) to accommodate higher burnups and increased enrichments Update agency PRA models to As needed to reflect changes to the as-built, Details of the plant support the 5 1 year5 as-operated plant (if any) for modifications agencys risk relevant plants/models evaluations 472 4 Here, core load means the replacement of a large proportion (e.g., 50 percent or more) of the core.

5 This would occur after approval of the associated licensing action.

16

473 Table A.7 PRA Activities for Higher Burnups and Increased EnrichmentsDeliverables*

Title Lead Time Safety Evaluation contributions for TRs and LARs related to TBD using fuels with higher burnups and increased enrichments 1 year before the first long-term Report that documents results and recommendations from a core load of higher burnup fuels SPAR pilot study and fuels with increased enrichments Updated guidance (e.g., risk-assessment standardization project Varies depending on the guidance changes) to support licensing and oversight functions documents that require for plants making modifications (if any) to accommodate higher modifications burnups and increased enrichments Updated agency PRA models to reflect changes to the as-built, As needed to support the agencys as-operated plant (if any) for relevant plants/models risk evaluations 474 475

• The technical lead is the NRR Division of Risk Analysis, Probabilistic Risk Assessment Oversight Branch.

476 17

477 A.4 Task 4: Developing Independent Confirmatory Calculation Capabilities 478 479 Independent confirmatory calculations are one of the tools that the staff can use in its safety 480 review of topical reports and license amendment requests. Confirmatory calculations provide 481 the staff insight on the phenomenology and potential consequences of transient and accident 482 scenarios. In addition, sensitivity studies help to identify risk significant contributors to the 483 safety analyses and assist in focusing the staffs review.

484 485 RG 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants 486 (LWR Edition), identifies the standard format and content of safety analysis reports for nuclear 487 power plants, and NUREG-0800, Standard Review Plan for the Safety Analysis Reports for 488 Nuclear Power Plants: LWR Edition, (SRP) identifies the criteria that the staff should use to 489 review licensee safety analyses. The NRC plans to continue to develop independent 490 confirmatory analysis tools that support robust safety evaluations and provide insights into 491 safety significant factors for burnup and enrichment extension. Vendor codes used to support 492 analysis of fuel above existing burnup and enrichment limits will likely be based on smaller data 493 sets than the data sets available for Zr-UO2 fuel below existing limits. This will result in greater 494 uncertainty in the results of the safety analyses and the margins to the specified acceptable fuel 495 design limits. For these reasons, confirmatory calculation capabilities will be critical for 496 generating confidence in the safety assessment of burnup and enrichment extension against all 497 applicable regulatory requirements (see Section A.2 for more details). A confirmatory code can 498 be used to independently quantify the impact of modeling uncertainties and support more 499 efficient reviews with the potential for fewer requests for additional information. Finally, the 500 experience and insights gained by developing an in-house code can be leveraged in reviews of 501 externally developed models and methods, thus making reviews more efficient and effective.

502 503 The staff identified four technical disciplines needing calculation capability development to 504 support TR/LAR safety reviews for burnup and enrichment extension: (1) fuel performance, 505 (2) thermal hydraulics, (3) neutronics, and (4) severe accidents. The NRC has developed a 506 suite of codes to analyze these disciplines, and they have been used successfully to support 507 regulatory decision-making. Further development of these codes is appropriate to ensure that 508 the NRC has the capability to analyze Zr-UO2 fuel above existing regulatory burnup and 509 enrichment limits. Having tools that the staff can use to analyze fuel with higher burnup and 510 increased enrichment will be particularly important because applicants will use computational 511 tools to demonstrate that they have met fuel safety acceptance criteria and because, in some 512 cases, the properties and models for fuel at higher burnup and increased enrichment within the 513 computational tools will be based on limited experimental data.

514 515 Code development activities for higher burnup and increased enrichment will be integrated and 516 sequenced, as appropriate, with activities for ATF described in Section Error! Reference 517 source not found. of the ATF Project Plan. In particular, the NRC will participate in PIRT 518 exercises for increased enrichment, perform scoping studies to identify code architecture and 519 model updates needed, modify the codes based on outcomes of the increased enrichment PIRT 520 and scoping studies, and perform assessments against available experimental data. Section 18

521 Error! Reference source not found. of the ATF Project Plan describes the approach NRC will 522 take to update its codes to support confirmatory analysis for higher burnup and increased 523 enrichment limits.

19