ML25044A331
| ML25044A331 | |
| Person / Time | |
|---|---|
| Issue date: | 02/22/2025 |
| From: | Walter Kirchner Advisory Committee on Reactor Safeguards |
| To: | David Wright NRC/Chairman |
| Wang W | |
| Shared Package | |
| ML25056A105 | List: |
| References | |
| Download: ML25044A331 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 Honorable David Wright Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
RULEMAKING ON INCREASED ENRICHMENT OF CONVENTIONAL AND ACCIDENT TOLERANT FUEL DESIGNS FOR LIGHT-WATER REACTORS
Dear Chairman Wright:
During the 722nd meeting of the Advisory Committee on Reactor Safeguards, February 5 through 7, 2025, we completed our review of the draft rulemaking on Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors and its associated draft regulatory guides. Our Regulatory Rulemaking, Policies, and Practices Subcommittee reviewed this matter on October 18, 2023, December 17 and 18, 2024, and January 16 and 17, 2025. During these meetings, we had the benefit of discussions with Nuclear Regulatory Commission (NRC) staff, representatives from industry, and the Electric Power Research Institute (EPRI). We also had the benefit of the referenced documents.
CONCLUSIONS AND RECOMMENDATIONS 1.
The draft proposed rule represents a significant achievement in the efforts to safely regulate the use of increased enrichment fuels and accident tolerant fuels (ATF) in light water reactors (LWRs). Consistent with the NRCs new mission statement, the proposed rule enables higher burnups up to 80 gigawatt days per metric ton of uranium (GWd/MTU) and is expected to support advances such as 24-month fuel cycles and potential power uprates. The draft proposed rule is performance based, and risk-informed as directed by the Commission.
2.
With the introduction of the concept of transition break size (TBS), the proposed rule presents a major change from regulatory precedent by providing a solid technical rationale for moving away from a double-ended guillotine break in the primary piping as the presumed defining design basis accident for LWRs. We support staff plans to work with stakeholders and examine broader impacts of this change.
3.
We offer comments and suggestions in the body of this letter, itemized in the Summary, to be considered in the development of the draft final rule package. These comments and suggestions are in four technical areas: TBS; loss-of-coolant accident (LOCA) fuel February 22, 2025 D. Wright performance analysis; fuel fragmentation, relocation, and dispersal (FFRD); and updated source terms and control room dose.
4.
The draft proposed rule should be issued for public comment.
BACKGROUND The U.S. Nuclear Regulatory Commission (NRC) is proposing amendments to its regulations to support the use of LWR fuels enriched to greater than 5 and up to 20 weight percent uranium-235 while continuing to provide reasonable assurance of adequate protection of public health and safety. These changes aim to enable the licensing and deployment of conventional and accident-tolerant fuel designs with increased enrichment, offering benefits such as improved fuel and reactor performance, enhanced safety margins, and longer fuel cycles.
The Commission staff requirements memorandum (SRM-SECY-21-0109) provided the following direction:
The rule should only apply to high-assay low-enriched uranium fuel, both for nonproliferation and safeguards reasons.
FFRD issues relevant to fuels of higher enrichment and burnup levels should be appropriately addressed and analyzed in the regulatory basis.
The staff should take a risk-informed approach when developing the rule and the associated regulatory basis and guidance.
The staff should work with stakeholders to identify and develop necessary regulatory guidance and technical bases to support effective and efficient licensing of increased enrichment applications.
In light of the above direction, the proposed rulemaking responds to technological advancements in the nuclear industry and seeks to establish efficient and effective licensing pathways while considering safety and security.
The rule recategorizes a large break LOCA as a beyond design basis event. This was one of five alternatives presented in the associated regulatory basis document. The staff also proposed revising or developing the following associated draft regulatory guides (DGs):
Transition Break Size o
DG-1428, (RG 1.258, Revision 0), Plant-Specific Applicability of Transition Break Size LOCA Fuel Performance o
DG-1261, Revision 1 (RG 1.222, Revision 0), Measuring Breakaway Oxidation Behavior o
DG-1262, Revision 1 (RG 1.223, Revision 0), Determining Post-Quench Ductility o
DG-1263, Revision 1 (RG 1.224, Revision 0), Establishing Analytical Limits for Zirconium-Based Alloy Cladding D. Wright Fuel Fragmentation, Relocation, and Dispersal o
DG-1426 (RG 1.225, Revision 0) An Approach for Risk-Informed Evaluation Process Supporting Alternative Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Reactors o
DG-1434 (RG 1.259, Revision 0), Addressing the Consequences of Fuel Dispersal in LWR Loss of Coolant Accidents Updated Source Terms and Control Room Design Criteria Draft Guide (10 CFR 50.67 and GDC-19) o DG-1425 (RG 1.183, Revision 2), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" The core of the rulemaking is based on previous draft rulemakings for emergency core cooling system performance during a LOCA. The previous Title 10 of the Code of Federal Regulations (CFR) 50.46a draft final rule established a TBS above which the LOCA is defined as beyond design basis, and the previous 10 CFR 50.46c draft final rule addressed fuel cladding performance during a design basis LOCA. The currently proposed voluntary rule, 10 CFR 50.46a, utilizes some of the TBS and previous draft 10 CFR 50.46c efforts.
In this proposed rule, analysis methods are permitted for LOCAs greater than the TBS that are consistent with current approaches used for analysis of anticipated transients without scram and station blackout postulated events. The proposed rule also requires that the LOCA analysis include the effects of FFRD in response to recent research results.
The draft proposed rule also establishes a graded risk-informed and performance-based control room dose-based acceptance criterion in accident analysis.
Finally, the Committee had no substantive comments on other conforming changes to regulation related to criticality accident and transportation packaging requirements.
DISCUSSION AND RECOMMENDATIONS The following sections discuss four technical areas and our recommendations therein:
transition break size; LOCA fuel performance analysis; fuel fragmentation, relocation, and dispersal; and updated source terms and control room dose. Broader impacts of the rule are also discussed.
Transition Break Size A key element of the proposed rulemaking is the application of a TBS, defining a flow area-based break threshold for primary loop piping (PLP), above which the pipe break probability is considered to be extremely low. The original TBS technical basis reflected in NUREG-1829 and NUREG-1903 (both circa 2008), was supported by the best available data and methodologies at the time. Additional operating experience and improved materials degradation knowledge in the intervening years is well-represented in the NRC staff White Paper on Continued Applicability of NUREG-1829. Likewise, the parallel White Paper on Continued Applicability of NUREG-1903 represents a similarly thorough treatment of recent D. Wright seismic-related experience and research. Furthermore, recent analytical results using the joint EPRI/NRC developed fracture mechanics-based code, xLPR, support the updated analysis.
Several NRC and EPRI evaluation activities that are either directly or indirectly relevant to the TBS analysis, broadly support and lend substantial engineering credibility to the original elicitation-based technical basis.
The TBS represents a reasonable and defensible demarcation line for the change in analytical treatment defined in the draft rulemaking for PLP locations above the TBS. The DG-1428, Plant-Specific Applicability of the Transition Break Size, provides practical guidance for an applicant to assess applicability of the TBS technical basis. Multiple assessment paths are described that may be used to leverage existing plant programs and prior licensing approvals to the extent possible to lower the plant implementation burden. The ability to leverage existing programs may differ between the pressurized water reactor (PWR) and boiling water reactor (BWR) fleets. For example, the PWR fleet can leverage leak before break (LBB) analyses in their applicability assessment, whereas the BWR fleet has not been approved for LBB.
The proposed rule also imposes requirements for a staff-approved risk-informed performance monitoring program for an inspection sample set representing the similar metal welds above the TBS within the PLP. The intent is to support the plant-specific applicability assessment, verify analysis conclusions over time, and seek to identify any novel (unexpected or new) degradation mechanisms. While we acknowledge the importance of performance monitoring as a key element of a robust degradation management program, the challenge is to devise a scheme that optimizes the information gained relative to the required levels of effort and occupational radiation dose. The draft proposed rule text has recently been revised from a more prescriptive inspection requirement to an NRC approved, risk-informed sample of the similar metal piping circumferential welds. While this would introduce additional regulatory risk if implemented on an individual plant basis, the intent is to encourage industry development of a technically robust yet optimized fleet-based performance monitoring approach. We encourage the staff to continue working with industry to develop such an approach that:
efficiently supports plant-specific TBS applicability assessment; provides an appropriate level of ongoing insight into the condition of piping welds above the TBS; avoids biasing the overall risk-informed in-service inspection program sampling away from more risk-significant locations; keeps radiological exposure as low as is reasonably achievable (ALARA);
minimizes regulatory uncertainty; and enhances operational efficiency.
LOCA Fuel Performance Analysis 10 CFR Part 50, Appendix K, and Section 50.46, specifies the methods and fuel performance criteria for LOCA analysis. Among the metrics are a maximum cladding temperature of 2200°F and an equivalent cladding oxidation less than 17%. In addition, 10 CFR 50.46 requires that a coolable geometry must be maintained during a postulated LOCA event. Considerable D. Wright uncertainty existed in the state of knowledge at the time of the original rule, justifying the requirement of significant margin in the thermal-hydraulic analysis. The performance (corrosion, ductility, and steam oxidation response) of the zirconium (Zr) alloys in use at that time, Zircaloy-2 and Zircaloy-4, also justified the additional margin.
Extensive research and development efforts over the past several decades have shown that during operation, the irradiated ductility and corrosion performance of the cladding is significantly degraded due to the effect of hydrogen absorption. The net result is a burnup-dependent degradation of clad ductility. Additionally, under some extreme conditions, the cladding oxidation rate transitions from parabolic behavior to less desirable linear behavior (breakaway oxidation) as a function of temperature.1 Linear behavior is the result of the oxide being less protective.
These cladding degradation phenomena have led to a proposed reduction in the allowable equivalent cladding oxidation as a function of the clad hydrogen content and hence burnup.
The staff has developed several regulatory guides that define processes and procedures that should be used to: (1) measure breakaway oxidation behavior (DG-1261, Revision 1),
(2) determine post-quench ductility (DG-1262, Revision 1), and (3) establish analytical limits for Zr-based cladding (DG-1263, Revision 1).
DG-1261 establishes recurring testing requirements that have the potential to be burdensome without a commensurate safety benefit. For example, the breakaway oxidation concern is based on test results from Russian E-110 cladding that has never been used in the U.S. At the same time, industry has developed advanced Zr alloys that have much better performance.
Testing of these advanced alloys under the same conditions as that for the E-110 cladding demonstrated that the extreme oxidation did not occur. Based on the evolution of alloy development of current generation cladding materials, it is very unlikely, given the extensive quality and process control in place for modern cladding, that unanticipated behavior would occur. Existing quality and fabrication process controls and specifications for cladding production should yield acceptably consistent material performance results without the need for frequent recurring laboratory oxidation testing. Such controls should include assuring a given material heat meets American Society for Testing and Materials and vendor specification requirements, and that overall fabrication is consistent with that of the same materials tested in the past.
Fuel Fragmentation, Relocation, and Dispersal Analytically, fuel cladding failures have historically been allowed for some design basis accident sequences, but the potential for fuel dispersal was not yet identified as an issue.
Recent research results suggest that at pellet average burnups exceeding 55 GWd/MTU the possibility exists for the ejection and dispersal of fine fuel fragments should the fuel cladding burst. Dispersal has at least the potential to influence fuel coolability. Draft regulatory guide DG-1434 has been developed to suggest paths forward in accounting for dispersal. The DG-1434 provides the results and analysis of recent testing to quantify the extent of dispersal assuming cladding burst. Based on data from tests performed at Studsvik National Laboratory in Sweden, the Halden Boiling Water Reactor in Norway, and the Severe Accident Test Station at the Oak Ridge National Laboratory, for locations with greater than 3% cladding strain, no 1 Parabolic oxidation kinetics means that the square of the oxide layer thickness grows linearly with time. Linear oxidation kinetics means the oxide layer thickness grows linearly with time.
D. Wright dispersal is assumed for burnups below 55 GWd/MTU; 100% dispersal is assumed above 80 GWd/MTU; and linear interpolation is used between 55 GWd/MTU and 80 GWd/MTU.
However, while test results suggest that dispersal is a significant possibility upon cladding burst at higher burnups, estimating the occurrence of cladding burst (and subsequent fuel dispersal) is fraught with uncertainty. The Committee has identified this issue in our letter report for our review of Research Information Letter (RIL) 2021-13. Additionally, the draft guide acknowledges this:
However, as noted previously, the amount of fuel that disperses during a test is governed by a number of factors, some of which may vary stochastically (such as the burst opening size). Furthermore, the test conditions may not account for all the conditions that may be present during a LOCA at a commercial power reactor. For these reasons, in RIL 2021-13, the NRC argued that it is important to compare the models to the amount of mobile fuel in the rod, since some of the mobile material that remained in the rod following the test could have been dispersed had the burst opening size been larger, or had the rod been subjected to forces greater than those experienced in the test.
Reduction in the uncertainty related to cladding burst and fuel dispersal to a manageable value would, in our judgement, require an extensive and extremely expensive research and development effort with no guarantee of adequate results. Given the complexity of the phenomena involved in FFRD, analytically demonstrating no burst resulting in dispersal is preferable for increased regulatory certainty.2 Recent results from best-estimate analyses suggest that a no burst resulting in dispersal criterion for LOCA analysis may be possible. However, there is a need for guidance related to what is a true best estimate as allowed under the rule for large break (greater than TBS)
LOCA analysis.
We are aware that EPRIs Alternate Licensing Strategy is currently undergoing parallel review by the staff, as an alternative to, rather than a replacement for, the draft rule. This alternative avoids the need to model the consequences of dispersal or perform a true best estimate LOCA analysis and thus may provide a greater degree of regulatory certainty in regard to FFRD. We urge the staff to complete this review on an expeditious basis.
Updated Source Terms and Control Room Dose The proposed regulatory guide (RG) 1.183, Revision 2 (DG-1425) provides revised guidance on LWR source terms in the following major areas:
Maximum Hypothetical Accident (MHA) LOCA source terms to containment with application to Zr-UO2, FeCrAl clad UO2, and Cr-coated Zr-UO2 fuels accounting for:
recent research results on severe accident phenomenology, damage progression, and source terms; increase of enrichment from 8 to 10%; and burnups up to 80 GWd/MTU; Credit for removal of fission products from BWR suppression pools via scrubbing in certain postulated severe events reducing airborne containment release fractions based 2 This term refers to no burst to high burnup fuel rods that are susceptible to dispersal.
D. Wright on both industry (Modular Accident Analysis Program or MAAP) and NRC (Modular Engineering Level Computer Code or MELCOR) detailed source term calculations; Adoption of graded risk-informed performance-based control room accident dose criterion; Clarification of the role of the MHA, with its fission product release to containment exceeding that of a design basis accident to evaluate the effectiveness of containment and other safety features; Considerations for evaluating the adequacy of defense in depth and safety margins in light of the new guidance; and Addition of flexibility in evaluating radiological consequences of postulated accidents.
The new source terms are more firmly rooted in data and take credit for actual measured fission product chemical forms and physical aerosol transport phenomena instead of simpler conservative approaches used in the past. These source terms support extension to new fuel cladding systems, higher burnups, longer (24 month) fuel cycles, and power uprates.
The control room dose values used as a metric to judge the acceptability of the control room design are reconsidered in the proposed RG 1.183, Revision 2 (DG-1425). The dose metrics in the proposed rule and regulatory guide revision align more closely with both national (10 CFR Part 20 and Environmental Protection Agency emergency planning requirements) and international radiation protection standards, and are supported by the technical rationale behind those standards. NRC Staff in the Offices of Nuclear Regulatory Research and Nuclear Reactor Regulation have performed a comprehensive review and independent evaluation of these modified standards and have demonstrated further that these changes have a strong technical basis, as discussed in References 15 and 16.
We find that the discussions in the preamble in the draft rule and in the proposed RG-1.183, Revision 2, about the historical context surrounding the development of the control room dose-based criteria are very informative. They should be captured by the NRCs knowledge management (KM) program.
The rule establishes the allowable accident analysis control room dose criterion at 10 rem total effective dose equivalent (TEDE), with the option to increase to 25 rem TEDE based on the plant risk profile. The associated guidance sets the allowable design basis accident control room dose criteria at 10 rem TEDE for some postulated accident events, while leaving the criterion for less severe postulated accident events at 5 rem TEDE. For the MHA LOCA, where the analysis is biased to assume a release larger than any of the postulated accident events, the metric would be on a sliding scale from 10 to 25 rem TEDE depending on the overall risk of the facility as measured by the core damage frequency (CDF) based on an individual plants risk assessment. This risk-informed and performance-based sliding scale was implemented to provide greater design and operational flexibility for plants with a low CDF and to incentivize plants to enhance the safety of the facility.
While we agree with the overall objective of the staff, there was no consensus among the members about the use of CDF as a metric. The use of a sliding scale up to 25 rem TEDE based on CDF deviates from the standard approach for risk-informed design changes since a D. Wright relationship between the change in dose design criterion and the change in public risk is not established. Some members appreciated the staffs attempt to develop a sliding scale based on an overall plant risk metric. Others felt it was not an appropriate application of CDF and that an MHA LOCA limit of 25 rem TEDE could be applied regardless of CDF. This divergence of opinion shows the challenge inherent in risk-informing deterministic design criteria that are not directly related to items modeled in a probabilistic risk assessment.
Members also noted the control room dose design criterion does not directly relate to the exposure that control room operators could be reasonably expected to receive in an accident scenario. The dose consequences of a beyond design basis severe accident may not be bounded by the MHA LOCA. In such cases, other layers of defense exist to protect the operators, and the actual dose received by an operator is managed by the emergency response organization. Designing a control room with minimal in-leakage supports these emergency response actions. While not specifically stated in the draft proposed rule language provided to the Committee, NRC staff intends to provide guidance to assess the impact of any change enabled by the rule on both probabilistic risk assessments and severe accident management.
We agree and observe that control room habitability in a severe accident scenario can be important to overall plant safety and should continue to be considered when making design changes that might degrade habitability.
Broader Impacts of the Rule The rulemaking is broad and its impact on all aspects of plant safety and operation may not be apparent until licensing actions that use the rule occur. Treatment of a LOCA as a beyond design basis accident has wide ranging impacts throughout the safety analyses for the plant and may require downstream changes not yet realized. Examples of such impacts include containment integrity, severe accident management, emergency planning, content of technical specifications, emergency operating procedures, operator training, and aging management.
Staff should continue to examine impacts beyond those they have already considered.
The broader impacts of the proposed rule changes can be seen as a series of trade-offs that must be assessed individually and collectively. We note that the proposed rule largely impacts PWRs. Its impact on BWRs is likely not as significant since they, for the most part, already operate on 24-month cycles. There is also a growing interest in power uprates across the industry. In any case, developing an efficient strategy for implementation will likely be critical.
Although no comprehensive list of potential impacts has been developed, continuing engagement with stakeholders should allow for the identification of areas to be vetted for improvement. During staff and industry presentations, the importance of timely interaction (workshops, tabletop exercises, etc.) was stressed. This future stakeholder engagement during the proposed rule comment period will be important for the staff in developing the final rule package. The introduction of the increased enrichment rule sets a good example for future policy development. It demonstrates a commitment to continuous improvement and adaptation in response to emerging challenges for the next generation of reactor technology.
SUMMARY
The draft proposed rule represents a significant achievement in the efforts to safely regulate the use of increased enrichment fuels and accident tolerant fuels in light water reactors.
Consistent with the NRCs new mission statement, the proposed rule enables higher burnups up to 80 GWd/MTU and is expected to support advances such as 24-month fuel cycles and D. Wright potential power uprates. The draft proposed rule is performance based, and risk-informed as directed by the Commission.
With the introduction of the concept of transition break size, the proposed rule presents a major change from regulatory precedent by providing a solid technical rationale for moving away from a double-ended guillotine break in the primary piping as the presumed defining design basis accident for LWRs. We support staff plans to work with stakeholders and examine broader impacts of this change.
As discussed above, we have the following comments and suggestions:
In the area of TBS, we recommend the staff work with industry to develop an approach to inspections that balances the increased burden of inspections (in terms of cost and radiation exposure) versus the need to support the plant-specific TBS applicability assessment.
In the area of LOCA fuel performance analysis, recurring testing of cladding by the fuel fabricator is an increased burden that may not be warranted given current quality controls on cladding fabrication.
From our perspective, a no burst resulting in dispersal approach is preferable to address fuel rods susceptible to FFRD given the uncertainties inherent in the underlying phenomena. At this point in time, this approach would provide more regulatory certainty.
EPRIs Alternate Licensing Strategy is currently undergoing parallel review by the staff as an alternative to, not a replacement for, the rule. We urge the staff to complete this review on an expeditious basis because this approach may provide a greater degree of regulatory certainty in regard to FFRD.
The rationale for the proposed new control room dose design criterion of 10 rem TEDE, including the proposed sliding scale up to 25 rem TEDE, is justified based on accepted radiation protection guidelines. We encourage the staff to continue to evaluate implications of design changes enabled by the new criterion to assess potential impact on the public risk.
The draft proposed rule should be issued for public comment.
D. Wright We look forward to interacting with the staff as appropriate as the draft final rule package is developed as a result of public comments.
We are not requesting a formal response from the staff to this letter report.
Sincerely, Walter L. Kirchner Chairman
Enclosures:
1.
List of Acronyms 2.
Additional Comments By ACRS Member Robert Martin REFERENCES 1.
US NRC, Predecisional Information to Support - ACRS Public Meeting - Draft Federal Register Notice to Support Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, January 13, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No.ML25013A080) 2.
U.S. NRC, Staff Requirements - SECY-21-0109 - Rulemaking Plan on Use of Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, March 16, 2022 (ADAMS Accession No. ML22075A103) 3.
U.S. NRC, Regulatory Basis Document for Public Comment, Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors, September 2023 (ADAMS Accession No. ML23032A504) 4.
U.S. NRC, DG-1261, Revision 1 (RG 1.222, Revision 0), Measuring Breakaway Oxidation Behavior, ACRS Version2 (ADAMS Accession No. ML24327A107) 5.
U.S. NRC, DG-1262, Revision 1 (RG 1.223, Revision 0), Determining Post-Quench Ductility, ACRS Version2 (ADAMS Accession No. ML24327A112) 6.
U.S. NRC, DG-1263, Revision 1 (RG 1.224, Revision 0), Establishing Analytical Limits for Zirconium-Based Alloy Cladding, ACRS Version 2 (ADAMS Accession No. ML24326A215) 7.
U.S. NRC, DG-1426 (RG 1.225, Revision 0), An Approach for Risk-Informed Evaluation Process Supporting Alternative Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Reactors, ACRS Version (ADAMS Accession No. ML24319A118)
Signed by Kirchner, Walter on 02/22/25 D. Wright 8.
U.S. NRC, DG-1428 (RG 1.258, Revision 0), Plant-Specific Applicability of Transition Break Size, ACRS Version (ADAMS Accession No. ML24341A159) 9.
U.S. NRC, DG-1434 (RG 1.259, Revision 0), Addressing the Consequences of Fuel Dispersal in Light-Water Reactor Loss-of-Coolant Accidents, ACRS Version (ADAMS Accession No. ML24320A013) 10.
U.S. NRC, DG-1425 (RG 1.183, Revision 2), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, ACRS Version (ADAMS Accession No. ML24304A864) 11.
U.S. NRC, NUREG-1829, Volume 1, Estimating Loss-of-Coolant Accident (LOCA)
Frequencies through the Elicitation Process, March 2008 (ADAMS Accession No. ML082250436) 12.
U.S. NRC, NUREG-1903, Seismic Considerations for the Transition Break Size, February 2008 (ADAMS Accession No. ML080880140) 13.
U.S. NRC White Paper, White Paper on Continued Applicability of NUREG-1829, November 13, 2024 (ADAMS Accession No. ML24205A015) 14.
U.S. NRC White Paper, White Paper on Continued Applicability of NUREG-1903, November 18, 2024 (ADAMS Accession No. ML24323A205) 15.
U.S. NRC, Control Room Design Criteria and Radiological Health Effects, June 2023 (ADAMS Accession No. ML23027A059) 16.
U.S. NRC, Method for Graded Risk-Informed Performance-Based Control Room Design Criteria Framework, December 2024 (ADAMS Accession No. ML24150A080) 17.
U.S. NRC, RIL 2021-13, Interpretation of Research on Fuel Fragmentation, Relocation, and Dispersal at High Burnup, December 2021 (ADAMS Proprietary Accession No. ML21313A110, Non-Public)
LIST OF ACRONYMS ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System ALARA As Low As Is Reasonably Achievable ASME American Society of Mechanical Engineers ATF Accident Tolerant Fuel ATWS Anticipated Transients Without Scram BWRs Boiling Water Reactors CDF Core Damage Frequency CFR Code of Federal Regulation DGs Draft Regulatory Guides EPRI Electric Power Research Institute FFRD Fuel Fragmentation, Relocation, and Dispersal GWd/MTU Gigawatt Days Per Metric Ton of Uranium KM Knowledge Management LBB Leak Before Break LOCA Loss of Coolant Accident LWRs Light Water Reactors MAAP Modular Accident Analysis Program MELCOR Modular Engineering Level Computer Code MHA Maximum Hypothetical Accident NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation PLP Primary Loop Piping PWR Pressurized Water Reactor RIL Research Information Letter RG Regulatory Guide SBO Station Blackout SRM Staff Requirements Memorandum TBS Transition Break Size TEDE Total Effective Dose Equivalent ADDITIONAL COMMENTS BY ACRS MEMBER ROBERT MARTIN Consistent with statements in the Committees letter addressing increased fuel enrichment, I support redefining the design-basis LOCA in terms of a Transition Break Size (TBS) for compliance with 10 CFR 50.46. This position aligns with modern understanding of pipe rupture probabilities and associated risks. However, while the Committee letter appropriately recommends that the staff consider the broader impacts of the rule, I strongly caution against extending the TBS LOCA into areas such as containment integrity, severe accident mitigation, and emergency response.
The large double-ended guillotine LOCA has historically served as a bounding event for regulatory considerationnot only as a physical limit on primary system failure but also to account for incompleteness in the nuclear safety case. While early regulators recognized the low likelihood of the large LOCA, they lacked the data to quantify risk and instead adopted a conservative approach to ensure reactor designs could withstand extreme failures. Decades of progress in nuclear safety, including advances in risk assessment and the recent insights into fuel performance at higher burnups, now provide a strong basis for updating regulations for normal and design-basis conditions. However, rare events in complex systems often arise from cascading interactions rather than independent failures, meaning probabilistic models based on incomplete data may underestimate true risk.
A hybrid nuclear safety approach remains essentialone that leverages risk insights where failure modes are well-characterized while retaining deterministic safeguards to address residual risks, i.e., unknown unknowns. In particular, regulatory assessment of containment integrity must remain a deterministic last-line of defense, engineered against the physical realities of low-probability, high consequence events. Moreover, shifting containment evaluations to a purely risk-informed basis could introduce unintended operational and maintenance incentives that were not originally considered.
In summary, while risk-informed methods can improve efficiency in well-characterized areas, credible residual risks remain that only deterministic safeguards can address at this time. As the NRC moves away from the large LOCA as a design-basis event, it may find that certain design features, e.g., containment, require stricter performance goalsjust as it did when adopting the ATWS (10 CFR 50.62) and SBO (10 CFR 50.63) rules, which required specific protections for beyond-design-basis events. As public engagement on the increased enrichment rulemaking continues, the NRC must carefully evaluate whether potential vulnerabilities, both known and unknown, warrant the addition of a future rule of similar magnitude to ensure that safety functions historically linked to the large LOCA remain adequately addressed.
SUBJECT:
RULEMAKING ON INCREASED ENRICHMENT OF CONVENTIONAL AND ACCIDENT TOLERANT FUEL DESIGNS FOR LIGHT-WATER REACTORS Accession No: ML25044A331 Publicly Available (Y/N): Y Sensitive (Y/N): N If Sensitive, which category?
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NRC Users or ACRS only or See restricted distribution OFFICE ACRS SUNSI Review ACRS ACRS ACRS ACRS NAME WWang WWang LBurkhart RKrsek MBailey WKirchner DATE 02/13/25 02/13/25 02/14/25 02/20/25 02/20/25 02/21/25 OFFICIAL RECORD COPY February 22, 2025