ML20219A702
| ML20219A702 | |
| Person / Time | |
|---|---|
| Issue date: | 08/04/2020 |
| From: | Michael Snodderly Advisory Committee on Reactor Safeguards |
| To: | Dimitrijevic V, Walter Kirchner Advisory Committee on Reactor Safeguards |
| Snodderly M | |
| References | |
| Download: ML20219A702 (3) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 August 4, 2020 MEMORANDUM TO:
Walter L. Kirchner, Chairman Vesna B. Dimitrijevic, Member NuScale Subcommittee Advisory Committee on Reactor Safeguards FROM:
Mike Snodderly, Senior Staff Engineer /RA/
Technical Support Branch Advisory Committee on Reactor Safeguards
SUBJECT:
ANALYSIS OF NRR RESPONSE TO ACRS LETTER ON NUSCALE AREA OF FOCUS - PROBABILISTIC RISK ASSESSMENT AND EMERGENCY CORE COOLING SYSTEM VALVE PERFORMANCE Attached is a copy of the July 1, 2020, Office of Nuclear Reactor Regulation (NRR) response to the June 1, 2020, ACRS letter on NuScale Area of Focus - Probabilistic Risk Assessment (PRA) and Emergency Core Cooling System (ECCS) Valve Performance. A copy of the Committees letter is also attached.
Committee Letter:
In its June 1, 2020 letter:
The Committee concluded that the NuScale design certification application (DCA) meets the 10 CFR 52.47(a)(27) requirement to include a description of the design-specific PRA and its results.
The Committee further concluded that a primary purpose of the PRA at the DCA stage is to inform the design to reduce risk. The PRA scope is sufficient to enable the discussion of risk results and insights, and the level of detail in the PRA is consistent with its intended uses in support of design certification; i.e., to identify design alternatives, operational vulnerabilities, and to provide risk-informed support for other programs. The Committee stated the risk insights CONTACT: Mike Snodderly, ACRS/TSB 301-415-2241
W. Kirchner and identified in Chapter 19 should not be considered final because there are omissions in the existing Final Safety Analysis Report (FSAR) that need to be properly reflected in the PRA. The Committee recommended that the resolution of the boron dilution issue needs to be evaluated to determine if these scenarios should be included in the PRA at the DCA stage. Such inclusion could impact the reported risk measures and the risk insights as presented in Chapter 19.
The Committee concluded that the risk measures of core damage frequency (CDF) and large release frequency (LRF), quantified in the PRA, suggest that the NuScale design meets the Commissions Safety Goals with large margins. However, recently identified design issues, underlying omissions, and uncertainties indicate that the large margins between CDF and LRF and safety goals cannot be substantiated at this time.
To promote identification of valid risk insights through the combined license (COL) process, the Committee provided recommendations on several other topics: ECCS valve performance and qualification; risk importance of the chemical and volume control system; errors of commission associated with reactor building crane operations; risk increase to single unit operation with multiple unit operation and buildout; steam generator integrity; post-accident combustible gas monitoring; and more rigorous treatment of sensitivities and uncertainties.
The Committee felt that the risk insights will be better supported when the COL applicant addresses the requirements of the NuScale design certification rule appendix to 10 CFR Part
- 52. This includes addressing the COL items, closing the Inspections, Tests, Analyses, and Acceptance Criteria items, updating the site and plant-specific PRA before fuel load, and, of particular interest with respect to the NuScale design, addressing the additional requirements and restrictions in the rule appendix to address design completeness.
The Committee could not reach a final conclusion on the safety of the NuScale design until the issue of the potential for a reactivity insertion accident due to boron dilution in the downcomer is resolved to our satisfaction.
EDO Response:
The NRR Office Director stated that he appreciated the time and effort that the ACRS has devoted to this review. The staff agreed with the Committees first conclusion that the DCA meets the 10 CFR 52.47(a)(27) requirement to include a description of the design-specific PRA and its results. With regard to the Committees second conclusion, the staff stated that as they finalize their review of the affected information associated with FSAR Chapter 19, the staff will reaffirm its safety findings regarding NuScales DCA.
In response to the third recommendation concerning the resolution of the boron dilution issue.
The staff stated that they will continue to evaluate the extent to which the design changes associated with boron redistribution affect the risk results and insights identified in FSAR Chapter 19. The staff stated that they will ensure that the PRA adequately reflects the applicants resolution of the boron redistribution concerns. The staff will also ensure that the PRA accounts for all known significant risk contributors and that the identification of risk insights is acceptable for the staff to make its safety findings as described in Section 19.0, Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:
LWR Edition.
In response to the fourth conclusion, the staff stated that they will continue to evaluate the NuScale design for consistency with the Commissions goals for CDF and LRF once the applicant has adequately addressed the impact of boron redistribution issues on the PRA.
W. Kirchner and In response to the fifth recommendation, the staff stated that they will continue to evaluate the extent to which the risk results and insights affect FSAR Chapter 19. The staff went on to say that consistent with the Part 52 process risk insights will be better supported when a future COL applicant addresses the requirements in the NuScale design certification appendix to 10 CFR Part 52, Licenses, certifications, and approvals for nuclear power plants, assuming that the NuScale design is approved for certification during the COL application phase, additional details on the design and operations will become available. The staff will review these details for applicability to the COL application PRA under COL Information Item 19.1-8. Under COL Information Item 19.1-8, the COL applicant will confirm assumptions and data supporting the PRA.
Finally, the staff acknowledged the Committees viewpoint as discussed in the sixth conclusion that the Committee could not reach a final conclusion on the safety of the NuScale design until the issue of the potential for a reactivity insertion accident due to boron dilution in the downcomer is resolved to the Committees satisfaction.
Analysis:
I recommend that the Committee accept the staffs response in light of the Committees July 29, 2020 letter report, Report on the Safety Aspects of the NuScale Small Modular Reactor. In this letter report, the Committee identified several potentially risk-significant items that are not completed at this time. The Committee requested the opportunity to review the qualification of emergency core cooling system valve performance, the identification of a successful recovery strategy to prevent potential reactivity insertion accidents associated with boron dilution sequences, and the updated probabilistic risk assessment.