ML19344A257

Summary Report - 667th Meeting of the Advisory Committee on Reactor Safeguards, October 2-4, 2019

ML19344A257
Person / Time
Issue date:	12/10/2019
From:	Riccardella P Advisory Committee on Reactor Safeguards
To:	Kristine Svinicki NRC/Chairman
Burkhart, L, ACRS
Shared Package
ML19344B949	List: ML19344A257
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Download: ML19344A257 (7)

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UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 December 10, 2019 The Honorable Kristine L. Svinicki Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUMMARY

REPORT - 667th MEETING OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, OCTOBER 2-4, 2019

Dear Chairman:

During its 667th meeting, October 2-4, 2019, the Advisory Committee on Reactor Safeguards (ACRS) discussed several matters and completed the following correspondence:

LETTER REPORTS Letter Reports to Chairman Svinicki from Peter C. Riccardella, Chairman, ACRS:

Advanced Boiling Water Reactor Design Certification Renewal, dated October 31, 2019, ADAMS Accession No. ML19305D117.

Review of Advanced Reactor Computer Code Evaluations, dated November 4, 2019, ADAMS Accession No. ML19302F015.

LETTER Letters to Margaret M. Doane, Executive Director for Operations (EDO), NRC, from Peter C. Riccardella, Chairman, ACRS:

Safety Evaluation of Topical Report ANP-10346P, Revision 0, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA, dated November 4, 2019, ADAMS Accession No. ML19308A004.

MEMORANDA Memoranda to Margaret M. Doane, Executive Director for Operations, NRC, from Scott W. Moore, Executive Director, ACRS:

Documentation of Receipt of Applicable Official NRC Notices to the Advisory Committee on Reactor Safeguards for October 2019, dated November 4, 2019, ADAMS Accession No. ML19316B574.

2 Regulatory Guides, dated November 4, 2019, ADAMS Accession No. ML19316B450, regarding no review of Draft Guide (DG) -1287, An Approach for Plant-Specific, Risk-Informed Decision making: Technical Specifications, which is the latest proposed revision to Regulatory Guide (RG)-1.177 of the same name. The memorandum did provide comments on terminology for the staffs consideration.

Memorandum to Ho K. Nieh, Director, Office of Nuclear Reactor Regulation (NRR), NRC, from Scott W. Moore, Executive Director, ACRS:

ACRS Review of NuScale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapters 7, 10, 11, and 17, dated November 19, 2019, ADAMS Accession No. ML19297D008.

HIGHLIGHTS OF KEY ISSUES

1. Advanced Boiling Water Reactor Design Certification Renewal The Committee met with representatives of the NRC staff and GEH to discuss the design certification renewal application and associated NRC safety evaluation reports.

On December 7, 2010, GEH requested the NRC to renew the ABWR design certification. The ABWR design certification rule, effective June 11, 1997, would otherwise expire at the end of a period of 15 years, or June 11, 2012. GEH applied for a design certification renewal on December 7, 2010. On July 20, 2012, staff identified proposed changes including Fukushima Near Term Task Force Recommendations. GEH provided the ABWR design control document (DCD), Revision 6, in response to staff requested changes. On June 28, 2019, the staff completed the SEs with no open items.

In total, 39 design items were reviewed and approved by the staff in supplemental SEs to NUREG-1503 or closed by letter. In addition to reviewing DCD, Revision 6, and responses to requests for additional information, the staff performed audits to resolve outstanding technical issues.

Committee Action The Committee issued a report to the Chairman on October 31, 2019, with the following conclusion and recommendation:

a. There is reasonable assurance that the ABWR, under the renewed design certification, can be constructed and operated without undue risk to the health and safety of the public.

b. We concur with the conclusions of the staffs supplemental renewal SEs to NUREG-1503, Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design, with no open items. The SEs should be issued, and the GEH application for the Design Certification Renewal of the ABWR should be approved.

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2. Review of Advanced Reactor Computer Code Evaluations The NRC staff developed a report in 2016, NRC Vision and Strategy: Safely Achieving Effective and Efficient Non-Light Water Mission Readiness, as the staff contemplated how to review and regulate a new generation of non-LWRs including their associated fuel cycles and waste forms. The report lays out six strategies to accomplish its goals and provides a set of Implementation Action Plans that identify specific, actionable tasks that can fulfill the strategies.

Over the intervening years, the Committee has provided letter reports on the vision and strategy document as well as several products of the Implementation Action Plansnon-LWR design criteria, functional containment performance criteria, the licensing modernization project (LMP, now called the Technology-Inclusive, Risk-Informed, and Performance-Based Approach to Inform the Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,) and siting for advanced reactors.

The subject of the Committees current review is a series of draft documents prepared by the staff on evaluating computer codes needed to conduct confirmatory analyses of non-LWR nuclear power plants. The draft documents reviewed are entitled, Code Assessment Plans for NRCs Regulatory Oversight of Non-Light Water Reactors, NRC Non-Light Water Reactor (Non-LWR) Vision and Strategy, Volume 1 - Computer Code Suite for Non-LWR Design Basis Event Analysis, NRC Non-Light Water Reactor (Non-LWR) Vision and Strategy, Volume 2 -

Fuel Performance Analysis for Non-LWRs, and NRC Non-Light Water Reactor (Non-LWR)

Vision and Strategy, Volume 3 - Computer Code Development Plans for Severe Accident Progression, Source Term, and Consequence Analysis.

The staff has completed a major step in addressing Strategy 2: Computer Codes. The four draft documents they shared with the Committee complete a major portion of their near-term action plan by identifying and assessing the available computer codes and databases. This activity is helping the staff develop their understanding of the technologies involved and the associated phenomena that will be encountered in reviewing non-LWR designs. Two final documents, Volume 4 on licensing and siting dose assessment codes and Volume 5 on fuel cycle topics, have not been completed. The gap analyses associated with Volumes 1-3 consider both knowledge gaps (fundamental physics and chemistry) and computer code gaps. Staff stated that most of the knowledge gaps are in the area of severe accidents. They have been comprehensive in identifying potential gaps, consistent with Phenomena Identification and Ranking Tables (PIRTs) that have been performed in recent years.

Knowledge gaps will need to be addressed by experiment and operating history, perhaps as interpreted by expert elicitation. Computer code gaps have been addressed in the current drafts. The staffs contractor developed a predictive capability maturity model (PCMM) to characterize the state of readiness of the computer codes. It generates maturity level scores over a set of six fundamental modeling and simulation elements for each reactor type. It has been useful for evaluating the level of effort expected to complete development of the computer codes for use in staff reviews.

4 Committee Action The Committee issued a report to the Chairman on this topic via letter dated November 4, 2019, with the following conclusions and recommendations:

a. The approach taken by the staff supports their readiness to review submittals for non-LWR designs of many different types. This approach can also help the staff understand the new reactor designs and associated phenomena.

b. Ideally, the tools for staff confirmatory analysis should be as independent as practical, validated, understood by the staff, and usable on the staffs computer resources.

c. The staff also needs to become sufficiently familiar with applicant codes to support timely reviews of submitted analyses.

d. The overview report should be revised to better explain how this approach integrates the evaluations discussed in Volumes 1-3 and to present a coherent strategy for evaluations. Four principles should underlie the strategy: simplicity, completeness, working the problem backwards starting with the source term, and scaling down the level of effort of licensing review proportionately as the hazard decreases.

e. The staff should perform pilot studies using relatively mature designs to illustrate how the analysis should proceed. They should consider a case using the licensing modernization project (LMP) and one that uses an alternative approach. This would increase confidence in the overall approach being pursued by staff and flush out any needed refinements.

3. Safety Evaluation of Topical Report ANP-10346P, Revision 0, ATWS-I Analysis Methodology for BWRs Using RAMONA5-FA The Committee met with representatives of the NRC staff and Framatome to review the subject Topical Report and associated safety evaluation.

The staff has previously approved multiple components of the RAMONA5-FA ATWS-I methodology as part of their review of the Monticello EFW license amendment request; therefore, the primary focus of the staff review was the aspects of this methodology that are novel to ensure applicability on a generic basis, as well as the integration of multiple methodologies developed at different times into a single approach for generic ATWS-I analyses.

The staff also reviewed the ATWS-I PIRT, experiment benchmarking, and an example plant application. The staff review followed key elements of the evaluation model development and assessment process outlined in Regulatory Guide 1.203, including: accident scenario description and phenomena identification and ranking; evaluation methodology; code assessment; uncertainty analysis; and documentation.

The staff has found the RAMONA5-FA methodology acceptable for ATWS-I calculations with seven limitations and conditions: the gap conductance sensitivity shall be reevaluated for new fuels; justification must be provided to demonstrate adequate margin in operator action timing; the assumptions employed in the analysis of record must be verified for core specific applications; transition cores must have additional verification; both turbine trip and recirculation

5 pump trip must be analyzed to determine the limiting ATWS-I event; plant-specific steam line and valve models must be verified; and plant-specific applications must justify the selected settings for RAMONA5-FA. The Committee concurs with these limitations and conditions.

Committee Action The Committee issued a report to the EDO on this Topical Report and associated staff safety evaluation via letter dated November 4, 2019, with the following conclusion and recommendation:

a. The RAMONA5-FA methodology to analyze anticipated transients without scram with instability (ATWS-I), when used in compliance with the seven limitations and conditions imposed by the staff, is acceptable for use in boiling water reactor (BWR) licensing applications.

b. The safety evaluation should be issued.

RECONCILIATION OF ACRS COMMENTS AND RECOMMENDATIONS The Committee considered the letter from the Director, Office of New Reactors, dated July 16, 2019, ADAMS Accession No. ML19184A545, in response to the Committees letter dated June 20, 2019, ADAMS Accession No. ML19171A323. The topic was review of Nuclear Energy Institute 96-07, Appendix D, Supplemental Guidance for Application of 10 CFR 50.59 to Digital Modifications, issued November 2018, and the U.S. Nuclear Regulatory Commissions Associated Draft Revision 2 to Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59 Changes, tests, and experiments. The Committee accepted the staffs response and looks forward to reviewing the final Revision 2 to RG 1.187 following resolution of public comments.

NUSCALE PHASE 5 DISCUSSIONS As documented in the memorandum from Scott Moore, ACRS Executive Director, to Ho Nieh, Director, NRR, the Committee decided that no further briefing by the staff is needed to support the Committees Phase 5 review as it pertains to Chapters 7, 10 and 17 of the NuScale design certification application. To complete their Chapter 11 Phase 5 review, the Committee will review the staffs evaluation of the acceptability of NuScales realistic and design basis failed fuel fraction as part of its Source Term area of focus review. The Committee will inform the Commission of its finding in this matter relative to the requirements of 10 CFR 52.53 in a subsequent letter.

The Committee identified its approach to the Phase 5 review in its letter dated September 25, 2019 (ADAMS Accession No. ML19269B682.

SCHEDULED TOPICS FOR THE 668th ACRS MEETING The following topics were placed on the agenda for the 667th ACRS meeting which was scheduled for November 6-9, 2019:

Radiation Embrittlement of Reactor Vessel Materials

6 FRAMATOMEs Topical Report, RAMONA5 for Anticipated Transient without Scram Brunswick Atrium 11 Fuel Transition and Application USAPWR Chapters 8, 18, and Advanced Accumulator Topical Report NUREG/KM-0013, Credibility Assessment Framework for Critical Boiling Transition Models Further discussion of various safety evaluation reports related to the NuScale design certification application review Sincerely,

/RA/

Peter C. Riccardella, Chairman

7 December 10, 2019

SUMMARY

REPORT - 667th MEETING OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, OCTOBER 2-4, 2019 Accession No: ML19344A257 Publicly Available (Y/N): _Y___

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NRC Users or ACRS only or See restricted distribution OFFICE ACRS SUNSI Review ACRS ACRS NAME LBurkhart LBurkhart SMoore PRiccardella (SMoore for)

DATE 12/02/19 12/02/19 12/10/19 12/10/19 OFFICIAL RECORD COPY