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Category:Meeting Agenda
MONTHYEARPMNS20211527, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports PMNS20211528, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports PMNS20211529, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports PMNS20211530, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports PMNS20211531, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports PMNS20211532, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports PMNS20211533, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports PMNS20211534, Meeting with NuScale Power, LLC on NuScale Topical Reports2021-12-0303 December 2021 Meeting with NuScale Power, LLC on NuScale Topical Reports ML21004A2372021-01-0404 January 2021 Standard Design Approval Teleconference ML20349D0382020-12-14014 December 2020 Standard Design Approval Teleconference ML20321A2912020-11-16016 November 2020 Standard Design Approval Teleconference ML20314A0402020-11-0909 November 2020 Standard Design Approval Teleconference ML20314A2382020-11-0909 November 2020 Standard Design Approval Teleconference ML20314A2392020-11-0909 November 2020 Standard Design Approval Teleconference ML20305A3112020-10-31031 October 2020 Standard Design Approval Teleconference ML20300A5582020-10-26026 October 2020 Standard Design Approval Teleconference ML20300A5602020-10-26026 October 2020 Standard Design Approval Teleconference ML20300A5612020-10-26026 October 2020 Standard Design Approval Teleconference ML20219A4112020-08-10010 August 2020 Summary of the May 28, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss Nuscale'S Standard Design Application Control Room Staffing Topical Report-0520-69851 ML20108F5082020-05-27027 May 2020 Summary of April 15 2020 Public Meeting on NuScale EPZ Sizing Methodology TR ML20140A1892020-05-19019 May 2020 Public Meeting Summary May 14, 2020, NuScale Boron Redistribution Issue ML20114E0192020-04-27027 April 2020 Summary of the April 15, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Emergency Core Cooling System Boron Distribution of the Design Certification Application ML20114E0392020-04-23023 April 2020 Meeting Summary for the April 21, 2020, Public Meeting to Discuss NuScale SDA Significance Review Process ML20072M7982020-03-19019 March 2020 Meeting Summary for the January 30, 2020, Public Meeting to Discuss NuScale SDA Regulatory Engagement Plan ML19282B1202019-10-12012 October 2019 Summary of January 9, March 5, April and August 1, 2019, Public Teleconference with NuScale Power, LLC, Regarding NON-LOCA TR RAI Responses (PROJ0769) ML19260G7352019-10-0707 October 2019 Summary of Public Meeting with NuScale to Discuss Response to RAI 9681 ML19240A3892019-10-0303 October 2019 Summary of January 8 and 29, 2019 and April 17, 2019, Public Teleconference with NuScale Power, LLC, Regarding Responses to Staff Request for Additional Information (52-048) ML19256B7002019-09-23023 September 2019 Summary of Public Meeting with NuScale to Discuss Test #107 and Test #47 in the Initial Test Program ML19260G5382019-09-23023 September 2019 Summary of 9-4-19 Public Meeting with NuScale to Discuss ITAAC ML19256B4942019-09-18018 September 2019 Summary of Public Meeting with NuScale to Discuss Radiation Shielding ITAAC ML19247C0342019-09-18018 September 2019 Summary of August 26 2019 Public Meeting with NuScale on Chapter 20 ML19252A0212019-09-12012 September 2019 August 27, 2019, Summary of Public Meeting with NuScale to Discuss Test #6, Test #47, and Test #50 in the Initial Test Program ML19255D2722019-09-12012 September 2019 Aug 26, 2019, Summary of Public Teleconference on NuScale Chapter 4 ML19228A2502019-08-27027 August 2019 Summary of Public Meeting with NuScale to Discuss Structural ITAAC ML19217A3082019-08-15015 August 2019 Summary of Public Meeting with NuScale to Discuss Initial Test Program Test #107 for the Remote Shutdown Workstation ML19212A6882019-08-12012 August 2019 Summary of Public Meeting with NuScale to Discuss Structural ITAAC ML19199A0152019-07-31031 July 2019 Summary of Public Meeting on Various ITAAC Issues ML19204A2772019-07-22022 July 2019 Discussion Topic July 22, 2019 Public Meeting to Discuss Interface Requirement Associated with the NuScale Design Certification Application ML19134A3632019-07-0808 July 2019 Summary of the April 9, 2019 Public Meeting to Discuss Accident Source Term (AST) Methodology and Post Exemption Associated with the NuScale Design Certification Application ML19176A2502019-07-0808 July 2019 Summary of June 17, 2019 Public Meeting with NuScale on MBDBE ML19151A3382019-06-0505 June 2019 Summary of Public Meeting with NuScale to Discuss ITAAC for the Module Protection System and Safety Display and Indication System ML18340A0132019-04-17017 April 2019 July 24, 2018, Summary of Category 1 Public Teleconference with Nuscale Power, LLC, Design Certification Application Section 3.7, Seismic Design, and 3.8, Design of Category I Structures. ML19084A1782019-04-16016 April 2019 Summary of Category 1 Public Teleconference with NuScale Power, LLC, to Discuss the NuScale Turbine Missile Analysis ML18333A3612019-04-11011 April 2019 & 08/28/18 Summary of Public Teleconference with Nuscale Power, LLC, Regarding Responses to Staff Requests for Additional Information ML19065A1242019-04-0505 April 2019 Summary of the February 14 and 27, 2019, Public Meeting to Discuss Accident Source Term (AST) Methodology Associated with the Nuscale Design Certification Application ML19085A0412019-03-26026 March 2019 02/13/2019 and 03/06/2019 Summary of Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Progress on Emergency Core Cooling System Valve Test Plans of the Design Certification Application ML19084A1362019-03-25025 March 2019 Summary of Meeting with NuScale Power, LLC to Discuss Nuclear Regulatory Commission Staff'S Options to Review Nuscale'S Design Certification Application, Tier 2, Chapter 20, Mitigation of Beyond-Design-Basis Events ML19017A1532019-02-0606 February 2019 Summary of Public Meeting with NuScale to Discuss Response to Request for Additional Information 8894 Question 03.11-15 ML19010A0932019-02-0404 February 2019 December 19, 2018, and January 9, 2019 - Summary of Public Meeting with NuScale to Discuss ITAAC Related to Electrical Penetration Assemblies ML18337A0192018-12-0606 December 2018 Summary of Category 1 Public Teleconference to Discuss Technical Specification Potential Phase 2 Open Items Associated with the NuScale Power, LLC Design Certification Application 2021-12-03
[Table view] Category:Meeting Summary
MONTHYEARML24004A0472024-01-31031 January 2024 U.S. Nuclear Regulatory Commission Summary of the November 16, 2023, Observation Public Meeting to Discuss NuScales Us600 Exemption Request ML20274A1082020-09-30030 September 2020 Summary of September 16, 2020 Public Meeting with NuScale Power, LLC to Discuss Operator Licensing Topics ML20216A5462020-08-19019 August 2020 Public Summary of the July 15, 2020 Pre-Submittal Meeting with NuScale, LLC, Regarding Its Plans to Submit a Topical Report Using Automated Test System ML20219A4112020-08-10010 August 2020 Summary of the May 28, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss Nuscale'S Standard Design Application Control Room Staffing Topical Report-0520-69851 ML20108F5082020-05-27027 May 2020 Summary of April 15 2020 Public Meeting on NuScale EPZ Sizing Methodology TR ML20114E2162020-05-0404 May 2020 Summary of the April 1, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Emergency Core Cooling System Boron Distribution of the Design Certification Application ML20114E0192020-04-27027 April 2020 Summary of the April 15, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Emergency Core Cooling System Boron Distribution of the Design Certification Application ML20114E0392020-04-23023 April 2020 Meeting Summary for the April 21, 2020, Public Meeting to Discuss NuScale SDA Significance Review Process ML20097C2292020-04-0707 April 2020 Summary of the March 9, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Emergency Core Cooling System Boron Distribution ML20072M7982020-03-19019 March 2020 Meeting Summary for the January 30, 2020, Public Meeting to Discuss NuScale SDA Regulatory Engagement Plan ML20072K5232020-03-11011 March 2020 Summary of the September 4, 2019, Public Teleconference with NuScale Power, Llc., Regarding Responses to Staff Requests for Additional Information (RAI) (52-048) ML19345E9362020-01-13013 January 2020 Summary of the December 9, 2019, Public Teleconference Meeting on Incorporation by Reference ML19302E4602019-10-29029 October 2019 July 22, 2019 Public Meeting - NuScale Interface Items for Penetrations and Ventilation Ducts ML19297C8772019-10-24024 October 2019 Summary of Public Meeting with NuScale to Discuss ECCS Test #47 ML19291A8872019-10-21021 October 2019 U.S. Nuclear Regulatory Commission Summary Report of Regulatory Close-out Audit of NuScale Power, LLC, Component Design Specification ML19282B1202019-10-12012 October 2019 Summary of January 9, March 5, April and August 1, 2019, Public Teleconference with NuScale Power, LLC, Regarding NON-LOCA TR RAI Responses (PROJ0769) ML19289C3982019-10-12012 October 2019 Summary of the January 9, 2019, March 5, 2019, April 23, 2019, and August 12, 2019, Public Teleconferences with NuScale Power, LLC ML19281B3982019-10-10010 October 2019 Summary of September 23, 2019 - Public Teleconference on NuScale Chapter 4 ML19281B2612019-10-0909 October 2019 Sept 25, 2019 - Public Teleconference on NuScale SDA ML19260G7352019-10-0707 October 2019 Summary of Public Meeting with NuScale to Discuss Response to RAI 9681 ML19240A3892019-10-0303 October 2019 Summary of January 8 and 29, 2019 and April 17, 2019, Public Teleconference with NuScale Power, LLC, Regarding Responses to Staff Request for Additional Information (52-048) ML19256B7002019-09-23023 September 2019 Summary of Public Meeting with NuScale to Discuss Test #107 and Test #47 in the Initial Test Program ML19260G5382019-09-23023 September 2019 Summary of 9-4-19 Public Meeting with NuScale to Discuss ITAAC ML19256B4942019-09-18018 September 2019 Summary of Public Meeting with NuScale to Discuss Radiation Shielding ITAAC ML19247C0342019-09-18018 September 2019 Summary of August 26 2019 Public Meeting with NuScale on Chapter 20 ML19252A0212019-09-12012 September 2019 August 27, 2019, Summary of Public Meeting with NuScale to Discuss Test #6, Test #47, and Test #50 in the Initial Test Program ML19255D2722019-09-12012 September 2019 Aug 26, 2019, Summary of Public Teleconference on NuScale Chapter 4 ML19242B8092019-08-30030 August 2019 Discussion Topic for August 13 and August 19, 2019, Public Meeting to Discuss AST Methodology and H2 and O2 Monitoring for the NuScale Design Certification Application ML19242B8802019-08-30030 August 2019 Discussion Topic for July 24, 2019, Public Meeting to Discuss Technical Specifications for the NuScale Design Certification Application ML19242B9932019-08-30030 August 2019 Discussion Topic May 14, 2019, Public Meeting to Discuss Accident Source Term (AST) Methodology Associated with the NuScale Design Certification Application ML19242C0162019-08-30030 August 2019 Discussion Topic May 8, 2019, Public Meeting to Discuss Post-Accident Sampling Exemption Request Associated with the NuScale Design Certification Application ML19242C4792019-08-30030 August 2019 Discussion Topics for the March 12 2019 Public Meeting to Discuss Neutron Absorbers ITAAC and COL Information Item 9.1-8 Associated with the NuScale Design Certification Application ML19228A2502019-08-27027 August 2019 Summary of Public Meeting with NuScale to Discuss Structural ITAAC ML19217A3082019-08-15015 August 2019 Summary of Public Meeting with NuScale to Discuss Initial Test Program Test #107 for the Remote Shutdown Workstation ML19212A6882019-08-12012 August 2019 Summary of Public Meeting with NuScale to Discuss Structural ITAAC ML19199A0152019-07-31031 July 2019 Summary of Public Meeting on Various ITAAC Issues ML19149A5052019-07-22022 July 2019 May 22 2019 NRC-NuScale Public Meeting Summary_Reduced Staffing Plan Validation Activities ML19171A1292019-07-17017 July 2019 Summary of Category 1 Public Teleconference with NuScale Power, LLC to Discuss of Section 3.9.4 Control Rod Drive System of the Design Certification Application ML19134A3632019-07-0808 July 2019 Summary of the April 9, 2019 Public Meeting to Discuss Accident Source Term (AST) Methodology and Post Exemption Associated with the NuScale Design Certification Application ML19176A2502019-07-0808 July 2019 Summary of June 17, 2019 Public Meeting with NuScale on MBDBE ML19178A3502019-06-27027 June 2019 Discussion Topics for April 09, 2019 Public Meeting with NuScale - Public Version ML19242B9342019-06-12012 June 2019 Discussion Topic for June 12, 2019, Public Meeting to Discuss Technical Specifications for the NuScale Design Certification Application ML19151A3382019-06-0505 June 2019 Summary of Public Meeting with NuScale to Discuss ITAAC for the Module Protection System and Safety Display and Indication System ML19087A2402019-05-0606 May 2019 Summary of March 5, 2019 Public Meeting on NuScale EPZ Sizing Methodology TR ML19106A0452019-04-24024 April 2019 Summary of Public Teleconference Held April 15, 2019, Regarding the Inspections, Tests, Analyses, and Acceptance Criteria in NuScale Power Llc'S Design Certification Application ML18340A0132019-04-17017 April 2019 July 24, 2018, Summary of Category 1 Public Teleconference with Nuscale Power, LLC, Design Certification Application Section 3.7, Seismic Design, and 3.8, Design of Category I Structures. ML18348A9442019-04-17017 April 2019 Summary of the Category 1 Public Teleconference with NuScale Power, LLC Design Certification Application Section 3.6.3, Leak-Before-Break Evaluation Procedures, and Section 3.9.4, Control Rod Drive System. ML18333A3612019-04-11011 April 2019 & 08/28/18 Summary of Public Teleconference with Nuscale Power, LLC, Regarding Responses to Staff Requests for Additional Information ML19065A1242019-04-0505 April 2019 Summary of the February 14 and 27, 2019, Public Meeting to Discuss Accident Source Term (AST) Methodology Associated with the Nuscale Design Certification Application ML19085A0412019-03-26026 March 2019 02/13/2019 and 03/06/2019 Summary of Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Progress on Emergency Core Cooling System Valve Test Plans of the Design Certification Application 2024-01-31
[Table view] Category:Memoranda
MONTHYEARML24004A0472024-01-31031 January 2024 U.S. Nuclear Regulatory Commission Summary of the November 16, 2023, Observation Public Meeting to Discuss NuScales Us600 Exemption Request ML20274A1082020-09-30030 September 2020 Summary of September 16, 2020 Public Meeting with NuScale Power, LLC to Discuss Operator Licensing Topics ML20199M1492020-08-19019 August 2020 Memo Transmitting Summary of the July 15, 2020 Pre-Submittal Meeting with NuScale, LLC, Regarding Its Plans to Submit a Topical Report Using Automated Test System ML20219A4112020-08-10010 August 2020 Summary of the May 28, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss Nuscale'S Standard Design Application Control Room Staffing Topical Report-0520-69851 ML20210M0772020-07-28028 July 2020 Memo - Category 1 Public Meeting - Audit Exit NuScale Emergency Core Cooling System Boron Redistribution Issue and Applicable Design Changes ML20160A2472020-07-28028 July 2020 Audit Summary Memo NuScale ECCS Boron Redistribution Issue ML20176A1582020-07-10010 July 2020 Audit Summary 2019 Memo NuScale Chapter 15 ML20108F5082020-05-27027 May 2020 Summary of April 15 2020 Public Meeting on NuScale EPZ Sizing Methodology TR ML20140A1892020-05-19019 May 2020 Public Meeting Summary May 14, 2020, NuScale Boron Redistribution Issue ML20114E2162020-05-0404 May 2020 Summary of the April 1, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Emergency Core Cooling System Boron Distribution of the Design Certification Application ML20114E0192020-04-27027 April 2020 Summary of the April 15, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Emergency Core Cooling System Boron Distribution of the Design Certification Application ML20114E0392020-04-23023 April 2020 Meeting Summary for the April 21, 2020, Public Meeting to Discuss NuScale SDA Significance Review Process ML20097C2292020-04-0707 April 2020 Summary of the March 9, 2020, Category 1 Public Teleconference with NuScale Power, LLC to Discuss the Emergency Core Cooling System Boron Distribution ML20072M7982020-03-19019 March 2020 Meeting Summary for the January 30, 2020, Public Meeting to Discuss NuScale SDA Regulatory Engagement Plan ML20076D8852020-03-16016 March 2020 Summary Report for the January 2019 Regulatory Audit of NuScale Chapter 20 ML20072K5232020-03-11011 March 2020 Summary of the September 4, 2019, Public Teleconference with NuScale Power, Llc., Regarding Responses to Staff Requests for Additional Information (RAI) (52-048) ML20059N6872020-03-0202 March 2020 Regulatory Audit Plan for NuScale DCA Chapters 6, 7, and 15 ML20054A0602020-02-28028 February 2020 Regulatory Audit Report for NuScale DCA Inadvertent Actuation Block (Iab) Operating Range Change ML20044E3522020-02-18018 February 2020 Memo to ACRS Rod Ejection Accident Methodology SER ML20049H1872020-02-18018 February 2020 Memo to ACRS Rea NON-LOCA TR SE ML20049H5442020-02-18018 February 2020 Letter to NuScale NON-LOCA TR SER ML20042E0242020-02-12012 February 2020 LOCA Topical Report Phase IV Audit Report - Public ML20042E0392020-02-12012 February 2020 NON-LOCA Topical Report Phases 3 and 4 Audit Report ML20010D1312020-01-16016 January 2020 Audit Summary Memo NuScale LOCA TR ML19345E9362020-01-13013 January 2020 Summary of the December 9, 2019, Public Teleconference Meeting on Incorporation by Reference ML19340A9712020-01-13013 January 2020 U.S. Nuclear Regulatory Commission Report of the Regulatory Follow-Up Audit Performed Between May 1, 2019, Through September 19, 2019, Regarding the Nuscale Stress and Fatigue Analysis of Major Design Components (Public) ML20006G4202020-01-0808 January 2020 ACRS Review of NuScale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapters 2, 8 and 12 ML19343C7562020-01-0808 January 2020 Generator Tube - Siet Audit Report (Public) ML20006G5752019-12-20020 December 2019 Proposed Recommendation for ACRS Review of NuScale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 8, Electric Power ML19331A3972019-12-20020 December 2019 U.S. Nuclear Regulatory Commission Report of the Regulatory follow-up Audit Performed Between 8/22/19 - 9/24/19, Regarding NuScale Power, Seismic Category I Equipment and Environment Qualification of Electrical Equipment Specifications ML19340A0152019-12-20020 December 2019 U.S. Nuclear Regulatory Commission'S Report of the Regulatory Audit Performed Between March 4, 2019, Through September 10, 2019, for Nuscale Power, LLC, Regarding Nuscale'S Power Module Leakage Flow Instability Analysis - Public Memo ML20006G4892019-12-19019 December 2019 Proposed Recommendation for ACRS Review of NuScale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 2, Site Characteristics and Site Parameters ML19340A0192019-12-19019 December 2019 U.S. Nuclear Regulatory Commission Report - Regulatory Audit Between 03/20/19 - 10/09/19 - Emergency Core Cooling System Valve Design Demonstration Testing and Follow-Up Items NuScale Power, LLC, Standard Plant Design Certification ML20006G7002019-12-19019 December 2019 Proposed Recommendation for ACRS Review of NuScale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 12, Radiation Protection ML19143A1042019-12-12012 December 2019 DCA - Safety Evaluation with No Open Items for Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors - Memo ML19168A1382019-12-12012 December 2019 LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 3, Design of Structures, Systems, Components and Equipment ML19169A2992019-12-11011 December 2019 Memo to ACRS Transmitting SER - Chapter 15 ML19169A1322019-12-10010 December 2019 Memo to ACRS Transmitting SER ML19165A2792019-12-0909 December 2019 Nuscale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 1, Introduction and General Discussion ML19168A1682019-12-0909 December 2019 Nuscale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 4, Reactor - Memo ML19168A2002019-12-0909 December 2019 Memo to ACRS Transmitting SER - Chapter 6 ML19169A3142019-12-0404 December 2019 LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 16, Technical Specifications - Memo ML19169A3322019-12-0404 December 2019 LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 20, Mitigation of Beyond-Design-Basis Events - Memo ML19169A2952019-11-26026 November 2019 Memo to ACRS Transmitting SER ML19319C2322019-11-26026 November 2019 U.S. Nuclear Regulatory Commission Summary Report of Regulatory follow-up Audit of Design Documents for NuScale Containment Isolation Valves and Reactor Safety Valves ML19270G6762019-11-22022 November 2019 Audit Summary for the Regulatory Audit of NuScale Power, LLC, Final Safety Analysis Report Chapter 15, Transient and Accident Analyses, (Memo) ML19169A2912019-11-22022 November 2019 Memo to ACRS Transmitting SER ML19169A1192019-11-22022 November 2019 Memo to ACRS Transmitting Phase 4 SER for Chapter 8 ML19162A2732019-11-15015 November 2019 LLC, Design Certification Application Memo - Safety Evaluation with No Open Items for Chapter 18, Human Factors Engineering ML19162A0962019-11-13013 November 2019 Nuscale Power, LLC, Design Certification Application - Safety Evaluation with No Open Items for Chapter 2, Site Characteristics and Site Parameters 2024-01-31
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October 3, 2019 MEMORANDUM TO: Samuel S. Lee, Chief Licensing Branch 1 Division of Licensing, Siting, and Environmental Analysis Office of New Reactors FROM: Rani L. Franovich, Senior Project Manager /RA/
Licensing Branch 1 Division of Licensing, Siting, and Environmental Analysis Office of New Reactors
SUBJECT:
SUMMARY
OF THE JANUARY 8, 2019, JANUARY 29, 2019, AND APRIL 17, 2019, PUBLIC TELECONFERENCES WITH NUSCALE POWER, LLC, TO DISCUSS THE REQUESTS FOR ADDITIONAL INFORMATION IN CHAPTER 15, TRANSIENT AND ACCIDENT ANALYSES, AND CHAPTER 19, PROBABILISTIC RISK ASSESSMENT AND SEVERE ACCIDENT EVALUATION, OF THE NUSCALE DESIGN CERTIFICATION APPLICATION (DOCKET NO.52-048)
On January 8, 2019, January 29, 2019, and April 17, 2019, representatives of the U.S. Nuclear Regulatory Commission (NRC) and NuScale Power, LLC (NuScale), held a public teleconference meeting. The purpose of the meeting was to discuss the NRC staffs Requests for Additional Information (RAI) Nos. 9512, 9504, 9507, 9483, and 9491, related to Final Safety Analysis Report (FSAR) Chapter 15, Transient and Accident Analyses; RAI 8840, related to FSAR Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation; and the wording of FSAR Section 19.2, Severe Accident Evaluation.
A complete copy of NuScales Design Certification Application is available on the NRC public Webpage at https://www.nrc.gov/reactors/new-reactors/design-cert/nuscale/documents.html. , Summary of the January 8, 2019, January 29, 2019, and April 17, 2019, Teleconference between the NRC Staff and NuScale, provides a summary of the topics discussed during the teleconference.
CONTACT: Rani L. Franovich, NRO/DLSE 301-415-7334
S. Lee 2 The agenda and list of meeting attendees are provided in Enclosures 2 and 3, respectively. The meeting notices are available in the NRCs Agencywide Documents Access and Management System, under Accession Nos. ML19007A046, ML19002A196, and ML19067A041.
Docket No.52-048
Enclosures:
- 1. Meeting Summary
- 2. Agenda
- 3. Attendees
ML19240A389 *via email NRC-001 OFFICE NRO/DLSE/LB1: PM NRO/DLSE/LB1: LA NRO/DSRA/SPRA: BC NRO/DSRA/SRSB: BC NAME RFranovich MMoore* MHayes* RKaras*
DATE 8/27/2019 10/03/2019 10/2/2019 10/1/2019 U.S. NUCLEAR REGULATORY COMMISSION
SUMMARY
OF JANUARY 8, 2019, JANUARY 29, 2019, AND APRIL 17, 2019 PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation Section 19.2, Severe Accident Evaluation On January 8, 2019, U.S. Nuclear Regulatory Commission (NRC) staff and NuScale Power, LLCs (NuScale) continued to discuss the language in Final Safety Analysis Report (FSAR)
Section 19.2, Severe Accident Evaluation. This discussion began November 13, 2018 (ML18331A062). The NRC staff reiterated its concern with NuScale assertions that the failure of the reactor pressure vessel (RPV) lower head is not physically realistic, an ex-vessel steam explosion is physically unrealistic, and failure of the containment lower head is not physically realistic. In response, NuScale offered the following:
- 1. NuScale proposed to replace physically unrealistic with cannot occur or does not occur because its analysis led to this outcome. The NRC staff asked NuScale to consider replacing absolute terms such as physically unrealistic and cannot occur or does not occur with probabilistic terms such as unlikely to reflect the phenomenological uncertainties identified by the staff. NuScale indicated it was disinclined to characterize failure of the RPV lower head as unlikely because the basis of its assertion was analytical, not probabilistic. The NRC staff also suggested the NuScale remove the word conservative from several passages in the FSAR because of phenomenological uncertainties not quantified in the analysis. NuScale indicated it would consider removing the word conservative.
- 2. NuScale considered revising the FSAR to acknowledge phenomenological uncertainties (critical heat flux, focusing effect, and intermetallic reactions) identified by the NRC staff and proposed to add language on page 19.2-20 of the FSAR to address uncertainties and an example of uncertainty such as critical heat flux.
- 3. With respect to an alternative acceptance criterion (i.e., that the NuScale design is unlikely to lead to a large release following a severe accident) for containment performance during severe accidents involving ex-vessel steam explosion, NuScale indicated its intent to revise the FSAR to align with the alternative acceptance criterion.
NuScale proposed to state that an ex-vessel steam explosion causing containment failure is unlikely to lead to a large release and to make conforming changes to reflect this statement in the FSAR discussion on ex-vessel steam explosion, including Table 19.1-32 on page 19.1-187. NuScale asked what metric the NRC staff is using to determine if a postulated release is large. The NRC staff responded it was using the 2.9 percent iodine release metric, which is NuScales large release metric for at-power accidents in FSAR Section 19.1.
- 4. With respect to in-vessel steam explosion and the NRC staffs confirmatory analysis, NuScale proposed to revise the FSAR to indicate its analysis assumed molten debris.
The NRC staff responded that the proposed change partially addresses this issue but falls short of acknowledging additional phenomenological uncertainty, including the Enclosure 1
possibility that corium temperatures could be higher than what was assumed in NuScales probabilistic analysis for in-vessel steam explosion. NuScale indicated it understood the NRC staffs concerns and proposed to revise the FSAR and replace assuming molten debris with assuming that pre-mixing was equivalent to that of molten debris, even though the debris was not hot enough to be molten.
Subsequent to the January 8, 2019, public meeting, and in response to issues raised therein, NuScale provided a mark-up of FSAR Chapter 19 via email on January 18, 2019.
On January 29, 2019, the NRC staff and NuScale met again to discuss the revisions to the language in FSAR Chapter 19. The NRC staff indicated that the FSAR Chapter 19 mark-up resolved the NRC staffs issues. In addition, the NRC staff proposed two clarifying changes to the FSAR Chapter 19 mark-up.
- 1. To delete the following passage in FSAR Section 19.1:
These phenomena are not included in the [containment event tree (CET)]
consistent with the approach taken in NUREG-1524 (Reference 19.1-64) in which phenomena that are judged to be physically impossible, vanishingly small, or very unlikely have a probability of less than 1E-3.
Because this probability is small with respect to the large release frequency (LRF) and the conditional containment failure probability (CCFP) safety goal of less than 0.1, such events are not explicitly included in the CET.
The NRC staff expressed concern with this passage in light of phenomenological uncertainties, which are addressed in FSAR Section 19.2. NuScale agreed to consider revising this passage and repeated its assertion that the containment vessel is sufficiently robust to withstand severe accident challenges. The NRC staff responded that it did not believe this deletion would detract from NuScales assertions.
- 2. To retain the following original language in FSAR Section 19.2 regarding hydrogen generation and control: In summary, over-pressurizing the NuScale containment vessel due to hydrogen combustion is physically unrealistic due to the very limited oxygen concentration before and after postulated severe accidents. The NRC staff indicated it believed the original language more clearly described NuScales hydrogen combustion analysis results. The phenomenological uncertainties that the NRC staff previously identified were associated with in-vessel retention and steam explosion and not with hydrogen combustion. NuScale agreed to consider restoring the original language.
Request for Additional Information No. 8840 In its response to Request for Additional Information (RAI) No. 8840, NuScale assumed that containment isolation is not required during various loss-of-coolant-accident (LOCA) events inside containment. This assumption is not reflected in the FSAR. The plant design is for containment isolation to actuate when there is a LOCA type event in containment as described in parts of the FSAR. However, since this is not modeled in the probalistic risk assessment, while potentially acceptable, the NRC staff believes the assumption should be captured in the FSAR (likely in the key assumptions table in Chapter 19).
2
On April 17, 2019, the NRC staff asked the following three questions about NuScales response to RAI No. 8840:
- 1. Did the evaluation consider operation of the Containment Evacuation System (CES) vacuum pumps continuing to take a suction on containment during LOCA inside containment events?
Response: There is no discussion in the FSAR about the CES system shutting down.
NuScale responded that its evaluation did not consider the CES vacuum pumps and that the running pumps would not impact the evaluation. NuScale agreed to submit a supplemental response to the RAI and describe the impact of the CES vacuum pumps on containment during a LOCA inside containment scenario.
- 2. Why was 100 degrees Fahrenheit (°F) assumed rather than 140 °F?
Response: The response states the simulation assumed a normal pool operating temperature of 100 °F. However, Technical Specification (TS) 3.5.3, Revision 2, states that the pool should be maintained at less than 140°F.
Does the analysis change significantly when pool temperature is assumed to be 140 °F?
Response: NuScale responded that the pools TS limit has been reduced from 140 °F to 110 °F and agreed to provide the rationale for assuming a pool temperature of 100 °F for its evaluation in the supplemental response to the RAI.
- 3. How much of the pool was assumed in the simulation that was referred to in the July 2018 RAI response?
Response: NuScale responded that the simulation assumed the LOCA event involving one module concurrent with decay heat load from the other shutdown modules. The simulation assumed the entire pool excluding the spent fuel pool was available to provide cooling. NuScale added that it had performed a sensitivity analysis on the evaluation, which assumed only one twelfth of the pool was available and showed no core damage in the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the module experiencing the LOCA event.
Chapter 15, Accident and Transient Analyses On January 29, 2019, the NRC staff and NuScale discussed the following RAIs related to Chapter 15.
RAI 9512, Question 15.04.03-2 The markup of Table 15.4-33, Key Inputs for Key Inputs for Limiting Control Rod Assembly Misalignment, shows that the nominal core inlet temperature of 487.4 °F, which corresponds to a low-biased RCS average temperature of 535 °F, is biased high by 10 °F. It is not clear to the NRC staff that the core inlet temperature is biased in a limiting manner, as it appears the net effect of the nominal (biased low) temperature plus the 10 °F bias corresponds to the nominal RCS average temperature of 545 °F. In other words, the combination of the nominal (biased-low) temperature and the applied bias appear to cancel out. Furthermore, the NRC staff notes that a biased-high RCS average temperature is typically limiting for minimum critical heat flux ratio (MCHFR). The NRC staff requested NuScale to explain the logic for the choice of the 3
nominal and biased core inlet temperature values, including why the nominal core inlet temperature does not correspond to the nominal RCS average temperature of 545 °F.
The NuScale responded that the control rod assembly misalignment analysis is a steady-state analysis that requires no interface with the transient code (NRELAP5). It is evaluated using the subchannel analysis methodology (VIPRE-01) with low RCS flow bias conditions, including the core inlet temperature corresponding to minimum RCS flow. Because the NuScale design is based on natural circulation, core inlet temperature and RCS flow are coupled for a given power level. In other words, the nominal 487.4 °F core inlet temperature corresponds to the normal core inlet temperature when RCS flow rate is biased low. Simultaneously assuming a core inlet temperature corresponding to nominal RCS flow and biasing RCS flow low would be unphysical because of the relationship between RCS flow and RCS temperature. The applicant also noted that differences exist in the treatment of RCS temperature between VIPRE-01 and NRELAP5 because RCS temperature inputs are specified differently in the two codes. NRELAP5 is based on an average RCS temperature control scheme, whereas the core inlet temperature is specified directly as a boundary condition in VIPRE-01.
RAI 9504, Question 15.04.07-3 The response to this RAI similarly states that the core inlet temperature bias corresponds to the minimum RCS average temperature range specified in FSAR Table 15.0-6. The NRC staff asked why a minimum RCS average temperature is used when a high bias is typically limiting for MCHFR. NuScale reiterated that the control rod assembly misalignment analysis is a steady-state analysis that requires no interface with the transient code (NRELAP5). It is evaluated using the subchannel analysis methodology (VIPRE-01) with low RCS flow bias conditions, including a consistent core inlet temperature corresponding to minimum RCS flow.
The NRC staff further noted that the response stated that NuScale identified a deficiency in the methodology for determining the limiting radial peaking augmentation factor. The NRC staff requested NuScale to please discuss the nature of the deficiency and asked if the deficiency affects any other events that utilize a radial peaking augmentation factor. NuScale clarified that it had made an incorrect assumption in the initial calculation. Specifically, NuScale had analyzed all possible misloaded fuel assembly configurations (misloads) at 25 percent power to identify those that are undetectable by the in-core instrumentation. The applicant had originally assumed that the misloads that were barely within the undetectable threshold (i.e.,
those closest to being detected) would result in the largest radial peaking augmentation factor and therefore only calculated the radial peaking augmentation factor for those cases. However, this assumption was not correct.
NuScale acknowledged the error and revised the calculation used to investigate all undetectable misloads and identify the limiting radial peaking augmentation factor. NuScale confirmed that the error applies only to this particular event and affects no other methodologies.
RAI 9507, Question 15.04.01-5 The response states that the initial RCS flow for the event corresponds to the minimum flow rate to keep the module protection system from actuating on the low RCS flow analytical limit of 1.7 ft3/s. However, based on an audit of the underlying calculation note, it appears the case for limiting MCHFR was run at a flow roughly two times the analytical limit. Please clarify the initial flow rate for the limiting case. NuScale responded that it intended to consider conditions down to the low RCS flow analytical limit (i.e., zero power) for the uncontrolled control rod withdrawal 4
from subcritical or low power analysis. However, the zero-power case isnt limiting because initial conditions, particularly initial power, drive the power overshoot. NuScale participants stated they gained a better understanding of the staffs RAI and would submit a supplemental response that includes the actual RCS flow rate used in the analysis.
RAI 9483, Question 15.01.01-7 The NRC staff found NuScales responses to the first and third staff potential issues in this question to be satisfactory but requested clarification regarding the potential for SG overfill to degrade decay heat removal system (DHRS) capability resulting from reduced condensation heat transfer area. As stated in the original RAI, the NRC staff observed that the highest steady-state SG level may not lead to the highest transient SG level and was therefore concerned about degraded DHRS capability should the level in both SGs grow as a result of a more limiting combination of initial conditions (i.e., initial conditions that maximize the transient SG level). In response to the NRC staffs second observation in the RAI, NuScale stated that a degraded or failed single DHRS loop will not prevent the other loop from performing as required.
However, the staff notes that the thermal-hydraulic analysis of the DHRS shows that DHRS performance falls rapidly between 75 and 90 percent SG level. In the overfill scenario discussed in the RAI, the level in both SGs was about 80 percent, meaning both DHRS loops could potentially be degraded.
Although the existing analysis of the overfill case showed that the DHRS was able to perform its safety function, if the transient SG level were to grow, DHRS capability could be further degraded in both trains, especially if tube plugging, fouling, and noncondensable gases were not previously considered in the analysis. For these reasons, the NRC staff requested NuScale to explain why SG overfill is not a concern for an increase in feedwater flow event, or any other event. NuScale responded that for this transient, it had calculated total heat removal through both DHRS trains. Although increasing SG level causes decreased DHRS heat removal, the total energy removed through both DHRS trains still exceeds the heat entering the SGs.
Therefore, the DHRS is still removing decay heat effectively, maintaining a decrease in primary and secondary pressures. Noncondensable gases are only important at lower-pressure conditions. NuScale added that it had initiated an Engineering Change Notice (ECN) to further investigate the overfill case by assuming additional tube plugging and other biases.
RAI No. 9491 The NRC staff informed NuScale that its response to RAI No. 9491 partially resolved concerns raised in the RAI and requested clarification to resolve the following issues:
- 1. According to the response to RAI No. 9491, discussion of several analyses (including an analysis of a decrease in reactor coolant inventory with decreasing pressure) was deleted from FSAR Section 15.9, Rev. 2. In its RAI response, NuScale stated that the decrease in reactor coolant inventory with decreasing pressure analysis is replaced with a discussion that notes how the Module Protection System (MPS) mitigates the occurrence of divergent flow oscillations. According to the applicants stability analysis methodology as described in Topical Report TR-0516-49417-P, Evaluation Methodology for Stability Analysis of the NuScale Power Module (the Stability TR), a depressurization event is analyzed to demonstrate the effectiveness of the MPS to actuate a trip and shutdown the reactor before unstable flow oscillations occur. The Stability TR provides such an analysis to illustrate that the protection of the exclusion region afforded by the MPS results in a trip and reactor shutdown prior to instability.
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However, the NRC staff noted that the Stability TR analysis does not reflect the final design of the NuScale power module and associated systems. NuScale agreed to submit a revised RAI response with a revised FSAR Section 15.9 to reinclude the depressurization event consistent with the methodology described in the Stability TR.
- 2. Section 15.9.3 of FSAR, Revision 2, states that reactor trip is not credited in the stability operational event analysis. However, the NRC staff noted that the long term stability (LTS) solution requires operation of the MPS to enforce an exclusion region. Therefore, the applicants LTS solution inherently credits a trip. The applicant agreed to submit a supplemental response to the RAI with revised language in the FSAR to accurately reflect the nature of the credit taken for the MPS trip. The NRC staff also requested that NuScale replace the word credited in section 15.9.3 statements such as, the reactor trip is not credited in the stability operational event analysis. The NRC staff indicated that simulated would be an adequate and a more accurate replacement for credited.
The applicant indicated it would consider this revision as well.
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U.S. NUCLEAR REGULATORY COMMISSION
SUMMARY
OF JANUARY 8, 2019, JANUARY 29, 2019, AND APRIL 17, 2019 PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC MEETING AGENDA Tuesday, January 8, 2019 Time Topic Speaker Draft Final Safety Analysis Rector (FSAR) 3:00 pm - 4:00 pm Section 19.2 Mark-up, Containment NuScale/NRC Performance Issues Tuesday, January 29, 2019 Time Topic Speaker Chapter 15 Requests for Additional Information 1:00 - 2:00 pm NuScale/NRC (RAI) 9512, 9504, 9507 and 9483 2:00 - 3:00 pm Chapter 19 RAI 8840, FSAR Section 19.2 NuScale/NRC 3:00 - 4:00 pm Chapter 15 RAI 9491 NuScale/NRC Wednesday, April 17, 2019 Time Topic Speaker 3:00 - 4:00 pm Chapter 19 RAI 8840 NuScale/NRC Enclosure 2
LIST OF ATTENDEES NuScale January 8, 2019 January 29, 2019 April 17, 2019 G. Becker G. Becker S. Bristol B. Bristol B. Bristol J. Curry A. Child S. Bristol P. Infanger J. Curry M. Byram E. Mullin B. Galyean A. Callaway R. Gamble T. Codington P. Infanger J. Curry E. Mullin Y. Farawila S. Weber B. Galyean R. Goff B. Hayden P. Infanger M. McCloskey E. Mullin K. Rooks D. Throckmorton S. Weber C. Williams NRC Staff January 8, 2019 January 29, 2019 April 17, 2019 H. Esmaili R. Franovich O. Ayegbusi R. Franovich R. Karas M. Hayes M. Hayes J. Schmidt O. Tabatabai J. Martin A. Siwy J. Schaperow R. Skarda Public January 8, 2019 January 29, 2019 April 17, 2019 S. Fields Enclosure 3