ML20072K523
| ML20072K523 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 03/11/2020 |
| From: | Rani Franovich NRC/NRR/DNRL/NRLB |
| To: | Michael Dudek NRC/NRR/DNRL/NRLB |
| Franovich R , 415-7443 | |
| References | |
| Download: ML20072K523 (7) | |
Text
March 11, 2020 MEMORANDUM TO: Michael I. Dudek, Chief New Reactor Licensing Branch Division of New and Renewed Licenses Office of New Reactor Regulation FROM: Rani L. Franovich, Senior Project Manager /RA/
New Reactor Licensing Branch Division of New and Renewed Licenses Office of New Reactor Regulation
SUBJECT:
SUMMARY
OF THE SEPTEMBER 4, 2019, PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC., TO DISCUSS CHAPTER 19, PROBABILISTIC RISK ASSESSMENT AND SEVERE ACCIDENT EVALUATION, OF THE NUSCALE DESIGN CERTIFICATION APPLICATION On September 4, 2019, representatives of the U.S. Nuclear Regulatory Commission (NRC) and NuScale Power, LLC (NuScale)., held a public teleconference meeting. The purpose of this meeting was to discuss NuScales responses to the NRC staffs Request for Additional Information (RAI) 9659 related to Chapter 19, Probabalistc Risk Assessment and 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 September 4, 2019, Teleconference between the NRC staff and NuScale, provides a summary of the topics discussed during the teleconference.
CONTACT: Rani L. Franovich, NRR/DNRL 301-415-7334
M. Dudek 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 (ADAMS), under Accession No. ML19217A096.
Docket No.52-048
Enclosures:
As stated cc w/enclosures: DC Listserv
ML20072K523 *via email NRR-106 OFFICE DNRL/NRLB: PM DNRL/NRLB: LA DRA/APLC: BC DNRL/NRLB: PM NAME RFranovich CSmith SRosenberg* RFranovich DATE 12/26/2019 01/13/2020 03/11/2020 03/11/2020 U.S. NUCLEAR REGULATORY COMMISSION PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC HELD ON SEPTEMBER 4, 2019 Meeting Summary Request for Additional Information (RAI) No. 9659 On September 4, 2019, the U.S. Nuclear Regulatory Commission (NRC) staff met with NuScale Power, LLC., (NuScale) to discuss NuScales April 11, 2019, response to RAI 9659 (see ML19101A453) and Revision 2 of the Final Safety Analysis Report (FSAR), Section 19.1.7.4, Insights Regarding Low Power and Shutdown for Multi-Module Operation In FSAR Section 19.1.7.4, NuScale described additional damage to operating modules if a module moved for refueling is dropped and strikes the top of an operating module. Revision 2 of FSAR Section 19.1.7.4 added additional pipe breaks may occur, leading to [chemical and volume control system] CVCS line break outside of containment.
The NRC staff explained that Standard Review Plan Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, Revision 3, states that the NRC staff will verify that the applicant has: (1) used a systematic process to identify accident sequences, including significant human errors, that lead to multiple module core damages or large releases; and (2) selected alternative features, operational strategies, and design options to prevent these sequences from occurring and demonstrate that these sequences are not significant contributors to risk.
The NRC staff summarized RAI 9659, which requested NuScale to explain how multi-module risk insights for the struck module are unaffected by the additional damage described in Revision 2 of FSAR Chapter 19. In its response, NuScale stated that a dropped module is unlikely to impact an operating module based on geometry. The steel module platform columns and cross braces minimize the potential for significant impacts on an operating module. The chemical and volume control system (CVCS) containment isolation valves (CIVs) are located on top of the containment vessel head and under the module platform. The probability of failure-to-close of the CVCS CIVs is judged by NuScale, to be small. NuScale compared the frequency of a module drop to a single CVCS line break frequency without performing a mechanistic assessment of potential to damage other systems, structures and components (SSCs) (e.g.,
containment flood and drain system (CFDS) line, CVCS CIVs).
During the September 4, 2019, public teleconference, the NRC staff expressed concern that the comparison may not be valid and asked NuScale if risk insights may be affected because a moving module dropped upon an operating module could lead to a significant large release sequence given NuScales risk profile. The NRC staff added that, based on the current design of the reactor building crane (RBC), associated operator error could be a significant risk contributor. NuScale did not dispute this. NuScale responded that the RBC probabilistic risk assessment (PRA) was used to develop a module drop frequency and was not broken into basic events. NuScale added that the RBC crane including its operational aspects, was scoped into the Design Reliability Assurance Program (DRAP). NuScale said that the DRAP will ensure that the RBC reliability is maintained, including operator reliability. The NRC staff agrees that DRAP is intended to provide reasonable assurance that risk-significant SSCs do not degrade to an unacceptable level during plant operations as stated in SECY-95-132, Policy and Technical Enclosure 1
Issues Associated With the Regulatory Treatment of NonSafetySystems (RTNSS) in Passive Plant Designs. However, the Severe Accident Policy Statement [50 FR 32138; August 8, 1985]
directs advanced reactors designers to complete a PRA to search for vulnerabilities and insights. The Severe Accident Policy Statement states, the Commission recognizes the importance of such potential contributors to severe accident risk as human performance.
The NRC staff noted that FSAR Table 19.1-67, Module Drop Initiating Events and Mitigating Features, presents module drop initiating events caused by operator errors, yet the FSAR does not mention the contribution of operator errors leading to module drop. The NRC staff added that there could be other operator errors leading to module drop that are not included in the RBC PRA (e.g., the failure of the crane operator to correctly connect the module lift adapter to the nuclear power module lifting fixture was not included in the RBC PRA).
The NRC staff performed a sensitivity study on the large release frequency considering the uncertainties in the module drop frequency and conditional probabilities of striking and damaging SSCs associated with an operating module (e.g., CFDS line, CVCS CIVs) and found that a dropped module could lead to a significant large release sequence given the potential SSC impacts. The NRC staff also looked at cutsets from the RBC probabilistic risk assessment and found that the operator errors could be risk significant based on TR-0515-13952-NP-A, Risk Significance Determination, which is NuScales approved methodology to determine risk significance. The NRC staff acknowledged that the RBC reliability assessment was performed to develop an initiating event frequency for a module drop, and a combined operating license (COL) applicant would validate the key assumptions and PRA data then update the COL PRA per COL Item 19.1-8.
NuScale agreed to revise FSAR, Table 19.1-71, Key Assumptions for the Low Power and Shutdown Probabilistic Risk Assessment. Specifically, NuScale agreed to document the key assumption(s) related to operator actions that were used to estimate the reactor building crane reliability and that these assumptions will be verified by the combined license applicant per COL Item 19.1-8. The NRC staff found the proposed approach to be a reasonable path forward.
No members of the public were in attendance.
U.S. NUCLEAR REGULATORY COMMISSION PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC HELD ON SEPTEMBER 4, 2018 Meeting Agenda Time Topic Speaker 1:00 p.m. Introductions NRC 1:00 p.m. - 3:30 p.m. FSAR Chapter 19 RAI 9659 NRC/NuScale 3:30 p.m. Adjourn All Enclosure 2
U.S. NUCLEAR REGULATORY COMMISSION PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC HELD ON SEPTEMBER 4, 2018 List of Attendees Name Organization J. Curry NuScale Power, LLC. (NuScale)
B. Galyean NuScale M. Melton NuScale R. Norris NuScale N. Wahlgren NuScale S. Weber NuScale C. Williams NuScale G. Cranston U.S. Nuclear Regulatory Commission (NRC)
R. Franovich NRC B. Green NRC T. Nakanishi NRC M. Pohida NRC
- No members of the public were in attendance.
Enclosure 3