ML18331A062
| ML18331A062 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 11/29/2018 |
| From: | Rani Franovich NRC/NRO/DLSE/LB1 |
| To: | Samson Lee NRC/NRO/DLSE/LB1 |
| Franovich R L/NRO/7334 | |
| References | |
| Download: ML18331A062 (7) | |
Text
November 29, 2018 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 NOVEMBER 13, 2018, PUBLIC TELECONFERENCES WITH NUSCALE POWER, LLC, TO DISCUSS THE WORDING OF NUSCALES CHAPTER 19, PROBABILISTIC RISK ASSESSMENT AND SEVERE ACCIDENT EVALUATION, OF THE NUSCALE DESIGN CERTIFICATION APPLICATION (DOCKET NO.52-048)
On November 13, 2018, 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) Nos. 9138 and 9043 related to Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, and the wording of Section 19.2, Severe Accident Evaluation, of the NuScale Final Safety Analysis Report.
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 November 13, 2018, 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 No. ML18296A108.
Docket No.52-048
Enclosures:
- 1. Meeting Summary
- 2. Agenda
- 3. Attendees
ML18331A062 *via email NRC-001 OFFICE NRO/DLSE/LB1: PM NRO/DLSE/LB1: LA NRO/DSRA/SPRA: BC NAME RFranovich MMoore* MHayes DATE 11/29/2018 11/27/2018 11/29//2018 OFFICE OGC NAME JMartin*
DATE 11/29/2018 U.S. NUCLEAR REGULATORY COMMISSION
SUMMARY
OF NOVEMBER 13, 2018 PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC NuScale Power, LLC Final Safety Analysis Report, Section 19.2, Severe Accident Evaluation During the meeting, the U.S. Nuclear Reglatory Commission (NRC) staff referred to NuScale NuScale Power, LLCs (NuScale) response to a request for additional information (RAI) No.
9138. The NRC staff described its independent assessment of NuScale reactor pressure vessel (RPV) and containment performance for severe accident challenges as presented in Final Safety Analysis Report (FSAR) Section 19.2. In this section, NuScale concluded that failure of the 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. The NRC staff presented the results of its independent review and questioned NuScales technical basis for these conclusions based on the following:
- 1. A pour of corium into the RPV lower plenum could lead to steam-explosion-induced containment upper head failure (i.e., alpha-mode failure). The NRC staff calculated the probability of NuScale containment upper head failure caused by in-vessel steam explosion using the method in NUREG/CR-5030, An Assessment of Steam-Explosion-Induced Containment Failure, February 1989. The NRC staff considered thermal-to-mechanical-energy conversion ratios reflecting more recent analysis and experiments and concluded that, while this failure mode is possible, an in-vessel steam explosion is unlikely to lead to containment upper head failure.
- 2. A pour of corium into the RPV lower plenum could lead to RPV lower head failure due to contact with corium. The NRC staff calculated heat fluxes to the RPV lower head using the method in NUREG/CR-6849, Analysis of In-Vessel Retention and Ex-Vessel Fuel Coolant Interaction for AP1000, August 2004. The heat flux from the ceramic corium pool independently calculated by the NRC staff is lower than the applicants analyzed value. The heat flux from the corium metal layer independently calculated by the staff is comparable to that in the applicants analysis. Based on uncertainties with the critical heat flux, the focusing effect of a metal layer on top of the molten core debris, and the potential for inter-metallic reactions (e.g., exothermic reaction of molten zirconium with the RPV lower head steel), the staff cannot confirm that failure of the reactor pressure vessel lower head is not physically realistic.
- 3. A pour of corium into the containment lower plenum could lead to steam-explosion-induced containment upper head failure. The staff reviewed results of its independent MELCOR confirmatory analysis and determined that at the time postulated containment upper head failure would occur, the iodine airborne in the RPV and containment would be lower than the applicants large release definition. Also, several meters of water would have accumulated in the containment and could provide scrubbing. Although the staff could not confirm that an ex-vessel steam explosion is physically unrealistic, the staff did determine that radiological release from steam-explosion-induced containment upper head failure is unlikely to be large based upon the low inventory of airborne iodine coupled with radionuclide scrubbing provided by water in the containment.
Enclosure 1
- 4. A pour of corium into the containment lower plenum could lead to containment lower head failure due to contact with corium. The staff reviewed results of the applicants MELCOR and MACCS analysis of module drop accidents, which involve severe core damage occurring with a failed containment lying on the pool floor. The maximum dose predicted by the staff from module drop severe accidents is below the applicants large release definition. The staff determined that a release from the containment lower head failure caused by contact with corium is unlikely to be large.
The staff stated that it does not seek more detailed analysis by NuScale. However, the phenomenological uncertainties and lack of more detailed experimental and analytical research preclude the staff from confirming the FSAR conclusions that the conditional containment failure probability is less than 0.1 assuming there is core damage, and the containment will provide a leak-tight barrier for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with no uncontrolled releases after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Nevertheless, failure of containment by ex-vessel steam explosion, or containment lower head failure by contact with corium, is unlikely to lead to a large release based on the staffs independent analysis.
The staff summarized that it cannot confirm certain statements in the FSAR such as Failure of the 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. Because any design certification rule would incorporate and approve the FSAR as written, the staff explained to NuScale that alignment between the language in the FSAR and the staffs safety evaluation report is desired to avoid future confusion regarding the technical basis for design certification.
The staff requested that NuScale consider revising the FSAR to acknowledge the phenomenological uncertainties identified by the staff. An alternative is to take exception to assertions in the FSAR that the staff could not confirm, which is less desirable because it complicates the technical basis for design certification.
The staff also conveyed that it is considering 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 involing ex-vessel steam explosion. Should an alternative basis for a staff safety finding be tenable, the NRC staff would seek a revision to the FSAR to align with the staffs safety evaluation report. The applicant agreed to consider the staffs preferred option, which would involve a potential supplemental response to RAI No. 9138.
The staff stated it hoped to have a clear path forward on either option by the end of November.
The applicant asked if there were any other concerns with the applicants analysis as described in the FSAR. The NRC staff responded that, for in-vessel steam explosion, the staffs confirmatory analysis a) included higher corium temperatures caused by late-phase melt progression uncertainties, and b) used a lower mechanical energy needed to fail the RPV upper head because it modeled the interaction between the slug and the RPV upper head as that of one solid object hitting another solid object.
2
U.S. NUCLEAR REGULATORY COMMISSION
SUMMARY
OF NOVEMBER 13, 2018 PUBLIC TELECONFERENCE WITH NUSCALE POWER, LLC MEETING AGENDA Time Topic Speaker 2:00 pm - 3:00 pm Language in NuScale FSAR Section 19.2 NRC/NuScale Enclosure 2
LIST OF ATTENDEES NuScale NRC Staff Public G. Becker K. Coyne None S. Bristol H. Esmaili A. Childs R. Franovich J. Curry M. Hayes B. Galyean I. Jung P. Infanger S. Lee E. Mullin J. Martin N. Wahlgren J. Schaperow S. Weber C. Williams Enclosure 3