ML20151S462
| ML20151S462 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 04/21/1988 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8804280334 | |
| Download: ML20151S462 (6) | |
Text
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NUCLEAR REGULATORY COMMISSION 3f ij WASHINGTON D. C. 20555 Apri1 21, 1988 Docket No. 50-601 APPLICANT:
Westinghouse Electric Corporation FACILITY:
RESAR SP/90
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS THE PRA FOR RESAR SP/90 On March 31, 1988, representathes of the NRC, Brookhaven National Laboratories (BNL), and Westinghouse met at the Westinghouse Energy Complex in Monroeville, Pennsylvania to discuss the probabilistic risk assessment (PRA) for the RESAR SP/90. Enclosure 1 is a list of attendees. is the agenda followed during the meeting.
The first part of the meeting concerned BNL's review of the analysis of core melt probability (commonly referred to as the "front end" portion of the PRA).
The applicant made a brief presentation concerning the safety systems of RESAR SP/90, with particular emphasis on the integrated protection system (IPS). The participants then discussed the concerns raised in the staff's March 21, 1988 draft SER. The remainder of the meeting concerned review results of the consequence analysis of the PRA (or the "back end" portion of the PRA). The following is a sumary of the staff's and BNL's concerns:
1.
The PRA did not adequately address the potential for dependent failure of the engineered safety features actuation system (ESFAS) during an ATWS event. Westinghouse indicated its intent to include appropriate analysis in the revised PRA to be submitted during the final design approval (FDA) review.
2.
During the development of the FRA, the design of the IPS was insufficiently complete to be appropriately modelled in the PRA.
In lieu of this model, Westinghouse used typical estimates of the reliability of Westinghouse reacter trip breakers to represent the reliability of the IPS. Westinghouse indicated its intent to provide the detailed model of the IPS in the FDA PRA.
3.
A recent study by BNL has shown that accumulators can be a large contributor to core melt frequency due to failure of the high pressure / low pressure interface, and subsequent LOCA release.
The RESAR SP/90 design includes four accumulators and four core reflood tanks for which this concern is ap>l1 cable. Westinghouse indicated the design of the core reflood tan <s (small (4") discharge line with an orifice in each line) is such that such a break would be a small break LOCA. Westinghouse indicated its intent to address this concern at the FDA stage of design since the issue is new and has not undergone full review by the NRC or industry.
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- 4. is a list of assumptions and success criteria for which insufficient supporting analysis was provided. Westinghoisse agreed' to provide this information as it evolved during the development of the facility design.
S.
BNL noted that the RESAR SP/90 PRA did not address concerns with the direct heating (DH) effect and H burning since the applicant utilizedtheMAAPcontainmentfa$1ureanalysis(whichdoesnot consider the DH effect).
6.
BNL was concerned that certain input to the CRAC2 code appeared to have been entered incorrectly. Westinghouse agreed to review the matter.
The staff and BNL indicated that the NRC would need to determine which, if 3
any, of the open issues would require resolution prior to -issuance of the PDA, and which could be resolved during the FDA review.
T
.a s nyon, Project Manager Standard zation and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation l
n
.. April 21, 1988
- 4. is a list of assumptions and success criteria for wHeh insufficient supporting analysis was provided. Westinghouse agreed to provide this information as it evolved during the development of the facility design.
5.
BNL noted that the RESAR SP/90 PRA did not address concerns with the direct heating (DH) effect and H, lure analysis (which does not burning since the applicant utilized the MAAP containment fai consider the DH effect).
t 6.
BNL was concerned that certain input to the CRAC2 code appeared to have been entered incorrectly. Westinghouse agreed to review the matter.
The staff and BNL indicated that the NRC would need to determine which, if any, of the open issues would require resolution prior to issuance of th( PDA, and which could be resolved during the FDA review.
t)rfginrti Signed By:
Thomas J. Kenyon, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects l
Office of Nuclear Reactor Regulation i
, Distribution:,,
TXenyon HBClayton (Region I)
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l ENCLOSURE I MEETING ATTENDANCE LIST RESAR SP/90 PRA MARCH 31, 1988 NAME AFFILIATION T. Vande Venne Westinghouse H.H. Shannon W NTSD S.S. Tsai W NTSD W.M. Schivley R NTSD-Nuclear Safety Tom Kenyon NRC, NRR David Sharp
.W NTSD Trevor Pratt UNL Tsong-Lun Chu BNL-T.L. Schulz W NTSD S. Sancaletar 9 NTSD
+
Bruce Cook 9 NTSD i
+
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1 Pv2M Agenda for March 31st MeetirxJ W/NRC/RG Di='immico en SP/90 PSS (8:30 AM - Oorx:lusicn)
- 1. Introducticms/ opening Remarks / Meeting Goals (W. M. Schivley)
- 2. Brief presentaticn of H SP/90 Safety Systems (T. van de Venne)
(approx. 45 min.)
- 3. Presentaticn of SP/90 IPS System Design (Bnce Cook)
(approx. 30 min.)
- 4. Dim'ianicn on IPS Model o What is a satisfactory model?
Millstcne PFA, WCAP-10271 o Potential dependant failure of ESTAS in an A33 What systems may be available upcn a total loss of vital AC?
o
- 5. Interfacing Systems IDCA through ammlators and refloodirg tanks 4
o R L amaaaamant for Calvert Cliffs - 5.34 x 10 per year i
- 6. Snmaaa criteria and a==,tions used in SP/90 PSS o Supportisq analysis needed l
- 7. Cenpariscn of H MAAP ard RE SICP calculations I
o Areas of agreement o Areas of disayswdLEiit
- Hydrogen generaticn and M = tion L
- Fissicn product releases
- Cuiw-a.h.te interacticms
- Centainment reirgse
- 8. Direct Ccotainment Heating
- - Itans for di-=icn as requested by RtI/NRC
ENCLOSURE (3)
SP-90 PSS SUPPORTING ANALYSIS NEEDED ABILITY TO USE RHR TO INJECT IN A SMALL LOCA WITH HIGH-HEAD SAFETY INJECTION SYSTEM FAILURE VERIFY THAT LOW PRESSui1 COMPONENT WILL NOT BE CHALLENGED IN AN INTERFACING SYSTEM LOCA IN THE RHR SUCTION LINE VERIFY THAT NO CORE DAMAGE WILL OCCUR IN THE FOLLOWING SCENARIO:
REACTOR OPERATING AT 100% POWER WHEN A LOSS OF MAIN
~~
FEEDWATER OCCURS, AND FAILURE TO TRIP THE REACTOR AUTOMATICALLY OR MANUALLY, AND FAILURE OF TURBINE TRIP USE OF FAN COOLERS FOR LONG-TERM COOLING AFTER A SMALL OR LARGE LOCA (RHR HXs NOT AVAILABLE)
EMERGENCY FEEDWATER SYSTEM SUCCESS CRITERIA IN AN ATWS - 2/4 j
PUMPS TO FOUR SGs
,i:
ATWS PRESSURE RELIEF SUCCESS CRITERIA - 3/3 SAFETY VALVES ANU
(
1/3 PORVs h
SUCCESS CRITERI A FOR LARGE LOCA - 2/3 ACCUMULATORS AND 5/8 CORE REFLOOD TANKS AND HHSI PUMPS SUCCESS CRITERI A FOR CONTAINMENT SPRAY - 1/4 OR 2/4 j
NUMBER OF PORVs NEEDED FOR BLEED AND FEED 0
.