ML20059J193
| ML20059J193 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 11/05/1993 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9311120079 | |
| Download: ML20059J193 (21) | |
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fij NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20556-0001
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November 5,1993 I
L Docket No.52-003 APPLICANT: Westinghouse Electric Corporation FACILITY:
AP600
SUBJECT:
SUMMARY
OF SENIOR MANAGEMENT MEETING TO DISCUSS THE FOCUSED PROBABILISTIC RISK ASSESSMENT FOR THE AP600 1
On October 26, 1993, representatives of the Nuclear Regulatory Commission (NRC) and the Westinghouse Electric Corporation (Westinghouse) met to discuss the results of the focused probabilistic risk assessment (PRA), and Westinghouse's proposed resolution to address the issue of regulatory-treatment of non-safety-related systems for the AP600. Attendees are listed in Enclosure 1. is Westinghouse's slide presentation.
Westinghouse opened the meeting stating that they used the evaluation process conceptualized in January 1993, that was finalized with the Electric Power Research Institute (EPRI) in May 1993. They developed a focused PRA that removed all credit for the mitigation functions of the non-safety-related systems. Then the designer evaluated the following:
Are the initiating event frequencies impacted by the non-safety-related systems in the PRA?
If yes, does the unavailability of non-safety-related systems significantly affect the initiating event frequencies?
If yes, does the initiating event significantly (> 1 percent) affect the core damage frequency?
Westinghouse indicated that the PRA addressed interdependencies, and that the design is such that the non-safety-related systems do not adversely impact the operation of the safety-related systems and make the consequences of the initiating event unacceptable.
Westinghouse stated that, even with no credit for the mitigation functions of all of the non-safety-related systems, the PRA demonstrated that the design of the AP600 still met the Commission's core damage frequency and large release goals of 1x10 and lx10, respectively.
4 Westinghouse proposed that additional regulatory oversight should be required during power operation for the diverse actuation system, and non-class IE dc and uninterruptible power supply systems to ensure their availability in the event of an anticipated transient without scram. Westinghouse proposed that the utility " periodically verify the availability of required functions [of these systems) during power operation."
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. November 5, 1993 In addition, Westinghouse proposed additional regulatory oversight for certain non-safety-related structures, systems, and components (SSCs) during shutdown operations. Westinghouse proposed that the utility " verify system availability [of these SSCs] prior to initiating reduced inventory operations" to address this concern during shutdown operations.
Westinghouse stated that they have determined that no additional regulatory oversight was required for those non-safety-related SSCs identified as important in the evaluation of at-power initiating event frequencies for transient events.
The staff noted that the PRA indicated that the non-safety-related systems affected the results of the PRA by a factor of approximately 100, although they are not needed to meet the Commission's safety goals.
This demonstrates the overall importance of the non-safety-related systems to the design.
The staff indicated that the process Westinghouse used is an acceptable approach to resolve this issue. However, the staff further indicated that it would have to review the assumptions and details of the PRA analyses and Westinghouse's regulatory oversight proposal to determine the acceptability of their proposal.
Thomas J. Kenyon, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
1 As stated cc w/ enclosures:
See next page l
DISTRIBUTION w/ enclosures:
Docket File PDST R/F DCrutchfield TKenyon PDR PShea DISTRIBUTION w/o enclosures:
AThadani, 8E2 RArchitzel T ravers WRussell, 12G18 MMalloy BDLiaw, 7D26 T1urley/FMiraglia ACRS (11)
JMoore, 15B18 TGody, Jr., EDO OFC LA:PDST:ADAR PM:PDST:ADAR SC:PDST:ADAR M NAME PShea Oh TKenyd'ni:k'I RArchitzel k z
DATE 11/6/93 11/(/93 ll/f/93 I
0FFICIAL RECORD COPY: RTNSS.TJK
i Docket No.52-003 Westinghouse Electric Corporation cc:
Mr. Nicholas J. Liparulo Mr. Victor G. Snell, Director f
Nuclear Safety and Regulatory Analysis Safety and Licensing
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Nuclear and Advanced Technology Division AECL Technologies Westinghouse Electric Corporation 9210 Corporate Boulevard P.O. Box 355 Suite 410 l
Pittsburgh, Pennsylvania 15230 Rockville, Maryland 20850 i
Mr. B. A. McIntyre j
Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit L
Box 355 Pittsburgh, Pennsylvania 15230 i
Mr. John C. Butler Advanced Plant Safety & Licensing i
Westinghouse Electric Corporation Energy Systems Business Unit i
Box 355 i
Pittsburgh, Pennsylvania 15230 Mr. M. D. Beaumont Nuclear and Advanced Technology Division l
Westinghouse Electric Corporation f
One Montrose Metro 11921 Rockville Pike j
Suite 350 l
Rockville, Maryland 20852 l
Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Mr. S. M. Modro EG&G Idaho Inc.
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Post Office Box 1625 Idaho Falls, Idaho 83415 1
1 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 Mr. Frank A. Ross i
U.S. Department of Energy, NE-42 j
Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874
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MEETING' ATTENDEES i
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OCTOBER 26, 1993 I
i HAME ORGANIZATION
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Tom Kenyon NRR/PDST Ashok Thadani NRR/DSSA Ralph Architzel NRR/PDST Bill Travers NRR/ADAR W. T. Russell NRR/ADT Tom Murley NRR Dennis Crutchfield NRR/ADAR Melinda Malloy NRR/PDST B. D. Liaw NRR/DE Andrea Sterdis Westinghouse-AP600 Brian McIntyre Westinghouse-AP600
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WESTINGHOUSE ELECTRIC CORPORATION PRESENTATION s
TO UNITED STATES NUCLEAR REGULATORY COMMISSION OCTOBER 26,1993 i
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AP600 IMPLEMENTATION OF THE RTNSS PROCESS l
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ANDREA L. STERDIS ADVANCED PLANT SAFETY AND LICENSING 0005els/2
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NRC Management Meeting AP600 Implementation of the Regulatory Treatment of Nonsafety-Related Systems Process l
26 October 93 i
AGENDA I
AP600 RTNSS Implementation l
Process Overview l
Focused PRA Process Results Evaluation of Baseline PRA Initiating Event Frequencies Process Results Deterministic Evaluations ATWS rule (10 CFR 50.62)
Loss of ac power rule (10 CFR 50.63)
Post-72 hour actions l
Containment performance Adverse systems interactions Seismic considerations Mission statements for the important nonsafety-3 related systems Proposed regulatory oversight I
i NRC review process / schedule November 8,1993 staff briefing
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NRC feedback i
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RTNSS Prccces Impismontction Complete RTNSS Evaluations by Screening Nonsafety Related Systems in Table 1.0-1 Against RTNSS Criteria v
Focused PRA Baseline PRA Initiating Events 10 CFR 50.63 Blackout 10 CFR 50.62 ATWS Containment Performance Adverse Systems Interaction Seismic Considerations Post 72 Hour Actions r
No Nonsafety-Related Systems Based on RTNSS Criterle, are Nonsafety Related g
SSCs important?
Important to RTNSS Process Yes Define R/A Missions and Propose Recommended Additional Regulatory Oversight u
industry Review of RTHSS Submittel e
v AP600 RTNSS Submittal (September 1993), WCAP-13856
prg FOCUSED PRA ANALYSIS
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Methodology identical to AP600 baseline PRA Baseline PRA initiating event frequencies maintained Includes at-power and shutdown conditions i
includes internal and external events (excluding seismic)
No credit for mitigating functions of nonsafety-related l
systems l
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gr,g NONSAFETY-RELATED SYSTEMS REMOVED
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Chemical and volume control r
Main feedwater Condensate Startup feedwater Normal residual heat removal AC electrical power (connection from grid and diesel)
Diverse actuation (including diverse indication)
Spent fuel pit cooling Circulating water / cooling tower Main steam Turbine building closed cooling water Chilled water Component cooling water s ru en air Hydrogen igniters 4
Other systems that rely on ac power l
i Non-class 1E de power and plant control system indirectly removed i
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SAFETY-RELATED SYSTEMS CREDITED
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Passive core cooling j
IRWST injection / containment recirculation l
Core makeup tank Accumulator Passive residual heat removal Automatic depressurization Passive containment cooling Containment isolation Class 1E dc and UPS power Protection and safety monitoring Reactor trip Engineered safeguards actuation Safety-related monitoring Reactor coolant pump trip Steam generator isolation W
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l FOCUSED PRA RESULTS
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TABLE P-10 PRA Study Core Damage Frequency Large Release BASELINE PRA (At power and 4.4E-07 2.1 E-08 shutdown events)
FOCUSED PRA 1.3E-05 5.2E-07 (At power and shutdown events)
GOALS 1.0E-04 1.0E-06 0005els/8
pr,gg INITIATING EVENT FREQUENCIES
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Baseline PRA IE frequencies maintained in the PRA Importance of nonsafety-related SSCs evaluated Evaluation includes at-power and shutdown initiating events i
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pr,ig INITIATING EVENT FREQUENCIES
_171 At-power transients Systems identified as important:
Main steam Main feedwater Condensate Plant control Main turbine Turbine control and diagnostics 0005els/10
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INITIATING EVENT FREQUENCIES
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Shutdown events i
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Normal residual heat removal
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Component cooling water 1
Service water Offsite power Main ac power j
Onsite standby power
- Provides shutdown power source flexibility l
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DETERMINISTIC EVALUATIONS
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10 CFR 50.62 (ATWS rule)
DAS actuation of PRHR and turbine trip required to comply the rule Nonsafety-related SSCs required (functions):
Diverse actuation system (turbine trip and PRHR actuation during power operation) f Non-class 1E dc and UPS system (support diverse j
actuation system and required actuation components to provide turbine trip and PRHR functions during j
power) l l
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DIhTERMINISTIC EVALUATIONS
_ 77 10 CFR 50.63 (Loss of all ac power rule) t No nonsafety-related SSCs required to comply with the rule Post-72 hour actions No permanently installed nonsafety-related SSCs relied upon to support post-72 hour actions 4
Containment performance No nonsafety-related SSCs relied upon fn> support containment performance assumptions in the PRA i
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gpr,g DETERMINISTIC EVALUATIONS "7
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Adverse systems interactions Functional Spatial l
i Induced human-intervention No nonsafety-related SSC functions are relied upon to l
j prevent the potential for the nonsafety-related SSCs to l
adversely interact with the safety-related SSCs l
l Seismic considerations No nonsafety-related SSCs relied upon to support the l
seismic margins evaluation 0005 1s114 l
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IMPORTANT NONSAFETY-RELATED SSCs
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10 CFR 50.62 (ATWS rule)
Diverse actuation system (turbine trip and PRHR actuation during power operation)
Non-class 1E dc and UPS system (support diverse L
actuation system and required actuation components to provide turbine trip and PRHR functions during power)
Periodically verify availability of required functions during power operation t
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IE frequencies (shutdown events)
Normal residual heat removal Component cooling water Service water j
Offsite power Main ac power l
Onsite standby power" Support shutdown decay heat removal during reduced reactor coolant system inventory conditions i
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inventory operations I
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