ML20056D423
| ML20056D423 | |
| Person / Time | |
|---|---|
| Issue date: | 06/02/1993 |
| From: | Joshua Wilson Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-669A NUDOCS 9308160139 | |
| Download: ML20056D423 (12) | |
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NUCLEAR REGULATORY COMMISSION
' s WASHINGTON. D.C. 20555&J01 June 2, 1993 Project No. 669 APPLICANT:
Electric Power Research Institute (EPRI)
PROJECT:
Advanced Light Water Reactors (ALWRs)
SUBJECT:
SUMMARY
OF MEETING BETWEEN NUCLEAR REGULATORY COMMISSION (NRC)
STAFF AND EPRI HELD ON MAY 20, 1993, IN ROCKVILLE, MARYLAND, CONCERNING REGULATORY TREATHENT OF NON-SAFETY SYSTEM (RTNSS) IN PASSIVE ADVANCED LIGHT WATER REACTORS A public meeting was held on May 20, 1993, at the NRC headquarters in Rockville, Maryland, to discuss the issue of RTNSS.
In the opening remarks, the NRC staff stressed that this was a noticed public meeting between the NRC staff and EPRI, not an advisory committee meeting, as defined by the newly-promulgated Federal Advisory Committee Act of 1993. A list of attendees and their affiliation is provided in Enclosure I.
The handouts used in the meeting are provided as Enclosure 2.
A copy of EPRI's May 13, 1993, submittal describing EPRI's proposed process for determining RTNSS, which has been marked-up to reflect the agreements reached between the staff and EPRI, is provided as Enclosure 3.
This was the latest iit a series of meetings between the NRC and EPRI to determine the level of regulatory oversight that is appropriate for those active non-safety systems in passive ALWRs that are determined to be risk significant. On May 13, 1993, EPRI submitted its response to the staff's comments on EPRI's proposed approach to determining which systems should receive what kind of regulatory oversight.
The staff concluded that EPRI's process for determining RTNSS described in the May 13, 1993 submittal, as modified at the meeting in response to the staff's comments, is
.n acceptable approach to resolving this _ issue. The staff stated that it will develop a Commission paper describing the proposed approach, and close out the approximately 40 issues in the draft safety evaluation report for the EPRI Requirements Document based on this process. The staff indicated that it will review the implementation of this process on individual applica-tions for an final design approval / design certification to ensure that the passive plant vendors follow this process satisfactorily.
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. June 2, 1993 A key element of this process is to ensure that the vendors use probabilistic risk assessment methods to examine reliability of the safety and non-safety systems and ensure that important risk insights are considered in the design.
(Original signed by)
James H. Wilson, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation i
Enclosures:
As stated cc w/ enclosures:
See next page DISTRIBUTION w/ enclosures:
Central File PDST R/F DCrutchfield PShea PDR TEssig SBajwa, 12G18 1
DISTRIBUTION w/o enclosures:
TMurley/FMiraglia WRussell, 12G18 JCalvo RBorchardt JNWilson RHasselberg JHWilson TKenyon WBeckner, 10E4 AEl-Bassioni, 10E4 DThatcher, 7E4 RJones, 8E23 JLazevnick, 7E4 MRubin, 8E23 GHsii, 8E23 JLyons, 8D1 CLi, 8D1 TPolich, 10A19 CHoxie, 11E22 ACRS (11)
JMoore, 15818 EJordan, MNBB 3701 RPerch, 8H7
- See previous concurrence OFC:
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OFFICIAL RECORD COPY:
DOCUMENT NAME: MTSUM520.JW 9
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i ALWR Utility Steering Committee EPRI Project No. 669 cc:
Mr. E. E. Kintner Chairman Utility Steering Committee Bradley Hill Road Post Office Box 682 Norwich, Vermont 05055 s-Mr. John Trotter Nuclear Power Division Electric Power Research Institute Post Office Box 10412 Palo Alto, California 94303 Mr. Brian A. McIntyre, Manager Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Post Office Box 355 Pittsburgh, Pennsylvania 15230 Mr. Joseph Quirk GE Nuclear Energy Mail Code 782 General Electric Company 175 Curtner Avenue San Jose, California 95125 Mr. Stan Ritterbusch Conibustion Engineering 1000 Prospect Hill Road i
Post Office Box 500
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Windsor, Connecticut 06095 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.
20585 Mr. Steve Goldberg Budget Examiner-725 17th Street, N.W.
Room 8002 Washington, D.C.
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LIST OF ATTENDEES AT MEETING WITH EPRI HELD IN ROCKVILLE, MARYLAND ON MAY 20, 1993 Name Affiliation W. Russell NRC J. Calvo NRC R. Borchardt NRC J. H. Wilson NRC T. Kenyon NRC W. Beckner NRC A. El-Bassioni NRC D. Thatcher NRC J. Lazevnick NRC R. Jones NRC M. Rubin NRC G. Hsii NRC J. Lyons NRC C. Li NRC T. Polich NRC C. Hoxie NRC R. Hasselberg
'NRC J.N. Wilson NRC J. Wheeler DOE J. Youngblood BNL E. Kintner EPRI J. DeVine Polestar E. Rumble Polcstar J. Trotter EPRI J. P. Berger EPRI S. Additon TENERA D. Chapin MPR T. Schulz Westinghouse B. McIntyre Westinghouse i
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Enclosure I l
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- RTNSS RESOLUTION PROCESS 2
W by Ma'rk Rubin Division of. Systems and Safety Analysis NRR-May 20,1993 I
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gel _\\lERAL_O_BSERVATIQN EPRI 5/13/93 RTNSS process mostly in line with staff position i
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Some minor areas of clarification reflected in staff comments This process forms an acceptable basis for resolution of RTNSS issue, t
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_ STAFF C_OMMENTS ON EPRl_ PROPOSE _D RTNSS PRO _C_E_SS 1.
"RTNSS Scope" Section should include:
" Item F. SSC Functions relied upon to maintain heat removal during shutdown operation."
2.
Suggest consistent use of the term " Comprehensive Baseline PRA" rather than
" Comprehensive PRA" and " Baseline PRA" interchangeably.
3.
Long-term safety issue regarding the design's ability to maintain core cooling beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> will be an area of extensive review, and needs to be evaluated through PRA in addition to deterministic methods.
Seismic events should be evaluated with PRA-based margins approach.
4.
Means of regulatory oversight (Step 6) should be proposed by the designer, and may include other elements such as:
O TS LCOs and administrative procedures such as shutdown configuration control, RAP, maintenance rule.
Deterministic requirements of risk significant SSCs included in tier 1 design certification rule including ITAAC.
F_OLLOWUP ACTl_ONS_
EPRI submittal of a revised URD incorporating the agreed RTNSS process will be a confirmatory issue in SER.
Staff will prepare a SECY paper and update URD SER based on the May 13, 1993, letter.
g Staff will review passive designs based on this process.
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funcnons and R/A missions identified in the RTNSS approach. Continued effort is required to develop examples of the types and levels of regulatory oversight for representative categories of risk significant SSC functions.
Scone:
The RTNSS basis is broadly applicable to those nonsafety SSCs which have risk significant functions, and are therefore candidates for regulatory oversight. The plant designer will identify these SSC functions, utilizing the following criteria:
i A.
SSC functions relied upon to meet beyond design basis deterministic NRC performance requirements: 10 CFR 50.62 for i
ATWS mitigation, and 10CFR 50.63 for loss of all ac power.
B.
SSC functions relied upon to resolve long term safety (beyond 72-hours) and to address seismic even&s. 60 t
SSC functions relied uporiffo} meet the Commission's Safety g C.
guidelines of core damage frequency ofless than 1.0E-4 per reactor year and large release frequency of less than 1.0E-6 per reactor year.
D.
SSC functions needed to meet the containment performance goal (SECY-93-087, issue I.J), including containment bypass, during severe accidents.
i E.
SSC functions relied upon to prevent significant adverse systems interactions.
Stens in the RTNSS Process for each design:
1.
Comorehensive,.PRA:
/6 The evaluation process starts with designer constructed comprehensive Level 3 PRAs prepared in accordance with the ALWR URD. These PRAs must include all appropriate internal and external events considering both power and shutdown operations. Seismic events will be evaluated by a margins approach. Adequate treatment of uncertainties,long term safety operation, and containment performance should be included.
Containment performance should be addressed with considerations for sensitivities and ' uncertainties in accident progression and inclusion of severe accident phenomena, including explicit treatment of containment bypass. Appropriate uncertainty distributions and mean values must be used for passive systems 2- '
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unavailabilities and for core damage frequency and large release frequency. Results of an adverse systems interaction study will also be considered in the PRA.
2.
Search for Adverse Systems Interactions:
The designers must provide a systematic evaluation of adverse systems interactions between the active nonsafety and passive systems. The results of this analysis should be used for design improvements to minimize adverse systems interaction, and also be factored into the PRA model.
3.
Focused PRA:
The designers should construct Focused PRAs. The Focused PRA is used to determine the R/A missions of nonsafety SSCs which are risk significant. There are two main considerations in constructing Focused PRAs.
First, the scope of initiating events and their frequencies are maintained in the Focused PRA as in the B seline PRA. As a result, nonsafety SSCs used to prevent the occurrence of initiating events will be subject to regulatory oversight applied commensurate with their R/A missions for prevention, as discussed in steps 4 and 5.
Second, the effect of nonsafety SSCs is removed from the comprehensive Level 3 PRA event tree logic. As a minimum, the defense-in-depth functions and their support such as ac power are removed. This is to determine if the passive safety systems, when challenged, can provide sufficient capability without nonsafety backup to meet the NRC safety goal guidelines of core damage frequency of 10-4 per year and large release frequency of 10-6 per year. The containment performance, including bypass, during a severe accident should also be evaluated. Nonsafety SSCs which remain in the Focused PRA model are subject to regulatory oversight based on their risk significance in steps 4 and 5.
4.
Selection of Important Nonsafety Systems:
The designers determine what combination (if any).of nonsafety-l SSCs are necessary to meet NRC regulations, safety goal guidelines and the containment performance goal objectives. This is done i
both for scope items Adand E where NRC regulations are the primary consideration and scope items C and D where PRA 1
methods are prevalent. Nonsafety SSC functions required to meet beyond design basis te uirements (item A), to resolve the long term i
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safety and seismic issues (item B), and to prevent significant adverse interactions (item E) are subject to regulatory oversight as discussed in step 5.
The Focused PRA is used to determine the nonsafety SSCs important to risk. This is done in two parts.
First, nonsafety SSCs needed to maintain initiating event frequencies at the baseline PRA levels will be identified from the PRA.
Second, the designers will,if needed, add the necessary success paths with nonsafety systems and functions in the " Focused PRA" in order to meet safety goal guidelines, containment performance goal objectives and NRC regulations. The designers can choose those systems needed by considering the factors for optimizing design impact and benefit of particular systems. All relevant issues which are addressed by PRA should be included in this evaluation.
PRA importance studies should be performed to assist in determining the importance of these SSCs. In principle, all nonsafety SSCs in the Focused PRA model needed to meet NRC requirements, the safety goal guidelines and containment performance goals are potentially subject to regulatory oversight, commensurate with their risk significance.
5.
Nonsafety System Reliability / Availability Missions:
The designer willidentify and document risk significant nonsafety systems functional R/A missions from the Focused PRA which are needed to meet the safety goal guidelines, containment performance goals, and other NRC requirements per Step 4. Steps 4, S and 6 should be iterated to optimize the selection of risk-significant nonsafety systems and their R/A missions.
6.
Reculatorv Oversicht Evaluation:
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based on that information will propose regulatory oversight 4
measures. This regulatory oversightinay include the following:
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calculations to determine that the design of these risk
- significant nonsafety SSCs satisfies the performance capabilities and R/ A missions identified.
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