ML19325E605

Provides Summary of Proposed Individual Plant Exam Program as Requested by NRC Generic Ltr 88-20,Suppl 1.Results of Program Will Be Documented to Meet NUREG-1335.Rept to NRC Expected to Be Submitted by Aug 1992

ML19325E605
Person / Time
Site:	Summer
Issue date:	10/31/1989
From:	Bradham O SOUTH CAROLINA ELECTRIC & GAS CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-1335 GL-88-20, NUDOCS 8911080101
Download: ML19325E605 (4)

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4.nglpgs; cam SCESO l October 31, 1989 I

l Document Centrol Desk l U. S. Nuclear Regulatory Commission i Washington, D.C. 20555  !

SUDJECT: Virgil C. Summer Nuclear Station )

Docket No. 50/395 i Operating License No. NPF-12 i Response To Generic Letter 88-20,  ;

Supplement 1  !

1 Gentlemen I I. INTRODUCTION i l

This letter provides a summary of the proposed Individual Plant Examination l' (IPE) program for the Virgil C. Summer Nuclear Station (VCSNS), as recuested by NRC Generic Letter 88-20. Supplement No. 1, dated August 29, 1989. It also acknowledges that South Carolina Electric & Gas Company (SCE&G) has received a copy of NUREG-1335 " Individual Plant Examination: Submittal

Guidance " Final Report (August, 1989). The results of the proposed IPE i

program will be documented to meet the NRC submittal guidance outlined in NUREG-1335.

i l The program description as summarized in this document includes the three i items specified by Supplement No. I to the generic letter, namely:

1. It identifies the method and approach for performing the IPE,

2. It describes the method to be used, and l 3. It identifies the milestones and schedules for performing the IPE and l submitting the results to the NRC.

While benefits are anticipated from the results of the IPE, SCE&G considers theperformanceoftheIPEtobeanevaluationthatexaminesevents(internal events) that are beyond the design bases of the VCSNS.

j II. METHOD AND APPROACH TO PERFORMING THE IPE STUDY The method selected for performing the IPE it the first method discussed in Section 4 of NRC Generic Letter 88-20, i.e., it is a Level I Probabalistic Risk Assessment (PRA) using current methods and information, plus a l

containment performance analysis that follows the general guidance given in l Appendix 1 to the generic letter. PRA procedures described in NRC NUREG/CR-l 2815 will be used as guidance in performing the Level I PRA and the evaluation of containment system performance will consider the most current severe accident phenomenological issues (as discussed in Appendix 1. Generic Letter 88-20).

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. - October 31, 1989 Page 2 of 4 SCE&G has assigned engineers and plant staff support personnel to provide engineering, plant operations, maintenance, training and licensing support for the IPE study. SCE&G has also contracted with the Westinghouse Electric Corporationanditssubsidiary,Fauske& Associates,Inc.(FAI)toperform and document the IPE study. Analyses performed during the IPE study will be documented in calculation note packages consistent with the requirements of 10CFR50, Appendix B.

The PRA study will be performed in terms of major tasks which outline the scope of the program. These tasks are listed below to provide a glimpse of the scope involved:

1. Project Management

2. Information Collection

3. Plant Data Collection & Database Development

4. Definition and Analysis of Initiating Events

5. Development of Plant Damage States

6. Support State Modeling and Quantification

7. Event Tree Development

8. System Notebook Design Documentation

9. Fault Tree Development

10. Systems Interaction Evaluation

11. Human Reliability Assessment

12. Fault Tree Quantification

13. Accident Sequence Quantifiction

14. Internal Flooding Assessment

15. Evaluation of Recovery Actions for Dominant Accident Sequences

16. Evaluation of System Analysis Results

17. Sensitivity and Importance Analysis

18. Impaired Containment Analysis

19. Containment System Analysis

20. Containment Structural Capability Assessment

21. Containment Event Tree Modeling

22. Source Term Analysis and Evaluation

23. Evaluation of IPE Results 24 IPE Project Documentation

25. Project Review III. DESCRIPTION OF LEVEL I PRA ANALYSIS METHOD

, The plant specific Level I PRA will be performed using the current design of l VCSNS as it exists after Refuel 5 (currently scheduled for March 1990).

Plant walkdowns will be performed to identify possible systems interactions.

This technique is beneficial for uncovering intersystem dependencies due to

( location of equipment. The plant / system walkdown will also aid in the search

for potential internal flooding events.

The Level I PRA will determine dominant accident sequences and core-damage frequency based on plant system and human-factor evaluations. The methods provided in the PRA Procedure Guide NUREG/CR-2815 will be used in performing the Level I PRA and will be supplemented with current PRA methods as needed.

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Page 3 of 4  ;

l Internal initiating events (including a loss-of-offsite power event) will be ,

analyzed in the performance of the Level 1 PRA. In addition, an internal  !

flooding event will be analyzed. AsdiscussedinNUREG-1335(pageC-16),

only power operation and hot-standby modes of plant operation will be  ;

considered in the IPE study.

The front-end analysis (i.e., the probabil'.ty of severe accidents, the resultant core damage frequency and plant damage states) will be performed using the large event tree, small fault tree approach. Both system level accident sequences and component / operator error level cutsets will be -

quantified and dominant contributors will be identified and ranked. This information will be used in the performance of sensitivity and importance '

analyses to gain insight for use in the vulnerability screening of plant systems.  !

IV. DESCRIPTION Of CONTAINMENT PERFORMANCE ANALYSIS METHOD The containment performance analysis and source term evaluations will be performed in a systematic way by employing a detailed containment event tree (CET) relevant to VCSNS design. The analysis will be performed using '

Appendix 1 of Generic letter 88-20 and NUREG-1335 (Section 2.2 and Appendix A) as guidance. l l

The containment event tree will be developed in a manner consistent with the generic letter. Specifically, the first node on the tree will characterize the likelihood of containment bypass, the second node will address containment isolation for the given accident initiator, the third (and perhaps fourth) node will address water systems that provide addition to the i containment as would be required to cool core debris, and the last node will focus on the capabilities for long term decay heat removal. Phenomenological issues will be incorporated into the event tree with specific attention given  ;

to issues of direct containment heating and possible attack of the containment liner. To quantify the magnitude of the radionuclide releases, the CET will be applied to each plant damage state and a set of CET end states will be produced. Should the CET end states (i.e., those representing dominant risk contributors) need to be quantified by plant specific analysis, the MAAP (3.08) computer code will be utilized.

V. IPE STUDY SCHEDULE AND MAJOR MILESTONES SCE&G has initiated the IPE study for VCSNS. The expected schedule for conducting the study is approximately 24 months. The target date for completion of the study is December 1991. The targeted major milestones of the IPE study are as follows:  :

IPE Project Plan Completed December, 1989 Quantification of Accident Sequences Completed June, 1991 Containment Performance Analysis Completed August, 1991 Draft IPE Report December. 1991 Submittal of Report to NRC August, 1992 r

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- - .- 6ocumentControlDesk i- - October 31, 1989 Page 4 of 4 I declare that the statements and matters set forth herein are true and correct to the best of my knowledge, information and belief.

Should you have any questions, please call at your convenience.

Very truly yours, I

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0. S. Bradham ARR/OSB: led c: D. A. Nauman/0. W. Dixon, Jr./T. C. Nichols, Jr.

W. A. Williams, Jr.

E. C. Roberts S. D. Ebneter J. J. Hayes, Jr.

General Managers C. A. Price K. E. Nodland R. B. Clary J. C. Snelson R. L. Prevatte J. B. Knotts, Jr.

B. S. Mullinax A. R. Rice D. Modeen - NUMARC NSRC NPCF RTS LTR 880020)

File 815.14)