ML20097A090

Forwards Outline of Proposed Scope & Schedule for PRA to Be Submitted as Part of Application for Preliminary Design Approval of Advanced PWR Power Block Design

ML20097A090
Person / Time
Site:	05000601
Issue date:	08/24/1984
From:	Rahe E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NS-EPR-2951, NUDOCS 8409130172
Download: ML20097A090 (7)

Text

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NS-EPR-2951 August 24, 1984

, Docket No. SIN-50-601 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation

,U.S. Ibclear Regulatbry ~ Commission Washington, D.C. 20555

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Subject:

Westinghouse Advanced Pressurized Water Reactor (MAIWR),

SE3AR-SP/90, PDA-Level PRA Scope / Schedule Attention: KF T. Eccleston, Project Manager, SSPB v A. C. Thadani, . Branch Chief, RRAB

Dear Mr. Denton:

R n L' In response 'to an NRC request made at a Westinghouse /NRC meeting held ' August 3, 1984, enclosed is an outline of the proposed scope and schedule for the Probabilistic Risk Assessment .(PRA)z to be sutznitted as part of the application for Preliminary Design Approval (PDA) of the HAfWR Nuclear Power Block design (i. e. , RESAR-SP/90 ) .

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L As discussed in,this meeting, it is inportant to have an unambiguous mutual 5 understanding and agreement as to wLat constitutes an appropriate level of detai1~ for the PRA for a preliminary design well in advance of this sutraittal.

Westinghousd appreciates the NRC's desire to be involved in the RAfWR FRA during its developnent so as to expedite the review of the resulting doctanentation in an attenpt to meet Westinghouse's desire to obtain the SER/PDA, referencable by a construction pennit applicant, in early 1986.-

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W.-H. R.-Denton Page Two Your prompt, response to this letter is requested. Please contact Douglas G.

Bevard (412/374-5597) of my staff should you require further information.

Very truly yours, E. P. Rahe, . , Manager Nuclear Safety Department HDB/kk-Enclosure .

-cc: C. O. Ihamas F. R. Miraglia, Jr.

8- R. Bernero

-D. Eiserhut R. Silver T. P. Speis

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ENCLOSURE Westinghouse Advanced (MAPWR) PDA PRA Scope / Schedule lhe MANR Probabilistic Risk Assessment (PRA) will address the intended NHC criteria in NUREG-1070 and will be completed and submitted in two phases. The report of the first phase to be delivered to the NRC in June,195 will cover the internal initiating events analysis and will address the requirenents for the PDA phase. The external events analysis of the second phase will be completed at a later date when the data required to perform it will be more mature. A comprehensive PRA study will be carried out for both phases. The internal initiating events report will contain the following sections:

1. Internal Initiating Event Analysis

'2. Accident Sequence Modeling

3. Plant Systems Analysis

4. Core Melt Quantification 5 Core and Contairunent Analysis

6. Consequence Analysis 7 Uncertainty. Analysis

8. Plant Risk Analysis To facilitate the NRC review of the final reports, the following meetings with the NRC PRA staff at the Westinghouse Nuclear Center are suggested:

1st Meeting: Novenber, 1984 Objectives:

1. Review the decision making processes involved in selection of design options and alternatives.

2. Discuss the methodology and tools being used and to demonstrate the codes.

2nd Meeting: February, 195 Objective:

1. Discuss the preliminary numerical results and the analysis of dominant risk contributors.

A detailed PRA PDA-level report outline follows.

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.WAIMR PRA PDA-Level Report Outline 1

1.0 INTERNAL INITIATING EVENT ANALYSIS 1.1 Internal Initiating Event Cr.tegorization 1.2 Internal Initiating Event a2antification

-2.0 ACCIDENT SEQUENCE M)DELING 2.1 Event Tree Guidelines 2.1.1 Event Tree Guidelines ,

2.1.2 Event Tree Node Definitions 2.1.3 Event Tree Success Criteria Definitions 2.1.4 Core Melt State Definitions 2.1.5 Node Success Criteria Definitions 2.1.6 Consequential Failure Model 2.1.7 Support State & del (See Note 1) 2.2 Event Tree Modeling 2.2.1 Transients 2.2.2 Loss of Offsite Power 2.2.3 Ste m Generator Tube Rupture 2.2.4 Ste aline Break 2.2.5 Small LOCA 2.2.6 Large LOCA 2.2.7 AIWS 2.2.8 Interfacing Systems LOCA 2.2 9 Vessel Failure 2.2.10 Total Loss of Auxiliary Cooling 3.0 PLANT SYSTEMS ANALYSIS 3.1 AC Power 3.2 Integrated Protection System 3.3 Service Water / Component Cooling Water Systes 3.4 Emergency Core Cooling System 35 Contaiment Sprays 36 Contairment Fans 3.7 F2nergency Feedwater Systen 3.8 Emergency Seal Cooling 3.9 Other Fault Trees 3.10 Fault Tree Guidelines 3.10.1 Fault Tree Mocel Guidelines 3.10.2 Test and Maintenance & del 3 10.3 Hunan Error Model for Fault Trees 3.10.4 Common Cause Model (See Note 2) 3.10.5 Data Bank (See Note 3) 3.11 Screening Model for Operator Actions in Event Trees (See Note 4)

4.0 CORE ELT QUANTIFICATION 4.1 Quantification of Event Tree Nodes 4.2 Quantification of Core Melt 4.3 - Analysis of Core Melt Contributors

-4.4 Uncertainty Analysis in Core Melt 5.0 CORE AND CONTAINENT ANALYSIS 5.1 Sequence Categorization 5.2 Core Melt and Contairrnent Analysis 5.3 Contairment Event Tree 5.4 Fission Product Source Term Analysis 5.5 Uncertainty Analysis 6.0 CONSEQUENCE ANALYSIS 6.1 Data Preparation 6.2 Consequence Analysis 63 Sensitivity Analysis f 6.4 Uncertainty Analysis 6.5 Liquid Pathways Analysis 70 UNCERTAINTY ANALYSIS (See Note 5) 8.0 H ANT RISK ANALYSIS (See Note 6) f i

The'second phase report will contain the following sections:

1. Seisnic Events Analysis

2. -Fire Events Analysis

3.  : Other External Events Analysis (See Note 7)

4. Satotage Analysis (See Note'8)

5. Plant' Risk From External Events

6. Plant. Risk From Internal and External Events (See Note 9)

c; NOTES:

1. Dependencies on major support state systems such as AC-power, SWS, CWS, ,.

etc. will be treated by the support state modeling approach. Bis approach 1 is already used in the Millstone Unit 3 Probabilistic Safety Study.

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2. Common cause will be included at least for active components, such as l pmps, MDV's, etc. other system specific major common cause sources, such I as strainers, haan tasks, . . will also be considered whenever applicable.

he common cause data will be taken from Atwood's work. (

3 he most current generic data base maintained by Westinghouse will be used.

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4. A screening model for operator actions will be used. It will be based on generic THERP modeling and Westinghouse emergency procedures.

5. . Propagation of uncertainty in core melt, core and contalment analysis, and consequence analysis will be done using the dominant accident sequences.

For this purpose the results of Sections 4.4, 5 5 and 6.4 will be used.

6. . Matrix approach in risk assembly process will be used for constructing the point estimate risk curves. Bis approach is the same as that of Millstone Unit 3 Probabilistic Safety Study.

7 Other external initiating events analysis may include relevant events such as wind, aircraf t, etc.

8. Sabotage analysis will be done as a stand alone analysis and its results will be kept separate from the total plant risk analysis. It is envisioned that at most conditional probabilities for a set of sabotage scenarios will be quantified.

9 Uncertainties in external events will be addressed in each analysis section and for the total plant risk, based on uncertainty analysis on the dominant accident sequences.