ML20211D525

Discusses 970916 Telcon W/Westinghouse Electric Corp Re AP600 Design Features Which Help Mitigate Containment Bypass Due to Multiple SG Tube Rupture Event.Resolution of Msgtr Issue Encl

ML20211D525
Person / Time
Site:	05200003
Issue date:	09/17/1997
From:	Huffman W NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9709290164
Download: ML20211D525 (4)

Text

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September 17, 1997 l APPLICANT: Westinghouse Electric Corporation -

PROJECT: AP600 f

SUBJECT:

SUf94ARY OF. TELEPHONE CONFERENCE WITH WESTINGHOUSE TO DISCUSS '

DOCUMENTATION THE STAFF NEEDS TO COMPLETE EVALUATION OF THE AP600 MULTIPLE STEAM GENERATOR TUBE RUPTURE RESPONSE On September 16, 1997, the Nuclear Regulatory Commission (NRC) staff (Huffman) and Westinghouse (Haag) conducted a telephone conference (telecon) concerning ,

the AP600 design features which help mitigate containment bypass due to a l multiple steam generator tube rupture event. The basis for this assessment-is i described in SECY-93-087, " Policy Technical, and Licensing Issuas Pertaining to Evolutionary and Advanced Light-Water Reactor Designs." In an NRC letter dated August 7, 1997,. additional information,(RAls) was requested from:

Westinghouse about its multiple steam generator tube rupture analysis report submitted on March 24. 1997. Westinghouse is preparing to respond to-the  !

staff's RAls and requested clarification on what the NRC-needed to resolve the  :

issue. The staff provided additional details on the documented information that it required.to resolve the multiple steam generator tube rupture issue during the telecon. The clarification the NRC provided to Westinghouse is in the attachment to this memorandum.

original signed by:

William C. Huffman, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003 gg QN

Attachment:

As stated cc w/att: 'See next page .

Q1STRIBUTION w/ attachments: .

Det t File PDST R/F TKenyon PUi BHuffman DTJackson-JSt sky GHsii, 0-0 E23 Alevin, 0-8 E23 ,

TCollins, 0-8. E23 SNewberry, 0-8 E02 -JFlack, 0-10 E4 DJ11]lIBUT10N w/o attachment:

SCollins/FMiraglia, 0-12 G18 BSheron, 0-7 D25 RZimmerman, 0-12 G18 JRoe- DMatthews TQuay W0aan, 0-5 E23 ACRS (11) JMoore 0-15 B18 ,

DOCUMENT NA ME  : A: TLC 9-16.MYR

1. ..% . ..n w. -.n w m m. w c con -*e .n-i no.ne. e - con -*ei .n.a-.av.n*. . r No con 0FFICE PM:PDST:DRPM D:PDST:DRPM l l NAME WCHuffman:sh.'tAwr TRQuay W i DATE 09//7/97 09/r1/97 -

, , ,9( 0FFICIAL RECORD COPY h h .

Westinghouse Electric Corporation Cocket No.52-003 cc: Mr. Nicholas J. Liparulo, Manager Mr. Frank A. Ross  !

Nuclear Safety and Regulatory Analysis U.S. Department of Energy, NE-4?

Nuclear and Advanced Technology Division Office of LWR Safety and Technology Westinghouse Electric Corporhtion 19901 Germantown Road P.O. Box 355 Germantown, MD 20874 Pittsburgh, PA 15230 '

Mr. Russ Bell Mr. B. A. McIntyre Senior Project Manager. Programs Advanced Plant Safety & Licensing Nuclear Energy Institute WrItinghouse Electric Corporation 1776 I Street, NW Energy Systems Business Unit Suite 300 Box 355 Washington, DC 20006-3706 Pittsburgh, PA 15230 Ms. Lynn Connor Ms. Cindy L. Haag Doc-Search Associates Advanced Plant Safety & Licencing Post Office Box 34 Westinghouse Electric Corporation Cabin John, MD 20818 Energy Systems Business Unit Box 355 Pittsburgh, PA 15230 Dr. Craig D. Sawyer, Manager Advanced Reactor Programs GE Nuclear Energy Mr. M. D. Beaumont 175 Curtner Avenue, MC-754 Nuclear and Advanced Technology Division San Jose, CA 95125 Westinghouse Electric Corporation One Montrose Metro Mr. Robert H. Buchholz 11921 Rockville Pike GE Nuclear Energy Suite 350 175 Curtner Avenue, MC-781 Rockville, MD 20852 San Jose, CA 95125 Mr. Sterling Franks Barton Z. Cowan, Esq.

U.S. Department of Energy Eckert Seamans Cherin & Mellott NE-50 600 Grant Streat 42nd Floor 19901 Germantown Road Pittsburgh, PA 15219 Germantown, MD 20874 Mr. Ed Rodwell, Manager Mr. S. M. Modro PWR Design Certification Nuclear Systems Analysis Technologies Electric Power Research Institute Lockheed Idaho Technologies Company 3412 Hillview Avenue Post Office Box 1625 Palo Alto, CA 94303 Idaho Falls, ID 83415 Mr. Charles Thompson, Nuclear Engineer AP600 Certification hE-50 19901 Germantown Road Germantown, MD 20874

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= RESOLUTION OF MSGTR-ISSUE:

1. Respond to RAIl440.676 regarding MAAP4 benchmark: l i

Provide a summary of MSGTR scenarios and important phenomena, and a summary of MAA14/NOTRUMP benchmark and results that can'be used to argue that important phenomena during a MSGTR can be reasonably analyzed with MAAP4. On the-SG secondary side, there may not be a MAAP4 benchmark. It would be necessary to discuss the simplified one-node model, the fundamental equations and constitutive models used to conservatively analyzed the secondary.

-side, and discuss why this is appropriate.

2. The argument for resolution of the issue:

a. Even for 5-tube rupture' case,-the secondary pressure never reaches the safety valve setpoints as long as the PORV is j operable and open to relieve.the pressure 4 This is evidenced in case SGS-cvs.

b. The probability of the PORV fails to opan is low though it is a non-safety related system:

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' Provide the PORV design features and measures to ensure the PORV will automatically open on demand, and provide an estimate of its failure probability. (RAI 440-683 b)-

( c. -In the low-probability event that. the PORV fails to open, the

- MSSVs will open at a higher pressure. The SG overfill protection automatically trip the CVS and SFW on hi-2 SG narrow range level, and the MSSV will release-steam rnly.. Therefore, the probability of MSSV failure to resent is smail:

The scenario is similar to SGS-evs, except that the MSSV open_at higher: pressure than the PORV. However, this case was not analyzed. 1The SGS-stk case assumes the MSSVs fail to reclose once it open, and the result shows the SG' overfill even after the CVS and SFW are isolated at- 50 minutes on hi-2 ' SG narrow range level. However, the analysis results showed that the SG is

-filled with water after about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, mad therefore water is releace through the MSSVs.

(A) Provide (1) an analysis (similar to case - SG5-evs except that the PORV is assumed 'to fail to open), or (2) an argument based on similar results-from-case SG5-cvs except for higher set pressure of the safety valves, to demonstrate that even if the PORV fails to open..the safety relief valve will open at a higher pressure than the PORV would, and will release steam and reclose as SG i pressure declines.

(B) -provide available data to support the argument that the safety valve will not stick open with steam release. (440,683 c) 3 Attachment

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d. Even if the MSSV fails to resent after it is acutated, the SGTR scenario turns into a through automatic ADS actuation, and no core damage would not occur, as shown in case SG5-stk. The maximum total release *.sould be limited to the initial activity in the RCS. '

• Provide an estimate of total release if the safety valve is stuck open, and the core melt does not occur.

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