ML20133G226

Requests That W Review Claim & Provide Detailed Justification as to Why Info in Rept WCAP-14811, Accident Spec & Phenomena Evaluation of AP600 Passive Containment Cooling Sys Considered to Be Proprietary W/In 30 Days

ML20133G226
Person / Time
Site:	05200003
Issue date:	01/14/1997
From:	Diane Jackson NRC (Affiliation Not Assigned)
To:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9701150238
Download: ML20133G226 (3)

Text

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,' [" -t UNITED STATES

, g g NUCLEAR REGULATORY COMMISSION

• WASHINGTON, D.C. 20665-0001

\ . . . . . p# January 14, 1997 Mr. Brian A. McIntyre, Manager Advanced Plant Safety and Licensing Westinghouse Electric Corporation Energy Systems Business Unit P.O. Box 355 Pittsburgh, PA 15230-0355

SUBJECT:

REQUEST FOR WITHHOLDING INFORMATION FROM PUBLIC DISCLOSURE FOR WESTINGHOUSE AP600 DESIGN LETTER OF DECEP.BER 19, 1996

Dear Mr. McIntyre:

By Westinghouse letter (NSD-NRC-96-4921), dated December 19, 1996, you submitted Westinghouse report WCAP-14811, " Accident Specification and Phenomena Evaluation for AP600 Passive Containment Cooling System." The material was provided via enclosure in only a proprietary version. You stated in your letter that a non-proprietary version would be submitted within 60 days.

In your letter and report, you did not identify information in the report which you consider proprietary information. The report, in its entirety, is identified as proprietary and you request that it be withheld from public disclosure in accordance with 10 CFR 2.790. The attached affidavit (AW 1053), executed by you on December 20, 1996, states that the information is owned and held in confidence by Westinghouse which should be considered exempt from mandatory public disclosure because it contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

We have reviewed your submittal and the material in accordance with 10 CFR 2.790 and, on the basis of Westinghouse's statements, have determined that all of the submitted information sought to be withheld does not contain i trade secrets or proprietary information. Therefore, you are requested to review your claim and provide detailed justification as to why the information is considered to be proprietary w/i 30 days of the date of this letter. The proprietary version must identify the information and the reasons that Westinghouse considers the material proprietary.

Withholding from public inspection shall not affect the right, if any, of persons properly and directly concerned to inspect the documents. If the need arises, we may send copies of this information to our consultants working in this area. We will, of course, ensure that the consultants have signed the appropriate agreements for handling proprietary information.

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M . Brian A.- McIntyre '

January 14, 1997  !

l If the basis for withholding' this information from public disclosure should l change in the future such that the information could then be made available i

for public inspection, you should promptly notify the NRC. You should understand that the NRC may have cause to review this determination in the future, for example, if the scope of a Freedom of Information Act request i

includes your withheld information. In all review situations, if the NRC makes a determination adverse to the above, you will be notified in advance of any public disclosure. -

Sincerely, original signed by:

Diane T. Jackson, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003 cc: See next page DISTRIBUTION:

' Docket File PDST R/F TMartin  !

PUBLIC DMatthews TQuay TKenyon DJackson WHuffman JSebrosky WDean, 0-17 G21 JMoore, 0-15 B18 EJordan, T-4 D18 MSiemien, 0-15 818 CBerlinger, 0-8 H7 i JKudrick, 0-8 H7 EThrom, 0-8 H7 l ACRS (11) l DOCUMEi(T NAME: A: PHEN n ...n.ew. u i OME2.PRP. i.,m.i c c .uc.n.ch 4 no.nce. c . con wem .n=hm.ou.ncio. r - wo .o,y 0FFICE- PM:PDST:DRPtt BC:SCSB:llSSA OGC 2 DPDST:DRPM l NAME DJackson V)( CBerlinger MPSiemien W D TRQuay # 4 DATE 01/@/97 () 01/ 7 /97 Auf 01// 3/97 01/lv/97

' 0FFICIAL RECORD COPY

Mr. Brian A. McIntyre Docket No.52-003 Westinghouse Electric Corporation AP600 i

l cc: Mr. Nicholas J. Liparulo Mr. Ronald Simard, Director Nuclear Safety and Regulatory Analysis Advanced Reactor Programs Nulcear and Advanced Technology Division Nuclear Energy Institute Westinghouse Electric Corporation 1776 Eye Street, N.W.

P.O. Box 355 Suite 300 Pittsburgh, PA 15230 Washington, DC 20006-3706 4

Mr. John C. Butler Ms. Lynn Connor Advanced Plant Safety & Licensing Doc-Search Associates Westinghouse Electric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 Box 355 Pittsburgh, PA 15230 Mr. James E. Quinn, Projects Manager LMR and SBWR Programs Mr. M. D. Beaumorit GE Nuclear Energy Nuclear and Advanced Technology Division 175 Curtner Avenue, M/C 165 Westinghouse Electric Corporation San Jose, CA 95125 One Montrose Metro

, 11921 Rockville Pike Mr. John E. Leatherman, Manager '

. Suite 350 SBWR Design Certification Rockville, MD 20852 GE Nuclear Energy, M/C 781 San Jose, CA 95125 Mr. Sterling Franks U.S. Department of Energy Barton Z. Cowan, Esq.

NE-42 Eckert Seamans Cherin & Mellott 4

Washington, DC 20585 600 Grant Street 42nd Floor Pittsburgh, PA 15219 Mr. S. M. Modro Nuclear Systems Analysis Technologies Mr. Ed Rodwell, Manager Lockheed Idaho Technologies Company PWR Design Certification Post Office Box 1625 Electric Power Research Institute Idaho Falls, ID 83415 3412 Hillview Avenue Palo Alto, CA 94303 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Mr. Charles Thompson, Nuclear Engineer Office of LWR Safety and Technology AP600 Certification 19901 Germantown Road U.S. Department of Energy Germantown, MD 20874 NE-451 Washington, DC 20585 4

i

MOVENFORCEMENT CONFERENCE MOVProgram History & Timeline l l 8/92 - Altran Calculation determines Core Deluge valves require 38,800 lbs. (PLTB) 3/93 - Actuator Capability (" Partial stall") test completed for Core Deluge valve to address PLTB 4/93 - NRC guidance issued - realistic valve factors 0.4 - 0.6 3/94 - GL 89-10 Sup. 6 - dynamic test results applied to grouped MOV's not tested 4/94 - IR 94-03 RG&E MOV Program well managed and executed 6/94 - EPRI MOV Performance Prediction Program released 3/95 - QA performed GL 89-10 close-out review 7/95 - RG&E submitted GL 89-10 and IR 95-06 IFI's close-out letter 8/95 - NRC issues GL 95-07 PLTB 2/96 - Comed pressure locking methodology /WOG PRESLOK 2/96 - RG&E response to GL 95-07 4/96 - ACTION Reports generated on Core Deluge valves 6/96 - NRC Request for Additional Informatiou GL 95-07 7/96 - NRC Inspection Manual revised on GL 89-10 close-out 7/96 - NRC Inspection to close-out GL 89-10 & identilles Core Deluge valve issues 8/96 - RG&E shuts dmvn plant 9/96 - RG&E responds to RAI GL 95-07 10/96 - NRC Inspection identifies two apparent violations 11/13/96 4

. .I MOVENFORCEMENT CONFERENCE Understanding of the Facts I I o Core Deluge Valve Design Control e Pcest Operability - Core Deluge Valves o Current Operability of MOVs 11/13/96 RG&E

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MOVENFORCEMENT CONFERENCE Understanding of the Facts l 1 Core Deluge Valve Design Control e ACTION Report Prepared by - Program Engineer  :

e ACTION Report Reviewed by - Lead Engineer e 10CFR 50.59 Preparer - System Engineer e 10CFR50.59 Reviewer - System Engineer e PORC Reviewed e NRC Questions & RG&E Initiates Plant Shutdown l

.c. u,1;'ee i

l MOVENFORCEMENT CONFERENCE Understanding of the Facts l

Past Operability - Core Deluge Valves e Lower DP on Reactor Side Disc under PL e Actual Stem Factor Based on Strain Gaged Stems (3 tests min) e Actual Motor Capability (Dynamometer tests) e Actual Gear Efficiency (NRC/INEL tests) e Realistic but Conservative (Not Bounding) Disc / Seat Friction Coefficient based on EPRI data v Core Deluge Valve History v EPRI Preconditioning Tests RG&E

MOVENFORCEMENT CONFERENCE Understanding of the Facts Core Deluge Valve Design Control .

o Plant Shutdown i

e Modified Valve Actuators

o Increased Margin 1

11/13 N RG&E 8

i MOV ENFORCEMENT CONFERENCE Understanding of the Facts I I  !

Core Deluge Valve Design Control e In-Situ Actuator Capability Testing in 1993 e Static Testing Core Deluge Valves 4/96 v Required Thrust Exceeds Available Thrust at Reduced Voltage e Technical Basis for Interim Use until 1997 v Justification uses 1993 In-Situ test Data v Valve Factor 0.3 used (Limitorque Manual)

RG&E t

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