ML17256A831

Submits Preliminary Evaluation of Capability of Safety & Relief Valves to Operate Under Expected Operating & Accident Conditions,Per NUREG-0578 & NUREG-0737,Item II.D.1. Evaluation Based on EPRI Valve Test Program Results

ML17256A831
Person / Time
Site:	Ginna
Issue date:	04/15/1982
From:	MAIER J E ROCHESTER GAS & ELECTRIC CORP.
To:	CRUTCHFIELD D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-0737, RTR-NUREG-578, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8204230351
Download: ML17256A831 (8)

Text

REGULATORY ORNATION DISVRISUTION SYl (RIDE)'AOCESSION NBR:8204230351, DOC.DATE: 92/00/t5 NOTARZZEDp,.NO DOCKET 0 FACILp50"244"'Robert Emmet Gfnna Nuclear'..Plantg'Un<<f<<t ii"Rochester

"-G 05000244 AUTH,N)Hg AUTHOR AFFILIATION HAIERr J~E~~Rochester Gas,5"Electric CorTps RECIP~NAME, RECIPIENT AFFIlIA'TION

'ORUTCHF IELD g 0 s Operatina Reactors Branch"5

SUBJECT:

'Subpf ts IPnel;imfnery, evaluation of qawabfli<<ty.

Of<<Safety 8 relief.-valves to operate" under egpeoted aperatfng'f accident"conditfopsipyr,NURE0~0578

'$NUREG (737pI<<tern

'$I~D~1'Evaluation-based on'.EPRI valve"test program result<<s.DISTRIBUTION iCODE:'AOASS

.<COPIES iRBDBIVED:LTR j iENCL Q SXZE:, TITLE: Response to NUREG.'737/NUREG 0660 YTHI Action Plan Rrrmts (OL's)NOTES;NRR/RL/SEP icy, 05000204 ,'RECXPIENT-IO CODE/NA'HE, ORB 05 ABC'01 INTERNAl!<<ELD IE/DEP.DIR I33, IE/DEP/EPLB NRR/DE/ADCSE "22 NRR/OE/ADSA 17 NRR/DHFS/DEPY?9 NRR/DL/ADL 16 NRR/DL/ORAB 18 NRR/DSI DIR 24 NRR/DSI/ADPS

?5 NRR/DS I/AEB NRR/DSI/RAB NRR/OST/ADT 32'GN1 FXTERNAL':

ACRST34'INPOUR'J,'STARNBS NRC PDR'02" NTI'S rCOPIES.LTTR ENC 7 1 1 1 1 1 1 1 1 1 1 1 10 1 1 1 1 IREOIPIENT-ID~CODE/NAME IE.1'2 XE/DEP/EPOS NRR/DE DER'21 NRR/DE/ADHQE "23 NRR/DHFSDIR 28 NRR/DL DIR 14 NRR/DL'/AOOR

.1'5 NRR/OSI AORS 2L7 NRR/DSI/ADGP I31 NRR/OSI'/ADRP

'26 NR R/D 8 I/678 B~N IR 30 EG"FILE'04 F<<EHA~REP, DI V LPDR 03 NSIC 05 tCOPIES'L"7T'R ENCL 1;1 1 1 1 a TOTAL NUMBER, OF COPIES REQUIRED'TTR

~'ENCl 4>>H tl, I I~~C)4>>gr>l)nl>>3 a 0(3$>>>>fIPif f>I f(,$">>('>>k tp'>>t(~'!p9 f>>t ft t t QCQ~mrna~~'I>><If ll fi~lt>>(H n t'it i.$>>H'0 LP (, it, lC Tii II>>t'I f 4 f(Vg"-3 fif PI I I y)J$~ll n t iiiiiiriiiiii i iiiiiiiiiiii Iiiii iI iiiiiiIiiiii t 4 t~Oe t t a ttts/t YORK ROCHESTER GAS AND ELECTRIC CORPORATION

~89 EAST AVENUE, ROCHESTER, N.Y.I4649 JOHN E MAILER vite President TCI.CttHONC

• tie*cooc 7io 546-2700 April 15, 1982 Director of Nuclear Reactor Regulation Attention:

Mrs Dennis M.Crutchfield, Chief Operating Reactors Branch No.5 U.S.Nuclear Regulatory Commission Washington, D.C.20555 RIRCB Vga~pa22)98<~

t0 O lrIKLIKKlstt tSttttts MlN49gg g

Subject:

PWR Safety and Relief Valve Test Program R.E.Ginna Nuclear Power Plant Docket No.50-244

Dear Mr.Crutchfield:

In accordance with the initial recommendation of NUREG 0578, Section 2.1.2 as later clarified by NUREG 0737, Item II.D.l and revised September 29, 1981, each Pressurized Water Reactor (PWR)Utility on or before April 1, 1982 was to submit information relative to the pressurizer safety and relief valves in use at their plant.Specifically this submittal should include a preliminary evaluation supported by test results which demonstrates the capability of the relief and safety valves to operate under expected operating and accident conditions.

This letter references reports provided to you by the PWR Safety and Relief Valve Test Program Subcommittee and provides additional information applicable to the Ginna Plant.Rochester Gas and Electric Corporation has actively pursued resolution of this issue through direct.participation in the Electric Power Research Institute (EPRI)valve test program.The primary objective of the Test Program was to provide full scale test data confirming the functionability of the primary system power operated relief valves and safety valves for expected operating and accident conditions.

The second objective of the program was to obtain sufficient piping thermal hydraulic load data to permit confirmation of models which may be utilized for plant unique analysis of safety and relief valve discharge piping systems.Relief valve tests were completed in August 1981,.and safety valve tests were completed in December 1981.The reports prepared for the EPRI Test Program and referenced in this submittal included:

'hh<q~]P QIJ<</p<<"/ly hf If rr If hf ,4 ROCHESTER GAS AND EL RIC CORP-DATE April 15, 1982 TO Mr.Dennis M.Crutchfield HEET NO.2 (1)"Valve Selection/Justification Re ort", December 1981 This report documents that the selected test valves represent all participating PWR safety and relief valves.(2)"Test Condition Justification Re ort", A ril 1982 and the (3)"Valve Inlet Fluid Conditions'or'ressurizer Safet and Relief'Valves xn Westin house Desi ned Plants", Februar 1982 These reports document the basis and justification of the valve test conditions for all participating PWR plants.(4)"Safet and Relief Valve Test Re ort", A ril 1982 This report provides evidence demonstrating the functionability of the selected test valves under the selected test conditions for all participating PWR plants.All of these documents have been received by Rochester Gas and Electric Corporation.

Additionally, they have been trans-mitted to you on behalf of the participating PWR Utilities from David Hoffman of Consumers Power Company, Chairman of the PWR Safety and Relief Valve Test Program Subcommittee in a letter dated April 1, 1982.Rochester Gas and Electric Corporation has undertaken review of these documents but the review, at present, is preliminary due to the preliminary nature of the test data received to date.Power Operated-Relief'alve (PORV)The PORVs at Ginna are: Manufacturer

-Copes Vulcan Model No.-Globe, D-100-160 Size S Description

-3"NPS, 316 w/Stellite Plug and 17-4PH Cage Quantity-,2t r, These valves have been tested by EPRI in a test configuration similar to that of the actual pressurizer configuration at the plant.The fluid conditions for the PORV's, while not specifically addressed in the Ginna Safety Analysis Report, were identified by Westinghouse and are discussed in the Fluid Conditions Justifica-tion Report, EPRI Report (3).The fluid conditions for the tests were selected on the basis of enveloping those fluid conditions for all plants utilizing the same valve type.Tests were performed both at the Marshall Steam Station and at the Wyle test facility (Phase III).A total of 20 tests were conducted utilizing steam, water, and steam-to-water transition fluid conditions.

These tests confirmed that the valve opened 8',+~ll'"/)t rt P 4 ROCHESTER GAS AND EL RIC CORP.DATE April 15, 1982 TO Mr.Dennis M.Crutchfield HEET NO.3 H.and closed on demand and that the valve suffered no damage that, would preclude future operations Specific information related to these tests is given in Section 4.6 of the"Safety and Relief Valve Test Report", EPRI Report (4)~Safet Valves The safety valves at Ginna are-.Manufacturer

-Crosby Valve and Gage Model No.-HB-BP-86 Size-4K26 Assembly No.-51249 Quantity-2 These valves have not been tested by EPRI.However safety valves similar in design and operation have been tested in a test configuration similar to that of the actual pressurizer configuration at the plant.Justification that the valves tested envelope those valves at Ginna is given in the Valve Justification Report, EPRI Report (1).Justification that the fluid conditions for the tests envelope the actual plant conditions is presented in EPRI Report (3).The safety valves have been shown to open and close.Certain areas of system performance are still undergoing review, specifically addressing the inlet piping, outlet piping and the valves.The Rochester Gas and Electric Corporation is actively involved with the Westinghouse Owners Group program and has additionally contracted Westinghouse to evaluate the specific Ginna system, including downstream piping.All relevant EPRI results will be employed in the evaluation.

Clarification of concerns and their impact on plant performance is expected prior to the July 1, 1982 information submittal date.Our preliminary evaluation of the EPRI work is that the valves tested represent the Ginna safety and relief valve designs and that the conditions tested envelope the range of expected operating and accident conditions for the Ginna Plant-The EPRI reports provide the evidence required by NUREG 0737, Item II.D.l.A which will be used to perform the final plant specific evaluations.

The September 29, 1981 USNRC letter requested that plant specific final evaluations be submitted by July 1, 1982'n order to meet that date, evaluations have been initiated.

Depending on the outcome of the evaluations it may be necessary to continue the evaluation beyond July 1, 1982-If a longer evaluation period is required, you will be notified.Very truly yours,~/u+.~~John E.Maier g l I@II r,~qr~rw, l'q g I II'I