ML20059F866

Confirms Info Provided by 931015 Telcon Re LCR 92-06, SRV Testing Requirements

ML20059F866
Person / Time
Site:	Hope Creek
Issue date:	10/29/1993
From:	Hagan J Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLR-N93170, NUDOCS 9311050132
Download: ML20059F866 (4)

Text

4 PubitC Service Doctoc and Gas Company Joseph J, Hagan Pubhc Service Eiecinc and Gas Cornpany P.O. Box 236, Hancocks Bridge. NJ 08038 609-3394200 t

va n ,uu.n . wmar orem. o OCT 231993  !

NLR-N93170 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

OCTOBER 15, 1993 TELECONFERENCE LICENSE CHANGE REQUEST 92-06  !

HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 ,

DOCKET NO. 50-354  !

The purpose of this letter is to confirm the information previously provided by PSE&G during an October 15, 1993 teleconference concerning License Change Request (LCR) 92-06, ,

Safety Relief Valve (SRV) Testing Requirements. Attachment 1 of this letter provides a list of those NRC questions and PSE&G responses requiring confirmation via this letter. Please note that this submittal does not change LCR 92-06 nor modify any of the conclusions stated in the Significant Hazards Consideration Evaluation previously submitted.

A copy of the this letter has been sent to the State of New Jersey.

Should you have any questions or comments on this transmittal, do not hesitate to contact us.

Sincerely,

/

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i l

i 0400 %

9311050132 931029 PDR

" SDDI

) I P ADOCK 05000354 I PDR -

a OCT 291993 )

l Document Control Desk -

NLh-N93170 Attachment C Mr. T. T. Martin, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. S. Dembek, Licensing Project Manager U. S. Nuclear Regulatory Commission .;

One White Flint North 11555 Rockville Pike >

Rockville, MD 20852  !

Mr. C. S. Marschall (SO9)

USNRC Senior Resident Inspector Mr. K. Tosch, Manager IV ,

NJ Department of Environmental Protection  !

Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 I

I t

1 i

ATTACHMENT 1  ;

OCTOBER 15, 1993 TELECONFERENCE QUESTIONS CONCERNING LICENSE CHANGE REQUEST 92-06, SAFETY RELIEF VALVE  !

TESTING REQUIREMENTS HOPE CREEK GENERATING STATION NLR-N93170 [

FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 NRC OUESTION 1:

What is the basis for believing that removal of the first stage ,

and reassembly with a refurbished first stage does not affect  !

valve performance, thereby making complete post-maintenance i testing of the valve assembly unnecessary?

PSE&G RESPONSE:  !

The pilot stage assembly contair s the pilot spring assembly.

Preload force on the pilot spring assembly transmits force to the ,

pilot disk assembly, thereby establishing the set-point. The set-point and blowdown adjustments are made at the pilot '

assembly. PSE&G's experience with SRV failures (see response to Question 4) has indicated that all of the as-found testing  !

failures were the result of pilot stage set-point failure. l The main stage consists of a main disc / piston assembly and l calibrated spring. The SRV set-point is not established or '

affected by the main disc / piston assembly. This stage of the SRV will be tested to confirm blowdown rate, reinstalled or replaced with spares that have been previously tested and stored in accordance with manufacturer's recommendations at least once '

every five years.

NRC OUESTION 2:

What actions are planned with the removed first stage, specifically, 1) is as-found testing of the first stage planned or is as-left testing of the refurbished first stage going to be performed, 2) if as-found testing is not to be done, how is the intent of OM-1 being met, and 3) if as-found testing is being done, what sample expansion plan is going to be followed?

PSE&G RESPONSE:

As-found testing of the first stage will be performed in addition to as-left testing. As a result of the as-found testing being performed, the intent of OM-1 will be met. The sample expansion plan will be conducted according to ASME Boiler & Pressure Vessel l Code,Section XI, IWV-3513, 1983 edition.

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Attacha2nt 1 LCR 92-06 Teleconference Questions NLR-N93170 .

NRC OUESTION 3: ,

Does the language of the proposed alternate testing with respect to the main (mechanical) portion mean that the entire assembly is set-point tested every five years?

PSE&G RESPONSE:

Yes, the entire assembly will be set-point tested every five years.

NRC OUESTION 4:

Could you please provide data from the last 3 refueling outages on set-point testing, including any failures?

PSE&G RESPONSE: .

i The following table provides information from refueling outages 2, 3, and 4:

VALVE SET-POINT R2 S/N TEST R3 s/N TEST R4 s/N TEST A 1130 357 1146a 365 1159* 346 1214*

8 1130 360 1136 370 1168* 357 1133 C 1130 346 1164* 348 1128 360 1195*

D 1130 350 1152* 352 1151* 366 1200* ,

E 1130 366 1182* 349 1141 350 1190* $

354 1141* 354 F 1108 1128* 359 1071*

G 1*?O 343 1149' 343 1195* 364 1175*

H 1108 368 1119 344 1127* 368 1126' J 1120 363 1048* 353 1187* 363 1110 K 1108 347 1133* 355 1115 355 1130*

L 1120 351 1152* 358 1131 367 1147*

M 1108 369 1115 369 1107 369 1158*

P 1120 3 64 1167* 345 1150* 351 1143*

R 1120 367 1128 356 1189* 356 1199*

Acceptance criteria for set-point tests are 2 1 % of set-point value (per Tech spec 3.4.2.1).

All data is "As found"

• = Failed Test S/N = Serial Nucer l Page 2 of 2