ML18153A465

LER 96-003-00:on 960512,Unit 2 Pressurizer Safety Valve as Found Lift Setting Out of Tolerance.Valve Was Reassembled & Lift Setting Was Established & Tested Satisfactorily. W/960607 Ltr

ML18153A465
Person / Time
Site:	Surry
Issue date:	06/07/1996
From:	Christian D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
96-296, LER-96-003-01, LER-96-3-1, NUDOCS 9606120014
Download: ML18153A465 (6)

Text

\ CA~~~UMY i AUTH.NAME

.e e REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9606120014 DOC.DATE: 96/06/07 NOTARIZED: NO FACIL:50-281 Surry Power Station, Unit 2, Virginia Electric & Powe AUTHOR AFFILIATION I

DOCKET#

05000281 CHRISTIAN,D.A. Virginia Power (Virginia Electric & Power Co.)

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 96-003-00:on 960512,Unit 2 pressurizer safety v~lve C as found lift setting out of tolerance.Valve was reassembled

& lift setting was established & tested sati$factorily. A W/960607 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR l_ ENCL TITLE: 50.73/50.9 Licensee Event Report (LER)~ Incident Rpt, etc.

_1 SIZE: ~ T E

NOTES: 05000281 G 0

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL R PD2-l PD 1 1 EDISON,G. 1 1 INTERNAL: ACRS AEOD/SPD/RRAB NRR/DE/ECGB NRR/DE/EMEB NRR/DRCH/HICB 1

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EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H- 2 2 NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 1 C NRC PDR 1 1 NUDOCS FULL TXT 1 1 NOTES: 1 1 u

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NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROo'M OWFN SD-S(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

. FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 27

10CFR50.73 Virginia Electric and Power Company Surry Power Station P. 0. Box 315 Surry, Virginia 23883 June 7, 1996 U.S. Nuclear Regulatory Commission Serial No.: 96-296 Document Control Desk SPS: BCB Washington, D. C. 20555 Docket Nos.: 50-281 License Nos.: DPR-37

Dear Sirs:

Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Surry Power Station Unit 2.

REPORT NUMBER 50-281 /96-003-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Management Safety Review Committee for its review.

Very truly yours, 1.

D. A. Christian Station Manager Enclosure pc: Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 M. W. Branch NRC Senior Resident Inspector Surry Power Station 12 OO1*.2 9606120014 960607 PDR ADOCK 05000281 S PDR

NRC FORty,366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION LICENSEE EVENT REPORT (LEA) AND RECORDS MANAGEMENT BRANCH (MNBB n14), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20~1. AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET. WASHINGTON, DC 20503.

FACILITY NAME (1) II PAGE(3)

Surrv Power Station, Unit 2 05000 - 281 1 OF 4 TTT1.E (4)

Unit 2 Pressurizer Safety Valve As-Found Lift Setting Out of Tolerance EVENT DATE 5) LER NUMBER (6 REPORT DATE 7) OTHER FACILmEs INVOLVED (Bl MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 05000*

FACILITY NAME DOCKET NUMBER 05 12 96 96 -- 003 -- 00 06 07 96 05000*

OPERATING THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR:(Check one or more) (11)

MODE (9) N 20.402(b) 20.405(c) 50. 73(a)(2)(iv) 73.71(c)

POWER 20.405(a)(1 )(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

LEVEL (10) 0% 20.405(a)(1 )(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER 20.405(a)( 1)(iii) 50.73(a)(2)(viii)(A) (Specify in Abstract below and X 50. 73(a)(2)(i) 20.405(a)(1 )(iv) 50.73(a)(2)(ii) 50. 73(a)(2)(viii)(B) in Text. NRC Form 366A) 20.405(a)(1 )(v) 50. 73(a)(2)(iii) 50.73(a)(2)(x)

LICENSEE CONTACT FOR THIS LEA 112)

NAME I TELEPHONE NUMBER (Including Area Code)

D. A. Christian, Station Manager (804) 357-3184 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT C13l CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS X AB RV C710 y SUPPLEMENTAL REPORT EXPECTED 14) I EXPECTED I MONTH I DAY I YEAR I YES mves comDlete EXPECTED SUBMISSION DATE) XI NO I SUBMISSION DATE (15)

I I I ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On May 12, 1996, with Unit 2 at Refueling Shutdown, setpoint testing revealed that the as-found lift setting for pressurizer safety valve (PSV) 2-RC-SV-2551 B (2379 psig) was out of tolerance. Technical Specification (TS) 3.1.A.3.b requires the as-found lift setting for the PSVs to be 2485 psig +/- 3% (2411 to 2559 psig). A Category II Root Cause Evaluation reviewed the previous test data for 2-RC-SV-2551 Band the methodology and conditions associated with the as-found testing. No discrepancies were identified relative to the conduct of the as-found testing and no adverse conditions or precursor problems were identified from the review of previous test data. To ensure internal integrity, 2-RC-SV-2551 B was disassembled and inspected by the valve vendor. The inspection revealed no mechanical abnormalities and verified that critical tolerances and the seating surfaces of the valve were acceptable. The valve was reassembled and the lift setting was established and tested satisfactorily. The as-left lift settings were within 2485 psig

+/- 1% as required by TS 3.1 .A.3.b. Based on the results of the reviews and inspections, the cause of this event could not be conclusively determined. This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B).

NRC FORM 366 (5-92)

e e NRC FORM *355 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104

• (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION LICENSEE EVENT REPORT (LEA) ANO RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, ANO TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET. WASHINGTON, DC 20503.

FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 161 PAGE 13)

YEAR SEQUENTIAL NUMBER REVISION NUMBER Surry Power Station, Unit 2. 05000 - 281 96 - 003 - 00 2 OF 4 TEXT (II more space is required, use additional copies of NRC Form 366A) (17)

1.0 DESCRIPTION

OF THE EVENT.

o*uring the 1996 Unit 2 refueling outage, the three pressurizer safety valves (PSV) were removed and sent to Wyle Laboratories for setpoint testing. The testing was performed in accordance with station testing procedures and the Surry Power Station ASME Section XI Testing Program for Pumps and Valves.

On May 12, 1996, with Unit 2 at Refueling Shutdown, the setpoint testing revealed that the as-found lift setting for PSV 2-RC-SV-2551 B (2379 psig) was out of tolerance. The results of subsequent valve lifts (performed for information) were within the as-found tolerance. Technical Specification (TS) 3.1.A.3.b requires the as-found lift setting for the PSVs [EIIS-AB,RV] to be 2485 psig +/- 3% (2411 to 2559 psig). This deviation from the allowable tolerance represents a condition prohibited by TS and, therefore,* is being

• reported pursuant to 10 CFR 50.73(a)(2)(i)(B).

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS An engineering evaluation was performed to determine the impact of the low as-found lift setpoint for PSV 2-RC-SV-2551 B. The primary concern associated with this condition is the potential for the PSV to open before a high pressurizer pressure trip (HPPT) could occur. To address this concern, the as-found HPPT setpoints were evaluated relative to the as-found PSV setpoint. Since the as-found HPPT setpoints were less than the as-tound PSV setpoint, engineering determined that a reactor trip on high pressurizer pressure would have occurred befor~ PSV 2-RC-SV-2551 B would have opened.

Based on the discussion above, the engineering evaluation concluded that the low as-tound lift setpoint for PSV 2-RC-SV-2551 B did not place the plant in an unanalyzed condition. Therefore, this event resulted in no safety consequences or significant implications and the health and safety of the public were not affected at any time.

NRC FORM 366A (5-92)

e e NRC FORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104

,a\ (5-92) - EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST:50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION LICENSEE EVENT REPORT (LER) AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 2055&0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF

{See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME 111 DOCKET NUMBER 121 LEA NUMBER 161 PAGE 131 YEAR SEQUENTIAL NUMBER REVISION NUMBER Surry Power Station, Unit 2 05000 - 281 96 - 003 - 00 30F4 TEXT (If more space is required, use additional copies or NRC Form 366A) (17) 3.0 CAUSE A Category II Root Cause Evaluation (RCE) was performed to determine the cause of the deviating condition. The RCE included a review of the methodology and conditions associated with the as-found testing, and a review of the previous test data for 2-RC-SV-25518. No discrepancies were identified relative to the conduct of the as-found testing and no adverse conditions or precursor problems were identified from a review of previous test data.

To ensure internal integrity, 2-RC-SV-2551 B was disassembled and inspected by the valve vendor, Crosby Valve and Gage Company. The inspection revealed no mechanical abnormalities and verified that critical tolerances and the seating surfaces of the valve were acceptable.

The PSVs currently used at Surry are a new design that have been in service for less than two operational cycles. Consequently, insufficient test data exists to enable the positive identification of any developing trends.

Based on the results of the reviews and inspections described above, the cause of this event could not be conclusively determined.

4.0 IMMEDIATE CORRECTIVE ACTION(S)

No immediate operational corrective actions were necessary since Unit 2 was at Refueling Shutdown.

A Deviation Report was submitted to document that the as-found lift setting for PSV 2-RC-SV-25518 was out of tolerance.

5.0 ADDITIONAL CORRECTIVE ACTIONS As described in Section 3.0, PSV 2-RC-SV-2551 B was disassembled and inspected by the vendor. Minor lapping was performed to restore the surfaces of the valve's disc insert and nozzle seats. The valve was reassembled and the lift setting was established and tested satisfactorily. The as-left lift settings were within 2485 psig +/- 1% as required by TS 3.1 .A.3.b.

NRC FORM 366A (5-92)

. e e t-.lRC FORM. 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (5-92)

• EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION LICENSEE EVENT REPORT (LEA) AND RECORDS MANAGEMENT BRANCH (MNB8 n14), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC ~ 1 . AND TO TiiE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 161 PAGE (3)

YEAR SEQUENTIAL NUMBER REVISION

• NUMBER Surry Power Station, Unit 2 05000 - 281 96 - 003 - 00 40F4 TEXT (H more space is required, use additional copies of NRC Form 366A) (17) 6.0 ACTIONS TO PREVENT RECURRENCE The results of PSV testing is monitored to identify any trends that may be indicative of valve degradation. When test results are unsatisfactory, Virginia Power's testing program requires that the testing be performed on a more frequent basis than that required by the ASME Section XI Testing Program for Pumps and Valves.

7.0 SIMILAR EVENTS During setpoint testing on September 19, 1995, the as-found lift setting for Unit 1 PSV 1-RC-SV-1551C was determined to be out of tolerance. The cause of this event was attributed to the test methodology used by the valve manufacturer to test the valve prior to installation. Licensee Event Report (LER) 50-280/95-008-00 reported this event pursuant to 10 CFR 50. 73(a)(2)(i)(B).

During setpoint testing on February 10, 1995, the as-found lift settings for, Unit 2 main steam safety valves 2-MS-SV-202C and 2-MS-SV-205A were determined to be out of tolerance. During setpoint testing on February 15 and 16, 1995, the as-found lift settings for Unit 2 PSVs 2-RC-SV-2551 B and 2-RC-SV-2551 C were determined to be out of the allowable+/- 1% tolerance. Note: the tolerance was subsequently changed to+/- 3%. The cause of this event was attributed to setpoint drift. Licensee Event Report (LER) 50-281/95-002-00 reported this event pursuant to 10 CFR 50.73(a)(2)(i)(B).

8.0 MANUFACTURER/MODEL NUMBER Crosby Valve and Gage Company Pressurizer Safety Valve 6K26 Model No. HB-86-BP 9.0 ADDITIONAL INFORMATION Unit 1 was operating at 100% power and was not affected by this event.

NRC FORM 366A (5-92)