ML18153C754

LER 91-009-00:on 910903,determined That Check Valve 2-RH-47 Not Full Flow Tested During Cycle 10 Refueling Outage.Caused by Personnel/Procedural Error.Procedure Changed & Testing Performed on 910908.W/910930 Ltr

ML18153C754
Person / Time
Site:	Surry
Issue date:	09/30/1991
From:	Kansler M VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
91-575, LER-91-009-02, LER-91-9-2, NUDOCS 9110070353
Download: ML18153C754 (6)

Text

~ ACCELERATED DliRIBUTION DEMONSTVTION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9110070353 DOC.DATE: 91/09/30 NOTARIZED: NO DOCKET#

FACIL:50-281 Surry Power Station, Unit 2, Virginia*Electric & Powe 05000281 AUTH.NAME AUTHOR AFFILIATION KANSLER,M.R. Virginia Power (Virginia Electric & Power Co.)

RECIP.NAM'E RECIPIENT AFFILIATION R

SUBJECT:

LER 91-009-00:on 910903,determined that check valve 2-RH-47 not full flow tested during Cycle 10 refueling I outage.caused by personnel/procedural error.Procedure changed & testing performed on 910908.W/910930 ltr. D DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR _l ENCL j_ SIZE:_S _ _ __ s TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

I NOTES:lcy NMSS/IMSB/PM. 05000281 A

RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 D BUCKLEY,B 1 1 INTERNAL: ACNW 2 2 ACRS 2 2 s AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFBlO 1 1 NRR/DLPQ/LPEBlO 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPBll 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 1 1 NRR/DST/SRXB 8E 1 1 ~~/D~iF~:-aP-1 1 1 RES/DSIR/EIB 1 1 RG~-*-**-~ 1 1 EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 R NOTES: 1 1 I D

s I

A D

D NOTE TO ALL "RIDS" RECIPIENTS:

s PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P!-37 (EXT. 20079) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 34 ENCL 34

Virginia Electric and Power Company Surry Power Station P. 0.Box315 Surry, Virginia 23883 September 30, 1991 U. S. Nuc:lear Regulatory Commission Serial No.: 91-575 Document Control Desk Docket No.: 50-281 Washingfan, D. C. 20555 License No.: DPR-37 Gentlemen:

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

REPORT NUMBER 91-009-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by the Corporate Management Safety Review Committee.

Very truly yours, Enclosure cc: Regional Administrator Suite 2:900 101 Marietta Street, NW Atlanta, Georgia 30323 9110070353 910930 F,_'DR Ar-11.J1"'_.1.*.*.

1 r rJ._,(1(10281 c:-

.::, PDR

NRC FORM 366 (6-89) e U.S. NUCLEAR REGULATORY COMMISSION e APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LERI COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER 12) I PAGE 131 TITLE 141 Surry Power Station, Unit 2 Io I s I o I o I o 12 I 8 1l 1 loF O I 4 Failure to Full Flow Test 2-RH-47 Due to Procedure Deficiency EVENT DATE (5) LER NUMBER 161 REPORT DATE 17) OTHER FACILITIES INVOLVED 18)

MONTH DAY YEAR YEAR f[:{ sEie~~~~~AL (j ~~~~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERIS)

THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUIREMENTS OF 10 CFR §: (Ch*ck on* or more of th* following/ 111)

OPERATING MODE IBl N

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60.73l*ll2llii) 50.73(all2lliiil LICENSEE CONTACT FOR THIS LER (121

- 50.731all2llvlill1Bl 50.731all2llxl NAME TELEPHONE NUMBER M. R. Kansler, Station Manager AREA CODE 8 10 I 4 3 I 51 7 I -1 3 1l 1 8( 4 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 CAUSE SYSTEM COMPONENT MANUFAC* REPORTABLE r=:r=::\:)):\\\\\\: CAUSE SYSTEM COMPONENT MANUFAC-TURER TO NPRDS :-:-:-:-:-:-:-:-:-:-:-:-:-:-:-:-:*:*:*:*:*:-:*:*:-:* TURER 11 111 11 I I I I I I I  ::li!i1:iiii ii i!il:::i ! 1!11!!1i::ii i :ii:li!i:: I I I I I I I

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I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAY YEAR EXPECTED I YES /If yes, compl*to EXPECTED SUBMISSION DA TEI ABSTRACT (Limit to 1400 spaces, i.e., spproximt1ttJly fiftt1t1n sing/6-spaca typewritten lines) 1161 rxi NO SUBMISSION DATE 1151 I I I On September 3, 1991, with Unit 2 at 59% reactor power, it was determined that check valve 2-RH-47 had not been full flow tested during the Unit 2 Cycle 10 refueling outage as required by the Unit 2 ASME Section XI Inservice Testing Program Revision 2. This determination was made during the station inservice inspection group's independent review of the implementation of the Surry ASME Section XI Program. This event occurred as the result of personnel procedural error in that the test procedure used to satisfy the full flow testing requirement contained an improper valve lineup. Following this discovery, the affected procedure was changed and full flow testing was satisfactorily performed on September 8, 1991 during a forced outage. Because the system had been verified capable of delivering design basis flow during the Unit 2 Cycle 10 refueling outage and subsequent valve testing determined that 2-RH-47 was operable, no safety implications were posed by this event. This missed surveillance test was determined to be a violation of Technical Specification 4.0.3 and is being reported pursuant to 10CFR50.73(a)(2)(i)(B).

NRC Form 366 (6-89)

ij NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3150-0104

• (6-89)

EXPIRES: 4/30/92 ES* TED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE,,REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME 11) DOCKET NUMBER 12) PAGE 13)

J LEA NUMBER 16)

Surry Power Station, Unit 2 TEXT /ff mom lf/NJt:e i.s n,quimd, use additional NRC Form .1151iA 'a) 117) 1.o DESCRIPTION OF THE EVENT On September 3, 1991, with Unit 2 at 59% reactor power, it was determined that check valve 2-RH-47 [EIIS-BP,V] had not been full flow tested during the Unit 2 Cycle 10 refueling outage as required by the Unit 2 ASME Section XI Inservice Testing (1ST) Program Revision 2. This determination was made during the station inservice inspection group's independent review of the implementation of the Surry ASME Section XI Program.

The Surry Unit 2 Residual Heat Removal (RH) system has a common discharge header which splits into separate discharge headers at RH loop isolation valves. Flow in one loop passes through a loop discharge check valve, 2-RH-47, before discharging to the B reactor coolant loop, whereas flow in the other loop discharges directly into the C reactor coolant loop. The normal RH system operating procedures specify both RH loop isolation valves be in the open position when the RH system is in operation.

I 1---~:TORCSLOOPB 1

2-RH-MOV-2720A 2-RH-47 FROMRH-_.....i 1------------ ro RCS LCXJP c 2-RH-MOV-27208 The requirement to full flow test 2-RH-47 was added to the 1ST Program during Revision 2 and was scheduled to be performed for the first time during the Unit 2 Cycle 10 refueling outage. An existing RH system test procedure, which verified full flow with both RH loop isolation valves open, was revised to include additional steps for full flow testing 2-RH-

47. It was intended that these steps shut the RH loop isolation valve to the C reactor coolant loop (2-RH-MOV-2720B), thereby directing full system flow through 2-RH-47. With full flow directed through 2-RH-47, an observed flow rate of greater than or equal to 4,000 gpm was NRC Farm 366A 16-89)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92 E* ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE' REPORT (LEA) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1"HE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

Surry Power Station, Unit 2 TEXT (ff mom /I/JIIC8 ,. fWIUimd,. UIS(J additional NRC Fann .11515A '*) (17) established as the acceptance

• criteria to indicate the valve had fully opened.

Due to a transposition error in valve mark numbers which occurred during preparation of the revision to the implementing procedure, these steps did not provide the proper valve lineup for directing system flow through 2-RH-47. The flow path specified isolated 2-RH-47 (shut 2-RH-MOV-2720A) and directed flow through the header discharging to the C reactor coolant loop. Due to the system configuration and the specified acceptance criteria, the acceptance criteria for this test could be satisfied with the improper lineup without actually full flow testing the valve. Consequently, when the test was performed during the Unit 2 Cycle 10 refueling outage, 2-RH-47 was recorded as having been full flow tested satisfactorily.

This occurrence was determined to be a violation of Technical Specification 4.-0.3 and is being reported pursuant to 10CFR50. 73(a)(2)(i)(B).

2 *O SIGNIFICANT SAFETY CQNSEOUENCES AND IMPLICATIONS No safety implications were posed by this event. Full flow had been verified with both RH loop isolation valves open and with only the RH loop isolation valve to the C reactor coolant loop open, during the Unit 2 Cycle 10 refueling outage. In addition, full flow testing performed on September 8, 1991 demonstrated that 2-RH-47 was operable. Since the RH system had been verified capable of meeting its design basis, the health and safety of the public were not affected.

3.0 CAUSE OF THE EVENT This event was caused by a personnel procedural error in that an improper flow path for conducting the full flow testing was specified during preparation of the implementing test procedure.

4. O JrMMEDIATE CORRECTIVE ACTION<Sl A station deviation was submitted and previous operations during shutdown conditions were reviewed. Based upon these reviews, which i:ncluded evaluation of the test data taken during the refueling outage, it was determined that required system flow rate was achieved and thus system operability was not affected.

NRC Form 366A (6-89)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION II (6-89) APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92 E* ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE' REPORT (LERI INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

LEA NUMBER (6)

,:,:-:,:,: SEQUENTIAL ,;,:,;,;,:, REVISION Surry Powe:r Station, Unit 2 YE,AR ::::::::::: NUMBER  ::,:::::::' NUMBER TEXT '" mon 8{J/JC9 a raquimd, UM additiotllll NRC Form 316&<1 's) (17) 5.0 ADDITIONAL CORRECTIVE ACTION<Sl The test procedure was changed and full flow testing of 2-RH-47 was performed satisfactorily September 8, 1991.

Station deviation trends were reviewed which indicated that the occurrence of personnel procedural deficiencies had declined approximately 61 percent over the past year. Additionally, this occurrence was discussed with quality assurance personnel who indicated that procedural deficiencies were rarely identified during the review of upgraded procedures. Consequently, it was concluded that this event was an isolated occurrence.

6. O ACTIONS TO PREVENT RECURRENCE The individuals involved in the preparation of the procedure change with responsibility for its technical content, were counseled on their preparation and review responsibilities.

The independent review of IST Program implementation will continue to completion.

7.0 SIMILAR EVENTS LER 89-019-00: Unplanned Engineered Safety Features Component Actuation, Feedwater Bypass Valve Closure and Main Feed Pump Trip, Due to Personnel Error and Procedure Deficiency.

LER 90-005-00: Containment Pressure Channel Rendered Inoperable and Not Placed in Trip Condition Due to Procedure Inadequacy.

LER 90-018-00: Unplanned Engineered Safety Features Actuation Due to Inadequate Procedures

8. O ADDITIONAL INFORMATION None.

NRC Form 366A (6-89)