ML17279A776

LER 87-033-00:on 871218,plant Tech Spec Sections 4.0.5 Re ASME Valve in-service Testing Requirements & 3.6.3 Concerning Containment Valve Operability Not Met.Caused by Personnel Error.No Corrective action.W/880115 Ltr

ML17279A776
Person / Time
Site:	Columbia
Issue date:	01/15/1988
From:	Powers C, Washington S WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LER-87-033, LER-87-33, NUDOCS 8801200325
Download: ML17279A776 (7)

Text

AC CELZMTKD DISHDBUTION 'EMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8801200325 DOC.DATE: 88/01/15 NOTARIZED: NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION WASHINGTON,S.L. Washington Public Power Supply System POWERS,C.M. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 87-033-00:on 871218,Tech Spec violation caused by missed

-'ASME valve operability surveillances due to personnel error.

W/8 ltr.

DISTRIBUTION CODE: IE22D COPIES RECEIVED:LTR 3- ENCL ASSIZE:

TITLE: 50.73 Licensee Event Report (LER), Incident Rpt, etc. D NOTES: 8 RECIPIENT COPIES RECIPIENT 'OPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD5 LA 1 1 PD5 PD 1 1 SAMWORTH,R 1 1 A INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 AEOD/DOA 1 1 AEOD/DSP/NAS 1 1 AEOD/DS P/ROAB 2 2 AEOD/DSP/TPAB 1 1 ARM/DCTS/DAB 1 1 DEDRO 1 1 NRR/DE ST/ADS 1 0 NRR/DEST/CEB 1 1 NRR/DEST/ELB 1 1 NRR/DEST/ICSB 1 1 NRR/DES T/MEB 1 1 NRR/DEST/MTB 1 1 NRR/DEST/PS B 1 1 NRR/DEST/RSB 1 1 NRR/DEST/SGB 1 1 NRR/DLPQ/HFB 1 1 NRR/DLPQ/QAB 1 1 NRR/DOEA/EAB 1 1 NRR/DREP/RAB 1 1 NRR/DREP/RPB 2 2 N SIB 1 1 NRR/PMAS/ILRB 1 1 EGF LE 02 1 1 RES TELFORD,J 1 1 B 1 1 RES/DRPS DIR 1 1 RGN5 FILE 01 1 1 EXTERNAL: EGS(G GROH, M 5 5 FORD BLDG HOY,A 1 1 H ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC HARRIS,J 1 1 NSIC MAYS,G 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR '6 ENCL 45

NRC Form 388 U.S. NUCLEAR REOULATORY COMMISSION (94>3)

APPROVED OMB NO. 3150-0104 EXPIRES: SI31ISS LICENSEE EVENT REPORT (LER)

DOCKET NUMBER (2> PACE 3 FACILITY NAME HI 1 uclear Plant - Unit 2 0 5 0 0 0 3 9 7 1 OF 0 4 Plant Technical Specification Violation Caused by llissed ASHE Valve Operability Surveillances Due to Personnel Error EVENT DATE (5) LER NUMBER (SI REPORT DATE (7) OTHER FACILITIES INVOLVED (8) fr W'EOUSNTIAL RSV~N OAY YEAR FACILITYNAMES DOCKET NUMBERISI MONTH DAY YEAR YEAR vrSS NUMBER NUMBER MONTH 0 5 0 0 0 21 887 8 7 0 3 3 0 00 115 88 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE REOUIAEMENTS OF 10 CFR (I. (Check oni Or more Of thi follovffnPI (11 OPERATING MODE (9) 20.402(bl 20.405(cl 50.73( ~ ) (2) (lvl 73.71(b)

POWER 20.406 (e l(1) Ill 50.38(cl(ll 50.7 3(e I (2)(v) 73.71(c)

LEVEL (10) 20.405(e l(1>(ill 50.35 (c) (2) 60,73( ~ )(2)(viI) OTHER ISpeclfy in AbtVict below ehd Ih Tint NIIC Form 20.405 (e) (1 I (illI 50.'73(el(2) II) 60.73(e)(2)(vill) IAI 34BAI 20A05(e) (1)(lv) 50.73(e l(2)(ill 50.73(e)(2)(vlii) IB) 20.405(el(1 l(vl 60,73(e)(2)BII) 60.73( ~ l(2) (el LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER AREA CODE S.L. Washin ton Com liance En ineer 50 93 77- 208 0 COMPLETE ONE LINE FOR EACH COMPONENT FAILUAE DESCRIBED IN THIS REPORT (13)

MANUFA(r REPORTABLE COMPONENT MANUFAC. ) PORTABLE CAUSE SYSTEM COMPONENT CAUSE SYSTEM TVRER TO NPRDS P<jQ~ Qe TVREA gjp~@

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SUPPLEMENTAL REPORT EXPECTF D (1 ~ I MONTH OAY YEAR EXPECTED SUBMISSION DATE V5)

YES flfyit, compliti EXPECTED SIIBkIISSION DATEI NO ABSTRACT ILlmlt to 1400 tPicit, I iv ePProxlmetely fifteen tlnPle tPete tyPiwrlttin lintel (15)

On December 16, 1987, while perforning an Operational Quality Assurance Surveillance, a Plant Quality Assurance engineer discovered a condition that potentially violated WHP-2 Plant Technical Specifications. The surveillance information was given to a Plant engineer who determined, on December 18, 1987 that WHP-2 Plant Technical Specification Sections 4.0.5, ASI1E valve in-service testing requirements, and 3.6.3, primary containment valve operability requirements, for the Post Accident Sampling System (PASS) had on two occasions not been net.

During the 1986 and 1987 Spring refueling and maintenance outages, the PASS Process Sample Radiation (PSR) primary containment isolation valve ASI1E in-service testing surveillance was not performed prior to Plant startup. The surveillance interval allowable by WHP-2 Plant Technical Specification 4.0.2 was exceeded and, therefore, changing to Operational Nodes 1, 2, or 3 from Operational trode 4 was not allowed by Technical Specification 3.6.3 since the PSR valves were by technical specification requirements inoperable. The two time periods when the Plant was not in compliance with Technical Specifications were from June 4, 1986 to July 12, 1986 and from August 1, 1987 to August 10, 1987.

The cause of this event is personnel error in that Plant Chenistry personnel assumed responsibility for the UPSR Valve Operability" surveillance without adequately understanding the surveillance requirements.

At the time of discovery, the Plant was in cor1pliance with these Technical Specification Sections and no immediate corrective action was required. Corrective actions to be taken to prevent recurrence of this event include: training for applicable personnel on the PSR Valve Operability Technical Specification surveillance requirements, a procedure revision that will include more guidance on surveillance completion requirements, and reinstruction on the requirements in the Plant procedure aoverning Technical Specification surveillance perfornance. 8801200325 880115 PDR ADOCK 05000397 WE-.~i

NRC form 366A U.S. NUCLEAR REGULA'TORY COMMISSION 0831 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO. 3150-0104 EXPIR ESI'8/31/88 6 ACII.ITYNAME 11) OOCKET NUMSER LTI LER NUMBER 161

,'2<'3 66ovENTrAL ~+I RIyrsroN YEAR NVMOER re NVMOIR Washington nuclear Plant - Unit 2 p s p p p 3 9 7 8 7 033 0 0 02oFO 4 TEXT //l /rroro Fpooo /F FOrrkor/, ooo oPOoro'on/I /YRC forrrr 3//54'Fl (ITI Abstract (Continued)

There is no safety significance associated with this event since subsequent surveillances verified the operability of these valves.

Plant Conditions a) Power Level - 94K b) Plant Mode - 1 (Power Operation)

Event Descri tion On December 16, 1987, while performing an Operational guality Assurance Surveillance, a Plant guality Assurance engineer discovered a condition that potentially violated WNP-2 Plant Technical Specifications. The surveillance information was given to a Plant engineer who determined on December 18, 1987 that WNP-2 Plant Technical Specification Sections 4.0.5, ASHE valve in-service testing requirements, and 3.6.3, primary containment valve operability requirements, for the Post Accident Sampling System (PASS) had on two occasions not been met.

The Plant PASS is designed for obtaining both primary liquid and gaseous samples. The PASS sample lines from the Primary Reactor Containment are required to be isolable and 20 Process Sample Radiation (PSR) valves provide this isolation function. These PSR valves are ASHE valves and, therefore, must meet the in-service testing requirements of Section XI of the ASHE 8oi ler and Pressure Yessel Code. The in-service testing and surveillance requirements for the PASS PSR primary containment isolation valves are satisfied by performance of Plant Surveillance Procedure PPH 7.4.0.5.51, RPSR Yalve Operability and PASS Operability." On January 10, 1986, the UPSR Yalve Operability and PASS Operability" surveillance was completed. It was scheduled to be completed again on March 23, 1986 by the Plant Scheduled Maintenance System (SHS - a computerized tracking system). On March 31, 1986, the SHS computer card (a notification of surveillance due) was turned in with the note that the surveillance was not required due to the Plant being in Operational Mode 4 "Cold Shutdown." ASME Section IWY-3416 does not require in-service testing for valves in systems not required to be operable; however, this section does require testing prior, to declaring the valve operable.

Technical Specification 3.6.3 requires the PSR primary containment isolation valves to be operable when the Plant is in Operational Modes 1, 2 and 3. Therefore, on June 4, 1986, when the Plant Operational Mode was changed from Hode 4, "Cold Shutdown," to Mode 2, "Startup,R the Plant was in violation of both Technical Specification Sections 4.0.5 and 3.6.3 until the surveillance was completed on July 12, 1986.

A second incident occurred under the same circumstances in the Spring and Summer of 1987. In this case, the surveillance was completed on April 7, 1987, skipped on July 7, 1987 due to being in Operational Mode 4, and completed on August 10, 1987.

Again, both Technical Specification Sections 4.0.5 and 3.6.3 were violated between August 1, 1987 when the allowable surveillance interval was exceeded and August 10, 1987 when the "surveillance was performed.

NRC FORM 3664 o U S GPO.1 985.0 524 538/455 19831

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NRC Form 388A U.S. NUCLEAR REGULATORY COMMISSION 19831 LICENSEE EVENT REPORT {LER) TEXT CONTINUATION AFFROVEO OMB NO. 3150WIOO EXFIRES: 8/31/88, FACILITY NAME 111 DOCKET NUMBER (ll LER NUMBER (Sl

Eel SEQUENTIAL PX ~I 4EVISIQN NVM ER ~

4VMOEA Washington Nuclear Plant - Unit 2 0 5 0 0 0 39 787 033 0 O3>>

TEXT llfmom Epooo /1 /oookNL voo ooU/o'os /VRC Fo/m 388A3/ I Ill The root cause of this event is personnel error in that Plant Chemistry personnel assumed responsibility for the UPSR Valve Operability" surveillance without adequately understanding the surveillance requirements. Responsibility for performing the PSR Valve Operability surveillance was transfer red from the Plant Operations Department to the Chemistry Section of the Plant Health Physics and Chemistry Department in April 1984. The transfer of responsibility was made because Chemistry personnel normally operated the PASS and the "PASS Operability" surveillance could be performed in conjunction with the UPSR Valve Operability" surveillance. This is the only ASME valve in-service test per'formed by the Chemistry Section. The SMS card showed the surveillance was required in Operational Modes 1, 2, or 3. Chemistry personnel did not understand the required Operational flodes on the SMS card signify when "PSR Valve Operability" surveillance requirements are applicable and not when the surveillance can be performed. Further, they did not understand that the surveillance had to be performed either within the allowable surveillance interval or prior to changing from Operational Modes 4 or 5 to Modes 1, 2 or 3. Because of this lack of understanding of the surveillance requirements, Plant procedure revisions which were instituted as a result of a similar event, described in LER 86-32, were not implemented. If they had been, the PSR Valve Operability would have been identified and tracked as requiring completion before the Plant Operational Mode could be changed. Plant procedures were not the cause of this event.

Immediate Corrective Action None, since the Plant was in Technical Specification compliance at the time of the discovery.

Further Evaluation and Corrective Action The other portion of the surveillance procedure "PASS Operability" is mandated by Plant Technical Specification Administrative Section 6.8.4.C and is described in the WNP-2 Final Safety Analysis Report Table 6.2-16.

Requirements for this surveillance have been met.

o Applicable Chemistry Department personnel will be trained on the requirements of Technical Specification Sections 4.0.5 and 3.6.3.

o The surveillance procedure will be revised to include applicable Operational Mode information.

o Plant Health Physics and Chemistry Department personnel will be re-instructed on Plant Procedure 1.5. 1, "Technical Specification Surveillance Testing Program."

Safet Si nificance There is no safety significance associated with this event because subsequent survei llances proved that the valves were operable during the two time periods when the survei llances were not performed within the Technical Specification time limits.

NRC FORM 3884 1983 I 4 U S.GFO.1988-0 82E 538/455

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NRC Form 358A U.S. NUCLEAR REGULA'TORY COMMISSION 1943)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROYED OMS NO. 315OWIOo EXPIRES: 8/31188 FACILITYNAME 11) DOCKET NUMBER 12) LER NUMSER 15) PAGE 13)

YEAR SEQUENzrAL REVISION NUM ER NUMSOR Washington Nuclear:Plant - Unit 2 o 5 o o o 3 9 7 8 7 033 00 04 OF 0 4 TEXT II/rrroro ffoco )f ~ )rr)orE rrff IrFPWo'oM)/4RC horn 3I)SA'f) 1)7)

Further, each PASS sample line has an inboard and outboard primary containment isolation valve so a single failure would not affect the Plant's system isolation capability. In case of an accident, the PASS serves as a backup to on-line Plant instrumentation in determining the extent of core damage. Also, the PASS is designed with redundant sample points so a single failure would not affect the plant's capability to obtain PASS samples. This event posed no threat to the safety of either the public or Plant personnel.

Similar Events LER 86-32 EIIS Information Text Reference EIIS Reference System Component Post-Accident Sampling System (PASS) BN Process Sample Radiation (PSR) Valves BN 63 NRC fORM 3OOA o U S GPO 1988% 824 538)<55 1943)

WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George 11'ashington Way ~ Richland, 11'ashington 99352 Docket Ho. 50-397 January 15, 1988 Document Control Desk U.S. Huclear Regulatory Commission Washington, D.C. 20555

Subject:

NUCLEAR PLANT HO. 2 LICEHSEE EVENT REPORT NO.87-033

Dear Sir:

Transmitted herewith is Licensee Event Report Ho.87-033 for the WHP-2 Plant.

This report is submitted in response to the report requirements of 10CFR50.73 and discusses the items of reportabi lity, corrective action taken, and action taken to preclude recurrence.

Very truly yours,

.M. Powers (H/D 927M)

WNP-2 Plant Manager CMP:sm

Enclosure:

Licensee Event Report No.87-033 cc: Hr. John B. Hartin, NRC - Region V Hr. C.J. Bosted, NRC Site (H/D 901A)

IHPO Records Center - Atlanta, GA Hs. Dottie Sherman, ANI Hr. D.L. Williams, BPA (H/D 399)