ML20043F829

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LER 90-007-00:on 900507,discovered That Drywell to Suppression Chamber Vacuum Breaker Surveillance Not Performed Prior to Startup in 1988.Caused by Misunderstanding of requirements.W/900606 Ltr
ML20043F829
Person / Time
Site: Pilgrim
Issue date: 06/06/1990
From: Basilesco G, Bird R
BOSTON EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
90-073, 90-73, LER-90-007, LER-90-7, NUDOCS 9006180197
Download: ML20043F829 (6)
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10 CFR 50.73-g  :,

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Pilgrim Nuclear Pour Station Rocky Hill Road ..

Plymouth, Massachusetts 02360 ..

June - 6, 1990 Ralph G. Bird -

BECo'Ltr.90-073

, Senior Vice President- Nuclear U.Si~ Nuclear Regulatory Commission

= Attn: Document Control Desk-

Hashington, D.C. 20555 Docket No. 50-293-License No. DPR-35

Dear Sir:

The: enclosed' Licensee Event Report (LER) 90-007-00, "Drywell to Suppression'

' Chamber Vacuum Breaker Surveillance Not Performed Prior to Startup-in 1988",

.is' submitted in accordance with 10 CFR Part 50.73.

Please do not hesitate to contact me if there are any questions regarding this report.

GJB/bal

Enclosure:

LER 90-007-00 cc:. Mr. Thomas T. Martin Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Rd.

. King of Prussia, PA 19406 R Sr. NRC Resident Inspector - Pilgrim Station Standard BECo LER Distribution ff 0

9006180197 900606 PDR ADOCK 05000293  !

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LICENSEE CONTACT POR THl4 LER H2) ]

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AeswCTw-, ux,-.ua.- ~ ww n..w,,. a. rr-n.. sm.o n s. ll During a recent performance review of surveillance 8.A.2 "Drywell to 4 Suppression Chamber Vacuum Breaker Leakage Rate Test (1.25 psig)", it was i

-discovered that when restarting from Refueling Outage (RF0) No. 7 in December 1 1988, surveillance 8.A.2 was not performed at the appropriate point in the startup process. The surveillance had expired during the outage for RF0 7 and oI should have Wn performed prior to reactor criticality in accordance with Technical Specificatios. The surveillance is required to be performed once per refueling outage and quarterly. The surveillance was performed during the outage in December 1987. The quarterly surveillance was satisfactorily '

performed shortly after Drywell inerting in March 1989.

The'cause has been determined to be a misunderstanding of Technical *

  • Specification surveillance requirements and an incorrectly scheduled surveillance in the Master Surveillance Tracking Program (MSTP).

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The Technical Specification surveillar a requirements have been clarified and  !

the MSTP review process has since beer vised to prevent the recurrence of mis-scheduled surveillances. In addition, a Technical Specification change is 4 being considered to remove the quarterly surveillance requirement, hence being consistent with Standard Technical Specifications.

The condition was determined to be reportable on May 7, 1990 during power operation. The reactor mode selector switch was in the RUN position and the reactor power level was approximately 100 percent. The reactor pressure was

'1035 psig with the reactor water temperature at 532 degrees Fahrenheit. This condition posed no threat to the health and safety of the public and this report is submitted in accordance with 10CFR50.73(a)(2)(1)(B).

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EVENT DESCRIPTION ,

I On April 27, 1990 during the recent mid cycle outage, surveillance 8.A.2, l "Drywell to Suppression Chamber Vacuum Breaker Leakage Rate Test (1.25 psig)" i reached its drop dead date. The surveillance is normally performed with i nitrogen with the Drywell inerted. On April 27, 1990 the surveillance was 1 performed with air because the Drywell had not yet been inerted and the j quarterly drop dead date would have been exceeded at 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br />. While >

investigating the requirements in the Master Surveillance Tracking Program 5 (HSTP) for the 4/27/90 test, it was discovered that surveillance 8.A.2 was not i performed prior to restarting from Refueling Outage (RFO) No. 7. The i surveillance was not performed prior to reactor criticality on December 30, ,

1988 and had expired during the outage.  :

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Technical Specification 3.7.A.".a requires that Primary Containment integrity be maintained at all times when the reactor is critical or when reactor water ,

temperature is above 212 degrees Fahrenheit and fuel is in the vessel.  !

Technical Specification 3.7.A.4.a specifies that when primary containment is f required, the Drywell-pressure suppression chamber vacuum breakers be  ;

operable. Operability is verified by the quarterly surveillance test (8.A.2)  !

as required by Technical Specification 4.7.A.4.a(2). The test was not i performed prior to reactor criticality or reactor water temoerature reaching  !

212 degrees Fahrenheit and therefore compliance with the Technical  ;

Specification was not achieved. j l

The surveillance was satisfactorily completed on March 13, 1989 at 1355 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.155775e-4 months <br /> with the mode switch in the RUN position, the Drywell inerted and reactor power level at approximately 25 percent. Failure and Halfunction Report I

90-135 was written to document this condition that was determined to be ,

L reportable on May 7, 1990. l The condition was discovered during power operation with the Reactor Mode  !

Selector Switch in the RUN position. The Reactor Vessel (RV) water temperature was approximately 532 degrees Fahrenheit with the RV pressure at 1035 psig. Reactor power level was approxima';ely 100 percent. ,

L 1 l CAUSE l

The cause of performing the surveillance later than required was that the [

surveillance was incorrectly scheduled on the HSTP. The cause of incorrectly ,

( scheduling the surveillance was a misunderstanding of Technical Specification  ;

surveillance requirement and a procedural weakness under the former HSTP '

l l process, Procedure 1.8, Xev. 4, "Haster Surveillance Tracking Program". The HSTP indicated that the surveillance be performed "on line" after the Drywell  ;

was inerted. The surveillance measures the decay rate of differential pressure between the Drywell and suppression chamber and therefore it is normally performed after a positive differential pressure has been established with nitrogen, when the Drywell is inerted. While the test can be performed using compressed air (e.g. 4/27/90 surveillance) it is normally performed "on-line" with nitrogen. It was not considered that during an extended ,

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shutdown, if the surveillance expired, it would need to be performed prior to i starting up (i.e. reactor criticality or reactor water temperature reaching ij 212 degrees Fahrenheit). The misconception existed that the test would be i' performed "on-line" after Drywell inerting was complete.

After RF0 7 ended, Pilgrim Station began a power ascension program that I consisted of a gradual increase in power operation to 100 percent. The ,

i program (TP 87-114) lasted several months with several shutdowns and startups  !

occurring. On March 12, 1989, at 1153 hours0.0133 days <br />0.32 hours <br />0.00191 weeks <br />4.387165e-4 months <br />, the mode switch was placed in the RUN position and Drywell inerting began. Inerting was completed at 2130 i hours (first Drywell inerting since cycle 6 operation). less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after placing the mode switch in RUN (in accordance rith Technical Specification 3.7.A.5.b). Shortly after inerting on March 13, 1989 at 1355 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.155775e-4 months <br />, surveillance 8.A.2 was satisfactorily performed. This was in accordance with the HSTP requirements that existed at the time.

Surveillance 8.A.2 had been performed just prior to a plant shutdown (start of 1 RF0 7) in April of 1986. Its next due date was scheduled for June 1986, but since the plant had remained in a shutdown condition, vacuum breaker operability was not required. The surveillance was rescheduled to be performed "on-line". It should have been rescheduled to " prior to startup",

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but since the o rveillance is normally performed on-line, it was not rescheduled appropriately. The HSTP review process that existed in 1986 did not identify this scheduling inadequacy. The process then did not require that an MSTP change receive a level of review currently required for an MSTP change. The 1986 change was reviewed by one operations engineer and the MSTP coordinator. The current process requires several reviews for each MSTP change (i.e. rescheduling),

i CORRECTIVE ACTION The MSTP process that existed during RF0 7 when the surveillance (8.A.2) was ,

rescheduled to "on-line", was reviewed. It was determine ( that the review l

L process that existed did not adequately review and schedule this change. The I

MSTP process has been revised (Procedure 1.8, Rev. 9) and now includes detailed reviews of each surveillance change. It h surveillance has not been routinely scheduled due to the plant being in a mo.de in which the surveillance is not required, the process requires a detailed review (including a review of the applicable Technical Specification requirements) before the surveillance is rescheduled. In addition, other controls exist in Procedure 2.1.1 "Startup from Cold Shutdown" to assure that as certain plant plateaus are reached (e.g.

criticality, startup, etc.), required surveillances are complete before exceeding that plateau. Therefore, the current controls are in place 'to assure that a surveillance would not be missed, as evidenced during the most recent performance review of surveillance 8.A.2 on 4/27/90. /

The misconception that existed to perform the test "on-line" after Drywell inerting is complete has been eliminated as a result of the 4/27/90 performance review. In the future, the surveillance will be performed using a

compressed air if the plant is preparing for startup and the surveillance is not current. The review processes in place will assure that the surveillance is appropriately rescheduled in the MSTP.

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1 A Technical Specification change is being considered to remove the quarterly l decay rate test surveillance requirement. Vacuum breaker operability is verified using other means including periodically exercising the valves and  ;

verifying position indicators and alarms, and performing several surveillances l during refueling outages including a differential pressure decay rate test l similar to the quarterly test. The quarterly surveillance requirement is not required in the standard Technical Specifications. The change will be considered as part of the ongoing Technical Specifications review (Long Term Plan Item No. 468).

SAFETY CONSEOUENCES This condition posed no threat to the public health and safety.

Procedure 8.A.2 was completed with satisfactory results on March 13, 1989.

The surveillance was performed with no corrective action required. The once per refueling surveillance (Procedure 8.A.2) of the vacuum breakers was performed with satisfactory results during the outage in December 1987. l l

Therefore, it is reasonable to conclude that the vacuum breakers were operable '

when required.

This report is submitted in accordance with 10CFR50.73(a)(2)(1)(B) because the quarterly surveillance of the Drywell to Suppression' Chamber vacuum breakers was not performed within three months prior to initial startup (December 30, 1988) from RF0 7.

SIMILARITY TO PREVIOUS EVENTS A review of Licensee Event Reports (LERs) issued since 1984 was conducted.

The review focused on reports submitted in accordance with 10 CFR 50.73(a)(2)(i) which involved a misunderstanding of Technical Specification requirements. The review identified similar instances reported in LERs 50-293/87-004-00, 90-001-00 and 90-006-00.

  • 87-004-00, " Improper Work Prioritization Resulting in Condition Prohibited by Technical Specifications". On February 18, 1987, at approximately 1610 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.12605e-4 months <br />, it was determined that the dry chemical fire suppression system associated with a piping trench below the "A" emergency diesel generator had been inoperable since December 21, 1986. Contrary to the plant Technical Specifications, the appropriate limiting condition for operation was not complied with

, from approximately 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on February 6, to 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br /> on u February 13, 1987, during which time the suppression system was required to be operable. The cause was failure of the utility personnel involved to recognize the Technical Specification requirements associated with the system.

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" Instrument Line Flow Check Valve Test" was inappropriately signed as completed on November 4, 1989 baseo on a previously written {

memorandum that indicated the two check valves were not required to  ;'

be functionally tested until the next refueling outage. As a result '

the T.S. surveillance interval was exceeded.

e LER 90-006-00 " Position of Primary Containment Isolation Valve not l Recorded Daily as Required by Technical Specifications". On March 30, 1990, it was determined that the position of one of the two inoperable, in-series, Primary Containment System isolation valves in '

the Head Spray piping had not been recorded daily as required by '

L T.S. 4.7.A.2.b.2. Valves M0-1001-60 and M0-1001-63 were placed in '

I the closed position with their respective circuit breakers open

'(de-energized)-in conjunction with work associated with Plant Design a

Change 86-20 which cut and capped the pipe. The cause was a '

misunderstanding relative to the operability status of the valves and applicable Technical Specification surveillance requirements.

ENERGY INDUSTRY IDENTIFICATTON SYSTEM (EIIS) CODES -

The EIIS codes for this report are as follows:

COMPONENTS CODES Breaker, Vacuum VACB 7 SYSTEMS l Containment Vacuum Relief System BF l

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