05000354/LER-1997-005, :on 940407,TS Surveillance Test Deficiency Was Identified Due to Personnel Errors & Review Process Failures.Performed Separate Evaluation of 10CFR50.59 & Engineering Performance Issues

:on 940407,TS Surveillance Test Deficiency Was Identified Due to Personnel Errors & Review Process Failures.Performed Separate Evaluation of 10CFR50.59 & Engineering Performance Issues

ML20140F548
Person / Time
Site:	Hope Creek
Issue date:	04/25/1997
From:	Karrick J Public Service Enterprise Group
To:
Shared Package
ML20140F500	List: 05000354/LER-1997-005, :on 940407,TS Surveillance Test Deficiency Was Identified Due to Personnel Errors & Review Process Failures.Performed Separate Evaluation of 10CFR50.59 & Engineering Performance Issues
References
LER-97-005, LER-97-5, NUDOCS 9705050051
Download: ML20140F548 (4)

LER-1997-005, on 940407,TS Surveillance Test Deficiency Was Identified Due to Personnel Errors & Review Process Failures.Performed Separate Evaluation of 10CFR50.59 & Engineering Performance Issues

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
3541997005R00 - NRC Website

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NRC FORM 366 U.S. NUCLEAR FitGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104

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EXPIRES 04/30/98

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F ACluTY NAME (1)

DOCKET ftUMBER (2)

PAGE (3)

Hope Creek Generating Station 05000354 1OF4 TITLE (4)

Opzration in a TS Prohibited Condition Due to Failure to Perform Monthly Flowpam Verification Surveillance Checks of Residual Heat Removal System Crosstie Valves.

EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQU AL 66 N MONTH DAY YEAR l

04 07 94 97 005 -- 00 04 25 97

* * "^"'

05000 OPERATING j

THIS REPORT IS SUBB41TTED PURSUANT TO THE REQUIREMENTS OF 10 CFR i: (Check one or more) (11)

MODE (9) 20.2201(b) 20.2203(a)(2)(v)

X 50.73(a)(2)(i)(B) 50.73(a)(2)iviii)

POWER 100 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(xi LEVEL (10) 20.2203(a)(2)(il 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(li) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER ms 20.2203(a)(2)(iii) 50.36(cH1) 50.73(a)(2)(v)

Specify in Abstract below

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• 20.2203(a)(2Hiv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER Onclude Area Codel i

John W. Karrick, Hope Creek LER Coordinator (609) 339-5298 l

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

C*USE SYSTEM COMPONENT MANUFACTURER R

RT E

CAUSE

SYSTEM COMPONENT MANUFACTURER R

R E

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SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR I

YES NO SUBMISSION (if yes, complete EXPECTED SUBMISSION DATE).

X DATE (16)

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

As a result of recent reviews of a safety evaluation for an April 1994 Residual Heat Removal (RHR) system design change, a Technical Specification (TS) surveillance test deficiency was identified.

The design change added a crosstie line with two valves installed between the discharge of the "A" and "C" RHR pumps. Neither of the two valves were required to be maintained locked closed, therefore, the nonthly Emergency Core Cooling Systems (ECCS) flow path verification requirements of TS 4.5.1.a.1.b should have been made applicable, but were not. This resulted in operation in a TS prohibited condition and is being reported pursuant to 10CFR50.73 (a) (2) (i) (B).

This condition was corrected on December 22, 1995, when a normally closed crosstie valve was locked closed.

The cause of this deficiency was the failure to identify the needed surveillance procedure revision during the 10CFR50.59 safety evaluation and the design change review process.

Personnel errors and review process failures lec to this deficiency.

4 Corrective actions include a separate evaluation of 10CFR50.59 and engineering l

performance issues, previous enhancements to the safety evaluation and design change l

processes, a design basis review of ECCS systems associated with 10CFR50.54 (f),

disciplinary action, and a re-assessment of the safety evaluation conclusions. There I

were no safety consequences associated with this deficiency.

970505C051 970425 PDR

/ DOCK 05000354 S

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!U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER 16)

PAGE (3)

NEn NE H::p:a Creek Generating Station 05000354 97 00 2

OF 4 005 TEXT (if more space is required, use additional copies of NRC Form 366A) (17)

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor (BWR/4)

Residual Heat Removal System (RHR) - EIIS Identifier {BO}

IDENTIFICATION OF OCCURRENCE l

Discovery date: March 25, 1997 Problem Report: 970225328 CONDITIONS PRIOR TO OCCURRENCE The plant was in OPERATIONAL CONDITION 1 (POWER OPERATION) at 100% of rated thermal power at the time of discovery.

No other structures, systems, or components were inoperable at the time of discovery that contributed to the event.

.D_ESCRIPTION OF OCCURRENCE E

i During Refuel Outage (RFO) 5 (March-April 1994), a plant design change (DCP 4EC-3411) was installed which added a crosstie line between the "A" and "C" RHR pumps' discharge.

Two valves, one manually operated (1BC-V570) and the other motor operated but electrically disconnected (1BC-V571), were installed in the crosstie line to maintain independence between the Low Pressure Coolant Injection (LPCI) subsystems.

The purpose of this modification was to provide for an alternate means of decay heat removal during reactor shutdown conditions.

Hope Creek Technical Specification (TS) 4.5.1.a.1.b requires the Emergency Core Cooling systems to be demonstrated operable at least once per 31 days, by verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

Reviews of the safety evaluation performed for DCP 4EC-3411 as a result of inquiries from the NRC Senior Resident Inspector, has revealed that from April 7, 1994, to December 22, 1995, neither of these crosstie valves met the criteria for exclusion from the surveillance requirements.

Therefore, these valves should have been included in HC.OP-ST.BC-0001(Q); the surveillance procedure designed to fulfill the TS 4.5.1.a.1.b requirements.

As a result, operation in a TS l

prohibited condition existed and is being reported pursuant to j

10CFR50. 73 (a) (2) (i) (B).

The safety evaluation for the "D" to "B" RHR crosstie modification identified the valve position discrepancy.

On December 22, 1995, 1BC-V571 was locked closed as part of the design change package closure; restoring compliance with TS 4.5.1.a.1.b.

There was no Corrective Action Program document written at that time to report the discrepancy.

As a result, no reportability review was performed and no LER was issued at that time.!

l

OU.S. NUCLEAR REGULATORY COMMISSION

• (*. 96)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER 16)

PAGE (3)

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"^a H:ps Creek Generating Station 05000354 97 oos oo 3

OF 4 l

TEXT (if more space is required, use additional copies of NRC Form 366A) (17)

APPARENT CAUSE OF OCCURRENCE The 10CFR50.59 safety evaluations performed in support of DCP 4EC-3411 for the A/C crosstie did not list TS 3.5.1 or 4.5.1 as having been reviewed for potential changes.

The failure to properly identify all potential TS impacted by the design change contributed to missing this surveillance requirement.

This failure is attributed to personnel errors and inadequate reviews of the 10CFR50.59 safety evaluation.

The design change process includes steps to identify the procedures impacted by the modification.

The procedure for the design change process used at the time DCP 4EC-3411 was issued required the system engineer to identify the procedures affected by the design change.

This review also failed to identify the need to add the crosstie valve (s) to the l

surveillance test and is considered a personnel error.

Improvements to the design change process since that time include an Operations department review for procedural impact.

ASSESSMENT OF SAFETY CONSEQUENCES

During the period of time that the surveillance checks were not performed l

for the A/C crosstie, the LPCI flow paths were not compromised.

Surveillance testing pursuant to TS 4.5.1.b.2, monthly LPCI full flow In-Service Tests, confirmed properly positioned crosstie valves.

In addition, since installation, the A/C crosstie flowpath, an abnormal lineup, has not been used.

Administrative controls were in place, which require two operators to verify the valve positions.

There is a difference between this modification and the design of other l

BWRs in that Hope Creek's crosstie valves' motor operators are not electrically connected.

This reduces the probability of inadvertent mispositioning during power operation.

There were no actual safety consequences associated with this event.

There was no impact on public health and safety.

i NRC FORM 36s %

U.S. NUCLEAR REGULATORY COMMISSION

• (* S6) r LICENSEE EVENT REPORT (LER) l TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER 4 6)

PAGE (3)

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H:p3 Creek Generating Station 05000354 97.. 005 -- 00 4

OF 4

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TEXT (if more space is required, use additional copies of NRC Form 366A) (17)

PREVIOUS OCCURRENCES

A review of previous LERs at Hope Creek revealed that LERs96-019, 96-021, and 95-033-00 through 95-033-14 (TSSIP LERs) involved inadequate surveillance testing due to procedural deficiencies.

The corrective actions from these events could not be expected to have prevented this event.

Corrective actions from a March 25, 1996, TSSIP identified deficiency with verification of primary containment penetrations (LER 95-033-04), did result in verifying all four RHR crosstie valves closed and adding them to the monthly surveillance procedure pursuant to TS 4.6.1.1.b.

This activity met the intent of the ECCS flowpath verification.

CORRECTIVE ACTIONS

1.

On December 22, 1995, 1BC-V571 was locked closed, restoring compliance with TS 4.5.1.a.1.b.

2.

Since the time the 10CFR50.59 evaluations were performed for this modification, the 10CFR50.59 training and qualification requirements have been increased.

Additional enhancements in the 50.59 process have also occurred and will continue to occur as prescribed by the i

Corrective Action Program.

No additional actions are deemed necessary at this time.

i 3.

A separate evaluation of 10CFR50.59 and Engineering performance issues has been initiated.

The evaluation will be completed by May 22, 1997.

4.

A review of other ECCS related design bases will be performed as part of Attachment 2 to the February 11, 1997, Hope Creek response pursuant to 10CFR50. 54 (f).

This review will detect deficiencies between plant configuration and design bases caused by design changes as well as other sources.

5.

PSE&G has evaluated performance deficiencies for personnel involved and implemented disciplinary actions as appropriate.

6.

The safety evaluations for the crosstie modifications will be re-assessed to validate the conclusions.

This action will be completed by May 16, 1997.

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