ML20006C157

LER 89-026-00:on 891105,reactor Scram from Full Power Occurred Due to Problems W/Turbine electro-hydraulic Control Sys.On 891220,contractor Notified Util That Valves Tested Did Not Actuate within Setpoint as required.W/900126 Ltr

ML20006C157
Person / Time
Site:	FitzPatrick
Issue date:	01/26/1990
From:	Fernandez W, Fish H POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JAFP-90-0086, JAFP-90-86, LER-89-026, LER-89-26, NUDOCS 9002060407
Download: ML20006C157 (6)

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_ PlJ. Box 41 . . .

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t _J. L 315 342 3640-is '

h hk - William Femander 11 Resident Manager d

LJ anuary 26, 1990 -

t 0 -JAFP-90-0086  :

n United-States Nuclear Regulatory Commission

~ Document' Control Desk

. Mail Station'P1-137 l Washington, D.C. 20555

REFERENCE:

DOCKET NO. 50-333  :

LICENSEE ~ EVENT REPORT: 89-026-00 Safety Relief Valve Setpoint y Drift 3 a

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Dear Sir:

This. Licensee' Event Report is submitted-in accordance with 10.CFR 50.73(a)(2)(1).

Questions _concerning this-report may be addressed to' .

Mr.-Hamilton. Fish at (315) 349-6013.  ;

Very truly yours, Q

(.g LLIAM FERNANDEZ l

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'Following a. reactor scram on November 5, 1989, two safety relief valves (SRVs) [AD] were removed for testing. The valve test facility provided written notice, received by the Authority on December 20, 1989, that both valves actuated at setpoints which deviated from the design point l by more than the +/-1% allowed by Technical Specifications. One valve lifted early at -1.4%. The other lifted at +4.7%. Evaluation of the possible cause awaits completion of disassembly and examination.

Evaluation of pressure relief capability shows operation would be acceptable with 2 of 11 SRVs inoperable and a setpoint tolerance of

+/-3%. Corrective action included replacing failed SRVs with

.recertified valves, continued participation in the BWR Owners' Group

.to resolve SRV issues, and submission to the NRC of proposed changes to Technical Specifications to take credit for excess installed SRV capacity.

LER-85-009,85-013, 87-004,88-004, and 88-010 are similar events involving SRV setpoint drift.

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g Description A reactor scram from full power occurred at 3:23 P.M. on November 5, 1989 which was attributed ~to' problems with the turbine electro-hydraulic control system [JJ)-(LER-89-020). Peak reactor pressure-reached-1082 psig during the turbine control' valve closure transient which was below the design setpoint for any of the safety relief valves [AD).

Safety relief valve (SRV) "F" had a design.setpoint of 1140 psig (58 psi ~above the transient pressure). The recorded increase in SRV

-exhaust pipe temperature for this valve indicates that it passed a l quantity of steam above that normally_ associated with pilot valve j leakage. However,.the absence of alarms from the acoustic monitor ,

combined.with the instrumented exhaust pipe temperature. rate of rise .!

indicated that the valve did not actually lift. This was confirmed by the exhaust pipe temperature which was below that associated with the ,

lifting of an SRV. 'l

. Safety relief valve "E" was mounted on the same line as SRV F". The.

SRV "E" design setpoint of 1105 was 27 psi above the peak pressure the transient. SRV "E" performed as expected and did recorded not lift. during"E" SRV would have been expected to lift before SRV "F" if.  !

the pressure had actually risen to the SRV "F" setpoint of 1140 psig. l i

The activating topworks mechanisms were removed from both SRV "E" and  !

"F" during the outage following the scram. The topworks were sent to a'  !

-contractor facility for testing. U Written notification from the contractor received on December 20, 1989' -[

informed the plant staff that the two SRVs which were tested did not 1 actuate within +/-1% of the nameplate setpoint as required by Technical l Specification 2.2.1.B. The initial set pressure observed-by the -

contractor at his facility were:

Difference

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Observed From i Pilot Nameplate Initial Specification Plant-Valve- Assembly Set Pressure Set Pressure In Number Serial No. (psig) _

(psig) PSI %_

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02RV-71E 1050 1105 1089 -16 -1.4%

02RV-71F 1087 1140 1194 +54 +4.7%

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0l0 $l3 oF] e 15-mea amawim During the pretest heat-up of the SRV "F" pilot assembly, the valve ,

began leaking with 100 asig being applied to the valve inlet. The

-leakage was so severe taat the test facility exhaust system was being overloaded. At the request of the test facility vendor, the Authority authorized immediate set pressure actuation without achievement of thermal equilibrium. Four set 3ressure actuations were ran. The severe leakage continued throughout the test. The following data was obtained from the tests of valve 02-RV-71F, Serial No. 1087, which had ,

a nameplate set pressure of 1140:

Observed Set Reseat i Run- . Pressure- Pressure Percent Delay No. (psig) (psig) Reseat (msec) Remarks

1 1194 822 68.8 -450 Leaking Severely.

2 1108 821 73.5 310 Leaking Severely

3. 1115 824 73.2 280 Leaking Severely 4 1108 -823 74.0 280 Leaking Severely

'Cause A detailed examination of the valve pilot assembly is required to determine possible causes. Vendor examination will be scheduled in the future. This LER will be updated when the results are known.

The= test results of SRV "F" (in which severe leakage was demonstrated) are consistent with the indicated leakage of the valve following the scram. The high lift pressure of the first run followed by a series of 1 lifts at a consistently lower setpoint is characteristic of a disc sticking problem.

Analysis-The observed setpoint of two SRVs deviated by more than 1% from the values specified in Technical Specification 2.2.1.B. Therefore, this

. event is reported under the provision of 10 CFR 50.73(a)(2)(1)(C) as a deviation from the plant's Technical Specifications.

The remote actuation-(operator demand) and automatic depressurization system (ADS) functions would not have been effected by this event. An -

evaluation to determine the effects of SRV setpoint drift was initiated as a result of earlier similar events (LER-87-004 and LER-88-004) and has been completed.

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relaxation of the +/-1% nominal valve nameplate secpoint tolerance to 9N +/'-3%, and o .

3 - operation with any 2 SRVs inoperaole, and l m ,

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setting all 11 SRVs atla single nominal nameplate setpoint.

• The results of a bounding eval ~uation show that continuous operation of -

.the plant would be. acceptable with a 50 psi margin to the American  ;

Society of MechanicaliEnSineers (ASME) Coda upset reactor _ vessel 4 pressure limit of-1375 psig during the limiting overpressure event with any 9Jof the.11~SRVs operable and with a common valve actuation- .

pressure of 1195 psig. .This analysis bounds the conditions. identified' '

by the SRV testing.

' Based on'the bounding evaluation. it is conc 1.uded that the setpoint.

drift of the valves did not represent any hazard. Plant response to ,

any;cf the accident conditions described in the Final Safety Analysis-

-Rep 9rt1(FSAR) would have been acceptable.

' C o r r e'e t i v e A c t i o n Immediate. Corrective Action: The valves were replaced with refurbiched- '1 and recertified valves. The failed SRVs will be refurbished and recertified for future installation. Although there was no indication of a failure of SRV "E", it was remlaced asi a conser vative arudent faction because it'was mounted on the same line as SRV 'F" which was

-leaking.

Long-Term Corrective Action:

1) A proposed change to Technical Specifications to allow continuous-operation of'the plant with SRV operations consistent with the new analysis was submitted to the NRC on December 20, 1989.

2)- The plant has previously modified SRVs by installing pilot valve discs made of a different material. This action is part of the.

Boiling Water heactor Owners' Group (BWROG) plan for correction of the Target Rock 2-Stabe.SRV drift problem. The plant has previously installed and operated with approximately hc1f of the valves containing the new pilot valve' discs made of PH13-0Mo.

Testing of these valves is scheduled during the spring 1990 refueling outage. Further modifications will be dependent on the results of this testing and the evaluation being performed by the BWROG SRV Setpoint Drift Fix Committec.

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fI 3) All SRVs (rather than half of the valves as specified by Technical Specifications) will continue to be subjected to test, inspection, 1

refurbishment, and recertification once each operating cycle until the test data, or the BWROG SRV Setpoint Drift Fix Committee recommendations indicate otherwise, or the Technical Specifications are amended.

Additional Information Failed Component Identification: __

- SRV Manufacturer: Target Rock Corp.

- Valve Model Number: 7567F

- Manufacturer NPRD Code: T020 -

LER-85-009,85-013, 87-004,88-004, and 88-010 are similar events which reported SRV setpoint drift. LER-89-020 describes the scram which initiated testing of the SRVs reported in this LER.

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