ML18095A421

LER 90-022-00:on 900717,observed That Relief Valve 1SJ167 Was Leaking Past Seat as Result of Investigation to Determine Cause of Decreasing Level in Accumulators.Caused by Equipment Failure.Valve Repaired on 900722.W/900815 Ltr

ML18095A421
Person / Time
Site:	Salem
Issue date:	08/15/1990
From:	MILLER L K, POLLACK M J Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-022-01, LER-90-22-1, NUDOCS 9008210155
Download: ML18095A421 (6)

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• Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear* Sir:

SALEM GENERATING STATION' LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 90-022-00 August 15, 1990 This IJicensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations lOCFR 50.73(a) (2) (i) (B). This report is required within thirty (30) days of discovery.

MJP:pc Distribution The Energy F>eop!e O n* r.r* '--' .l. v .,) Sincerely

yours, General Manager -Salem Operations

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FACILITY NAME 111 Salem Generating Station -Unit 1 I DOCKET NUMIEll 121 I . """! 13l o 15 Io IO Io 12 17 I 2 1 loF 01 5 TITLE 141 Tech. -Spec. Action Statement::3.:0.".3 Entry In Support of Maint. EVENT DATE 181 LER NUMBER 191 REPORT DATE 171 OTHER FACILITIES INVOLVEO Ill MONTH QAY YEAR YEAR ::::,:::::

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NUMBER m DAY YEAR FA.Cl LITY NAMES DOCKET NUMBERISI ol 1 2 I 2 9 o 91 o -d 2 I 2 -ol o o Is l Is 9 I o Ol'ERATINO THIS REPORT 11 IU!IMITTED PURSUANT TO THE REOUIREMENTI OF 10 CFll §: (0..Ck o,,. or mon of 111* fo//owln1111111

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,42 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE 0E;c1m1ED IN THllii REPORT 1131 SYSTEM COMPONENT MANUFAC. TUA ER CAUSE SYSTEM COMPONENT MANUFAC. TUR ER I I I I I I I I I I I I I I I I I I I I I I I I I I I I EXPECTED MONTH DAY SUllMISSION DATE 1151 I YES (If yn, compl*l'I EXPECTED DATE) I I ABSTRACT (Limit ro 1400 -. I.* .. *pproxim11'1/y flhHn"lfn11lo-11>>eo typowr/fl'ln

/inn! 11111 On 7/17/90, Operations Department personnel observed that the 1SJ167 relief.valve (Cold Leg Safety Injection Line Relief Valve) was leaking past its seat (to the Pressurizer Relief Tank). This was identified as a result of investigations to determine the cause of decreasing level in Nos. 12 and 14 Accumulators (i.e., less than 1 gpm). Apparently, the 12 and 14SJ144 check valves were leaking by. These check valves are in the Safety Injection System lines which feed each RCS cold leg. The Accumulators use the same line *for injection during a design basis accident thus providing the force to cause leakage past the 1SJ167 valve seat. On 7/22/90 it was decided to repair the 1SJ167 valve at power and at 0909 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.458745e-4 months <br /> (after reducing power to 60%), Tech. Spec. Action Statement 3.0.3 was entered. Repair of the 1SJ167 valve required isolation of the Cold Leg Safety Injection line. Tech. Spec. Action Statement 3.0.3 was entered since the Tech. Spec. 3.5.2 Action Statements (applicable in Modes l, 2, and 3) do not address the actions to be taken if cold leg injection capability, *via the Safety Injection Pumps, is not operable.

The root cause of this* event has been attributed to an equipment failure. The 1SJ167 valve was leaking past its seat which had required its removal. Removal of the valve identified that the valve's seat was cut. On 7/22/90 at 0940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br />, upon completion of the 1SJ167 valve repair and reopening the intermediate SI cold leg injection flowpath, Tech. Spec. Action Statement 3.0.3 was exited. The 12SJ144 and 14SJ144 check valves* will be inspected and repaired (as requi!ed) during the refueling outage. NRC Form 3111 19.all I e e LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 90-022-00 2 of 5 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse

-Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as {xx} IDENTIFICATION OF OCCURRENCE:

Technical-Specification Action Statement

3.0.3 entered

in support of required maintenance Event Date: 7/22/90 Report Date: 8/15/90 This report was initiated by Incident Report Nos.90-490 and 90-506. CONDITIONS PRIOR TO OCCURRENCE:

Mode 1 Reactor Power 60% -Unit Load 655 MWe DESCRIPTION OF OCCURRENCE:

On July 17, 1990 at 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />, during normal power operation (i.e., 100% power), Operations Department personnel observed that the 1SJ167 relief valve (Cold Leg Safety Injection Line Relief Valve) was leaking past its seat. This was identified as a result of investigations to determine the cause of decreasing level in Nos. 12 and 14 Accumulators (i.e., less than 1 gpm). The Accumulator level was made up ensuring compliance with the requirements of the Technical Specifications and the plant design basis. Apparently, the 12 and 14SJ144 check valves were leaking by. These check valves are in the Safety Injection System lines which feed each RCS_ cold leg. The Accumulators use the same line for injection during a design basis accident thus providing the force to cause leakage past the 1SJ167 valve seat. -On July 22, 1990 it was decided to repair the 1SJ167 valve during power operations.

Prior to initiation of the valve repair, reactor power was reduced to 60% as a conservative measure. At 0909 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.458745e-4 months <br /> (that day), Technical Specification Action Statement 3.0.3 was entered and valve repair was initiated.

Repair of the 1SJ167 valve required isolation of the Cold Leg Safety Injection line fBQl. Technical Specification Action Statement 3.0.3 was entered since the Action Statements for Technical Specification 3.5.2 (applicable in Modes 1, 2, and 3) do not address the actions to be taken if cold leg injection capability, via the Safety Injection Pumps, is not operable.

Technical Specification

3.5.2 states

e LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station Unit 1 DESCRIPTION OF OCCURRENCE:

DOCKET NUMBER 5000272 (cont'd} LER NUMBER 90-022-00 PAGE 3 of 5 "Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of the following .injection systems:" Technical Specification*3.5.2.b states: "One OPERABLE safety injection pump and associated flow path capable or taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and; 1. Discharging into each RCS cold leg, and; initiation, 2. Discharging into its two associated RCS hot legs." Technical Specification Action Statement

3.0.3 states

"When a Limiting Condition for-is not .met except. as provided in the associated ACTION requirements, within one hour action shall be initiated to* place the unit in a MODE in which the specification does not apply by placing it, as in: 1. At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Where corrective measures are completed that perm.it operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition of Operation.

Exceptions to these requirements are stated in the individual specifications." APPARENT CAUSE OF OCCURRENCE:

The root cause of this event has been attributed to an equipment failure. The 1SJ167 valve was leaking past its seat which had required its removal. Removal of the valve identified that the valve's seat was cut. ANALYSIS OF OCCURRENCE:

The ECCS incorporates the use of high head safety injection (Centrifugal Charging Pumps),

head safety injection pumps (SI Pumps} and low head safety injection pumps (Residual Heat Removal Pumps). There are two (2) of each type of pump where either pump will provide 100% of the flow requirements during ECCS operation.

The discharge from the SI Pumps merge to a single pipe which feeds all four. (4} Reactor Coolant System (RCS) {ABI cold legs. The 1SJ167 valve provides pressure relief to protect the injection flow path e e LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE ANALYSIS OF OCCURR_E_N_C_E_:

from overpressurization.

Discharge from the 1SJ167 valve is directed to the Pressurizer Relief Tank (PRT) when system pressure exceeds 1750 psi. As stated previously, the 1SJ167 valve was leaking past its seat. The observed flow rate had been calculated to range between 0.5 gpm at 150 psi to 3 gpm at 1500 psi. The highest expected system pressure is 1500 psi when the SI pumps are on recirculation with the Refueling Water Storage Tank (RWST) {which is during an SI) . The worst case SI event involves a small break Loss-Of-Coolant Accident (LOCA) when RCS pressure does not rapidly decrease thereby maintaining maximum leakage past the 1SJ167 valve. As RCS pressure decreases, and intermediate SI begins, the three (3) gpm valve leakage will be compensated by additional pump flow. The pump performance during recirculation is 30 gpm to the RWST at a developed head of 1510 psid. The 1SJ167 valve leakage flow added to the RWST flow will result in a reduced pressure of 8.5 psid max. As discussed in the Updated Final Safety Analysis (UFSAR) , Section 15 "Accident Analysis, any break of sufficient size requiring SI will reduce the RCS pressure to below 1500 psi. The 8.5 psi reduction in discharge injection is with respect to the overall SI requirements.

The total reduction in safety injection flow is estimated to be less than 1.5 gpm at run out (full flow test) conditions.

Based on the last full flow test, performed during the last outage, the reduction in total safety injection flow due to the leakage past the 1SJ167 valve will not violate the minimum Technical Specification required flow. The valve manufacturer (Crosby) was contacted.

Crosby stated that the valve leakage will not* affect the valve relief setpoint (i.e .. , it will not cause premature lifting}.

• As addressed above, the 1SJ167 valve leakage did not adversely affect the design operability of the intermediate head SI flowpath.

Therefore, the leakage did not affect the health or safety of the public. However, the leakage was a concern in terms that it may increase with time. Therefore, it was decided to replace the valve. This required entry into Technicai Specification Action Statement 3.0.3. The Action Statement was exited after thirty-one (311 minutes upon restoration of the SI Pump flowpath.

As stipulated by regulatory guidance, entry into Technical Specification Action Statement 3.0.3 is reportable in accordance with Code of Federal Regulations lOCFR 50.73(a) (2) (i) (B). Prior to initiating valve repairs, detailed discussions were conducted, with the Nuclear Regulatory Commission (resident and region) , which addressed the intent to perform valve repair at power. These discussions included a summary of proposed actions.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION salem Generating Station Unit 1 FAILURE DATA: 1SJ167 Valve: Crosby Valve Co. Valve Model JMAK-BS DOCKET NUMBER 5000272 3/4"x1" Nozzle Type Relief Valve CORRECTIVE ACTION: LER NUMBER 90-022-00.

PAGE 5 of 5 On July 22, 1990 at 0940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br />, upon completion of the repair of the 1SJ167 valve and reopening the intermediate SI cold leg injection flowpath, Technical Specification Action Statement 3.0.3 was exited. On June 4, 1990, the 12SJ144 and 14SJ144 check valves had successfully completed surveillance procedure SP{0)4.4.6.3, "Emergency Core Cooling System -ECCS Subsystems".

This procedure verifies the integrity of RCS boundary valves. Although some leakage was noted, by the surveillance, it was within the allowable limits as specified by Technical Specifications.

These check valves will be inspected and repaired (as required) during the next refueling outage. MJP:pc SORC Mtg.90-107 General Manager Salem Operations