05000272/LER-1983-008, Supplemental LER 83-008/03X-1:on 830311,pressurizer Code Safety Valves 1PR3,1PR4 & 1PR5 Exceeded Lift Set Pressure Range & Had Heavy Seat Leakage.Partly Caused by Difference in Test Vs Actual Operating Conditions

Supplemental LER 83-008/03X-1:on 830311,pressurizer Code Safety Valves 1PR3,1PR4 & 1PR5 Exceeded Lift Set Pressure Range & Had Heavy Seat Leakage.Partly Caused by Difference in Test Vs Actual Operating Conditions

ML20077G918
Person / Time
Site:	Salem
Issue date:	07/13/1983
From:	Frahm R, Zupko J Public Service Enterprise Group
To:	Murley T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
LER-83-008-03X, LER-83-8-3X, NUDOCS 8308090353
Download: ML20077G918 (5)

LER-2083-008, Supplemental LER 83-008/03X-1:on 830311,pressurizer Code Safety Valves 1PR3,1PR4 & 1PR5 Exceeded Lift Set Pressure Range & Had Heavy Seat Leakage.Partly Caused by Difference in Test Vs Actual Operating Conditions

Event date:
Report date:
2722083008R00 - NRC Website

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74 75 REPORT DATE EVENT DESCRIPTION AND PROBABLE CONSEQUENCES O'o Iol2l l Pressurizer Code Safety Valves lPR3,1PR4, and 1 PRS were tested for lift set pressure g g,3, g and seat leakage by Wyle Laboratories during the period of November 2-13, 1982. All g

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ggg,yg 1 PRS - 2546 psig (36 psig over). Engineering evaluation revealed that the setpoint y lOlHl l l 7 8 9 80 SYSTE\1 CAUSE CAUSE COYP. VALVE CODE CODE SUBCODE COMPONENT CODE SUBCODE SUSCODE Q

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]g 43 l C l 7 l1 l 0 [g 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h l i l o l l All three valves were repaired and retested satisfactorily, then reinstalled. During l li ;i l l subsequent testing of Salem Unit 2 valves it was revealed that the setpoint deviation l

, 7 land leakage were apparently due to the testing methods utilized. The valves will be l g l tested again during the next refueling; support will be provided to improve testing l g l methods. l 7 8 9 80 ST S *. POWER OTHER STATUS ISCO RY DISCOVERY DESCRIPTION

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80 5 NAME OF PREPARER PHONE:

O PSEG Public Service Electric and Gas Cornpany P.O. Box E Hancocks Bridge, New Jersey 08038 Salem Generating Station July 27, 1983 Dr. Thomas E. Murley Regional Administrator USNRC Region 1 631 Park Avenue King of Pruscia, Pennsylvania 19406

Dear Dr. Murley:

LICENSE NO. DPR-70 DOCKET NO. 50-272' REPORTABLE OCCURRENCE 83-008/03X-1 SUPPLEMENTAL REPORT Pursuant to the requirements of Salem Generating Station Unit No. 1 Technical Specifications, Section 6.9.1.9.b, we are submitting supp121aental Licensee Event Report for Reportable Occurrence 85-008/03X-1.

. Sincerely yours, 1

' WY J. M. Zupko, Jr.

General Manager -

Salem Operations RF:kl CC: Distribution P

The Energy People lll

% 2189 GOMt 11-81

Report Number: 83-008/03X-1 Report Date: 07-13-83 Occurrence Date: 03-11-83 Facility: Salem Generating Station, Unit 1 Public Service Electric & Gas Company Hancocks Bridge, New Jersey 08038 IDENTIFICATION OF OCCURRENCE:

Pressurizer Code Safety Valves - Inoperable This report was initiated by Incident Report 83-052.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 5 - Rx Power 0% - Unit Load 0 MWe.

DESCRIPTION OF OCCURRENCE:

Pressurizer Code Safety Valves 1PR3, lP R4, and lPR5 were tested for lift set pressure and seat leakage by Wyle Laboratories during the period of November-2-13, 1982. All valves lifted in excess of the 2485 psig + 1% pressure range specified in Technical Specification 3.4.2.2. All valves displayed heavy seat leakage. The actual lift pressures were: 1PR3 - 2564 psig (54 psig over), IPR 4 - 2532 psig (22 psig over), 1PR5 -

2546 psig (36 psig over). These valves had been tested, repaired, and retested during the first refueling outage in 1979. They were within specification when reinstalled. No Reactor Coolant System (RCS) pressure transients occurred during previous power operation which resulted in actuation of the safety valves; the integrity of the RCS and redundant fission product barriers was maintained.

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE:

Subsequent investigation revealed that leakage observed through the valves could be attributed partly to the difference in test versus actual operating conditions. During testing, the valve was directly placed above the pressurizing chamber and was sub-jected to temperatures close to 500 F. In the plant, a long water loop seal precedes eacg valve, and each valve body inlet temperature is less than 200 F. The valve internals are of a type that provides leak tightness during plant operation and are not designed to be tight against high temperature steam.

This was substantiated by nitrogen tests where a Salem Unit 2 valve tested leak tight (while during tests at the same setpoint with steam the valve leaked) . A steam test foglowed the nitrogen test, with the body temperature limited to 200 F. The valve was successfully lift tested with no subsequent seat leakage. It was therefore tentatively assumed that the setpoint variations were related to the elevated test temperatures utilized.

I LEI 83-008/03X-1 -k:

fd*1 - 3 ANALYSIS-OF OCCURRENCE:

s The pressurizer code safety valves operate. to prevent the RCS from being pressurized above its safety limit of 2735 psig.

Each' safety valve is designed to relieve 420,000 pounds per-hour of: saturated. steam at.the valve setpoint.- The relief

. capacity of(a single safety valve'is" adequate to relieve any overpressure condition which could occur during-shutdown. In

'the event ~ that no safety valves are. operable, an operating-RHR loop, connected to the RCS, provides overpressure relief scapability and will. prevent RCS overpressurization.. In.

. addition, the ~ Overpressure. Protection ' System .provides a diverse

, means of protection against RCS overpressurization'atLlow tem-peratures.

During operation, all' pressurizer code safety valves must be operable to prevent the RCS from being pressurized above its safety limit of:2735 psig. The combined-relief capacity of' all of .these valves is . greater than the maximum surge rate resulting from a complete loss of load. assuming no reactor trip '

until the first Reactor Protective . System trip setpoint is reached (i.e., no credit is taken for a direct reactor trip.on the loss of. load) and also assuming no operation of the power operated valves or steam dump valves.

Evaluation'of the potential impact of the possible safety; valve-setpoint variation on plant performance during the, analyzed transient was performed. The evaluation S-C-R200-NSE-193 states:

Valve opening pressures in excess of assumed were generically evaluated'by-. Westinghouse. 'The limiting overpressurization' transient evaluated for a four-loop unit is the loss of external load and/or . turbine -

'E trip without immediate re' actor trip.- For this event, safety valve functioning is not required if the ' reactor l ._ trips on.high pressurizer pressure. If the reactor does L

not, trip until the second protection. grade trip (over-temperature delta-T), a valve opening delay time of approximately two seconds' would still provide acceptable overpressure protection for the reactor coolant system and all components.would be exposed to a pressure within 110 percent of the system design pressure.

Although set pressure deviation varied from 22 psig to

~

54 psig (0~.8 to 2.2%) .above the allowable band, effective relief would~have started at-the 22 psig shift above allowable (2,485 + 1%, i.e., 2,510 psig). A two second delay.in safety ~ valve openings as analyzed by Westing-house would be-the limiting case as opposed to 22 psig to 54 psig, setpoint deviation at Salem-Unit 1 based'on-the rate of! pressure rise from Salem FSAR for the accident condition analyzed.

I I

LER 83-008/03X-1 '

ANALYSIS OF OCCURRENCE: (continued)

Based on the generic evaluation as above, it is judged that upward deviation of Salem 1 pressurizer safety valve setpoint during operation did not present a safety concern.

Apart from the qualitative difference between the test and operating condition, a leaky valve does not pose any safety concern, as long as the Unit Technical Specification governing the allowable RCS leakage is not violated.

As noted, no pressure transient was involved and the integrity of multiple fission product barriers was maintained. Finally, the problems apparently involved testing methods and not an actual variation in valve setpoints. The occurrence therefore constituted no undue risk to the health or safety of the public. Due to the potential for operation in a degraded mode permitted by a Limiting Condition for Operation the event is reportable in accordance with Technical Specification 6.9.1.9b.

CORRECTIVE ACTION:

The valve manufacturer, in conjunction with Wyle engineers, refur-bished the valves by ultrasonic cleaning, lapping the seating surfaces and reestablishing the ring positions. Subsequent re-testing indicated the opening pressures to be within the allowable tolerance and no leakage was observed.

The refurbished and retested valves have shown acceptable performance.

The valves will be tested again during the next refueling. At that time support will be provided to insure that testing performed more closely models actual valve operating conditions.

FAILURE DATE:

Crosby Valve and Gage Co.

Pressurizer Safety Valve Part No. HB-86-BP Prepared by R. Frahm 4

//

V Generald4ana"ger -

Salem Operations SORC Meeting No.83-094