05000333/LER-1995-001, :on 950106,test Results for Five of Six Valves Exceeded 1% Tolerance Allowed by TS for Valve Actuation. Caused by Initial Lift Setpoint Drift Set High for All Pilot Mechanisms.Replaced SRV Pilot Assemblies
| ML20082L838 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/13/1995 |
| From: | Mulcahey E POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20082L831 | List: |
| References | |
| LER-95-001, LER-95-1, NUDOCS 9504210325 | |
| Download: ML20082L838 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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) 10 CFR 50.73(a)(2)(1) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(x) |
| 3331995001R00 - NRC Website | |
text
NRC FORM 366 U.S. IRJCLEAR KeALATORY CIDOIISSION APPRDWED BT CBS 50. 3150-0104 (5-92)
EMPI K s 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITN
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FACILITY NAE (1)
DOGET ItseER (2)
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FitzPatrick Nuclear Power Plant 05000333 01 OF 04 TITLE (4)
R: actor Safety Relief Valve Setpoint Drift EYENT DATE (5)
LER IRABER (61 REPORT DATE (7)
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DAY Yr.AR YEAR FACILITT HAT DOCKET IILDIBER 01 06 95 95 001 01 04 13 95 05000 TN is REPORT IS SLBMITTED PLRSUANT T TE REEJIREENTS OF 10 CFR 1: (Check one or more' (11)
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LEVEL (10) 20.405(a)(1)(11) 50.36(c)(2) 50.73(a)(2)(vil)
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NAME TELEPHONE NUMBER (Include Area Code)
Mr. Eric Mulcahey, Senior Licensing Engineer (315) 349-6324 CupFtETE 010E LINE FOR EACM CtBFONENT FAILL5tE DESCRIBED IN TNIS KP(RT (13)
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ASSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten Lines) (16)
During the 1994/95 refueling outage, the pilot assemblies for six safety relief valves were removed for testing and recertification.
On 1/6/95 the Authority received notification from the test facility that the test results for five of the six valves exceeded the 1 percent tolerance allowed by Technical Specifications for valve actuation.
Setpoint drift ranged from 1.05% to 11.28%
A plant specific analysis performed previously envelopes the as-found setpoints.
This analysis determined that setpoint drift greater than that found would have no significant safety impact on vessel overpressure margin, thermal limits, or Emergency Core Cooling system performance.
Corrective action included replacing the SRV pilot assemblies with recertified assemblies composed of a material that we believe will reduce seat bonding and continued participation in the BWR Owners Group addressing setpoint drift.
LER numbers94-002, 92-016,90-018, 89-026,88-010, 88-004,87-004, 85-013, and 85-009 are similar events involving SRV setpoint drift.
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pWC FORM 366A U.S. MACLEAR REGALAT(RT EXB9llSSI(El APPA0 WEB SY Ge W. 3150-0106 (5-92)
EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COWLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
LICENSEE EVENT REPORT (LER)
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Jcmes A. FitzPatrick Nuclear YEAR 02 OF 04 Power Plant 05000333 95 01 001 TEXT (If more sonee is realfred. use additional cooles of NRC Form 366A) (IT)
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Event Descriotion During the 1994/95 refueling outage, the actuating mechanisms (pilots) from eix safety relief valves (SRVs) [SB] were removed and sent to a test facility for testing, refurbishment and recertification.
On January 6, 1995, the Authority received facsimile notification that five of the pilot nachanisms had actuated outside the 1 percent setpoint tolerance that is required by Technical Specification 2.2.1.B.
The initial set pressure observed for the SRV pilots were:
Observed Deviation Pilot Nameplate Initial From Plant Assembly Set Pressure Set Pressure Nameplate Valve No.
Serial No.
(PSIG)
(PSIG)
Percentaae l
02RV-71G 1052 1140 1206 5.79%
i 02RV-71C 1053 1140 1173 2.89%
O2RV-71A 1045 1140 1152 1.05%
02RV-71B 1110 1140 1167 2.37%
02RV-71J 1218 1140 1147 0.61%
02RV-71K 1051 1090 1213 11.28%
Each pilot mechanism are tested four times.
For each of the five pilot mechanisms that were out of tolerance, the initial actuation was the anximum out of tolerance.
02RV-71J was never out of tolerance.
Testing results for the remaining five SRVs will be reported under LER-95-006.
Cause
The pilot mechanisms are tested then disassembled, inspected and repaired (as needed) prior to recertification.
Because the initial lift setpoint drift was high for all pilot mechanisms, it in possible that there was some pilot disc to seat corrosion induced bonding.
This observation is eupported by subsequent test runs which showed uniformly lower lift actpoints.
All six pilot mechanisms required refurbishment because of leakage observed at normal operating pressure both before and after the test runs.
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NRC FORM 3664 U.S. NUCLEAR REGULATORY CUfetISSION APPN0WED BY G5 NO. 3150-0104 (5-92)
EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.
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FitzPatrick Nuclear 05000333 95 0
03 OF 04 Power Plant 001 TEXT (If more space is reouired use additional copies of NRC Form 366A) (IT)
Analysis The observed setpoint of five of six SRV pilot mechanism deviated by more than 1 percent from the values specified in Technical Specification 2.2.1.B.
Technical Specification Amendment 217, issued on September 28, 1994, modifies SRV performance limits to provide a single nominal setpoint for all valves to 1110 psig, a setpoint tolerance of 3 percent, and to allow for two SRV's to be inoperable during continuous power operation.
This occurrence is, however, being reported under the provision of 10CFR50. 73 (a) (2) (i) (B) as an operation of the plant in a condition prohibited by the Technical Specification that were in effect during the p st cycle.
The remote actuation (operator demand) and automatic d pressurization system (ADS) functions would not have been effected by this event.
An analysis to determine the effects of SRV setpoint drift was initiated as a result of earlier similar events (LER-88-004 and LER-94-002) cnd has been completed.
This analysis considered plant operation with two ADS SRVs inoperable and catablished an upper bound for the remainder of the SRVs.
The analysis chowed that continuous operation of the plant would be acceptable with nine SRVs actuating at 1195 psig.
The acceptance criteria for this analysis was a 50 psi margin to the ASME code upset reactor vessel pressure limit of 1375 psig during the limiting overpressure event.
Additionally, the cnalysis confirmed that setpoint drift of nine SRVs to the 1195 psig limit would not adversely affect the following:
High Pressure Coolant Injection (HPCI) [BJ) system Reactor Core Isolation Cooling (RCIC) [BN] system o
Primary Containment [NH] integrity Fuel Thermal Limits Emergency Core cooling System (ECCS)/ Loss of Coolant Accident e
(LOCA) performance
Corrective Actions
1.
The pilot assemblies were replaced with refurbished and recertified assemblies prior to startup following the maintenance outage.
The removed pilot assemblies will be refurbished and recertified for future installation.
2.
All SRVs, rather than half as specified in the Technical Specifications, will continue to be subjected to test, refurbishment and recertification once each operating cycle.
MRC FDRM 366A U.S. mrt m REGA.ATORY CX3ellSSION APPROWB BY 3 5 HD. 3150-0106 (5-92)
EXPIRES 5/31/95 ESilMATED BURDEN PER RESPONSE TO COMPLY WITH TH18 INFORMATION COLLECTION REQUEST: 50.0 HRS.
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FitzPatrick Nuclear YEAR sEQUE AL RE Power Plant 05000333 04 OF 04 95 01 001 TEXT (If more space is reauired, use additional cooles of NRC Fore 366A) (17) 3.
Four currently installed pilots (S/N's 1050, 1062, 1088 and 1217) have platinum alloy pilot discs which is the latest effort by the BWR owner's group that provides an alternate disc material in an effort to limit disc to seat bonding due to corrosion.
In addition, pilot S/N 1047 was retrofitted with a platinum alloy disc and has been in operating service for seven months.
As found testing of this pilot is expected to be done during Spring, 1995.
This test will provide information as to effectiveness of the disc material changes.
4.
The Authority will continue its participation in the BWR owners group to address the SRV setpoint drift issue.
Additional Information
Failed Component Identification:
Manufacturer:
Target Rock Corp Model Number:
7567F-010 NPRDS Manufacturer Code: T020 NPRDS Component Code:
Valve
Similar Events
LER-85-009,85-013, 87-004,88-004, 88-010,89-026, 90-018,92-016 and 94-002, are similar events which reported SRV setpoint drift.
Reason for Update:
To change the EIIS Code for Safety Relief Valves from [AD] to [SB] and change the location of pilot assembly
Also added a j
reference to LER-95-006 which reports the testing results for the remaining five SRVs.
1 l