05000333/LER-2001-005
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3332001005R00 - NRC Website | |
U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) LER NUMBER (6) DOCKET (2) PAGE (3) 01 00 05 James A. FitzPatrick Nuclear Power Plant
Event Description:
On June 19, 2001, while the plant was operating at 100 percent power, Engineering determined that two Safety Relief Valve (SRV) [S13] pilots removed during Cycle 14 had as-found setpoints in excess of the 1145 psig +/- 3 percent (i.e., 1111 to 1179 psig). The as-found allowed tolerance of 1145 psig +/- 3 percent is required per Technical Specification (TS) 4.6.E. Both SRVs exceeded the high limit of 1179 psig. One of the eleven pilots (serial 1047) was removed during Forced Outage (FO)145 in August 2000. The remaining ten were removed during R014. The removed SRV pilots were tested at Wyle laboratories during June 6 through June 12, 2001. The results from these tests were reported to FitzPatrick by Wyle Laboratories on June 19, 2001.
Test Results:
Pilot Serial Number Plant Valve Number As-Found Setpoint Pass/Fail (pass unless otherwise noted) 1062 02RV-71A 1147 1088 02RV-71B 1268 Fail 1192 02RV-71C 1155 1218 02RV-71D 1150 1056 02RV-71E 1157 1047 02RV-71F 1164 1050 02RV-71G 1127 1193 02RV-71H 1157 1217 02RV-71J 1251 Fail 1194 02RV-71K 1152 1196 02RV-71L 1161 Technical Specification 3.6.E.1 only requires nine operable SRVs. The Specification states in part:
During reactor power operating conditions and prior to startup from a cold condition, or whenever reactor coolant pressure is greater than atmosphere and temperature greater than Since only two pilot valves exceeded the allowable tolerance, the safety/relief mode of nine valves remained operable, satisfying this specification and assuring adequate overpressure protection in all cases.
U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) LER NUMBER (6) DOCKET (2) PAGE (3) 01 00 05 James A. FitzPatrick Nuclear Power Plant Event Description (continued):
Technical Specification Table 3.2-7, however, requires a reduction in the ATWS High Pressure setpoint trip from discussed in Analysis, below). This setpoint remained at setpoint out-of-tolerance was not known until setpoint testing was performed. This report is therefore being made under 10 CFR 50.73(a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications," due to the higher than allowable ATWS High Pressure setpoint trip given the number of inoperable SRVs.
Cause:
SRV setpoint drift is caused by corrosion bonding between the SRV pilot disc and seat (Cause Code B). With a bond forming between the pilot disc and seat, more pressure is needed to raise the pilot disc off the seat. As the normal balance of pilot assembly spring force and steam pressure force necessary to lift the pilot disc corresponds to the nominal setpoint of the SRV, the pilot disc to seat bond results in a higher pilot setpoint.
An oxygen rich environment in the pilot assembly, due to the radiolytic breakdown of water to hydrogen and oxygen, causes the corrosion bonding. Oxygen accumulates in the area of the pilot disc because the pilot assembly is a high point on the main steam [SB] line.
Analysis:
Two events are analyzed in determining the adequacy of overpressure protection; the limiting anticipated transient, MSIV Closure with Flux Scram; and the limiting unanticipated transient, the Pressure Regulator Failure Open (PRFO) with failure to scram (ATWS). The out-of-tolerance condition reported by this LER did not compromise overpressure protection for either analyzed event.
The reported condition therefore is of minor safety significance.
The limiting anticipated transient, the MSIV Closure with Flux Scram, is analyzed as part of the Cycle Reload Analysis using conservative assumptions and setpoints. The SRV setpoints used for the Cycle 14 Reload Analysis were nine SRVs opening at 1195 psig, with two valves out of service. The as-found setpoints given above are less than those assumed (that is, nine SRVs opened at psig and the other two were not entirely out of service); therefore, the Cycle 14 Reload Analysis is bounding and further evaluation of the MSIV Closure with Flux Scram is not required. Further, the as found condition satisfies the Technical Specification 2.2.1.B Limiting Safety System setting requirements for SRVs, which are:
At least 9 of the 11 reactor coolant system safety/relief valves shall have a nominal setting of 1145 psig with an allowable setpoint error of +/-3 percent.
FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) James A. FitzPatrick Nuclear Power Plant 01 05 00 Analysis (continued):
The limiting unanticipated transient, the Pressure Regulator Failure Open (PRFO) with failure to scram (ATWS), was analyzed in conjunction with the ELLLA (Extended Load Line Limit Analysis) project in GE-NE-A42-00137-2-01, ATWS Overpressure Analysis for FitzPatrick, dated March 2000. The ATWS analysis uses more nominal assumptions and setpoints, because the ATWS is not part of the plant design basis. The ELLLA analysis report lists eleven specific SRV setpoints used in the analysis. It also states that SRV out of service cases were analyzed by assuming the lowest setpoints from the list did not open. Using this methodology, satisfactory analytical results were obtained for limiting cases with two SRVs out of service and the ATWS High Pressure actuation setpoint at as-found setpoints given above are less than those assumed in the analysis; that is, the nine satisfactory SRV setpoints were all less than the third lowest setpoint used in the analysis (1170 psig), and the two out of tolerance SRVs were not entirely out of service. Therefore, the ELLLA ATWS Analysis is bounding and further evaluation of the PRFO ATWS is not required.
The above notwithstanding, Technical Specifications have not been revised consistent with this latest analysis. Thus, although current analysis demonstrates satisfactory results with zero, one, or two SRVs out of service and the ATWS High Pressure actuation setpoint at Specifications Table 3.2-7, Note 3 still states the following:
The ATWS Reactor Pressure High Recirculation Pump Trip setpoint shall be when either zero or one SRVs are out of service. The setpoint shall be two or more SRVs are out of service.
The failure of two SRVs therefore constitutes an "operation or condition which was prohibited by the plant's Technical Specifications.
Extent of Condition:
All of the SRVs are susceptible to setpoint drift due to pilot disc to seat bonding. This is an industry issue that has been the subject of both NRC and BWROG generic assessment.
A BWROG recommended modification to provide pressure switch actuation of the SRVs was installed during R014. This modification provides an electric actuation of SRV pilot valves based upon a pressure switch setpoint. This provides a diverse, redundant method of SRV actuation which is not susceptible to pilot disc-seat bonding. As such, this modification will mitigate and limit the extent of condition to one part of a diverse SRV actuation methodology.
FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) 05000333 YEAR SEQUENTIAL NUMBER REVISION NUMBER 5 � OF � 5 James A. FitzPatrick Nuclear Power Plant 01 05 00 Corrective Actions Ongoing Prior to this Report:
1. All 11 pilot assemblies were removed and replaced with steam certified assemblies.
2. All SRV pilot assemblies will continue to be tested and replaced each operating cycle.
3. A BWROG recommended modification to provide pressure switch actuation of the SRVs was installed during R014.
Corrective Actions for this Event:
1. A Technical Specification change will be requested to incorporate the most current analysis into the Table 3.2-7 requirements for the ATWS Reactor Pressure High Recirculation Pump Trip setpoint.
(Scheduled Completion Date: 09/30/01) Safety System Functional Failure Review This event did not result in a safety system functional failure in accordance with NEI 99-02, Revision 1.
Similar Events:
LERs99-003, 98-002,95-006 Revision 1,95-001, 94-005,94-002, 92-016,90-018, 89-026,88-010, 88-004,87-004, 85-013,85-009 Failed Component Identification:
Manufacturer: � Target Rock Corporation Model Number: � 7567F-10 NPRDS Manufacturer Code: � T020 NPRDS Component Code: � Valve
Reference:
1. GE-NE-A42-00137-2-01, "ATWS Overpressure Analysis for FitzPatrick," dated March, 2000