05000366/LER-2007-006
| Docket Number(S)Sequential Revisionmonth Day Year Year Month Day Yearnumber Number 05000 | |
| Event date: | 04-17-2007 |
|---|---|
| Report date: | 06-13-2007 |
| 3662007006R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
On April 17, 2007, Unit 2 was at 100 percent power. On that day it was identified during bench testing that two Safety Relief Valves (SRVs) (EIIS Code SB) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit. At the conclusion of bench testing, a total of five of the eleven SRVs were identified as having setpoint drift in excess of the TS limit. The setpoint for each of the eleven SRVs is 1150 +/- 34.5 psig. The following is a tabulation of the test results for the eleven SRVs:
MPL Number Pilot Serial Number As-Found Lift Pressure Percent Drift 2B21-F013A 1189 1250 108.7 2B21-F013B 306 1169 101.7 2B21-F013C 1231 did not lift n/a 2B21-F013D 313 1212 105.4 2B21-F013E 304 1169 101.7 2B21-F013F 1006 1178 102.4 2B21-F013G 310 1158 100.7 2B21-F013H 1010 did not lift n/a 2B21-F013K 1007 1273 110.7 2B21-F013L 1228 1149 99.9 2B21-F013M 1004 1159 100.8 These valves were removed from service during the Spring 2007 refueling outage and replaced with like kind valves that were serviced and tested in accordance with plant procedures.
CAUSE OF EVENT
The cause of the SRV setpoint drift exceeding the allowable plant TS limit is corrosion-induced bonding between the pilot disc and seating surface.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable per 50.73(a)(2)(i)(B) because an event occurred which is prohibited by Technical Specifications. Specifically, multiple test failures of the SRVs is defined as reportable in NUREG-1022, Revision 2, dated October 2000, in section 3.2.2, example 3, titled "Multiple Test Failures.
The 11 SRVs, which are located on the four main steam lines within the drywell between the reactor vessel and the inboard main steam isolation valves (MSIV EIIS Code SB), are required during Modes 1, 2, and 3 to limit the peak pressure in the nuclear system such that it will not exceed the applicable ASME Boiler and Pressure Vessel Code limits for the reactor coolant pressure boundary. Per TS Surveillance Requirement 3.4.3.1, the valves are tested in accordance with the In-service Testing Program to verify the safety function lift setpoints are within the specified limits. The safety relief valves must accommodate the most severe pressurization transient which, for the purposes of demonstrating compliance with the ASME Code limit of 1375 psig peak vessel pressure, has been defined as a closure of all MSIVs with a failure of the direct reactor protection system trip from the MSIV position switches; the reactor ultimately shutdowns from a high neutron flux trip. Analysis of this event using the as-found bench test results for SRV actuation pressures has demonstrated that the resultant peak pressure was within the ASME Code limit.
Furthermore, the plant Technical Specifications overpressure safety limit of 1325 psig dome pressure must be met during normal operations and for anticipated operational occurrences (A00s). The analysis of the as-found test results also showed that for the MSW Closure A00 with the MSIV position switches providing the reactor protection system trip, the resultant dome pressure was within the plant Technical Specifications Safety Limit.
In addition, a non-credited electrical actuation system was installed in 1993 to ensure proper actuation of the SRVs. This system provides a redundant, independent method (i.e., electrical signal) to actuate the SRVs. During the run cycle the redundant electrical system was available.
The system was procured to Class 1 E environmental and seismic standards, and is deemed highly reliable.
Based on this analysis, it is concluded that this event had no adverse impact on nuclear safety.
CORRECTIVE ACTIONS
All eleven pilot valves have been replaced with refurbished pilot valves.
Each of the eleven pilot discs from the valves removed for testing will be replaced with a pilot disc made from Stellite 21 material. Implementation will be tracked under the corrective action program.
Evaluation of additional actions to further improve SRV performance will be tracked under the plant's corrective action program.
ADDITIONAL INFORMATION
Other Systems Affected: No systems other than those already mentioned in this report were affected by this event.
Other Systems Affected: None Failed Components Information:
Master Parts List Number: 2B21-F013 EIIS System Code: SB Manufacturer: Target Rock Reportable to EPIX: Yes Model Number: 7567F Root Cause Code: B Type: Relief Valve EIIS Component Code: RV Manufacturer Code: T020 Commitment Information: This report does not create any new permanent licensing commitments.
Previous Similar Events:
results from the previous three outages where multiple SRV setpoint drift had occurred. Corrective actions for that LER focused on ensuring proper reporting of SRV setpoint drift was performed.