05000366/LER-2009-001
| Edwin I. Hatch Nuclear Plant Unit 2 05000 366 1 Of 4 | |
| Event date: | 03-12-2009 |
|---|---|
| Report date: | 05-04-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3662009001R00 - 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 March 12, 2009 at approximately 4:00 pm EDT, Unit 2 was in refuel mode at 0 percent power. On that day, it was determined that during bench testing at an independent testing facility five Safety Relief Valves (SRVs) (EIIS Code SB) experienced setpoint drift that exceeded the allowable plant Technical Specifications (TS) limit above the setpoint value.
All eleven SRV's were tested. Of those eleven, five failed the as found testing. The following is a tabulation of the test results for the five SRVs that failed the as-found test:
MPL Number Pilot Serial Number As-Found Lift Pressure Percent Drift 2B21-F013A 302 1193 103.7 2B21-F013B 315 1209 105.1 2B21-F013D 314 1200 104.3 2B21-F013H 307 1204 104.7 2B21-F013M 1005 1187 103.2 These valves were removed from service during a planned refuel outage and replaced with like kind valves that were serviced and tested in accordance with plant procedures.
CAUSE OF EVENT
The initially identified cause of the SRV setpoint drift above the setpoint value 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 (TS). 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 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 � accordance with the In-service Testing Program to verify the safety function lift setpoints are within the specified limits.
The impact of the "as found" setpoints for these safety relief valves was analyzed using the most severe pressurization transient which, for the purposes of demonstrating compliance with the ASME Code limit of 1375 psi 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 the tested SRV's concluded that there would have been at least a 50 psi margin to the 1375 psi ASME Boiler and Pressure Vessel Code overpressure limit. Even though this transient is not an anticipated operational occurrence (A00), the analysis demonstrates that even under the extreme conditions assumed adequate margin to the ASME Code limit of 1375 psi still exists.
The plant Technical Specifications overpressure safety limit of 1325 psi dome pressure must be met during normal operations and for anticipated operational occurrences (AOOs). The analysis of the as-found test results also showed that there is approximately a 35 psi margin to the 1325 psi Tech Spec Safety Limit during the limiting MSIVF event in Hatch-2 Cycle 20.
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 lE 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 pilot valves have been replaced with refurbished pilot valves which have been certified to actuate within 11.5 psi of the setpoint and have disc made from stellite 21 material.
Each of the 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.
Any additional actions to further improve SRV performance will be tracked under the plant's corrective action program.
ADDITIONAL INFORMATION
Other S stems Affected: None .■■■■ PRINTED ON RECYCLED PAPERNRC FORM WM (9-2007) U.S. NUCLEAR REGULATORY COMMISSIONNRC FORM 366A� LICENSEE EVENT REPORT (LER)�(9.2007) CONTINUATION SHEET r� 1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGE 2009� -� 001� -� 0 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:
Corrective actions for that LER, replacement of discs were implemented but discs made of stellite 21 for the Unit 2 SRV's were not available for all of the replaced discs and thus could not have prevented the current event.
Corrective actions for that LER, replacement of discs with stellite 21 discs, were not yet implemented for the Unit 1 SRV's and thus could not have prevented the current event.
Corrective actions for this LER, replacement of discs with stellite 21 discs, were not yet implemented for the Unit 2 SRV's and thus could not have prevented the current event.
described results from the previous three outages where multiple SRV setpoint drift had occurred. Corrective actions for this LER focused on ensuring the proper reporting of SRV setpoint drift was� erformed.
�