05000366/LER-2011-002
| Edwin I. Hatch Nuclear Plant Unit 2 | |
| Event date: | 07-05-2011 |
|---|---|
| Report date: | 08-30-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3662011002R00 - 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 July 5, 2011, at approximately 1000 EDT, Unit 2 was at 100 percent rated thermal power (RTP) when the "as-found" testing results of the 2-stage main steam safety relief valves (SRVs) were received which indicated that eight of eleven SRVs (EIIS Code SB) had experienced setpoint drift which resulted in their allowable Tech Spec limits of 1150 +/- 34.5 psig (± 3 percent) being exceeded.
The following is a tabulation of the test results of the eleven SRVs:
MPL Number Pilot Serial Number As-Found Lift Pressure Percent Drift 2B21-F013A 1003 1194 103.83 2B21-F013B 1011 1183 102.87 2B21-F013C 1009 1195 103.91 2B21-F013D 312 1207 104.96 2B21-F013E 1227 1276 110.96 2B21-F013F 311 1271 110.52 2B21-F01 3G 1188 1179 W2.52 2B21-F013H 1190 1243 108.09 2B2 1 -F013K 305 1177 102.35 2B21-F013L 1008 1309 113.83 2B21-F013M 301 1226 106.61 These eleven valves were removed from service during the Spring 2011 refueling outage and preemptively replaced with 3-stage SRVs that had been properly setup and tested at Wyle Laboratories prior to installation.
CAUSE OF EVENT
The root cause of the SRV setpoint drift is attributed to corrosion-induced bonding between the pilot disc and seating surface. This conclusion is based on previous root cause analyses and the repetitive nature of this condition at Hatch and in the industry. In General Electric (GE) service information letter (SIL) 196, Supplement 16, GE determined that condensation of steam in the pilot chamber of Target Rock 2-stage SRVs can cause oxygen and hydrogen dissolved in the steam to accumulate. As steam condenses in the relatively stagnant pilot chamber, the dissolved gases are released. In a volume such as the pilot chamber which is normally at approximately 1000 psig and temperature of 545 degrees F, the total pressure consists primarily of water vapor partial pressure because 544.6 degrees F is the saturation temperature at 1000 psig. This wet, hot, high-oxygen atmosphere can be very corrosive and can increase the likelihood of corrosion-induced bonding of the pilot disk to its seat. It was also noted that proper insulation minimizes the accumulation rate of non-condensable gases and the steady-state oxygen partial pressure. Despite improvements made in maintaining the integrity of insulation for the previously installed 2-stage SRVs the corrosion-induced bonding continued to occur as evidenced by the test results from this most recent outage.
This event is reportable in accordance with (iaw) Title 10 of the Code of Federal Regulations (CFR), Part 50.73(a)(2)(i)(B) because an event occurred which is prohibited by the Technical Specifications (TS). Specifically, an example of multiple test failures is given in NUREG 1022, Revision 2, "Event Reporting Guidelines 10CFR50.72 and 50.73" which describes the sequential testing of safety valves. This example notes that "Sometimes multiple valves are found to lift with set points outside of technical specification limits.
NUREG 1022 further notes that "discrepancies found in technical specifications surveillance tests should be assumed to occur at the time of the test unless there is firm evidence, based on a review of relevant information (e.g., the equipment history and the cause of failure) to indicate that the discrepancy occurred earlier. However, the existence of similar discrepancies in multiple valves is an indication that the discrepancies may well have arisen over a period of time, and the failure mode should be evaluated to make this determination." Based on this guidance and the fact that the development of the corrosion occurred over a period of time of plant operation, the determination was made that this "as found" condition is reportable under the reporting requirements of 10CFR50.73(a)(2)(i)(B).
There are eleven (11) SRVs located on the four main steam lines within the drywell (ElIS Code NH) between the reactor pressure vessel (ElIS Code AD) and the inboard main steam isolation valves (MSIV EIIS Code SB). These SRVs are required to be operable 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. The SRVs are tested iaw TS surveillance requirement 3.4.3.1 in which the valves are tested as directed by the In-Service Testing Program to verify lift setpoints are within their specified limits to confirm they would perform their required safety function of overpressure protection. The SRVs 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 by an event involving the closure of all MSIVs with a failure of the direct reactor protection system trip from the MSIV position switches with the reactor ultimately shutting down as the result of a high neutron flux trip (a scenario designated as MSIVF). This MSIVF event analysis was performed by the Nuclear Fuels Department for the H2C21 "as-found" condition of the SRVs. The results from this analysis showed a small increase in peak pressures relative to the Hatch-2 Cycle 21 reload licensing analysis (RLA) results. The higher peak pressures were due to the fact that eight of the eleven SRVs opened at pressures higher than that which was assumed in the RLA. It should be noted that in this analysis, the larger actual valve bore size was used in the calculations for nine of the valves rather than the smaller bore size which was conservatively assumed in the RLA. Therefore, higher steam flow capacities than those assumed in the RLA were used in this analysis for those nine valves. Based on the analysis, the calculated minimum margin to the 1375 psig ASME Boiler and Pressure Vessel Code overpressure limit for peak vessel pressure would have been 27.7 psig and the minimum margin to the 1325 psig Tech Spec Safety Limit for the reactor steam dome pressure would have been 2.9 psig during an MSIVF event during Cycle 21 operation. Therefore, the analysis of the "as found" test would have continued to perform its required safety function if called upon in its "as found" condition. Therefore, this event had no adverse impact on nuclear safety.
CORRECTIVE ACTIONS
All eleven 2-stage SRV pilot valves were preemptively replaced with new 3-stage SRV pilot valves as the long term corrective action. The use of the 3-stage SRVs is regarded as an industry- wide solution for the corrosion-induced bonding phenomenon which has been a historic industry issue since the early 1980s.
ADDITIONAL INFORMATION
Other Systems Affected: None Failed Components Information:
Master Parts List Number:2B21-F013A, B, C, D, E, F, G, H, K, L, M 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:
actions included refurbishment of the pilot valves and included the replacement of the pilot discs with discs made from Stellite 21 material. Additionally, the insulation surrounding each SRV was upgraded to improve resistance to corrosion-induced bonding. These were the same actions that were taken following similar failures reported in LER 2-2009-001, since improved results had been seen to some degree in the industry for at least one operating cycle when these actions were implemented.
Multiple examples of SRV setpoint drift occurred and were also reported in LERs 2-2008- 004, 1-2008-002, 2-2007-006 and 1-2006-003. These instances of SRV setpoint drift occurred due to like causes which have been noted to be similar to those of the ongoing industry issues with these type SRVs. In each of these cases SNC concluded that the issue, and this assertion will be confirmed during the performance of future "as found" testing during the next scheduled refueling outage.