05000321/LER-2014-003
| Edwin I. Hatch Nuclear Plant Unit 1 | |
| Event date: | 05-07-2014 |
|---|---|
| Report date: | 7-7-2014 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3212014003R00 - NRC Website | |
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1. FACIUTY NAME 3. PAGE 2. DOCKET 6. LER NUMBER
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 May 7, 2014, at approximately 0837, Unit 1 was at 99.9 percent rated thermal power (RTP) when the report of the "as-found" testing results of the 2-stage main steam safety relief valves (SRVs) were received which indicated that five of eleven SRVs (EIIS Code SB) had experienced setpoint drift during the previous operating cycle which resulted in their allowable TS surveillance requirement (SR) 3.4.3.1 limits of 1150 4- 34.5 psig (+1- 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 (psig) Peicent Drift 1621-F013A 1007 1175 2.17 1621-F013B 1010 1172 1.91 1821-F013C 1003 1223 6.35 1621-F013E 1226 1181 2.70 1621-F013F 1011 1168 1.57 1821-F013G 314 1181 2.70 1621-F013H 312 1166 1.39 1821-F013J 304 1204 4.70 1621-F013K 301 1198 4.17 1621-F013L 306 1201 4.43 All eleven valves were removed from service during the spring 2014 refueling outage and replaced with 3-stage SRVs. The 3-stage SRVs have a modified pilot that helps reduce the possibility of inadvertent lift and leak by.
The 2-stage SRVs installed on Unit 2 prior to the 2013 refueling outage utilized platinum coated pilot discs. Even though these valves are identical to the valves that were used in Unit 1 Cycle 26, there is confidence that the 2-stage SRVs currently installed on Unit 2 will function reliably for the remainder of the cycle. This is based on the recent that the SRVs were successfully tested prior to installation and that new platinum coated pilots were installed. Plans are to replace the existing 2-stage SRVs on Unit 2 with 3-stage SRVs during the next scheduled refueling outage as a long term corrective action.
CAUSE OF EVENT
The root cause of the SRV setpoint drift is attributed to corrosion-induced bonding between the pilot disc and its seating surface. This conclusion is based on previous root cause analyses and the repetitive nature of this condition at Plant 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 MO FORM 3460.102.2014) 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 a 1000 psig pressure and a 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.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable in accordance with Title 10 of the Code of Federal Regulations (CFR), Part 50.73(a)(2)(i)(B) because an event occurred which is prohibited by TS surveillance requirement (SR) 3.4.3.1. Specifically, an example of multiple test failures is given in NUREG-1022, Revision 3, "Event Reporting Guidelines 10 CFR 50.72 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 states in the example that "discrepancies found in TS 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 10 CFR 50.73(a)(2)(i)(B).
There are eleven SRVs located on the four main steam lines within the drywell (EllS Code NH) in between the reactor pressure vessel (EllS Code AD) and the inboard main steam isolation valves (MSIV, ENS 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 in accordance with TS surveillance requirement 3.4.3.1 in which the valves are tested as directed by the In-Service Testing Program to verify lift set points 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 011375 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 for the Unit 1 Cycle 24 "as-found" condition of the SRVs showed that there was adequate margin to the vessel dome pressure and ASME vessel overpressure limits. Another analysis was performed which compared the Cycle 26 "as-found" SRV measured opening pressures to those of the Cycle 24 reload licensing analysis (RLA). The results from this analysis showed a decrease in peak pressures due to the fact that eight of the eleven SRVs opened at pressures lower than those which were measured in the Cycle 24 RLA.
Based on this comparison, recognizing that SRV opening pressures used in the Cycle 24 RLA were generally much larger than the Cycle 26 measured pressures, and since the Cycle 24 RLA showed sufficient margin to the Technical Specification 2.1.2 dome pressure safety limit and the ASME Pressure Vessel Code overpressure limit, the results of the Cycle 24 remain bounding. Therefore the peak pressure at the bottom of the vessel remained below the ASME Boiler and Pressure Vessel code limit, and the peak reactor pressure vessel dome pressure remained within the Tech Spec Safety Limits.
Additionally, a highly reliable, though non-credited, electrical actuation system serves as a redundant, independent method to actuate the SRVs. During Cycle 26 this redundant electrical logic system was fully functional.
Based on the analyses performed, the overpressure protection system 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 and was of very low safety significance.
CORRECTIVE ACTIONS
The 2-stage SRVs with platinum-coated pilot discs were removed from Unit 1 during the 2014 refueling outage and replaced with 3-stage SRVs that have a modified pilot. 3-stage SRVs typically do not exhibit set point drift due to their design. The modified pilots will help reduce spurious openings and leak-by due to system vibration.
Additionally, current plans are to replace the remaining 2-stage Unit 2 SRVs with 3-stage SRVs with the same modified pilots during the 2015 refueling outage.
ADDITIONAL INFORMATION
Other Systems Affected: None Failed Components Information:
Master Parts List Number: 1B21-F013C, D, J, K, L Manufacturer: Target Rock Model Number: 7567F Type: Relief Valve Manufacturer-Code: T020 EIIS System Code: SB Reportable to EPIX: Yes Root Cause Code: B EIIS Component Code: RV Commitment Information: This report does not create any licensing commitments.
PREVIOUS SIMILAR EVENTS:
of the 2-stage SRVs with 2-stage SRVs whose pilot discs had undergone a platinum surface treatment which was considered at that time to be the long term fix for this corrosion bonding issue.
of the 2-stage SRVs with 3-stage SRVs during the Unit 2 Spring 2011 refueling outage which was considered at that time to be the long term fix for this corrosion bonding issue. Subsequent to that outage the 3-stage SRVs exhibited signs of unacceptable leakage which resulted in two separate outages that involved changing out four SRVs during the first outage and the remaining seven SRVs during the subsequent outage in May 2012. The 3-stage SRVs were replaced with 2-stage SRVs containing pilot discs that had undergone the platinum surface treatment.
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.
NRC FORM 36(0102-2014y