05000354/LER-2004-009

LER-2004-009, As Found Values for Safety Relief Valve Lift Setpoints Exceed Technical Specification Allowable

Docket Number
Event date:	11-19-2004
Report date:	1-17-2005
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3542004009R00 - NRC Website

PLANT AND SYSTEM IDENTIFICATION

General Electric — Boiling Water Reactor (BWR/4) Main Steam — EIIS Identifier {SB)* Safety Relief Valves - EIIS Identifier (--/RV)* *Energy Industry Identification System {EIIS} codes and component function identifier codes appear as {SS/CCC}

IDENTIFICATION OF OCCURRENCE

Event Date: November 19, 2004 Discovery Date: November 19, 2004

CONDITIONS PRIOR TO OCCURRENCE

Hope Creek was in cold shutdown for the twelfth refueling outage (RF12). No structures, systems, or components were inoperable at the time of discovery that contributed to the event.

DESCRIPTION OF OCCURRENCE

On November 9, 2004, Engineering personnel received the initial results of the Main Steam Safety Relief Valves (SRV){SB/RVI (Target Rock Model 7567F) setpoint testing required by Technical Specification 4.4.2.2. That report documented the failure of SRV B to meet TS 3.4.2.1 limit of +/- 3%. On November 19, 2004 additional test results were received. This testing revealed that following Hope Creek Cycle 12 run, two additional SRVs failed to meet the TS limit.

Upon completion of testing, a total of five SRVs experienced setpoint drift outside of the Technical Specification 3.4.2.1 limit of +/- 3%, (values listed below). Action "a" of TS 3.4.2.1 specifies "With the safety valve function of two or more of the above listed fourteen safety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Valve ID As Found TS Setpoint Acceptable Band % Difference (Psig) (psig) (Psig) F013A 1192 1130 1096 -1163 5.5% F013B 1171 1130 1096 —1163 3.6% F013C 1207 1130 1096 —1163 6.8% F013D 1184 1130 1096 —1163 4.8% F013F 1156 1108 1075 —1141 4.3%

CAUSE OF OCCURRENCE

The apparent cause for the "B", "D", and "F" SRV setpoint failures is corrosion bonding/sticking of the pilot disc. PSEG Nuclear has continued to experience as-found setpoint failures on SRVs even with the industry recommended coating (IBRD) installed on the pilot disc. The initial lift being out of specification, with subsequent lifts within specification and/or the initial failure of the stick test, is an indication of the pilot disc sticking. The other two SRV ("A' and "C") setpoint failures are still under investigation. The preliminary assessment indicates that the apparent cause may be misalignment of parts causing the high setpoint.

A review of LERs for the two prior years at Hope Creek and Salem was performed to determine if a similar event had occurred. There was a similar event during the last Hope Creek refueling outage when eight SRVs were found out of the TS required limits of +/- 3%. This event was reported as LER 354104-003-00. Corrective actions were not successful to prevent recurrence.

SAFETY CONSEQUENCES AND IMPLICATIONS

A bounding analysis was performed and documented in NEDC-32511P, "SAFETY REVIEW FOR HOPE CREEK GENERATING STATION SAFETY/RELIEF VALVE TOLERANCE ANALYSIS." This analysis supported the increase in allowable Technical Specification (TS) setpoint drift from + 1 percent to + 3 percent. An individual SRV upper limit setpoint of 1250 psig with 13 SRVs available out of a total of 14 was assumed in the calculation. The calculated peak vessel pressure at the bottom of the reactor vessel was 1331 psig. This provides a margin of 44 psig to the ASME upset limit of 1375 psig. In addition, loads on SRV discharge piping were reanalyzed. The analysis established an allowable percentage increase for each SRV line such that the allowable stresses would not be exceeded. Four of the five valves met their individual acceptance limits. The "A" SRV valve exceeded its limit of +3% with an as-found setpoint of 5.5%.

involved as a result of these valves exceeding the allowable tolerance and the analysis is bounding. Therefore, the public health and safety was not affected.

As a follow-up, however, the discharge piping analysis contained in NEDC-32511P will be re-assessed to determine the impact of the high setpoint of the "A" SRV. NEDC-32511P identifies a maximum increase in the nominal setpoint of "A" SRV to be 3%, without exceeding allowable stresses. The "A" SRV lifted at 5.5% above nominal setpoint. This requires a re-evaluation of the piping due to the potential for overstressing the line if the SRV A had lifted. There is no present operability concern due to the replacement of this pilot assembly with a fully tested spare.

Based on the above and because none of the SRV exceeded the 1250 psig analyzed limit, there was no impact to the health and safety of the public.

A review of this event determined that a Safety System Functional Failure (SSFF) has not occurred as defined in Nuclear Energy Institute (NEI) 99-02.

CORRECTIVE ACTION

• The pilot assembly for each of the failed SRVs was replaced with a fully tested spare assembly.

• The five SRVs will be disassembled and inspected to determine the cause of the high lift points.

• Reactor water chemistry will be evaluated for its potential to support the corrosion bonding of the pilot disc seating surfaces.

COMMITMENTS

This LER contains no commitments.