05000354/LER-2015-004
| Hope Creek Generating Station | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3542015004R01 - NRC Website | |
the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person Is not required to respond lo, the information collection.
PLANT AND SYSTEM IDENTIFICATION
General Electric — Boiling Water Reactor (BWR/4) Main Steam — EllS Identifier {SB}* Safety Relief Valves — EllS Identifier {SB/RV}* *Energy Industry Identification System {EllS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE
Event Date: June 2, 2015 Discovery Date: June 2, 2015
CONDITIONS PRIOR TO OCCURRENCE
When the reports of the 'as-found' results were received, Hope Creek was in Operational Condition (OPCON) 1 at approximately 100 percent rated thermal power. No other structures, systems or components that could have contributed to the event were inoperable at the time of the event.
DESCRIPTION OF OCCURRRENCE
During the nineteenth refueling outage (H1R19) at Hope Creek Generating Station (HCGS), all 14 Main Steam Safety Relief Valves (SRV) pilot stage assemblies {SB/RV} were removed and tested at NWS Technologies. The SRVs are Target Rock Model 7567F two-stage SRVs. During the period from June 2, 2015 through June 10. 2015, HCGS received the results of the 'as-found' set pressure testing required by Technical Specification (TS) Surveillance Requirement (SR) 4.4.2.2. A total of ten of the 14 SRV pilot stage assemblies had setpoint drift outside of the required TS 3.4.2.1 tolerance values of +/-3% of nominal value.
The 'as-found' test results for the ten SRVs not meeting the TS requirements are as follows:
Valve ID As Found TS Lift Setting Acceptable Band (psig) % Difference (psig) (psig) Actual F013C 1216 1130 1096.1 — 1163.9 7.61% F013F 1240 1108 1074.8 — 1141.2 11.90% F013G 1208 1120 1086.4 — 1153.6 7.86% F013H 1148 1108 1074.8 — 1141.2 3.60% F013J 1161 1120 1086.4 — 1153.6 3.66% F013K 1161 1108 1074.8 — 1141.2 4.80% F013 L 1165 1120 1086.4 — 1153.6 4.00% F013 M 1207 1108 1074.8 — 1141.2 8.90% F013P 1221 1120 1086.4 — 1153.6 9.00% F013R 1169 1120 1086.4 — 1153.6 4.38% the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to Impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person Is not required to respond to, the Information collection.
2015 - 004 -01 specified code safety valve function lift setting, within a tolerance of +/- 3%. 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 />." Therefore, this is a condition reportable under 10 CFR 50.73(a)(2)(i)(B) as an Operation or Condition Prohibited by TS.
The extent of condition for this event is to expand the scope of the SRV Group 1 valve testing, per ASME OM Code Section 1-1320 for Class 1 Pressure Relief Valves. However, since all 14 SRV pilot stage assemblies were removed and replaced with tested spares during the refueling outage (H1R19), the extent of condition scope was satisfied.
CAUSE OF EVENT
The cause of the setpoint drift for the ten SRV pilot stage assemblies is attributed to corrosion bonding between the pilot disc and seating surfaces, which is consistent with industry experience. This conclusion is based on previous cause evaluations and the repetitive nature of this condition at HCGS and within the BWR industry.
SAFETY CONSEQUENCES AND IMPLICATIONS
All 14 SRVs were operable during Cycle 19 and there were no events during that cycle that required operation of the SRVs. All SRVs lifted well below the Safety Limit, providing reasonable assurance that accident analysis conclusions would remain valid. The industry has recognized that corrosion bonding occurs during the operating cycle. Once an SRV lifts, the corrosion bond breaks and subsequent openings occur very close to the set point as demonstrated during testing.
Two technical evaluations were performed to assess the aggregate safety-significance of the 10 SRVs with out of tolerance initial lift setpoints and determine whether the condition would have had an adverse effect on the safety function of the valves or other affected system structures and components (SSCs). One technical evaluation looked at 1) the Reactor Pressure Vessel (RPV) over-pressure design function of the valves; 2) the impact of higher relief setpoints on other safety systems (i.e., HPCI, RCIC, and SLC); and 3) fuels considerations. The second technical evaluation looked at stress related issues (down-comer piping, supports, spargers, and torus loads).
The evaluations concluded that the as-found condition was bounded by margins which exist in current Hope Creek design analyses; thus, the aggregate effect of this condition has no Safety Significance. In all cases, the RCS would have remained within allowable limits, and safety-related systems relied upon during high-pressure events (HPCI, RCIC and SLC) would have functioned sufficiently in accordance with the station's design bases had an accident or limiting transient occurred during Cycle 19. Fuel limits were not adversely affected by this condition. Evaluation results of the stresses on down-comer piping, supports, spargers, and the torus loads showed satisfactory results.
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0113112017 hours. Reported lessons learned are Incorporated Into the licensing process and fed back to Industry. Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet e-mall to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to Impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the Information collection,
SAFETY SYSTEM FUNCTIONAL FAILURE
A review of this condition and the associated evaluations determined that a Safety System Functional Failure (SSFF) as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," did not occur.
PREVIOUS EVENTS
A review of events for the past four years at Hope Creek was performed to determine if similar events had occurred.
Similar events-occurred during the 2012 (H1R17) and 2013 (H1R18) Hope Creek refueling outages when multiple SRVs were found out of the TS required limits of +/- 3%. These events were reported as LER 354/2012-004-01(six inoperable SRVs) and LER 354/2013-007-00 (five inoperable SRVs).
CORRECTIVE ACTIONS
1. All 14 SRV pilot stage assemblies were removed and replaced with pre-tested, certified spare pilot valves (H1R19).
2. Evaluate options for the replacement of the currently installed Target Rock two-stage SRVs with a design that eliminates setpoint drift events exceeding +/-3% and improve SRV reliability. The replacement schedule will be developed after a suitable valve is identified.
COMMITMENTS
There are no regulatory commitments contained in this LER.