05000354/LER-2001-006
| Hope Creek Generating Station | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3542001006R00 - NRC Website | |
DOCKET (2)
UMBER
PLANT AND SYSTEM IDENTIFICATION
General Electric — Boiling Water Reactor (BWR/4) *Energy Industry Identification System (EIIS) codes and component function identifier codes appear as {SS/CC}
IDENTIFICATION OF OCCURRENCE
Event Date: October 10, 2001 Discovery Date: October 10, 2001
CONDITIONS PRIOR TO OCCURRENCE
The plant was in OPERATIONAL CONDITION 3 (hot shutdown) for Hope Creek's 10th refueling outage.
No structures, systems, or components were inoperable at the time of the occurrence that contributed to the event.
DESCRIPTION OF OCCURRENCE
At approximately 0332, on October 10, 2001, while performing a primary containment walk down at the beginning of the 10th refueling outage, plant personnel observed a leak on the 'A' Reactor Recirculation {AD} Pump {PMP} suction pipe elbow taps. The observed leak was producing a 3-4' spray with the reactor vessel pressure at approximately 300-400 psig. The leak was located where a one-inch pipe is welded to the 28-inch suction line of the "A" recirculation pump. Further investigation revealed that the leak was coming from the weld area and was, therefore, a through-wall leak of the reactor coolant system pressure boundary.
Upon discovery of the nature of the leak, the control room operating crew entered Technical Specification (T.S.) 3.4.3.2 "Reactor Coolant System — Operational Leakage." T.S. 3.4.3.2 requires that with any pressure boundary leakage, the unit be placed in 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 />. At the time the leakage was identified, Hope Creek Station was already in Mode 3 (HOT SHUTDOWN) with the reactor coolant system at a pressure of approximately 300 to 400 psig. The unit achieved COLD SHUTDOWN on October 10, 2001 at approximately 0917 hours0.0106 days <br />0.255 hours <br />0.00152 weeks <br />3.489185e-4 months <br /> well within the Technical Specifications requirement.
DOCKET (2)
MAOER
DESCRIPTION OF OCCURRENCE (cont'd) This event was reported in accordance with the requirements of 10CFR50.72(b)(3)(ii), and it is being reported in accordance with 10CFR50.73(a)(2)(ii).
APPARENT CAUSE OF OCCURRENCE
The apparent cause of the leak was attributed to weld failure due to vibration-induced fatigue of the weld. The fatigue induced failure was most likely caused by the second natural frequency of the piping with the accelerometer weight being resonant with the five vane passing running frequency of the "A" Recirculation pump.
In 1990, following a few similar weld failures at Hope Creek, PSEG commissioned an independent contractor to review the stress levels in the recirculation piping system. The stress levels were reported to be satisfactory; however, in 1991, the recirculation system was instrumented with accelerometers.
As a result of the data collected modifications to the system were performed, although, the data collected did not indicate that the fatigue stress levels were above the endurance limit of the material.
The testing was completed, but the accelerometer(s) was not removed. The accelerometers themselves caused a resonance condition to occur in the piping, which led to the identified failure.
SAFETY SIGNIFICANCE AND IMPLICATIONS
There were no actual consequences and no impact to the health and safety of the public or plant personnel. The condition did not result in a pipe break; and there was no radioactive release.
The through wall leak was discovered while the reactor was shutdown during a drywell inspection prior to plant cool down in support of RF10. However, assuming the leak existed prior to plant shutdown and had remained undetected the overall leakage into the drywell would not have exceeded the Technical Specification integrated leakage rate limit. Any leakage of radioactivity into the drywell in either gaseous or liquid form would have been contained by the drywell systems as per design and any subsequent release of this leakage through the plant radwaste systems would have been well within Technical Specification effluent limits.
,NRC FORM 366A � U.S. NUCLEAR REGULATORY COMMISSION (6-1998) DOCKET (2) SAFETY SIGNIFICANCE AND IMPLICATIONS (cont'd) Had leakage from the undetected crack approached the unidentified leakage limit, it would have been detected by the drywell leak detection system, and the actions required by Technical Specifications would have resulted in the detection of the source and its correction.
Had the crack progressed to a complete line failure, the resultant loss of coolant (LOCA) event would have been bounded by the small break LOCA analysis in the Updated Final Safety Analysis Report (UFSAR).
PREVIOUS OCCURRENCES
A review of events over the past two years identified no reportable events due to vibration induced fatigue failure of welds.
CORRECTIVE ACTIONS
1. The accelerometers were removed during the outage.
2. Station personnel walked down all recirculation lines for any other failure indications.
3. The equipment around the area of the leak was inspected to ensure no damage from leak impingement.
4. Performed radiographic examinations (RT's) on other extrados lines and penetrant tests (PT's) on all other susceptible welds — for extent of condition. These examinations were satisfactory.
5. The cracked weld and the affected section of pipe were removed and replaced.
6. The potential to develop an ISI weld inspection plan for RF11 outage will be evaluated by the PSEG Engineering department.
7. This event will be included in the PSEG Operating Experience program for potential improvements in our procedures or processes.
COMMITMENTS
The corrective actions cited in this LER are voluntary enhancements and do not constitute commitments.