05000354/LER-2005-003
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No. | |
| Event date: | 06-07-2005 |
|---|---|
| Report date: | 09-29-2005 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3542005003R01 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric — Boiling Water Reactor (BWR/4) Reactor Coolant System {AB}* Residual Heat Removal {BO}* *Energy Industry Identification System {EllS} codes and component function identifier codes appear as (SS/CCC)
IDENTIFICATION OF OCCURRENCE
Event Date: June 7, 2005 Discovery Date: June 7, 2005
CONDITIONS PRIOR TO OCCURRENCE
Hope Creek was in operational condition 1 at 100% power prior to the event. The B residual heat removal (RHR) pump was out of service for maintenance at the start of the event. There was no other equipment out of service that impacted this event.
DESCRIPTION OF OCCURRENCE
On June 7, 2005, at 1351 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.140555e-4 months <br />, the operating crew entered the abnormal operating procedure for drywell leakage. At 1411 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.368855e-4 months <br /> with drywell pressure at 0.18 psig, the drywell leak detection system alarmed. At 1412 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.37266e-4 months <br /> with drywell pressure at 0.27 psig and continuing to rise, reactor power was reduced to 80% at the direction of the control room supervisor. At 1413 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.376465e-4 months <br /> with drywell pressure at 0.40 psig, the reactor recirculation pumps were reduced to minimum speed and the reactor mode switch was locked in shutdown to scram the reactor. Technical Specification (TS) 3.4.3.2 was entered for unidentified leakage greater than 5 gpm. TS 3.4.3.2 requires the reduction of leakage to within limits in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or to be in at least hot shutdown within the next 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 following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
At 1421 hours0.0164 days <br />0.395 hours <br />0.00235 weeks <br />5.406905e-4 months <br /> on June 7, 2005 with drywell pressure at 0.47 psig, the operating crew commenced the lowering of reactor pressure to 600 psig. At 1426 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.42593e-4 months <br />, drywell pressure was at 0.46 psig and drywell leakage was determined to be greater than 10 gpm. At 1427 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.429735e-4 months <br /> with reactor vessel pressure control being provided by the electro-hydraulic control (EHC) system, reactor water level 8 was reached. At 1429 hours0.0165 days <br />0.397 hours <br />0.00236 weeks <br />5.437345e-4 months <br />, drywell floor drain leakage was determined to be approximately 10.5 gpm with drywell pressure at 0.43 psig. At 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br />, reactor vessel water level 3 scram level was reached due to the reactor vessel pressure reduction. Reactor vessel pressure was at 600 psig. The operating crew further reduced pressure to 550 psig. During the pressure reduction, reactor vessel water level reached level 8.
At 1434 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.45637e-4 months <br /> with reactor vessel pressure at 550 psig, drywell floor drain leakage was approximately 11.3 gpm and drywell pressure was 0.40 psig. At 1437 hours0.0166 days <br />0.399 hours <br />0.00238 weeks <br />5.467785e-4 months <br />, an Unusual Event (UE) was declared due to drywell floor drain leakage being greater the 10 gpm. At 1438 hours0.0166 days <br />0.399 hours <br />0.00238 weeks <br />5.47159e-4 months <br />, the operating crew commenced the cool down of the reactor in accordance with procedures. Reactor vessel level was being maintained at +35°. At 1442 hours0.0167 days <br />0.401 hours <br />0.00238 weeks <br />5.48681e-4 months <br />, drywell floor drain leakage was approximately 14.6 gpm with drywell pressure at 0.37 psig. At 1458 hours0.0169 days <br />0.405 hours <br />0.00241 weeks <br />5.54769e-4 months <br />, drywell floor drain leakage was approximately 11.2 gpm with reactor pressure at 500 psig. At 1522 hours0.0176 days <br />0.423 hours <br />0.00252 weeks <br />5.79121e-4 months <br />, drywell floor drain leakage was steady at 9.9 gpm. At 1618 hours0.0187 days <br />0.449 hours <br />0.00268 weeks <br />6.15649e-4 months <br />, reactor pressure was reduced to 300 psig. At 1651 hours0.0191 days <br />0.459 hours <br />0.00273 weeks <br />6.282055e-4 months <br />, the operating crew reset the scram signal. The position indication for the F050A residual heat removal (RHR) (BON) check valve was lost at 1802 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.85661e-4 months <br />.
At 1823 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.936515e-4 months <br /> the B RHR pump was restored from maintenance and retested satisfactorily. At 2344 hours0.0271 days <br />0.651 hours <br />0.00388 weeks <br />8.91892e-4 months <br /> with reactor pressure less than 82 psig, the A RHR train was declared inoperable for shutdown cooling due to questions with the failure of the valve position indication. Preparations were made for drywell entry and placing B RHR in shutdown cooling.
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1-2001) FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) DESCRIPTION OF OCCURRENCE (cont'd) At 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> on June 8, 2005, a report from the drywell stated that the leakage was coming from the F050A RHR check valve position indicator. At 0335 hours0.00388 days <br />0.0931 hours <br />5.539021e-4 weeks <br />1.274675e-4 months <br /> B RHR was placed in service for shutdown cooling. The plant entered Operational Condition 4 (cold shutdown) at 0456 hours0.00528 days <br />0.127 hours <br />7.539683e-4 weeks <br />1.73508e-4 months <br />. The UE was terminated at 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br /> on June 8, 2005.
This event is being reported in accordance with 10CFR50.73(a)(2)(i)(A), "the completion of any nuclear plant shutdown required by the plant's Technical Specifications," 10CFR50.73(a)(2)(iv)(A), "any event or condition that resulted in manual or automatic actuation of ...reactor protection system (RPS) including: reactor scram or reactor trip," and 10CFR50.73(a)(2)(ii)(A), "any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
CAUSE OF OCCURRENCE
The cause of the reactor coolant system leak was the result of an approximately 280 degree circumferential crack in the position indicating tube for the F050A RHR check valve. The through wall leak was caused by the vibration of the attraction sleeve (which is on the end of actuator rod) in the presence of the switch magnetic force resulting in the attraction sleeve fretting/wearing through the position indicating tube. The attraction sleeve is a harder material than the indicating tube. The design of the actuator rod allows the attraction sleeve to move relative to the actuator rod. The design includes a magnetic switch that holds the attraction sleeve against the wall of the indicating tube.
As a result, vibration can cause the attraction sleeve to wear into the actuator rod pin, thereby allowing it to wear through the indicating tube.
PREVIOUS OCCURRENCES
A review of LERs for Hope Creek for the past two years identified a reportable occurrence for a through wall reactor coolant system leak. LER 354/2005-002-00 reported a through wall leak of the 'B' reactor recirculation system decontamination port. The through-wall crack in the decontamination connection was determined to be fatigue initiation and propagation as caused by the following: presence of an original subsurface weld defect and presence of a decontamination port geometry that experienced damaging high cycle vibration during certain recirculation system operating conditions. The corrective actions associated with LER 354/2005-002-00 were the modification of the 'A' and 'B' reactor recirculation system decontamination port geometry to change the natural frequency of the decontamination port and a review of the connections to the reactor recirculation loops to determine if other similar connections were susceptible to the failure experienced with the decontamination port. These corrective actions would not have prevented the current event.
In addition LERs 354/2003-007 and 354/2004-006 identified reportable occurrences associated with component/system design. LER 354/2003-007 identified that the reactor feed pump oil system 'keep fill' line was undersized contributing to the trip of the reactor feed pump and low reactor water level scram. LER 354/2004-006 identified that the high pressure coolant injection (HPCI) was unable to deliver the required flow at the required pressure for design basis accidents. Original design documents had recommended that the 2 HPCI injection line orifices be rebored to a larger size but this recommendation had not been acted upon. The corrective actions associated with these events consisted of plant modifications and design document reviews and would not have prevented the current event.
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On June 7, 2005, drywell floor drain leakage increased to greater than 10 gpm. As a result an Unusual Event (UE) was declared. The operating crew took appropriate actions to bring the plant to a controlled cold shutdown condition.
Reactor water level was maintained above level 2 (initiation signal for emergency core cooling injection) and reactor pressure control was maintained throughout the event. Even if a complete severance of the F050A RHR check valve position indicating tube were to occur, this failure would not result in an uncontrolled reactor depressurization. Based on the above, 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. Hope Creek was brought to a cold shutdown condition following the identification of the reactor coolant system leak from the position indicating tube of the F050A RHR check valve.
CORRECTIVE ACTION
1. The F050A and F050B RHR check valves were modified to remove the position indicator tubes. These modifications were completed prior to the plant startup.
2. Six other check valves contain the same position indicator tube as the F050A and F050B RHR check valves.
The position indicating tubes for these six check valves underwent ultrasonic testing with no indications of wear as experienced on the F050A RHR check valve. These inspections were completed prior to the plant startup.
3. An engineering review was performed for components and sub-components communicating with the pressure boundary of the loop 'A' and 'B' reactor recirculation piping or connected piping systems that were susceptible to system vibration and fretting/wear vulnerabilities. The identified components were assessed as adequate with respect to the pressure integrity function.
COMMITMENTS
This LER contains no commitments.