05000277/LER-2004-001
| Peach Bottom Atomic Power Station, Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| 2772004001R00 - NRC Website | |
Unit Conditions Prior to the Event Unit 2 was in Mode 1 and operating at approximately 43% power. As a result of an increase in the air in-leakage into the condenser, entry into the Operational Transient (OT) procedure for low condenser vacuum had been performed. This entry resulted in reactor power being reduced to 43%. At the time of the event, leak testing to determine potential sources of the air in-leakage was in progress. The leak testing activities did not contribute to the scram event. There were no structures, systems or components out of service that contributed to this event.
Description of the Event
Unit 2 was manually scrammed at approximately 1510 hours0.0175 days <br />0.419 hours <br />0.0025 weeks <br />5.74555e-4 months <br /> on 2/22/04 as a result of decreasing Main Condenser vacuum. The Operational Transient procedure (0T-106) for low Condenser vacuum had been entered at 1424 hours0.0165 days <br />0.396 hours <br />0.00235 weeks <br />5.41832e-4 months <br /> due to a large increase in the air in-leakage into the condenser. Main Condenser vacuum was 27.1" Hg at the time OT-106 was entered. A power reduction had been performed to approximately 43% power with a resultant Condenser vacuum of approximately 25.5" Hg. A briefing was conducted and the reactor was manually scrammed in accordance with the OT procedural direction since Condenser vacuum was not restored to above the procedurally required 26.2" Hg value (i.e. the vacuum necessary to protect the low pressure turbine last stage buckets when operating at low turbine-generator load). The automatic scram set point for low Condenser (EIIS:
COND) vacuum is 23.0" Hg. As a result of the manual scram, reactor water level decreased, as expected, to the reactor water level three set point resulting in Primary Containment Isolation System (PCIS) (EIIS: JM) Group H and HI isolations.
Residual heat in the reactor was removed via the no teal heat sink (i.e. Condenser) using bypass valves.
Condenser capability for heat removal was maintained throughout the event.
The 'C' Reactor Feed Pump discharge control valve (EHS: FCV) exhibited sluggish behavior at approximately 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br />. This resulted in slightly exceeding the specified reactor water level control band (i.e. level reached +32" versus a band of-1-10" to +30"). This condition was promptly detected and resolved by operations personnel. No high water level set points were reached. This level control instrumentation is not safety related.
Appropriate regulatory notifications were completed by 1842 hours0.0213 days <br />0.512 hours <br />0.00305 weeks <br />7.00881e-4 months <br /> on 2/22/04 to report the scram and resultant PCIS isolations.
Analysis of the Event
as a result of this event. All control rods inserted on the III PCIS isolations initiated as expected when reactor water reactor water level set point.
the size of the leak in the Reactor Feed pump turbine exhaust within the capacity of the steam jet air exhausters and therefore, set point would not have been reached. Therefore, there was no plant could have been lost. Normal heat removal systems were ' A ._ significant.
in the, 2A Reactor Feed Pump Turbine Exhaust Expansion manufactured by Tube Turns). Subsequent to the unit for. the, source of the Condenser vacuum reduction. The had developed in the expansion joint.
Reactor. Feed Pump Turbine Exhaust expansion joint, XJ- with,safis factory results. The expansion joints for the other for leakage and no additional problems were noted. Other contribute to condenser in-leakage were examined and/or of the expansion joint is being performed. Additional will bepumued.in accordance with the Corrective Action are planned for the next refueling outage.
identified involving a failure of a reactor feed pump turbine There were no actual safety consequences reactor scram signal. Group II and Group level decreased to the level-three (lo-level) Subsequent analysis has determined that expansion joint bellows (EIIS: XJ) was the automatic low Condenser vacuum concern that the normal heat sink for the used for plant shutdown.
The event was determined to not be risk
Cause of the Event
The cause of the event was due to a breach Joint bellows (46"x46", Model 83-1983-C, shutdown, inspections were
- performed inspection revealed that a small opening
Corrective Actions
Temporary repairs were. made W.. the. 2A 2170A. Post repair testing was performed Reactor Feed Pump Turbines were tested pressure boundary components that could tested for leakage. A Additional, analysis underlying causes and corrective aetions program. Long term repairs to the. expansion; Previous Similar Occurrences There were no previous similar LERs exhaust joint.