05000293/LER-2003-003
| Event date: | 06-01-2003 |
|---|---|
| Report date: | 07-29-2003 |
| 2932003003R00 - NRC Website | |
FACILITY NAME (
PILGRIM NUCLEAR POWER STATION
DOCKET NUMBER (2) 1:6000-293 LER NUMBER (6)
BACKGROUND
The preferred offsite power source for Pilgrim Station is the 345 kV transmission lines 342 and/or 355 via the Startup Transformer. The secondary offsite power source Is the 23 kV distribution system via the Shutdown Transformer.
During power operation when the Main Transformer Is energized by the Main Generator, the Pilgrim Station 4.18 kV auxiliary power distribution system (APDS) is normally powered from the Unit Auxiliary Transformer.
The APDS provides power to nonsafety-related and safety related loads. The nonsafety-related electrical loads are powered from 4,16 kV Buses Al through A4 and the respective 480 volt load centers and motor control centers. The safety-related electrical loads are powered from 4.16 kV Buses A5 and/or A6 and the respective 480 volt load centers and motor control centers.
During plant shut down conditions when the 345 kV transmission lines are energized, the APDS is powered from the 345 kV transmission system by the Startup Transformer via the respective 345 kV switchyard circuit breakers. During plant shut down conditions when the 345 kV lines are not energized, the safety-related 4.16 kV Buses (A5/A6), and respective load centers and motor control centers, are designed to be powered by the respective emergency diesel generator (EDG 'AI/EDG LB') or the 23 kV distribution system via the Shutdown Transformer (Buses A5 and/or A6) or the Station Blackout Diesel Generator (Bus AS or Bus A6).
Just prior to the event the following conditions existed. The preferred and secondary offsite power sources were energized, the switchyard ringbus was intact, the EDGs and the Station Blackout Diesel Generator were in standby service. The Main Transformer was energized by the Main Generator with the APDS being powered from the Main Transformer via the Unit Auxiliary Transformer (UAT). The reactor was operating at 100% power (1998 MWt) with the reactor mode selector switch in the RUN position. The reactor vessel pressure was normal, about 1035 psig, with the reactor water at the saturation temperature for that pressure. The reactor water level was normal, at about +26" (narrow range).
EVENT DESCRIPTION
On June 1, 2003 at about 0850 hours0.00984 days <br />0.236 hours <br />0.00141 weeks <br />3.23425e-4 months <br />, an unplanned automatic reactor protection system scram signal and scram occurred while at 100% reactor power (1998 MWt). The scram signal resulted in the automatic Insertion of the control rods that were in a withdrawn position at the time of the event. The reactor core display panel and the plant computer (EPIC) call-rods function indicated all control rods were fully inserted.
The scram signal was initiated by the closing of the turbine steam control valves. The closing of the turbine steam stop valves also resulted in an automatic scram signal FACILITY NAME (1) DOCKET NUMBER (2) - � LER NUMBER (6) PAGE (3) PILGRIM NUCLEAR POWER STATION 1:6000-293 2003 � 003 � 00 3 oaf 6 The event was initiated by a trip of the unit auxiliary transformer (UAT) differential relay that resulted In a trip of the main generator lockout relay. The trip of the lockout relay Included the automatic transfer of the source of 4.16 kV power for the APDS from the UAT to the Startup Transformer, automatic opening of the 345 kV switchyard circuit breakers that connect the main transformer to the switchyard, automatic opening of the main generator field breaker, and automatic trip of the master trip solenoid (MTS-1) that Is part of the turbine mechanical-hydraulic control system.
The trip of MTS-1 Included the automatic closing of the turbine steam control valves and stop valves.
Collectively, the three turbine steam bypass valves have a capacity for diverting 25% of the rated main steam flow to the main condenser. The steam flow from the reactor vessel exceeded the capacity of the bypass valves and the consequence was the expected, relatively rapid increase in the reactor vesseVmain steam pressure.
The reactor vessel pressure increase and insertion of the control rods each contributed to a decrease In the reactor water void fraction (shrink). The decrease In the void fraction resulted in a decrease in the reactor water level. The reactor water level decreased to about —10" (narrow range). The decrease in reactor water level, to less than the low water level setting of about +12" (narrow range), resulted in the expected automatic actuation of the Primary Containment Isolation Control System UPC'S Groups 2 (Sampling System) and 6 (Reactor Water Cleanup System)] and Reactor Building Isolation Control System (RBIS).
Meanwhile, the pressure increase resulting from the closing of the turbine steam valves (control valves and stop valves) resulted in the automatic opening of three of the four main steam relief valves for the pressure relief function. Post trip review verified each relief valve that actuated demonstrated a normal opening and reseat within the valves' specification and that the response of all four relief valves was in accordance with design.
After the main steam relief valves closed, the turbine steam bypass valves provided the reactorlmain steam pressure control function.
The operation of the bypass valves resulted In a gradual decrease in the reactor vessel pressure and consequent increase In the reactor water void fraction (swell). The maximum reactor water level that occurred was about +52". The high pressure coolant Injection (HPCI) system turbine control system received a trip signal as a result of the high water level condition. The HPCI high water level trip function is nonsafety-related, functions to protect the HPCI turbine from water, and can be manually reset when the reactor water level is less than the trip setting or is automatically reset if a reactor water low-low level condition occurs.
The high water level condition also resulted in an automatic close signal to the normally closed reactor core isolation cooling (RCIC) system turbine steam supply valve. The high water level dosing function for the valve Is nonsafety-related, functions to protect the RCIC turbine from water. The valve can be manually opened when the reactor water level is less than the high water level setting or is automatically opened if a reactor water low-low level condition occurs. The RCIC turbine steam supply valve remained closed as designed.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Control room licensed operator response to the event included the following. The full Insertion of the control rods was verified. The PCIS and RBIS actuations were verified. A reactor water level band of 20" to 40" was established by 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />.
The HPCI turbine high water level trip was manually reset at 1040 hours0.012 days <br />0.289 hours <br />0.00172 weeks <br />3.9572e-4 months <br />.
The NRC Operations Center was notified of the event In accordance with 10 CFR 50.72 at 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on June 1 2003.
CAUSE
The direct cause of the scram was the automatic closing of the turbine control valves. The closing of the turbine control valves (and stop valves) was the result of the automatic actuation of the turbine master trip solenoid (MTS-1). The trip of MTS-1 was the result of the automatic trip of the generator lockout relay. The trip of the lockout relay was the result of the automatic trip of the UAT differential relay. The trip of the differential relay was the result of an electrical fault In the UAT.
The direct cause of the event was the failure of the conductor that connects to the phase 'A' of the UAT (X-winding). The failure was internal to the transformer. The UAT phase 'A' winding powers the phase 'A' portion of Pilgrim's 4.16 kV APDS. The UAT was installed during original plant construction (c. 1970).
The UAT Is a three phase, oil-cooled, 60 Hz, type SL power transformer manufactured by the Westinghouse Electric Corporation. The nameplate data is as follows: Serial Number RCR-23161, Class OA/FA/FDA, insuldur insulation, L. Spec. 674190, (primary) 23000 volts Delta winding (H winding).
(secondary) 4160/2400 volts Wye winding (X-winding, Y-winding), 5-position load tap changer. Tap number 4 was the working tap position at the time of the fault.
CORRECTIVE ACTION
Corrective actions taken Included the following:
- The Pilgrim Station electrical system was aligned such that the APDS is powered from the Startup Transformer during power operation, pending repair or replacement of the Unit Auxiliary Transformer.
- The Unit Auxiliary Transformer was disconnected and shipped offsite for evaluation and repair or replacement.
Corrective actions planned include the following:
- Repair or replacement of the Unit Auxiliary Transformer.
- Review of a vendor evaluation to establish additional actions necessary for early detection to eliminate the potential of a future failure of the UAT.
NRC Foirn-166A FACILITY NAME (I) _ DOCKET NUMBER (2) LER NUMBER (6) _ PAGE (31 I REVISION PILGRIM NUCLEAR POWER STATION 05000-293 2003
SAFETY CONSEQUENCES
The event posed no threat to public health and safety.
The maximum reactor power experienced during the event was about 101% and decayed as expected.
The plant systems responded as designed to the transient. The preferred and secondary sources of offsite power remained energized. The EDGs and Station Blackout Diesel Generator remained available during the event. The Core Standby Cooling Systems (HPCI System, Automatic Depressurization System, Residual Heat Removal System, Core Spray System) and the RCIC System remained available during the event to provide makeup water or core cooling If necessary.
The maximum reactor pressure that occurred was less than the Technical Specification limit Including tolerance for the relief valves and safety valves. The response of the main steam relief valves was in accordance with design. The opening or one or more relief valves is an expected occurrence resulting from a load rejection at 100% power.
The minimum reactor water level that occurred was about —10' (narrow range). The level was above the low-low water level (about —461 for automatic actuation of the Core Standby Cooling Systems and automatic actuation of the Group I portion of the PCIS. The level was also above the level (about —127") corresponding to the top of the active fuel zone.
REPORTABILITY
This report was submitted in accordance with 10 CFR 50.73(a)2)(iv) because the actuation of the RPS was not planned.
SIMILARITY TO PREVIOUS EVENTS
A review was conducted of Pilgrim Station Licensee Event Reports (LERs) submitted since 1984. The review identified previous scrams resulting from load rejections but identified no previous events that were caused by the Unit Auxiliary Transformer. The review Identified two similar LERs. LER 89-010-00 reported a failure (fault) of one of the feeder cables that connect the Startup Transformer to the APDS; the failure occurred white shut down when the SUT was powering the APDS and was caused by cable jacket damage during original construction installation (c. 1970). LER 97-004-01 reported a failure (fault) of the Main Transformer that occurred while the transformer was powering the APDS, in a backfeed configuration, while shut down; the failure was caused by a degraded transformer winding condition or static electrification.
FACILITY NAME (1)
PILGRIM NUCLEAR POWER STATION
DOCKET NUMBER (2) 05000-296 LER NUMBER (6) ENERGY INDUSTRY IDENTIFICATION SYSTEM (El IS) CODES The BS codes for the report are as follows:
COMPONENTS � CODES Circuit Breaker, AC � 52 Relay, Differential, Protective � 87 Relay, Locking-Out � 86 Solenoid (MTS-1) � _ � SOL Transformer (UAT) � XMFR
- Valve (Control, Stop, Bypass) � V Valve, Relief � (RV-203-3A/BIC/D) � RV
SYSTEMS
Engineered Safety Features Actuation � EA (RPS, PCIS, RB1S) High Pressure Coolant Injection � BJ Main Generator Output � EL Main Steam � SB Medium Voltage Power � EA Switchyard � FK Turbine Steam Bypass Control � J1