05000293/LER-2007-001
| Docket Number | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2932007001R00 - NRC Website | |
FACIUTY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
BACKGROUND
The Reactor Water Cleanup System (RWCU) functions to purify the reactor coolant water by continuously removing a portion of the reactor recirculation flow from the suction side of a recirculation pump, sending the removed flow through filter demineralizer units to undergo mechanical filtration and ion exchange processes, and returning the processed fluid back to the reactor via the feedwater line. The major equipment of the system consists of the two cleanup recirculation pumps, regenerative and non regenerative heat exchangers, and two filter demineralizer units with supporting equipment. Water from the reactor first enters the regenerative heat exchanger, then the non-regenerative heat exchanger, and finally the pumps. The water cooled by the heat exchanger then flows through two parallel filter demineralizer units for the removal of impurities. Normal routing of flow from the filter demineralizer units is then through the shell side of the regenerative heat exchanger where it is heated by the incoming (untreated) coolant, to the temperature range of reactor feedwater. It is then returned to the reactor through the Feedwater System.
Valve MO-1201-85 is a valve in the RWCU inlet line that draws suction from the A loop of the reactor recirculation system. MO-1201-85 is not a containment isolation valve or a reactor pressure vessel isolation valve.
During the operating cycle, a slow increase in equipment and floor sump inleakage had been observed.
Pilgrim entered these observations into the Corrective Action Program and created a comprehensive action plan under the Operational Decision Making Issue (ODMI) process to ensure proper focus and resources were applied to determining the likely sources of unidentified RCS leakage, monitoring and trending it's status, and developing various thresholds and trigger points for specific response actions.
EVENT DESCRIPTION
At midnight on 3/17/07, Operations personnel observed an increase in the Drywell floor sump inleakage rate. The calculated floor sump inleakage rate had climbed from 0.58 gpm to 0.65 gpm. Over the next several hours the rate of unidentified RCS leakage continued to increase. At 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br /> on 3/17/07, a decision was made to commence an orderly shutdown of the unit. Shutdown commenced at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />.
During the shutdown, RCS unidentified leakage continued to increase and at 1655 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.297275e-4 months <br /> a manual scram was initiated in accordance with PNPS procedures because administrative limits were reached. Prior to the manual reactor shutdown, leakage did not reach the applicable Technical Specification limits presented in TS 3.6.C.1. At 1657 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.304885e-4 months <br />, all control rods were verified to be fully inserted.
As expected, the scram caused a decrease in the void fraction in the reactor vessel water, which in turn caused the indicated water level in the reactor vessel to decrease. Indicated water level decreased to about -12 inches (narrow range level). The decrease in reactor water level to less than +12 inches resulted in the expected automatic actuation of the Primary Containment Isolation Control System (PCIS) and Reactor Building Isolation Control System (RBIS).
The PCIS actuation resulted in the automatic closing of the Primary Containment system (PCS) Group 2 and Group 6 (RWCU) isolation valves that were open.
The RBIS actuation resulted in the automatic start of the Standby Gas Treatment system (SGTS) trains 'A' and� and automatic closing of the Reactor Building/Secondary Containment system (SCS) trains 'A' and 'B' supply and exhaust ventilation dampers.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) At about 1733 hours0.0201 days <br />0.481 hours <br />0.00287 weeks <br />6.594065e-4 months <br />, the Reactor Protection System (RPS) was reset. While all rods remained fully inserted, the SDIV Hi Level bypass switch was mispositioned due to a human performance error at about 1735 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.601675e-4 months <br /> generating an RPS signal. RPS was reset at about 1735 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.601675e-4 months <br />. The RBIS was reset, the SGTS was returned to normal standby service, and the reactor building ventilation system was returned to service.
During a drywell entry at 1235 hours0.0143 days <br />0.343 hours <br />0.00204 weeks <br />4.699175e-4 months <br /> on 3/18/07, it was determined that the source of the increased floor sump inleakage was the packing on MO-1201-85. Valve MO-1201-85 is a valve in the RWCU inlet line that draws suction from the A loop of the reactor recirculation system. MO-1201-85 is not a containment isolation valve or a reactor pressure vessel isolation valve.
CAUSE
The root cause of the increase in unidentified leakage was the failure of the valve packing for Reactor Water Clean-up (RWCU) valve MO-1201-85, which is located inside primary containment in the drywell and is not accessible during power operation. The failure was due to inadequate preventive maintenance of the valve packing.
CORRECTIVE ACTION
Corrective actions taken included the following:
- On 3/18/07, it was determined that the source of the increased floor sump leakage was leakage through the packing on MO-1201-85.
- On 3/19/07, the valve was re-packed, post-work tested and returned to service.
- Individuals were coached on the use of human performance tools to address the error that occurred while resetting the RPS signal.
Corrective actions planned include the following:
- Developing a packing adjustment procedure and scheduling preventive maintenance for packing adjustments to critical valves in the drywell and other inaccessible areas using guidelines published by EPRI.
These planned corrective actions may be supplemented or modified in accordance with the corrective action process.
_ _ .FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
SAFETY CONSEQUENCES
The decrease in the reactor vessel water level was expected in response to the scram due to shrinkage in the water level caused by steam void collapse. The consequent actuations were the expected designed responses to a low reactor vessel water level condition, i.e. less than about +12 inches (narrow range).
During the event, the lowest reactor vessel water level that occurred was approximately -12 inches (narrow range). This level is significantly greater than the core standby cooling systems set point (approximately - 46 inches), and the level corresponding to the top of the active fuel zone (approximately -127 inches).
All safety systems performed as designed. This event did not result in a challenge to fuel limits or release of radioactive material above the expected normal operating level. The packing leak on the MO-1202-85 valve was not an RCS pressure boundary leak. MO-1201-85 is not a containment isolation valve or a reactor pressure vessel isolation valve. The packing leak had no adverse effect on any safety function.
Technical Specification limits for RCS leakage were not exceeded.
The event posed no threat to public health and safety.
REPORTABILITY
This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv)(A) as a condition that generated containment isolation signals affecting containment isolation valves in more than one system.
SIMILARITY TO PREVIOUS EVENTS
A review was conducted of Pilgrim Station LERs submitted since 2003. The review focused on similarities involving an automatic or manual scram(s) in response to increased drywell leakage. The review identified no similar events.
PILGRIM NUCLEAR POWER STATION
FACILITY NAME (1) 05000-293 DOCKET NUMBER (2) LER NUMBER (6) � ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES The EIIS codes for this report are as follows:
COMPONENTS
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CODES
Valve (MO-1201-85) � V
SYSTEMS
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CODES
Containment isolation control system (PCIS/RBIS)� JM Engineered safety features actuation system (RPS, PCIS, RBIS)� JE Plant protection system (RPS)� JC Reactor building (SCS)� NG Reactor water cleanup system (RWCU)� CE Standby gas treatment system (SGTS)� BH