05000339/LER-2001-001
| Docket Number | |
| Event date: | 01-19-2001 |
|---|---|
| Report date: | 02-16-2001 |
| 3392001001R00 - NRC Website | |
1. DESCRIPTION OF EVENT
On January 19, 2001, Unit 2 was in Mode 1 at 99 percent power (TAVG power coastdown).
At approximately 0615 hours0.00712 days <br />0.171 hours <br />0.00102 weeks <br />2.340075e-4 months <br />, a Reactor Coolant System (RCS) leak rate test was performed.
The results indicated an increase in identified RCS (System — AB) leakage to .8038 gallons per minute (gpm). The abnormal procedure for increased primary plant leakage was entered.
By 0840 hours0.00972 days <br />0.233 hours <br />0.00139 weeks <br />3.1962e-4 months <br /> identified leakage had increased to 2.6315 gpm. At 0944 hours0.0109 days <br />0.262 hours <br />0.00156 weeks <br />3.59192e-4 months <br />, a containment entry was made to determine the source of the leakage. At the time of the entry identified RCS leakage had increased to 8.0154 gpm. The containment entry team determined that the stuffing box on the C reactor coolant loop bypass valve (2-RC-MOV- 2587) (System — AB, Component — ISV) was the source of the leak as indicated by increased temperatures on the packing leak off line. By 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />, identified leakage was greater than the Technical Specification (TS) limit of 10 gpm, actual leakage was 10.0015 gpm, and applicable actions of TS 3.4.6.2 were initiated.
At approximately 1145 hours0.0133 days <br />0.318 hours <br />0.00189 weeks <br />4.356725e-4 months <br /> on January 19, 2001, operators began to ramp Unit 2 offline. At this time a Notification of Unusual Event (NOUE) was declared in accordance with the EPIP 1.01, Emergency Manager Controlling Procedure, Tab B8 Unit Shutdown Required by TS, for exceeding the 10 gpm TS limit for RCS leakage. The actual leakage recorded at the time was 10.0015 gpm. At 1155 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.394775e-4 months <br /> notification to state and local governments was completed.
At 1215, hours a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report was made to the NRC Operations Center in accordance with 10 CFR 50.72(a)(1)(i).
On January 19, 2001, at 1735 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.601675e-4 months <br />, Mode 2 was entered and subsequently at 1747 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.647335e-4 months <br /> Mode 3 was entered with the reactor shutdown. The TS action to be in hot standby (i.e., Mode 3) within six hours was exited. The TS action to be in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> was still applicable. A containment entry was made at 2113 hours0.0245 days <br />0.587 hours <br />0.00349 weeks <br />8.039965e-4 months <br /> to backseat the C reactor coolant loop bypass valve to stop the leakage. Once the valve was positioned on its backseat, at approximately 2145 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.161725e-4 months <br />, leakage stopped and a leak rate test was performed to confirm the identified leakage was in fact from the C reactor coolant loop bypass valve. The NOUE was terminated at 2323 hours0.0269 days <br />0.645 hours <br />0.00384 weeks <br />8.839015e-4 months <br /> on January 19, 2001, when RCS leakage decreased to less than 10 gpm. Identified leakage was measured at .0839 gpm. The TS action to be in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> was exited.
2. SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
No significant safety consequences resulted from the C reactor coolant loop bypass valve leaking past the valve stem packing material since the leakage was contained and drained to the primary drain transfer tank (PDTT) inside the containment building. The valve stem stuffing box has packing material above and below a lantern ring with the lantern ring position at the drain line to the PDTT (System — WD, Component — TK). The packing material above the lantern ring prevented leakage to the containment (System — NH) atmosphere. As such, there was no release of radioactive material. The health and safety of the public was not affected at any time during this event.
3. CAUSE
The increase in identified RCS leakage was the result of the C reactor coolant loop bypass valve leaking past the valve stem packing material. Each reactor coolant loop has an eight inch bypass line equipped with an isolation valve (e.g., C loop bypass valve 2-RC-MOV- 2587) which is closed during normal loop operations. Each loop bypass valve is equipped with a leak off line to divert any valve stem leakage to the primary drain transfer tank. The cause of the stem packing material failure below the lantern ring is attributed to aging. The packing material for the C reactor coolant loop bypass valve was last replaced in 1982.
4. IMMEDIATE CORRECTIVE ACTIONS
A NOUE was declared in accordance with Emergency Plan Implementing Procedure 1.01, Emergency Manager Controlling Procedure, Tab B8, RCS leakrate requiring plant shutdown per TS 3.4.6.2. State and federal notifications were made. Unit 2 was placed in hot standby (Mode 3) in accordance with TS. Containment entry confirmed the C reactor coolant loop bypass valve as the source of the leakage. The C loop bypass valve was placed on its backseat and the leakage was isolated. All applicable TS actions were entered and exited as required.
5. ADDITIONAL CORRECTIVE ACTIONS
The C reactor coolant loop bypass valve stem packing material was replaced. Subsequent leak rate testing verified leakage to be 0.0839 gpm. Both the A and B reactor coolant loop bypass valves were inspected to ensure there was no active leakage. As a precautionary measure, the packing material for the A and B reactor coolant loop bypass valves is expected to be replaced during the next scheduled refueling outage beginning in March 2001.
The following conditions were experienced during Unit 2 shutdown and subsequent startup.
During source range channel functional testing the pre-amp test circuit response for Source Range Nuclear Instrumentation Detector, N-31, was degraded. The TS action was entered to restore the channel to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open reactor trip breakers within the next hour. Following repairs and testing, N-31 was declared operable and returned to service.
During start-up, prior to entering Mode 2, functional testing of the auto stop oil pressure switch was performed. While making adjustments to the pressure switch, the Solid State Protection System (SSPS) train B Input Bay "1" 120 VAC supply was lost due to a blown fuse. As a result of the blown fuse, the SSPS inputs to the Loop Stop Valves, auto stop oil, reactor coolant pump bus undervoltage, reactor coolant pump bus under frequency, and 5. ADDITIONAL CORRECTIVE ACTIONS (continued) reactor coolant pump breaker auxiliary contact were affected creating a train disagreement.
Due to the station being in hot shutdown (mode 3) these signals were blocked by P-7 and P- 8 permissives. The fuse was replaced and the testing was completed satisfactory.
At 92 percent power while ramping the unit online the individual rod position indication for Rod B8 - D bank was not moving commensurate with the other indicators for the D bank as the rods were moved in the outward directions. An incore trace confirmed that Rod B8 was in fact moving and the problem was with indication only. A signal conditioning card was replaced and Rod B8 indication returned to normal.
6. ACTIONS TO PREVENT RECURRENCE
An engineering evaluation is currently in progress. Corrective actions identified by the evaluation will be implemented as required. Historically, leakage past the valve stem packing material for the A, B, and C reactor coolant loop bypass valves is rare. Work history associated with these valves indicate they were last repacked as follows: 2-RC-MOV-2585 in 1986, 2-RC-MOV-2586 in 1982, and 2-RC-MOV-2587 in 1982.
7. SIMILAR EVENTS
Unit 1 LER, N1-1991-011-00, documents a Unit 1 shutdown due to a reactor coolant system leak from a three quarter inch upper disc pressurization line for the B cold leg loop stop valve. Report date May 11, 1991.
Unit 2 LER, N2-1991-011-00, documents a Unit 2 shutdown due to a reactor coolant system leak from the RHR inlet isolation valve stem packing material. Report date November 3, 1991.
8. ADDITIONAL INFORMATION
Unit 1 was operating in Mode 1, at 100 percent power, and was not affected by this event.
C reactor coolant loop bypass valve component information:
Mark Number 2-RC-MOV-2587 Manufacturer Rockwell — Edwards Model Number 7517(CF8M)JMY Description � 8 inch motor operated isolation valve