05000410/LER-2001-007
| Event date: | 12-15-2001 |
|---|---|
| Report date: | 08-23-2002 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 4102001007R01 - NRC Website | |
I. Description of Event
On December 15, 2001, at 0432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br />, with Nine Mile Point Unit 2 (NMP2) at approximately 100 percent power, radiation alarms were received on drywell particulate radiation monitors, 2CMS*CAB10A-2 and 2CMS*CAB10B-2. At 0505 hours0.00584 days <br />0.14 hours <br />8.349868e-4 weeks <br />1.921525e-4 months <br />, drywell floor drain leakage rate was observed to rise from 0.22 gallons per minute (gpm) to 0.4 gpm. The leakage rate remained the same until approximately 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, when it rose to 1.67 gpm. Chemistry samples taken from the drywell floor drain tank indicated the presence of short-lived radioisotopes indicative of leakage of reactor coolant.
Limiting Condition for Operation (LCO) 3.4.5.b. of Technical Specification (TS) 3.4.5, RCS Operational LEAKAGE, restricts the maximum unidentified leakage from the reactor coolant (recirculation) system (RCS) to 5 gpm in Modes 1, 2, and 3, while LCO 3.4.5.d. restricts the maximum increase in unidentified leakage within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in Mode 1 to 2 gpm. If either of these limits is exceeded, Actions A and B of TS 3.4.5 require the leakage to be reduced to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and if this condition is not met, Action C requires the plant to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
At 1745 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.639725e-4 months <br />, it was decided to commence an orderly plant shutdown due to the adverse trend in primary containment (drywell) parameters. At 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, the control room staff commenced shift turnover while continuing to monitor drywell leakage. The control room staff was briefed relative to the pending shutdown and a strategy was developed for actions to be taken in the event that the leakage continued to rise. The strategy developed was to remove the unit from service via a controlled shutdown, and to initiate a manual scram if the leak rate reached four (4) gpm. At 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />, with the leak rate at 1.87 gpm, an orderly plant shutdown was commenced. The plant shutdown was commenced before reaching any of the leakage limits specified in TS LCO 3.4.5. At 1919 hours0.0222 days <br />0.533 hours <br />0.00317 weeks <br />7.301795e-4 months <br />, upon noticing a leakage rate of 2.27 gpm (greater than allowed by LCO 3.4.5.d), TS 3.4.5 Action B was entered. At 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br />, it was identified that drywell floor drain leak rate had risen in a step change to 5.87 gpm (greater than allowed by LCO 3.4.5.b), and, therefore, TS 3.4.5 Action A was entered. At 2046 hours0.0237 days <br />0.568 hours <br />0.00338 weeks <br />7.78503e-4 months <br /> on December 15, 2001, with reactor core thermal power at approximately 60 percent, the reactor was scrammed by placing the mode switch in shutdown. The systems associated with the scram functioned as required, with one exception:
valve 2FWS-LV55B (high-pressure low-flow) in the feedwater system, which was being utilized post-scram for reactor water level control, failed to open automatically or manually following the securing of the associated feedwater pump 2FWS-P1B.
Reactor water level was then controlled with low-pressure low-flow bypass valve 2CNM-LV137, in accordance with the applicable plant procedure. Following the scram, the drywell floor drain leakage trended downward during plant depressurization and compliance with LCO 3.4.5 leakage limits was restored. At 2053 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.811665e-4 months <br /> on December 15, 2001, TS Action A associated with the 5 gpm leakage rate limit was exited. At 2149 hours0.0249 days <br />0.597 hours <br />0.00355 weeks <br />8.176945e-4 months <br /> on that same day, TS Action B associated with the 2 gpm leakage rate increase limit was exited.
An inspection of the drywell identified packing leakage from gate valve 2RCS*MOV18A in the RCS as the cause of the drywell leakage. The immediate corrective actions consisted of repacking valve 2RCS*MOV18A to stop the leakage, checking three other similar gate valves in the RCS for leakage (no leaks found), and retorquing the packing gland nut on all four gate valves as a precautionary measure. The failure of valve 2FWS-LV55B was traced to a malfunctioning positioner and lockup assembly. This assembly was replaced and the valve tested satisfactorily.
Plant restart and reactor criticality occurred on December 17, 2001.
II. Cause of Event
The cause of this event was determined to be the failure of packing on valve 2RCS*MOV18A. The primary cause of the packing failure was that management direction was inadequate to ensure a sufficient commitment to implementation of the valve packing program. Consequently, the knowledge level of personnel implementing the program was deficient. A contributing cause was inadequate management follow-up of equipment deficiencies, in that, the need for timely repair was not identified. Opportunities to repair the valve were therefore not utilized.
III. Analysis of Event
This event is reportable in accordance with 10CFR50.73(a)(2)(iv)(A), which requires a report for any event or condition that resulted in manual actuation of the reactor protection system, including reactor scram. Additionally, this event is reportable according to 50.73(a)(2)(i)(A), which requires a report for the completion of any plant shutdown required by the plant TS.
The RCS takes suction on the reactor pressure vessel (RPV) and discharges back to the RPV. Each of the two RCS pumps, 2RCS*P1A and P1B, has a 24 inch Anchor Darling gate valve, 2RCS*MOV10A and 10B, respectively, on the suction side, and a 24 inch Anchor Darling gate valve, 2RCS*MOV18A and 18B, respectively, on the discharge side.
These valves are open during normal plant operation and their primary purpose is to isolate the RCS pump for maintenance.
During Refueling Outage Number 6 in Spring 1998, the stuffing boxes/packing glands for all four valves were modified from a three tier gland assembly to a single packing gland (simpler arrangement). These four valves were also then repacked using the Chesterton packing program. This program specified an inadequate gland stress, and this deficiency was not identified until the causal analysis for the current event was performed.
In September 2000, an inspection during a planned plant outage identified a packing leak of about 20 drops per minute from valve 2RCS*MOV18A. Subsequent inspections identified no leakage, and, therefore, this situation was accepted as is. During a planned outage in March 2001, the packing leak was noted again. The corrective action attempted was to manually back seat the valve, which did not change the leakage rate. However, subsequent inspections again indicated that the leakage had stopped. Based on this, it was concluded that the packing would not leak further as the plant heated up and resumed power operation. Thus, an opportunity was missed to generate an Action Request for valve repair during the next available outage. Later, in July 2001, an Action Request was initiated for repacking valve 2RCS*MOV18A during Refueling Outage Number 8 (RF08) in Spring 2002, and this item was added to the RFO8 work scope.
The December 15, 2001, plant shutdown was commenced before reaching any of the leakage limits specified in TS LCO 3.4.5, the systems associated with the scram performed as required with the exception of valve 2FWS-LV55B, no release of drywell leakage to the outside environment occurred, the drywell inspection did not identify any other leaking valves, and the leakage problem was corrected prior to plant restart. The failure of valve 2FWS-LV55B to operate post scram did not significantly impact the ability to control reactor water level post scram.
Nine Mile Point Nuclear Station, LLC performed a probabilistic risk analysis for the manual scram event and determined that it is not risk significant per NRC guidance.
IV. Corrective Actions
1. Valve 2RCS*MOV18A was repacked to stop the leakage.
2. Valves 2RCS*MOV10A, 2RCS*MOV10B, and 2RCS*MOV18B were checked for leakage (no leaks found).
3. The packing gland nut on all four of the above valves was retorqued as a precautionary measure.
4. The malfunctioning positioner and lockup assembly in valve 2FWS-LV55B was replaced and the valve tested satisfactorily.
5. Valves 2RCS*MOV10A, 2RCS*MOV10B, and 2RCS*MOV18B were repacked during RF08.
IV. Corrective Actions (cont'd) 6. Explicitly establish a position with responsibility for valve packing technical requirements, field implementation, and craft oversight.
7. Develop a site packing procedure that provides technical criteria for selection of packing material and field application.
8. Review previous valve repacking activities to identify applications that are inappropriate, and initiate actions to correct as appropriate.
V. Additional Information
1. Failed Components:
a) A 24-inch gate valve, 2RCS*MOV18A, manufactured by Anchor Darling, used to isolate RCS pump 2RCS*P1A for maintenance.
b) An automatic level control valve, 2FWS-LV55B, manufactured by Valtek, Inc., used to control reactor water level during low power operation.
2. Previous similar events:
a) Licensee Event Report (LER) 97-06 documents an event where NMP2 experienced a rapid rise in unidentified drywell leakage at 95 percent power. As a result, the reactor was placed in cold shutdown.
This particular problem was traced to a leaking flexible hose that served as the drain line to a flow control valve in the RCS.
b) In December 1998, a packing leak was noted on 2RCS*MOV18B. This valve was, therefore, repacked during Refueling Outage Number 7 in Spring 2000. (This event did not result in a LER.) 3. � Identification of components referred to in this LER:
Components � IEEE 805 System ID IEEE 803A Function Reactor Coolant (Recirculation) System � AD N/A Reactor Protection System JC N/A Feedwater System SJ N/A Control Rod Drive System AA N/A Radiation Monitoring System IL N/A Leak Monitoring System IJ N/A Valve AD FCV, ISV Valve SJ LCV Pump AD P Control Rod AA ROD Radiation Monitors IL MON Alarms (Radiation) IL RA Reactor Pressure Vessel AD RPV Switch JC HS Drain IJ, AD DRN