05000461/LER-2009-004
| Docket Number Sequential Revmonth Day Year Year Month Day Year None 05000Number No. | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
| Initial Reporting | |
| ENS 45390 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown |
| 4612009004R00 - NRC Website | |
PLANT OPERATING CONDITIONS
Unit: 1 Event Date: 9/30/09 Event Time: 0340 hours0.00394 days <br />0.0944 hours <br />5.621693e-4 weeks <br />1.2937e-4 months <br /> Central Daylight Time Mode: 3 (Hot Shutdown) Reactor Power: Zero Percent
DESCRIPTION OF EVENT
On 9/29/09, at 1544 hours0.0179 days <br />0.429 hours <br />0.00255 weeks <br />5.87492e-4 months <br />, with the unit in Mode 1 (Power Operation) and reactor power at 96.9 percent, operators began noting indications of a steam leak in the Drywell. Drywell pressure increased from 0.68 to 0.81 pounds per square inch gage (psig), Drywell temperature increased from 105 to 106 degrees Fahrenheit (F), the Drywell cooler differential temperature high alarm [ALM] actuated, Drywell Cooling Heating Ventilating and Air Conditioning [VB] chiller [CHU] amps increased from 50 to 56 amps, and fission product monitor [IJ] [MON] particulate and iodine indications were trending up. In response to indications of a steam leak, at 1555 hours0.018 days <br />0.432 hours <br />0.00257 weeks <br />5.916775e-4 months <br /> operators entered the abnormal coolant leakage procedure and evacuated containment.
At 1610 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.12605e-4 months <br />, the fission product monitor alarmed high with particulates at 7000 counts per minute. In response to the alarm, operators entered the abnormal release of airborne radioactivity procedure.
By 1616 hours0.0187 days <br />0.449 hours <br />0.00267 weeks <br />6.14888e-4 months <br />, Drywell floor drain leakage rate had increased from 0.25 gallons per minute (gpm) to 3.4 gpm, exceeding the limit of 2 gpm increase in the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period while in Mode 1 specified in Technical Specification (TS) 3.4.5, Reactor Coolant System (RCS) Operational Leakage. In response to the increased leakage, operators entered Required Action B.1 of TS 3.4.5 to verify within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> that the source of the unidentified leakage is not service sensitive type 304 or type 316 austenitic stainless steel.
At 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />, with the unit at 96.9% power, operators initiated an orderly plant shutdown due to the elevated Drywell leakage rate.
At 1850 hours0.0214 days <br />0.514 hours <br />0.00306 weeks <br />7.03925e-4 months <br />, the station was unable to determine that the source of the increase in unidentified leakage was not service sensitive type 304 or type 316 austenitic stainless steel; therefore operators entered Required Actions C.1 and C.2 of TS 3.4.5 which require the reactor to be in Mode 3 (Hot Shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 (Cold Shutdown) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
At 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />, Drywell floor drain leakage (unidentified leakage) was 3.3 gpm and stable.
At 1946 hours0.0225 days <br />0.541 hours <br />0.00322 weeks <br />7.40453e-4 months <br />, Clinton Power Station (CPS) completed a 4-hour notification to the NRC Operations Center in accordance with 10 CFR 50.72(b)(2)(i) via Event Notification 45390 for the initiation of a nuclear plant shutdown required by the plant's Technical Specifications.
At 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on 9/30/09, operators placed the reactor mode switch [HS] in startup & hot standby and the unit entered Mode 2 (Startup). The reactor was sub-critical at 0202 hours0.00234 days <br />0.0561 hours <br />3.339947e-4 weeks <br />7.6861e-5 months <br />.
At 0340 hours0.00394 days <br />0.0944 hours <br />5.621693e-4 weeks <br />1.2937e-4 months <br /> on 9/30/09, operators placed the reactor mode switch in the shutdown position to complete the reactor shutdown, inserting a planned manual scram in accordance with the unit shutdown procedure, and the unit entered Mode 3. The plant entered Mode 4 at 1834 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.97837e-4 months <br /> on 9/30/09.
Following plant shutdown, a walkdown of the Drywell identified steam leakage at Reactor Core Isolation NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER �NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (9-2007) LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET
� � � � REV � � � � YEAR NUMBER NO.�Clinton Power Station, Unit 1 05000461 3� OF� 4 Cooling (RCIC) [BN] steam line inboard isolation valve [ISV] 1E51-F063. The steam leakage from valve 1E51-F063 was verified to be a packing leak. No other significant steam leaks were identified during the initial or subsequent walkdowns of the drywell.
No other inoperable equipment or components directly affected this event.
Issue Report 972235 was initiated to investigate and correct this issue.
CAUSE OF EVENT
This event has two root causes. The first root cause is the valve stem of valve 1E51-F063 is off-center with the stuffing box with potential to cause packing side loading and accelerated loss of packing load. While packing the valve during this unit shutdown, maintenance technicians identified the stem was off-center with the stuffing box. Further scale measurements indicated the off-set condition was present in the same direction at the yoke-to-operator interface. If a valve stem is not centered in the stuffing box, when the valve is stroked, the stem will side load the packing set. Side loading will cause uneven stresses in the packing set and can be a contributor to accelerated packing failure. Additional detailed measurements should be obtained to confirm and correct the source for the off-set condition. In 2004, packing was installed in 1E51- F063 while the valve bonnet was in the shop before the bonnet was installed on the valve. If the source of the condition is the valve operator alignment, then the condition could not have been identified in 2004 when the packing was installed since the operator was not in place during that packing activity.
The second root cause is work instructions did not require the packing in 1E51-F063 to be torqued to the as left value from the original installation. In February 2004, the valve was modified to eliminate the leak-off line to provide more reliable valve stem packing due to the chronic valve packing leakage the valve has experienced. An additional work order was performed for this valve in February 2006 to verify packing torque based on a recommendation from another Exelon site to confirm the packing torque was within tolerance after one fuel cycle. Work instructions for the torque confirmation activity in 2006 did not include information on the specific torque value for the packing. The instructions were "do not exceed torque values listed on the Journeyman Worksheet," (that is, 32 to 39 foot-pounds). Using the plus or minus 2 foot-pounds tolerance of the torque wrench, the as-left value was 30 foot-pounds versus the as-found value of 29 foot-pounds.
SAFETY ANALYSIS
This event is reportable under the provisions of 10 CFR 50.73(a)(2)(i)(A) due to completion of the nuclear plant shutdown required by plant TS 3.4.5.
This event had minimal safety significance. The reactor was shut down safely and maintained in a safe shut down condition. The packing leak was less than the total allowed leakage of 30 gpm over the previous 24 hour period provided in the limitations of TS 3.4.5. Although the steam leak added heat load to the ventilation system and condensed water volume to the floor drain system, the leakage was within the design capabilities of the reactor coolant system inventory makeup, the drywell floor drain system, and the drywell ventilation system. There was no release of radioactive material to the environment. Operators continually monitored the leakage during this event.
This event report does not identify any safety system functional failures.
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CORRECTIVE ACTION
The packing in valve 1E51-F063 was replaced and the proper torque was applied.
The cause for valve 1E51-F063 valve stem being off-center with the stuffing box will be investigated and corrected.
Other valves were modified in February 2004 to eliminate the leak-off lines to provide more reliable valve stem packing. The packing torque applied to these additional valves will be verified and corrected as necessary.
PREVIOUS OCCURRENCES
None
COMPONENT FAILURE DATA
Manufacturer: Anchor Darling Valve Company Nomenclature: 8-inch, 600-pounds per square inch, Flex Wedge Gate Valve Manufacturer Model Number: 93-14582