05000296/LER-2010-001
| Docket Numbermont Sequential Revday Year Year Month Day Yearh Number No. N/A N/A | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2962010001R00 - NRC Website | |
I. PLANT CONDITION(S)
At the time of discovery, Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3 were at approximately 100 percent power (3458 MVVT) and unaffected by the event.
II. DESCRIPTION OF EVENT
A. Event:
On May 16, 2008 during Unit 3 Cycle 13 refueling outage, SRV pilot cartridges were installed with setpoints in accordance with TS.
On February 27, 2010, operations personnel entered a planned Manual Scram in accordance with plant procedures to end Unit 3 Cycle 14 operation and begin the Unit 3 Cycle 14 refueling outage.
On April 20, 2010, the Tennessee Valley Authority (TVA) determined that 8 of the 13 Safety Relief Valves (SRVs) [SB], removed from Unit 3 following Cycle 14 operation, mechanically actuated at pressures greater than 3 percent above their Technical Specification (TS) setpoints, and thus were inoperable for an unknown time frame during Cycle 14 operation. Unit 3 TS Limiting Condition for Operation (LCO) 3.4.3 requires twelve (12) SRVs to be operable in reactor Modes 1, 2, and 3. With one or more required SRVs inoperable (i.e., less than 12 SRVs are operable), TS 3.4.3 Action A requires the unit to be placed 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 in 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 />. Since 8 of the 13 SRVs were found during test to have actuated above their TS setpoints, it is concluded that Unit 3 operated with inoperable SRVs longer than allowed by TS 3.4.3 Action A.
TVA is submitting this report in accordance with 10 CFR 50.73(a)(2)(i)(B), as any operation or condition prohibited by the plant's TS.
B. Inoperable Structures. Components. or Systems that Contributed to the Event:
None
C. Dates and Approximate Times of Maior Occurrences:
May 16, 2008, at 0551 hours0.00638 days <br />0.153 hours <br />9.11045e-4 weeks <br />2.096555e-4 months <br /> � During the Unit 3 Cycle 13 refueling outage, SRV pilot cartridges, with setpoints in accordance with TS, were installed. The date and time provided is for satisfactory completion of testing (verification of setpoints in accordance with TS) for Unit 3 Cycle 14 operation.
February 27, 2010, at 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> �Operations personnel entered a planned Manual Scram in accordance with plant procedures to end Cycle 14 refueling outage.
April 20, 2010, at 1434 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.45637e-4 months <br /> � TVA determines that the as-found lift setpoints in 8 SRVs exceeded the allowable TS values during the Unit 3 Cycle 14 operating cycle.
D. Other Systems or Secondary Functions Affected
None
E. Method of Discovery
The out-of-tolerance lift setpoints were identified during the performance of BFN Surveillance Procedure 0-SR-3.4.3.1.b, "Bench Test Relief Valves As-Found," at Wyle Laboratories located in Huntsville, Alabama.
F. Operator Actions
None
G. Safety System Responses
None
III. CAUSE OF THE EVENT
A. Immediate Cause
The immediate cause for this condition is an undetectable out-of-tolerance high-lift setpoint drift condition on the SRVs, which existed for longer than allowed by the TS.
B. Cause
The root cause of this condition is a generic industry issue, SRV pilot valve disc-seat corrosion bonding. A metal oxide film that develops during normal reactor operation results in a bonding between the seat and the disc and adds resistance to the pressure needed to open the relief valve.
C. Contributing Factors
None
IV. ANALYSIS OF THE EVENT
The condition being reported is the operation of Unit 3 in a manner prohibited by the TS. The as-found valve lift set points following Unit 3 Cycle 14 operation are summarized. in the following table.
Unit 3 Cycle 14 As Found Lift Setpoints(1) Valve Position Serial SRV TS 1st test/dev. 2nd test/dev. 3rd test/dev.
Number Setpoint 3-PCV-001-0004 1038 1155 1168/1.1% 1167/1.0% 3-PCV-001-0005 1017 1145 1170/2.2% 1158/1.1% 1160/1.3% 3-PCV-001-0018 1232 1145 1166/1.8% 1146/0.1% 1157/1.0% 3-PCV-001-0019 1256 1135 1133/-0.2% 1137/0.2% '0 40A„.
3-PCV-001-0022 1037 1145 - 1155/0.9% 1144/-0.1% 3-PCV-001-0023 1077 1135 1153/1.6% 1142/0.6% 1138/0.3% ,0 ...., 3-PCV-001-0030 1253 1145 1161/1.4% 1149/0.3% 3-PCV-001-0031 1260 1135 a 1160/2.2% 1154/1.7% 3-PCV-001-0034 1259 1135 1133/-0.2% 1134/-0.1% 1138/0.3% 3-PCV-001-0041 1263 1155 1169/1.2% 1153/-0.2% 1147/-0.7% 3-PCV-001-0042 1039 1155 E#v 1166/1.0% 1160/0.4% Unit 3 Cycle 14 As-Found Lift Setpoints° Valve Position Serial SRV TS 1st test/dev. 2nd test/dev. 3rd test/dev.
Number Setpoint 3-PCV-001-0179 1019 1155 1186/2.7%:
3-PCV-001-0180 1262 1155 ::".' AM; 1174/1.6% 1164/0.8°k (1) The shaded values indicate test results outside the TS required 3 percent tolerance.
The Unit 3 SRVs are Target Rock Model 7567F two-stage safety/relief valves. The valve is a leak tolerant valve; however, it exhibits significant in-service setpoint drift due to corrosion bonding to the pilot disc to seat. The pilot valve seats are constructed from erosion and wear resistant Stellite 6B. The Stellite alloy develops a hard, metal-oxide corrosion layer on the pilot disc. When placed in an operating environment typical of a Boiling Water Reactor, the steam exposed surfaces can oxidize forming a surface corrosion film. This corrosion forms a bond between the valve seat and disc. The bond adds to the resistance of the setpoint adjustment spring pressure necessary to open the valve and increases the pressure required to actuate the valve. Generally, once the pilot valve is actuated the corrosion bond is broken, the subsequent lift setpoint is within the TS required tolerance.
V. ASSESSMENT OF SAFETY CONSEQUENCES
The safety consequences of this event were not significant. A reactor vessel overpressure evaluation performed for Unit 3 using the Unit 3 Cycle 14 Reload ASME Overpressure and Plant Transient Analysis at 3458 MWT demonstrates compliance with the ASME upset limit of 1375 psig for peak vessel pressure and dome pressure Safety Limit of 1325 psig. The evaluation of the as-found data from the Unit 3, Cycle 14, SRVs realized a peak reactor vessel pressure of 1311 psig in the vessel lower plenum and a maximum steam dome pressure of 1277 psig. The event resulted in a peak vessel pressure of 1396 psig in the vessel lower plenum, which demonstrates compliance with the ASME Service level C Limit of 1500 psig. As such, the pressure relief safety objective of the SRVs was satisfied during the operating cycle.
TVA has previously installed an electronic logic, which automatically opens the SRVs as appropriate during pressurization transients. The electronic logic, although not safety related utilizes high-quality instrumentation that has proven to be very reliable. This electronic logic largely negates the impact on safety presented by this condition.
Therefore, TVA concludes that there was no significant reduction in the protection of the public by this event.
VI. CORRECTIVE ACTIONS
A. Immediate Corrective Actions
All SRV pilot cartridges were replaced during the Unit 3, Cycle 14 refueling outage. Prior to installation, each of the replacement cartridges demonstrated a lift setpoint within the TS requirements during bench testing.
B. Corrective Actions to Prevent Recurrence
To resolve the setpoint drift problem, TVA has installed platinum coated pilot valves in the discs on the thirteen Unit 3 SRVs.
VII. ADDITIONAL INFORMATION
A. Failed Components
None
B. PREVIOUS LERS ON SIMILAR EVENTS
There have been previous similar events on Unit 1 (LER 259/2008-003-00) and Unit 2 (LER 260/2009-003-00).
C. Additional Information
Corrective action document for this report is PER 226627.
D. Safety System Functional Failure Consideration:
This event is not a safety system functional failure according to NEI 99-02.
E. Scram With Complications Consideration:
This event did not include a reactor scram.
VIII. COMMITMENTS
None