05000366/LER-2009-002
| Docket Number Sequential Revmonth Day Year Year Month Day Year 05000Number No. | |
| Event date: | 03-13-2009 |
|---|---|
| Report date: | 05-04-2009 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3662009002R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
On March 13, 2009, Unit 2 was in the Refuel mode with fuel in the reactor vessel. Personnel were evaluating local leak rate testing (LLRT) on the valves and penetrations which comprise the Primary Containment (EIIS Code NH) bypass leakage isolation valves. These valves are classified as Secondary Containment (HIS Code NG) bypass valves since the piping is routed outside the secondary containment boundary. It was determined that the "A" feedwater line, which is assumed to be a water filled line for 30 days after the initiation of a design basis accident (DBA), had a tested leakage value that would result in the line losing its water seal prior to the thirty day time frame following initiation of a DBA. The piping has a path which exits the secondary containment boundary. Therefore, the leakage from this penetration would become a source of bypass leakage when the water seal is lost. This LLRT is performed using water and the leakage criteria assume the leakage through the penetration will be water. When a comparable air leakage rate is calculated using the primary containment pressure at the time the water seal is lost, the leakage rate would exceed the allowed bypass leakage of 0.009 La (L sub a). This penetration is not included in the list of bypass leakage penetrations since the line is normally filled with water. However, recognizing that the line exits secondary containment and presents a potential leakage path if the line loses its water seal, it is appropriate to consider this line as a bypass leakage line when the valve leakage is above the allowable limit.
As a result, the above condition exceeds the 0.009 La (L sub a) bypass leakage limit.
During the investigation of this event the basis for the LLRT leakage acceptance criteria for penetrations 2T23-X009A and 2T23-X009B which includes 2B21-F010A, 2B21-F010B, 2B21-F077A, and 2B21-F077B was re-evaluated and determined to require revision to a lower allowable leakage limit. When reviewing previous outage test data, where this updated allowable leakage limit would have been applicable, it was determined that in March of 2007 the "as left" leakage for 2B21-F010B was left greater than the updated leakage limit. As a result the tested leakage through this one valve was above the allowed leakage limit to ensure compliance with the 0.009 La (L sub a) leakage criteria of the Technical Specifications. The "as left" leakage value for the companion valve (2B21-F077B) in this penetration was below the allowable leakage limit. Therefore the overall penetration leakage was within the allowable leakage limit. 10 CFR 50 Appendix J, Option B requires that when starting the reactor from a refueling outage the maximum pathway leakage of the primary containment penetrations must not exceed the established limit. With the one valve exceeding the updated allowable leakage limit the criteria was not met.
CAUSE OF EVENT
The cause of the feedwater outboard valve test failure (only the "A" line was affected) was misalignment caused by internal wear, and missing bearing cover lock pins (this is valve 2B21-F077A).
The cause of the feedwater inboard valve test failures (the "A" and "B" lines were affected) was misalignment caused by excessive clearance between the hinge pin and the disc and, the hinge pin adjustment had changed over the operating cycle.
� PRINTED ON RECYCLED PAPERNRC FORM 366A (9-2007) The cause of the failure to meet the technical specification requirement of 0.009 La (L sub a) for reactor start up maximum pathway leakage was an incorrect calculation.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable per 10 CFR 50.73 (a)(2)(ii)(A) because an event occurred which resulted in one of the plant's principal safety barriers being degraded. Specifically, the primary containment isolation function involving secondary containment bypass valves was found to not satisfy the leakage requirements of the Technical Specifications. In addition a reportable condition exists in that the reactor was operated with one valve having a leakage value above the allowable leakage rate which results in a condition prohibited by the technical specifications and is reportable per 10 CFR 50.73(a)(2)(i)(B).
The function of the primary containment is to isolate and contain fission products released from the reactor primary system following a design basis accident (DBA) and to confine the postulated release of radioactive material. The primary containment consists of a steel vessel which surrounds the reactor primary system and provides a barrier against the uncontrolled release of radioactive material to the environment. Some leakage from the primary containment is assumed to occur, although the majority of the leakage is assumed to be released into the secondary containment. The total allowable leakage rate for the primary containment is designated L. (L sub a) and is equal to 1.2 percent by weight of the contained air volume per day, most of which is assumed to occur within the secondary containment where it would be treated by the Standby Gas Treatment system (SBGT) (EIIS Code BH) before being released at an elevated point through the Main Stack (EIIS Code VL). However, some small amount of leakage is assumed to occur outside secondary containment where it is released without being treated by the SBGT system. Valves located in primary containment penetrations whose pipes lead outside the secondary containment are potential sources of such untreated leakage, so these valves are termed "secondary containment bypass valves". Since the atmospheres in such areas would not be filtered by the SBGT system, the allowable leakage through these valves is specifically addressed by the Technical Specifications, and is limited to a total of 0.009 of La (L sub a). The leakage rates measured in this event were greater than this amount.
The Final Safety Analysis Report (FSAR) for Plant Hatch Unit 2 designates the DBA as the break of a Reactor Recirculation system (EIIS Code AD) pipe which results in the rapid depressurization of the reactor vessel to the primary containment. However, the FSAR analysis shows that, for such an accident, resulting peak fuel cladding temperatures would be less than those required to produce damage to the fuel. The plant specific SAFER/GESTR analysis for this accident scenario shows that no damage to the fuel cladding would occur. Therefore, by this analysis, the only radioactive materials present in the released coolant would be those already present due to normal operation and the small additional amount of contaminated or activated crud released from vessel internals and primary system piping during the initial stages of the transient. This analysis applies to all operating conditions.
CORRECTIVE ACTIONS
2B21-F077A, the outboard valve was rebuilt using new parts, and the missing bearing cover locking pins were replaced.
2B21-F010A and 2B21-F010B were repaired by performing a modification to account for the excessive clearance between the hinge pin and the disc and to prevent the hinge pin adjustment from changing.
� PRINTED ON RECYCLED PAPERNRC FORM 366A (9-2007) 4 The calculation for the allowable leakage limit through these two penetrations has been updated.
ADDITIONAL INFORMATION
Other Systems Affected: No systems other than those already mentioned in this report were affected by this event.
Failed Components Information: None Failed Components Information:
Master Parts List Number:22B21-F010A & F010B EIIS System Code: SJ Manufacturer: Rockwell International Reportable to EPIX: Yes Model Number: 970 Root Cause Code: X Type: Valve, Shutoff EIIS Component Code: SHV Manufacturer Code: R344 Master Parts List Number:
22B21-F077A EIIS System Code: SJ Manufacturer: Atwood and Morrill Reportable to EPIX: Yes Model Number: CL 900 Root Cause Code: X Type: Valve, Shutoff EIIS Component Code: SHV Manufacturer Code: A585 Commitment Information: This report does not create any permanent licensing commitments.
Previous Similar Events:
valves failed the LLRT testing. Corrective actions for that event did not take into account the potential impact of not maintaining the internal tolerances and dimensions identified in this event and did not identify the incorrect length of the hinge pin on the inboard valve.