05000366/LER-2005-002
| Edwin I. Hatch Nuclear Plant - Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii) |
| 3662005002R01 - 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 3/2/2005 at 1500 ET, Unit 2 was in the Refuel mode with fuel in the vessel and the reactor cavity flooded for refueling operations. At that time, engineers and technicians were performing Local Leak Rate Testing (LLRT) on valves 2E51-F008 and 2E51-F007 (EIIS Code BN) for penetration number 10. The plants Technical Requirements Manual (TRM) table T7.0-1 identifies primary containment penetrations and references applicable notes for the penetration barriers. Note 28 for Table T7.0-1 is used to identify penetrations that are required to meet the leakage requirements for Secondary Containment bypass penetrations. The present version of the TRM does not reference Note 28 for penetration number ten.
However, Corporate Licensing has determined that the piping configuration for this penetration should be included as part of Secondary Containment bypass leakage. The Unit 2 Technical Specifications surveillance requirement (SR) 3.6.1.3.10 addresses the leakage restrictions for Secondary Containment bypass valves.
Secondary Containment bypass valves have specific leakage rates established in the plant's Technical Specifications to ensure that the assumptions of the safety analysis are met. The maximum leakage rate allowed for all of the Secondary Containment bypass valves is 0.009 L. (or approximately 544 ACCM) per SR 3.6.1.3.10.
Leakage through 2E51-F007 and 2E51-F008 was measured at 662 ACCM, which exceeded the allowable leakage established by the plant's Technical Specifications for Secondary Containment bypass leakage.
These valves were disassembled and repaired. Both of the seats of these valves were observed to be relatively flat (i.e., the seating surface was larger than what is normally desirable) and this was considered to be caused by normal wear. The valves were repaired by installing new wedges.
CAUSE OF EVENT
The cause of the 2E51-F007 and 2E51-F008 valves leakage exceeding the allowable leakage established by the plant's Technical Specifications for Secondary Containment bypass leakage is considered to be most likely normal wear.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable per 10 CFR 50.73 (a)(2)(ii) because an event occurred which resulted in the degradation of one of the plant's principal safety barriers. Specifically, the RCIC penetration number ten exceeded the allowable leakage established by the plant's Technical Specifications.
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 La and is equal to 1.2 percent by weight of the contained air volume per day. For Plant Hatch Unit 2, this equates to a total allowable leakage of 61,000 ACCM, most of which is assumed to occur within the Secondary Containment where it will be treated by the Standby Gas Treatment System (EIIS Code BH) before being released at an elevated point through the Main Stack (EIIS Code VL).
Additionally, there is also some amount of "Secondary Containment bypass" leakage 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 treated by the SBGT system, the allowable leakage through these valves is specifically addressed by the Technical Specifications, and is limited to a total of 544 ACCM. The 662 ACCM leakage through 2E51-F007 and 2E51-F008 exceeded the allowable leakage established by the plant's Technical Specifications for Secondary Containment bypass leakage.
The allowable leakage for Secondary Containment bypass valves was established using conservative licensing basis evaluation methods in accordance with NRC Regulatory Guide 1.3. These methods conservatively assume that the postulated accident results in fuel damage with 100 percent of the core noble gas activity and 50 percent of the iodine activity released. Consequently, the actual measured leakage of the valves identified in this report would likely have resulted in exceeding the values set forth in 10 CFR 100 during a postulated design basis accident that assumes fuel damage per NRC Regulatory Guide 1.3.
The Final Safety Analysis Report (FSAR) for Plant Hatch Unit 2 designates the Design Basis Accident (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 even if additional failures are postulated, such as failures of certain power supplies and certain low pressure emergency core cooling systems. 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.
Based on this analysis contained in the FSAR, it is concluded that this event did not result in any adverse impact on nuclear safety. This analysis applies to all operating conditions.
CORRECTIVE ACTIONS
Valves 2E51-F007 and 2E51-F008 were repaired by installing new wedges.
The information necessary to identify penetration ten as a potential bypass leakage path will be added to the Unit 2 Technical Requirements Manual. This will be done by August 15, 2005.
ADDITIONAL INFORMATION
No systems other than those already mentioned in this report were affected by this event.
Failed Component Information:
Master Parts List Number: 2E51-F007&F008 EIIS System Code: BN Manufacturer: Powell Reportable to EPIX: Yes Model Number: 73600 Root Cause Code: X Type: Valve, Shutoff EIIS Component Code: SHV Manufacturer Code: P305 Previous Similar Events: No events have been reported in the past two years in which the plant exceeded the Technical Specification limits for Secondary Containment bypasses leakage.
Commitment Information: This report does not create any permanent licensing commitments.