05000366/LER-2010-001
| Docket Number Sequential Revmonth Day Year Year Month Day Year 05000Number No. | |
| Event date: | 03-10-2010 |
|---|---|
| Report date: | 04-30-2010 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3662010001R00 - 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 10, 2010 at 20:00 EST, Unit 2 was at approximately 2799 CMWTh, 99.8 percent power.
Earlier that day Operations personnel were performing the Reactor Core Isolation Cooling (RCIC, EIIS Code BN) operability procedure. During that evolution an annunciator indicated that the RCIC barometric condenser pressure was high. Subsequent investigation determined that the vacuum pump discharge check valve, 2E51-F028, was stuck closed. A review of the system operability was performed and since the barometric condenser is not required for RCIC to fulfill the design function for the system mission time it was determined that RCIC was operable. This valve performs a second function of primary containment isolation valve (PCIV, EIIS Code NH). The cause of the valve sticking was unknown and thus the valve must be considered inoperable and the appropriate Technical Specification should be entered for the inoperable Primary Containment Isolation Valve. This was not identified until after the action time for the applicable Technical Specification had expired; therefore, a condition prohibited by the Technical Specification existed.
CAUSE OF EVENT
The cause of this event was the failure to review the PCIV operability from both a system and a component level. This resulted in the failure to identify the need to enter the plant Technical Specifications for the inoperable PCIV.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable under the provisions of 10 CFR 50.73(a)(2)(i)(B) Any operation or condition which was prohibited by the plant's Technical Specifications. Specifically, the PCIV was inoperable and the compensatory actions were not taken within the allowed timeframe.
The function of the Primary Containment (EIIS Code NH) is to isolate and contain fission products released from the reactor primary system (EIIS Code AD) 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 (EIIS Code NG). The total allowable leakage rate for the Primary Containment is designated "L sub a", and is equal to 1.2 percent by weight of the containment air volume per day. The leakage that occurs within the secondary containment is treated by the Standby Gas Treatment System (EIIS Code BH) before being released at an elevated point through the Main Stack (EIIS Code VL).
� The failed valve, the RCIC vacuum pump discharge check valve, is within the secondary containment boundary. In addition the valve discharges into the top of the Torus (EIIS Code NH) and terminates below the water line. The Torus is postulated to remain water filled post accident. Therefore this valve does not communicate with the gas atmosphere within the Torus.
Primary Containment leakage criteria were 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.
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 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. In addition since this valve communicates with Primary Containment through a pipe that is submerged in the Torus communication with a gaseous release is not postulated. Realistically, therefore, the 10 CFR 100 off site dose limits would likely not have been exceeded had an actual event occurred.
Based on this analysis contained in the FSAR, it is concluded that the RCIC valve failure being reported did not result in any adverse impact on nuclear safety. This analysis applies to all operating conditions.
CORRECTIVE ACTIONS
Corrective action is to incorporate this example into the Operation training program.
The failed valve has been replaced and tested with an acceptable PCIV.
ADDITIONAL INFORMATION
Other Systems Affected: None Failed Components Information:
Master Parts List Number: 2E51-F028B EIIS System Code: BN Manufacturer: Rockwell International Reportable to EPIX: Yes Model Number: 3674T Root Cause Code: X Type: Valve, Shutoff EIIS Component Code: SHV Manufacturer Code: R344 PRINTED ON RECYCLED PAPERNRC FORM 366A (12007) 4 U.S. NUCLEAR REGULATORY COMMISSIONNRC FORM 366A� LICENSEE EVENT REPORT (LER)�(9-2007)
CONTINUATION SHEET
2010 -��001 -��0 Commitment Information:
This report does not create any new permanent licensing commitments.
Previous Similar Events:
Technical Specification. In that event the manual action, realignment of a suction source, was not taken and the Technical Specification also was not entered. It is similar to this event in that proper entry into the Technical Specification as not made. The corrective action for this event focused on revision of a procedure to correct the specific event by requiring the manual action to be taken.
Therefore the corrective action would not have prevented the current event.
� PRINTED ON RECYCLED PAPERNRC FORM 366A (9-2007)