ENS 41442
ENS Event | |
|---|---|
02:00 Feb 25, 2005 | |
| Title | Potential Breach of Secondary Containment During Seismic Event. |
| Event Description | The licensee provided the following information:
This event notification is being made to report an event that could have prevented fulfillment of the safety function to mitigate the consequences of an accident IAW 10CFR50.72(b)(3)(v)(D). The potential for creation of an actual hole through Secondary Containment (SC) via the Plant Service Water (TSW) system high point reactor building auto vents (TSW-AV-1A and TSW-AV-1 B) exists if a seismic event occurs and the Seismic Category 2 TSW loop-seal piping outside of Secondary Containment is breached (e.g., because of a pipe break) and has drained. Neither of the above conditions presently exists. However, a Secondary Containment breach could occur as a result of a single passive failure (i.e., pipe break in the TSW loop-seal piping described above). TSW-AV-1A and TSW-AV-1B are designed to open automatically when neither TSW system pump is operating to break the vacuum condition that would otherwise exist in the piping. When TSW-AV-1A and TSW-AV-1B are open in vacuum breaker mode, the resultant hole size into the TSW system piping would exceed the allowable Secondary Containment cumulative hole size [ of 32 square inches total], if the loop seal were also breached as described above. Due to this condition Secondary Containment was declared INOPERABLE at 1800 PST 2/24/05. As a compensatory measure to prevent exceeding the allowable Secondary Containment cumulative hole size, one of the two TSW reactor building auto vents (TSW-AV-1B) was isolated by closure of a manual valve (TSW-V-55B). This action was completed at 1828 PST 2/24/05 and the Secondary Containment was declared OPERABLE. One TSW auto vent is sufficient to perform the vacuum breaker function. This condition was found by the licensee's System Engineer. The extent of condition and long term corrective action is under review by licensee. Licensee will inform the NRC Resident Inspector.
The following information was obtained from the licensee via facsimile (licensee text in quotes): On 2/25/05, Columbia Generating Station reported (ref: EN 41442) a condition that was discovered while reviewing service water (TSW) system design documents. During the review, it was determined that Seismic Category II TSW piping in the turbine (TG) and radwaste (RW) buildings could rupture and drain during a seismic event. This event would result in an inoperable secondary containment (SC) because TSW system high point vent valves located within the SC would automatically open when the piping drained after rupturing. Such an event would allow direct communication between the SC atmosphere and the TG/RW atmospheres and exceed the leakage rate assumed in Columbia's accident analysis. There are two principal accidents in Columbia's safety analysis for which SC is credited as a mitigating system. These are the Loss Of Coolant Accident (LOCA) and the Fuel Handling Accident (FHA). The SC performs no active function in response to either of these limiting events, however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis, and that fission products entrapped within the SC structure will be treated prior to discharge to the environment. Assuming a seismic event of the magnitude (0.25g) of the Safe Shutdown Earthquake (SSE) occurred and ruptured the TSW piping causing SC to become inoperable, it is beyond Columbia's safety analysis to postulate a release of radioactive material beyond Part 100 limits for that event. This is because analysis shows that the SSE will not in itself cause a LOCA or an FHA and Columbia's design and licensing bases do not assume a LOCA or FHA coincident with a seismic event. In the event that the SSE made SC inoperable, the Technical Specifications (LCO 3.6.4.1.B) require that the plant be in mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Since this is achievable, it is reasonable to conclude that, in the event previously reported, plant shutdown can be accomplished without radiological release and within the completion time of the action required by the Technical Specifications. The discussion in the guidance document (NUREG 1022) for reporting under Part 50.72(b)(3)(v) states the level of judgment for reporting under these criteria is a reasonable expectation of preventing fulfillment of a safety function. It also states that the intent of the criteria is to capture those events where there would have been a failure of a safety system to properly complete a safety function regardless of whether there was an actual demand. This discussion however, as the Part 50.72(b)(3)(v) and Part 50.73(a)(2)(v) criteria state, apply to safety functions of systems or structures that are needed to control the release of radioactive material, because safe shutdown of the plant without radiological release is assured post SSE (sans SC), as described in Columbia's design and licensing bases, SC would not be needed to control the release of radioactive material and therefore the reporting criteria is not met. The licensee will be voluntarily submitting a Licensee Event Report as a method of information sharing with the rest of the industry. The licensee has notified the NRC Resident Inspector. Headquarters Operations Officer notified R4DO (Howell). |
| Where | |
|---|---|
Columbia Generating Station  Washington (NRC Region 4) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | |
| Time - Person (Reporting Time:+2.7 h0.113 days <br />0.0161 weeks <br />0.0037 months <br />) | |
| Opened: | Richard Meyers 04:42 Feb 25, 2005 |
| NRC Officer: | John Knoke |
| Last Updated: | Apr 5, 2005 |
| 41442 - NRC Website
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Unit 2 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (100 %) |
| After | Power Operation (100 %) |