05000458/FIN-2012003-05
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Finding | |
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| Title | Failure to Specify Manual Actions for Safety Relief Valve Operations During a Station Blackout Event |
| Description | The inspectors identified a non-cited violation of 10 CFR 50.63, Loss of All Alternating Current, paragraph (a) (2), which states, in part, The reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. The capability for coping with a station blackout of specified duration shall be determined by an appropriate coping analysis. Licensees are expected to have the baseline assumptions, analyses, and related information used in their coping evaluations available for NRC review. Specifically, from November 1985 to May 17, 2012, the licensee failed to specify actions while AC power is unavailable to ensure that safety relief valves provided sufficient capacity and capability to ensure appropriate containment integrity is maintained during a station blackout event. This violation has been entered into the corrective action program as Condition Report CR-RBS-2012-03376. The inspectors determined that failure to specify actions for safety relief valve operation in procedures in accordance with NUMARC-8700 was a performance deficiency. The finding was more than minor because it adversely affected the procedure quality attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the station blackout coping procedures did not specify actions that would ensure the heat capacity temperature limit for the suppression pool would not be exceeded during the station blackout coping period. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the inspectors determined that the Mitigating Systems Cornerstone was affected because the finding could cause degradation of core decay heat removal. Using Table 4a from the Phase 1 worksheet, the inspectors determined that the finding represents a loss of safety function; therefore, a Phase 2 analysis was necessary. However, the inspectors determined that a Phase 2 analysis was not sufficient to assess significance because of the complexity of the finding. Therefore, a Phase 3 analysis was necessary. The result of the Phase 3 analysis determined that the change in core-damage-frequency (ACDF) for the performance deficiency was 2.4E-7 or very low safety significance (Green). The senior reactor analyst determined that the change in large-early-release-frequency (ALERF) was 4.8E-8 or very low safety significance (Green). No cross-cutting aspect was identified because the most significant contributor was not indicative of current licensee performance |
| Site: | River Bend  |
|---|---|
| Report | IR 05000458/2012003 Section 4OA5 |
| Date counted | Jun 30, 2012 (2012Q2) |
| Type: | NCV: Green |
| cornerstone | Mitigating Systems |
| Identified by: | NRC identified |
| Inspection Procedure: | |
| Inspectors (proximate) | K Clayton G George L Carson S Garchow J Melfi A Barrett G Larkin V Gaddy M Runyan R Latta C Alldredge |
| INPO aspect | |
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Finding - River Bend - IR 05000458/2012003 | |||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||
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Finding List (River Bend) @ 2012Q2
Self-Identified List (River Bend)
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