05000346/FIN-2010003-02
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Finding | |
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| Title | Licensee-Identified Violation |
| Description | Technical Specification 3.3.13, Steam and Feedwater Rupture Control System Actuation, requires channels 1 and 2 of each logic function to be operable in Modes 1, 2, and 3. Contrary to the requirement above, the licensee operated with a steam and feedwater rupture control system (SFRCS) condition that was prohibited by TS 3.3.13. On March 10, 2010, the licensee identified that the SFRCS could unexpectedly re-energize with the low steam line pressure block initiated following a short duration loss of power due to a loss of off-site power (LOOP) and emergency diesel generator start. This condition could result in auxiliary feedwater (AFW) being supplied to a SG affected by a rupture. The issue was entered into the CAP as CR 10-73067. Corrective actions were completed that addressed the cause of the condition, which was to change the SFRCS logic to ensure that a power-on-reset occurs anytime 48 VDC power is lost. SFRCS test procedures were also revised to ensure that the system properly responds to a short duration loss of power. The inspectors determined that the operation of SFRCS in a condition prohibited by technical specifications was a performance deficiency that was more than minor because the issue affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This condition did not screen out in Phase 1 of the SDP because there was a potential loss of a safety function for greater than the technical specification allowed outage time. The significance of this condition was evaluated by the Region III Senior Reactor Analyst (SRA) using an exposure time of 1 year, the maximum allowed by the SDP. The SRA estimated the risk significance of this event using the frequency of a steam line break inside containment (5.7E-4/yr), the conditional probability of a consequential LOOP following a reactor trip (5.3E-3), and the mitigating functions provided in the Davis-Besse Risk Informed Phase 2 Notebook for the Main Steam Line Break. For the AFW mitigating function, the SRA assumed the isolation of AFW was nonfunctional. The dominant sequence involved failure of the main steam isolation valve (MSIVs) to close and operator failure to stop makeup injection. The resultant delta core damage frequency (CDF) was conservatively estimated at 1E-8, meaning the finding was of very low risk significance (Green). The SRA also reviewed the licensees risk analysis of this issue, which also concluded that the risk was of very low safety significance. |
| Site: | Davis Besse  |
|---|---|
| Report | IR 05000346/2010003 Section 4OA7 |
| Date counted | Jun 30, 2010 (2010Q2) |
| Type: | NCV: Green |
| cornerstone | Mitigating Systems |
| Identified by: | Licensee-identified |
| Inspection Procedure: | |
| Inspectors (proximate) | B Palagi J Rutkowski D Passehl A Wilson R Walton R Russell J Cameron E Stamm |
| INPO aspect | |
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Finding - Davis Besse - IR 05000346/2010003 | |||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||
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Finding List (Davis Besse) @ 2010Q2
Self-Identified List (Davis Besse)
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