05000456/FIN-2013008-02
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Finding | |
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| Title | Inaccurate/Incomplete Information For Exemption Request From 10 CFR 50.60 |
| Description | The inspectors identified a finding of very low safety significance (Green) and an associated Severity Level IV Violation of 10 CFR 50.9 Completeness and Accuracy of Information, for the licensees failure to provide information to the NRC that was complete and accurate in all material respects. Specifically, in Letter RS-05-103 License Amendment Request Regarding Reactor Coolant System Pressure and Temperature Limits Report and Request for Exemption from 10 CFR 50.60, the licensee stated that WCAP-16143 provides a valid basis for changing the reactor pressure vessel (RPV) closure head flange limit and maintains the relative margin of safety commensurate with that which existed at the time the 10 CFR Part 50, Appendix G requirement was issued. However, the analysis documented in WCAP-16143 demonstrated adequate vessel margins based upon the original closure head flange configuration and did not represent the modified closure head configuration (53 head studs) applicable to the Unit 2 reactor vessel. Therefore, this analysis did not provide a valid basis for changing the Unit 2 RPV closure head flange limits in 10 CFR Part 50, Appendix G. The licensee entered this issue into the Corrective Action Program (AR 01558067), performed an operability evaluation, and was evaluating several options for corrective measures. The corrective actions under consideration by the licensee included: completing a calculation to validate the Westinghouse Electric vendor letter, revision to WCAP-16143, installation of a 54th head stud, submittal of a license amendment request with a revised WCAP-16143, or negate the existing exemption methodology and return to the pressure temperature limit curves based upon 10 CFR Part 50, Appendix G requirements. The inspectors determined that this issue was more than minor because it adversely affected the Barrier Integrity Cornerstone attribute of Design Control. The inspectors also answered yes to the More-than-Minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically, the inspectors determined that this issue was more than minor because, if left uncorrected, the failure to provide complete and accurate information for the Unit 2 vessel head stud configuration could have resulted in non-conservative pressure temperature limit curves that allowed operation in an unacceptable region that would increase the possibility of vessel failure during a pressurized thermal shock event. The inspectors performed a Phase I SDP screening using IMC 0609, Attachment 0609 Appendix A, Exhibit 3-Barrier Integrity Screening Questions, and selected the box under the Reactor Coolant System Boundary (e.g., pressurized thermal shock issues), which required a detailed risk-evaluation. A Region III Senior Reactor Analyst performed a detailed risk-evaluation of this finding. A potential increase in the probability for vessel failure would exist if the plant was operated in the unacceptable pressure temperature regions and a pressurized thermal shock event occurred. Based on the licensee and supporting vendor assessments which concluded that no substantial increase in vessel stresses will occur due to operation with 53 head studs, the driving force for crack propagation (e.g., K1) will remain essentially unchanged. However, to bound the delta risk-evaluation, it was assumed that the initiating event frequency for a reactor vessel failure increased by 10 percent. From the Braidwood Standardized Plant Analysis Risk Model Version 8.21, the initiating event frequency for reactor vessel failure from any cause was 1E-7/yr. Core damage is expected to occur if reactor vessel failure occurs. The exposure time for the finding was the maximum of one year. Thus, a bounding risk-assessment yields a delta risk of 1E-8/yr. Therefore, based on the detailed risk-evaluation, this finding is of very low risk significance (Green). Because the failure to provide complete and accurate information to the NRC had the potential to impede or impact the regulatory process, the finding was also evaluated in accordance with NRC Enforcement Policy for traditional enforcement. This violation was similar to an example of a Severity Level III violation identified in Section 6.9.c.1 of the NRC Enforcement Policy. However, after consideration by NRC management, and with the approval of the Director of the Office of Enforcement, it was determined that a Severity Level IV Cited Violation was appropriate. This decision was based upon the very low safety significance (Green) of the associated finding. The inspectors concluded that no cross-cutting aspect was applicable as the performance deficiency was not reflective of current performance because the issue was in excess of three years old. |
| Site: | Braidwood  |
|---|---|
| Report | IR 05000456/2013008 Section 1R08 |
| Date counted | Dec 31, 2013 (2013Q4) |
| Type: | Finding: Green |
| cornerstone | Barrier Integrity |
| Identified by: | NRC identified |
| Inspection Procedure: | IP 71111.08 |
| Inspectors (proximate) | B Boston B Palagi D Betancourt J Benjamin J Ellegood J Mcghee K Barclay M Holmberg M Perry R Jicklinga Garmoeb Metrow D Hills M Holmberg V Meghani |
| INPO aspect | |
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Finding - Braidwood - IR 05000456/2013008 | ||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||
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Finding List (Braidwood) @ 2013Q4
Self-Identified List (Braidwood)
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