05000275/FIN-2009005-05
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Finding | |
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Title | Less than Adequate Change Evaluation to the Facility as Described in the Final Safety Analysis Report Update |
Description | The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the facility as described in the Final Safety Analysis Report Update. In October 2009, the inspectors identified that the replacement reactor head contractor used incorrect damping values in the replacement head design. The contractor was unable to demonstrate that the design met ASME Code using the damping values specified in the plant design basis. On November 5, 2009, the licensee incorporated the new damping values and revised the method for determining the seismic response spectra. Using NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, Revision 1, the inspectors concluded that these changes resulted in a departure from a method of evaluation described in the Final Safety Analysis Report Update establishing the facility design bases. The licensees 50.59 evaluation, Licensing Basis Impact Evaluation LBIE 2009-021, Integrated Head Assembly, was less than adequate to conclude that prior NRC approval was not required for the changes. The licensee entered this issue into their corrective action program as 50276288. The failure of Pacific Gas and Electric to perform an adequate 10 CFR 50.59 evaluation prior to changing the facility as described in the Final Safety Analysis Report Update is a performance deficiency. The inspectors evaluated this issue using the traditional enforcement process because the performance deficiency had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors concluded that the issue was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The inspectors also evaluated this issue using the Significance Determination Process. The inspectors concluded that the violation affected the Initiating Events Cornerstone because the change potentially decreased the structural integrity of the control rod drive mechanism reactor coolant pressure barrier and screened Green because assuming worst case degradation, the finding would not result in exceeding the technical specification limit for reactor coolant system leakage nor have a likely effect on other mitigation systems resulting in a total loss of their safety function. The inspectors concluded that the violation was a Severity Level IV because the issue screened Green under the Significance Determination Process. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate the original problem associated with the replacement reactor head design such that the resolutions address causes and extent of conditions, as necessary P.1(c) |
Site: | Diablo Canyon ![]() |
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Report | IR 05000275/2009005 Section 4OA5 |
Date counted | Dec 31, 2009 (2009Q4) |
Type: | TEV: Severity level IV |
cornerstone | Initiating Events |
Identified by: | NRC identified |
Inspection Procedure: | |
Inspectors (proximate) | G Guerra M Brown M Peck G George G Miller S Makor D Graves N Greene |
CCA | P.2, Evaluation |
INPO aspect | PI.2 |
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Finding - Diablo Canyon - IR 05000275/2009005 | ||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||
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Finding List (Diablo Canyon) @ 2009Q4
Self-Identified List (Diablo Canyon)
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