The inspectors reviewed a self-revealing non-cited violation of
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of maintenance personnel to implement procedures associated with
foreign material exclusion controls while performing maintenance activities on safety-related 120Vac inverter equipment. Specifically, between October 2009 and April 2012, maintenance personnel failed to follow Procedure SO123-FO-1, Site
Foreign Material Exclusion Control Program, Revision 6, and Procedure SO123-I-1.18,
Foreign Material Exclusion (
FME) Control, Revision 18, to prevent the introduction of a metal air filter frame that was left inside an energized electrical cabinet. This issue was entered into the licensees corrective action program as Nuclear Notification NN 201958287. The performance deficiency is more than minor, and therefore a finding, because it is associated with the
Mitigating Systems Cornerstone attribute for human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, maintenance personnel failed to prevent the introduction of foreign material into the energized electrical cabinet of inverter 2Y004. The resident inspectors performed the initial significance determination for the inverter finding. The inspectors used the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of safety function. A Region IV senior reactor analyst performed a Phase 2 significance determination and attempted to use the pre-solved worksheet from the Risk Informed Inspection Notebook for the San Onofre Nuclear Generating Station, Revision 2.01a. However, the pre-solved worksheet did not include this inverter. Therefore, the analyst performed a bounding Phase 3 significance determination. The bounding change to the core damage frequency was 1.1E-8/yr, and therefore, determined to be of very low safety significance. The small population of affected equipment (included in the
probabilistic risk assessment model) helped to minimize the safety significance. The contributing core damage sequences involved a seismic event and a consequential failure of
auxiliary feedwater flow control and bypass valves to one
steam generator. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because maintenance personnel failed to have an appropriate threshold for identifying issues associated with a degraded air filter and its impact to
foreign material exclusion controls to ensure there would be no adverse impact to system operability
