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05000259/FIN-2012007-062012Q1Browns FerryLicensee-Identified Violation10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. BFN procedure NPG-SPP-03.1.9, Rev. 0002, which is a subset of the sites corrective action procedure NPG-SPP-03.1, Rev. 0002, stated, in part, that a PER cannot be closed that has identified a degraded or non-conforming condition until the corrective actions to resolve the degraded or non-conforming condition are completed. Contrary to the above, the licensee closed two PERs (177130, 243955) that were generated during the sites NFPA 805 transition process, based on the implementation of compensatory measures. The permanent corrective action for these nonconformances (transition to NFPA 805) has not been completed. Using IMC 0609, Attachment 4, Phase 1, Initial Screening and Characterization of Findings, inspectors determined the violation was of very low safety significance (Green) because it was not a design or qualification deficiency, did not result in the loss of any system safety function and was not risk significant due to seismic, flooding or severe weather. This violation was documented in the licensees corrective action program as PER 503024.
05000259/FIN-2012007-042012Q1Browns FerryFailure to Identify and Correct Deficiencies Associated with Safe Shutdown InstructionsThe inspectors identified a Green non-cited violation of 10 CFR 50 Appendix B, Criteria XVI, Corrective Action, for the licensees failure to assure conditions adverse to quality associated with the establishment and implementation of four new Safe Shutdown Instructions (SSI) were promptly identified and corrected. Specifically, the inspectors identified instances where previously identified issues with SSIs were either not entered into the corrective action program, corrective actions were not implemented, or the corrective actions were ineffective in addressing the identified issue. The licensee entered this finding into the corrective action program (PER 505551) and adequate procedural guidance was restored following licensee procedure revisions, training and demonstration to inspectors that operators had acquired an adequate level of proficiency to implement the new SSIs. This finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and it affected the cornerstone objective of protection against external events, such as fire, to prevent undesirable consequences. The finding was assigned a Low degradation rating and screened as very low safety significance (Green) in step 1.3.1 of IMC 0609 Appendix F, attachment 1, Application of Fire Protection SDP Phase 1 Worksheet. This finding was directly related to the cross-cutting aspect of Thorough Evaluation of Identified Problems in the Corrective Action Program component of the Problem Identification and Resolution area because the licensee did not thoroughly evaluate identified problems such that the resolutions addresses the causes and extent of conditions of the issues.
05000259/FIN-2012007-032012Q1Browns FerryFailure to Implement Appropriate Safe Shutdown InstructionsThe inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish procedures appropriate to the circumstances for combating plant fires. Specifically, four new Safe Shutdown Instruction (SSI) were established which contained multiple procedural deficiencies. The licensee entered this finding into the corrective action program (PER 507721) and adequate Safe Shutdown Instructions were restored following procedure revisions. This finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and it affected the cornerstone objective of protection against external events such as fire to prevent undesirable consequences. The finding was assigned a Low degradation rating and screened as very low safety significance (Green) in step 1.3.1 of IMC 0609 Appendix F, attachment 1, Application of Fire Protection SDP Phase 1 Worksheet. The team determined the cause of this finding was directly related to the crosscutting aspect of Work Coordination in the Work Control component of the Human Performance area because the licensee did not adequately incorporate actions to address the impact of the work on different job activities and the need for work groups to maintain interfaces with offsite organizations, and communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance. This contributed to the failure to identify deficiencies with the new SSI procedures prior to procedure implementation.
05000259/FIN-2012007-022012Q1Browns FerryFailure to Establish Adequate Compensatory Measures for Non- Conforming Fire BarriersThe inspectors identified a Green NCV of Browns Ferry Operating License Conditions 2.C(13), 2.C(14) and 2.C(7), for Units 1, 2, and 3, respectively, for the licensees failure to establish adequate compensatory measures for non-conforming fire barriers, in accordance with the approved fire protection program (FPP). Specifically, the licensee failed to establish continuous fire watches for non-conforming fire barriers in the Intake Pumping Station (IPS), after discovering that the barriers were not credited in the sites approved FPP. The licensee initiated PER 509589 to document this condition and enter it into the corrective action program. The licensee also established a continuous fire watch, in accordance with the FPR. The licensees failure to establish adequate compensatory measures for non-conforming fire barriers, as required by their approved fire protection program, is a PD. The finding is more than minor because it is associated with the Reactor Safety Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and it affects the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events. Using the guidance of IMC 0609, Appendix F, Fire Protection Significance Determination Process, inspectors determined that the PD represented a finding of very low safety significance (Green). Inspectors determined that the cause of this finding has a cross-cutting aspect in the Corrective Action Program component of the Problem Identification and Resolution (PI&R) area, in that it was directly related to the licensee not thoroughly evaluating problems, such that the problem was properly classified and evaluated for operability
05000259/FIN-2012007-012012Q1Browns FerryFailure to Follow NRC Commitment Management ProcedureThe inspectors identified a Green finding (FIN) for the licensees failure to follow procedure NPG-SPP-03.3, Rev.001, NRC Commitment Management. Specifically, the procedure states, in part, that each responsible organization ensures commitment implementation/completion occurs as scheduled. Contrary to this requirement, the licensees commitment to verify the accuracy and adequacy of completed Inspection Procedure (IP) 95002 corrective actions had not been performed adequately. The licensee entered this issue into the corrective action program as PERs 510126 and 510161. The performance deficiency (PD) associated with this finding was the failure of licensee personnel to follow procedures regarding managing NRC commitments. The finding is greater than minor because, if left uncorrected, the finding would have the potential to lead to a more significant safety concern. Specifically, the failure to assess the adequacy of corrective actions can lead to problems not being properly corrected. Using Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, the finding was determined to have a very low safety significance (Green) because the finding did not result in a loss of system safety function, an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross cutting aspect in the area of Human Performance because the licensee did not ensure supervisory and management oversight of work activities associated with the commitments made to the NRC, which resulted in the commitments not be tracked or monitored to ensure completion.
05000259/FIN-2012007-072012Q1Browns FerryLicensee-Identified Violation10 CFR 50.72(b)(3)(ii)(B) states, in part, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Additionally, 10 CFR 50.73(a)(2)(ii)(B) requires licensees to submit a Licensee Event Report (LER) within 60 days after discovery of any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Contrary to the above, on February 5, 2011, the licensee identified that they had failed to recognize that six unanalyzed conditions discovered during the sites NFPA 805 transition process were reportable conditions (see Section 4OA5 of this report). Consequently, the licensee failed to make an eight-hour report as required by 10 CFR 50.72, and submit LERs within 60 days, as required by 10 CFR 50.73. This finding was considered as traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The NRC has characterized this violation as a Severity Level IV NCV in accordance with Section 6.9 of the NRC Enforcement Policy. This violation was documented in the licensees corrective action program as PERs 505749, 505750, 505751, and 505752. Additionally, the licensee made an eight-hour report, and at the time of the exit, planned to submit LERs for the unanalyzed conditions.
05000250/FIN-2011005-022011Q4Turkey PointFailure to maintain TSC habitabilityThe licensee identified an Apparent Violation (AV) of 10 CFR Part 50.54(q), for failure to follow and maintain in effect emergency plans which require that adequate emergency facilities and equipment to support the emergency response are provided and maintained. Specifically, during the periods from December 4, 2010 to July 13, 2011, and from October 10 to October 28, 2011, the licensee failed to maintain a fully functional Technical Support Center when portions of its ventilation system were removed from service without compensatory measures. As a result, had the facility been required, personnel assigned to respond in the TSC would not have been protected from radiological hazards that would occur in some accidents. The licensee documented this issue in their corrective action program as AR 1701357. The finding was more than minor because it affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The Emergency Preparedness cornerstone was affected in that during the time the Technical Support Center was not functional, it did not meet 10 CFR 50.47(b)(8) Planning Standards program elements in that personnel assigned to the TSC during an emergency may not have been protected from radiological hazards. This finding was evaluated in accordance with Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Section 4.8 and Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, and determined to be a finding of low to moderate safety significance (White) because there was a loss of the planning standard. The two events, December 2010 to July 2011, and October 2011, were assessed as a single finding with a common performance deficiency. The cause of the finding is related to the Problem Identification and Resolution cross-cutting area, in that the licensee did not thoroughly evaluate problems with the TSC ventilation system as necessary, including properly classifying, prioritizing, and evaluating for operability and reportability, conditions adverse to quality.
05000250/FIN-2011005-012011Q4Turkey PointFailure to Correct Valve Deficiency Results in Both Headers of Intake Cooling Water InoperableA self-revealing non-cited violation of 10 CFR 50 Criterion XVI was identified when the licensee failed to repair a degraded butterfly valve in the Unit 3 intake cooling water system. On August 11, 2011, failure of this valve led to a loss of intake cooling water (ICW) flow to the component cooling water heat exchangers. The licensee documented the failure in their corrective action program as AR 01680272 and initiated a cause investigation. An NRC special inspection of this occurrence was documented in NRC Inspection Report 05000250/2011013. The licensees failure to take prompt corrective actions for a degraded valve, though it had been identified in 2007 as vibrating excessively, was a performance deficiency. This performance deficiency was considered more than minor because it could be reasonably viewed as a precursor to a significant event, the loss of all intake cooling water. A Senior Reactor Analyst in a Phase 3 risk assessment, determined the increase in risk to either unit was of very low risk significance i.e., Green. Unit 3 risk was assessed because the event occurred on that unit; however Unit 4 risk was also assessed because the same vulnerability existed on the ICW valves on that unit (e.g., similar design, maintenance history, etc.). The main contributors to the low risk results were: 1) the recovery probability of the ICW system, given the extended time available to operators before a RCP seal LOCA could occur; and 2) the multiple redundant sources available to cool the core should the CCW system fail. The dominant core damage scenarios were valid demands for a reactor trip followed by the failure to recover ICW proceeding to a RCP seal LOCA and core damage. The inspectors determined that the cause of this finding was related to the Problem Identification and Resolution cross cutting area when the licensee failed to take appropriate corrective action to address safety issues (valve fluttering) in a timely manner, commensurate with the safety significance.
05000250/FIN-2011005-032011Q4Turkey PointFailure to make a required 8 hour NRC report for major loss of emergency assessment capabilityThe inspectors identified an Apparent Violation of 10 CFR 50.72(b)(3)(xiii) when a major loss of emergency assessment capability was not reported to the NRC within 8 hours. The TSC ventilation system was identified as being in a degraded condition from December 4, 2010 until July 13, 2011, affecting the habitability of the TSC for emergency responders, and the occurrence was not reported. The issue was identified to the licensee by the inspectors after review of NRC Event Notification 47387. The finding was more than minor because it impacted the NRCs regulatory process, which relies on certain events being properly reported to the NRC. Because this finding impacted the regulatory process, it was evaluated using traditional enforcement and is being considered for escalated enforcement action in accordance with NRCs Enforcement Policy. No cross-cutting aspect associated with this issue was identified.
05000261/FIN-2011003-022011Q2RobinsonInadequate Seismic Analysis for Installation of Safety Related Cable Trays and ConduitThe inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to perform an adequate seismic analysis during the plant modification of the 125VDC Battery Chargers. Specifically, the interface evaluation for installation of the safety-related, Battery Charger, cable tray and conduit failed to consider the seismic interaction with the adjacent air-handling unit structure. Subsequent review and analysis determined that the modification introduced a degraded/nonconforming condition which does not affect operability. The licensee documented the issue in Nuclear Condition Report 458971 and initiated actions for a plant modification. The failure to perform an adequate seismic analysis for the installation of the safetyrelated cable trays and conduit is a performance deficiency. This performance deficiency is associated with the design control attribute of the Mitigating System Cornerstone. It is more than minor since it is similar to Inspection Manual Chapter 0612, Appendix E, Example, 3.a, in that the seismic analysis for the cable trays and conduits require revision and modification to the air handling unit structural supports to correctly resolve the seismic concerns. In accordance with IMC 0609 (Table 4a), Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency which resulted in a loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the performance deficiency occurred in 1991 and does not represent current licensee performance.
05000261/FIN-2011003-032011Q2RobinsonRefueling Water Storage Tank Inoperable While On PurificationThe inspectors identified a NCV of Technical Specification (TS) 3.5.4 Refueling Water Storage Tank (RWST), which required the RWST to be operable in modes 1 through 4. The licensee failed to comply with the TS Action Statements when the RWST was rendered inoperable by placing the non-seismically qualified purification loop in operation. Upon discovery the licensee promptly restored the RWST to operable status by removing the purification loop from service, put administrative controls in place to prevent use of the purification loop, and initiated Action Request (AR) 452093 to evaluate the event. Use of the non-seismically qualified Spent Fuel Pool Demineralizer System for purification of the Refueling Water Storage Tank was determined to be a performance deficiency. This action rendered the RWST inoperable and the licensee failed to comply with the required action statement for an inoperable RWST. The finding is more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Specifically, during a seismic event the purification piping could break and cause a loss of inventory in the RWST. Significance Determination Process (SDP) Phase 1 screening determined that this finding was within the mitigating systems cornerstone and was potentially risk significant due to a seismic external event and therefore required a Phase 3 SDP analysis. A phase 3 risk assessment was performed by a regional SRA using the NRC SPAR model. An exposure period of 213 days was utilized as this represented the worst case one year exposure period determined using the RWST purification history data. No recovery credit was assumed in the analysis. The non-seismic RWST purification piping and the dedicated shutdown diesel generator were assumed to fail at the same seismic input as that assumed for a loss of offsite power. The dominant sequence was a seismically induced loss of offsite power leading to a station blackout with failure of the emergency power system and failure to recover offsite power or the emergency diesel generators. Subsequent battery depletion and operator failure to control the turbine driven auxiliary feedwater pump would lead to core damage. The risk was mitigated by the low probability of a seismic event and the failure probability of the emergency diesel generators. The analysis determined that the risk increase of the performance deficiency was an increase in core damage frequency less than 1E-6/year a GREEN finding of very low safety significance. The cause of the finding was directly related to the conservative assumptions aspect in the Decision Making component of the Human Performance area because during a previous review of this evolution the licensee did not demonstrate the proposed action was safe in order to proceed. Instead the licensee could not find a requirement to show it was unsafe and concluded placing the RWST on purification was acceptable.
05000335/FIN-2011003-012011Q2Saint LucieFailure to Comply with Design Drawing Results in Main Steam Vent Line Failure and Subsequent TransientA self-revealing finding of very low safety significance was identified following a rapid downpower and manual reactor trip of Unit 2 on May 16, 2011. Specifically, the licensee failed to comply with an approved design drawing during installation of a one-inch vent line which resulted in a fatigue failure of the vent line. No violations of NRC requirements were identified because the location of the vent line was downstream of the main steam isolation valve and was classified as non-safety related. The licensee entered the issue into the Corrective Action Program as Action Request (AR) 1651817. The finding was more than minor because it resulted in a rapid downpower and manual reactor trip. The finding was associated with the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as at power operations. Using NRC Inspection Manual Chapter 0609.04, Significance Determination Process (SDP) Phase 1 Initial Screening and Characterization of Findings, Table 4a for the Initiating Events Cornerstone, the finding was determined to be of very low safety significance (Green) because it was a transient initiator but did not increase the likelihood that mitigation equipment would not be available. This finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current plant performance. Specifically, the performance deficiency occurred in 2005 or earlier
05000261/FIN-2011003-012011Q2RobinsonRainstorm Results in Flooding of the Power BlockOn May 27, 2011, a heavy rainstorm was not successfully managed by the sites engineered rainwater management features. This resulted in water run-off into the protected area, backing up of storm drains and water intrusion into the power block, Auxiliary Building and other support buildings. Additional review by the NRC is required following the completion of the licensee\\\'s root cause investigation. The review will determine whether this issue represents a performance deficiency. This issue is identified as URI 05000261/2011003-1, Rainstorm Results in Flooding of the Power Block.
05000261/FIN-2011002-042011Q1RobinsonRefueling Water Storage Tank Operability While On PurificationAn Unresolved Item is being opened to provide for additional inspection in response to an NRC identified issue regarding Refueling Water Storage Tank (RWST) operability with the purification loop in operation. The inspectors noted on March 8, 2011, that the RWST purification loop had been in operation for approximately 14 hours. The piping and components of the purification loop are shown on plant drawings to be beyond the seismic qualification boundary for the RWST. The licensee had previously reviewed this issue using AR 422778 in late 2010 and determined it was acceptable to place the RWST on purification without declaring the RWST inoperable. The inspectors questioned the basis for that conclusion. The licensee removed the RWST from purification and put administrative controls in place to prevent use of the purification loop until the issue is resolved. The licensee is continuing to evaluate the use of the RWST purification loop and the impact on operability of the RWST. Additional review by the NRC is required following the completion of the licensees evaluation. This review will also determine whether this issue represents a performance deficiency. The issue will be identified as URI 05000261/2011002-4, Refueling Water Storage Tank Operability While On Purification
05000361/FIN-2010006-092010Q2San OnofreFailure to Establish Goals And Monitor for A(A) Auxiliary Feedwater TrainsTwo examples of a noncited violation of 10 CFR 50.65(a)(1) were identified involving the failure to monitor the unavailability time associated with equipment failures which were maintenance induced. The first example involved maintenance inadvertently bending the fuse holder contacts such that there was a loose connection on the power supply on the turbine-driven auxiliary feedwater pump resulting in its failure. The second example involved the failure to perform maintenance associated with a condensate storage tank isolation valve resulting in its failure during in-service testing. In both cases, if the licensee had assessed the unavailability time due to the maintenance induced failures, the systems would have exceeded the 10 CFR 50.65(a)(2) monitoring criteria, necessitating the systems to be placed in 10 CFR 50.65(a)(1) goal setting. The licensee\'s corrective actions included evaluating its procedures to prevent recurrence, and re-evaluating these systems to determine the impact of accounting for unavailable time. This finding is more than minor because it affects the equipment performance attribute of the Mitigating Systems Cornerstone per Inspection Manual Chapter 612, Appendix 8. Using Inspection Manual Chapter 0609, Phase 1, \"Initial Screening and Characterization of Findings,\" the inspectors determined the finding to be of very low safety significance (Green) because they did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event The cause of the finding was determined to have a crosscutting aspect in the area of human performance. Specifically, personnel failed to use a formal decision making process to determine how to count unavailable hours for the maintenance rule.
05000361/FIN-2010006-102010Q2San OnofreFailure to Identify and Correct Use of Deficient Relays

The inspectors identified a noncited violation of 10 CFR Part 50, Appendix 8, Criterion XVI, \"Corrective Action,\" in that, from October 2008 to April 2010, the licensee failed to promptly identify and correct potentially degraded motor-driven relays in safety-related systems and components. Specifically, after identifying a degraded relay affecting an emergency diesel generator, the licensee replaced all similar relays in the other diesel generators but failed to evaluate the use of these potentially degraded relays in other safety-related systems. The licensee entered this issue into the corrective action program as Nuclear Notification 200146292, and developed a plan to replace the 62 degraded relays that were installed in other safety-related equipment

This finding was more than minor because it impacted the equipment performance 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. Using Inspection Manual Chapter 0609.04, Phase 1, initial screening and Characterization of Findings,\" the inspectors determined the finding to be of very low safety significance (Green) because it did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of human performance associated with the decision-making component, in that the licensee did not use conservative assumptions in making decisions about the extent of condition.

05000361/FIN-2010006-122010Q2San OnofreFailure to Maintain Design Basis InformationThe inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, \\\"Design Control\\\" in that the licensee failed to translate design basis information into procedures for the turbine-driven auxiliary feedwater pump steam admission valves. Specifically, the licensee did not translate into procedures the design requirements to manually close and gag the valves within 30 minutes in response to high energy line breaks, a fire in the auxiliary feedwater pump room, or a steam generator tube rupture event. This issue was entered into the licensee\\\'s corrective action program as Nuclear Notification 200887620. Immediate actions included posting a leveraging device for operators to use should it be necessary, training operators, and scheduling lubrication of the valves. The finding is more than minor because it impacted the Mitigating Systems Cornerstones and its objective to ensure the availability and reliability of equipment that responds to initiating events. The analyst screened the issue to more than one cornerstone due to its effect on early release (steam generator tube rupture), fire protection, and mitigating systems (high energy line break). The analyst performed a Phase 3 analysis that considered the effects of a high energy line break in the pump room, a steam generator tube rupture, and fires in the pump room and auxiliary feedwater pipe tunnel. The analyst determined that the combined significance of these scenarios was a delta- core damage frequency of 5.E-9/yr and a delta- large early release frequency of 1.6E-9/yr. Therefore, the violation was determined to be of very low safety significance (Green). The inspectors determined that cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program. Specifically, the licensee had previous opportunities to identify this problem when the valve was removed from the in-service testing program and when they evaluated relevant external operating experience.
05000361/FIN-2010006-052010Q2San OnofreControl Room Operators\' Failure to Adhere to Conduct of Operations Procedural RequirementsThe inspectors identified a noncited violation of Technical Specification 5.5.1.1.a involving the failure of control room operators to follow San Onofre Procedure S0123-0-A1, \"Conduct of Operations.\" These included failures to: implement alarm response procedure place-keeping, announce alarms to the control room supervisor, stop conversations when an alarm annunciated and cleared, perform three-way communication during pre-job briefing, review the summarize, anticipate, foresee, evaluate and review questions during a pre-job brief, review the prerequisites of a procedure prior to use, and remain cognitive of the re-activity change evolution by a control room supervisor. This issue was entered into the licensee\'s corrective action program as Nuclear Notification 200871332, and operations management immediately began actions to institute a recovery plan to improve operator performance. The finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of human performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. As a result, the issue was of very low safety significance (Green). The finding has a crosscutting aspect in the area of human performance associated with the work practices because the licensee did not ensure supervisory and management oversight of work activities.
05000361/FIN-2010006-062010Q2San OnofreFailure to Provide Adequate Procedure for Boron Dilution ActivitiesThe inspectors reviewed a self-revealing noncited violation of Technical Specification 5.5.1.1.a involving the failure to maintain adequate instructions in San Onofre Procedure S023-3-2.4, \"RCS Purification and De-borating Ion Exchanger Operation,\" Revision 21 to control borating of ion exchangers. The failure to maintain an adequate procedure resulted in an unplanned power reduction by control room operators. This issue was entered into the licensee\'s corrective action program as Nuclear Notification 200721702. Immediate corrective actions included revising the procedure and operator crew training. The finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of human performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. As a result, the issue was of very low safety significance (Green). The finding has a crosscutting aspect in the area of human performance associated with the work practices because licensee supervisory personnel did not ensure activities associated with re-activity control were performed in a controlled manner such that nuclear safety was assured.
05000361/FIN-2010006-082010Q2San OnofreFailure to Maintain Written Procedures Covered In Regulatory Guide 1.33The inspectors identified a cited violation of Technical Specification 5.5.1.1.a, involving the failure to maintain adequate written procedures. Specifically, as of April 23, 2010, the licensee\\\'s controls over its backlog of procedure change requests associated with plant modifications were inadequate to prevent licensee personnel from using outdated procedures with known technical errors in the plant. The performance deficiency of failing to control the backlog of procedure changes, such that procedures with known technical errors were in use in the plant were previously identified by the NRC on two occasions and were documented as noncited violations 05000361; 05000362/2009003-09 and 2009009-02. Because the licensee failed to restore compliance within a reasonable time after the previous noncited violations were identified, this violation is being cited in a Notice of Violation in accordance with Section Vl.a.1 of the NRC\\\'s Enforcement Policy. This finding was entered into the licensee\\\'s corrective action program as Nuclear Notification 200888919. The licensee\\\'s corrective action included immediate actions to administratively suspend these procedures until they could be revised and to evaluate changes needed to its program to prevent recurrence. The failure to maintain procedures covered by Regulatory Guide 1.33 is a performance deficiency. The finding is of more than minor significance because, if left uncorrected, the failure to maintain and control procedures would have the potential to lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Phase 1,\\\"Initial Screening and Characterization of Findings,\\\" the finding was determined to have a very low safety significance because the finding did not result in a loss of system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component, because problems were not thoroughly evaluated, such that the resolutions addressed the causes and extents of condition. This includes properly classifying and prioritizing conditions adverse to quality.
05000361/FIN-2010006-032010Q2San OnofreLack of Preventive Maintenance Results in Valve Failure and Inoperable Condensate Storage TankThe inspectors identified a noncited violation of Technical Specification 3.7.6, which requires, in part, that Condensate Storage Tank T-120 be operable. Specifically, the tank isolation valve 2HV5715 had been inoperable for a period greater than the allowed outage time of seven days while Unit 2 was in Modes 1, 2, and 3. The valve isolates nonseismic piping from the tank and is required to be manually closed within 90 minutes following a seismic event. The licensee had not performed preventive maintenance on the valve resulting in the valve failing to close during an in-service test on January 26, 2010. This finding was entered into the licensee\'s corrective action program as Nuclear Notification 200765235. The licensee\'s corrective actions included repairing the isolation valve. This finding is more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Phase 1, \"Initial Screening and Characterization of Findings,\" a Phase 2 analysis was performed because the condensate storage, Tank T-120, was inoperable greater than that allowed in technical specifications. Phase 2 analysis resulted in a potential greater than Green issue therefore, a Phase 3 was performed. The analyst performed a Phase 3 using San Onofre seismic information and fragility data associated with the piping that could not be isolated because of the failed condition of valve 2HV5715. The frequency of a seismic event that would cause a pipe break and drain tank T-120 was estimated to be 2.7E-5/yr. Given a seismic event that causes a loss of offsite power (nearly 100 percent of seismic events that rupture the piping would also cause a loss of offsite power), operators are compelled by procedure to cool down and initiate shutdown cooling. The amount of water that is protected with valve 2HV5715 failed to open, which includes inventory from tank T-121 and water below the break line in tank T-120, given that operators close the working manual isolation valve within 30 minutes, is more than what is needed to get to shutdown cooling in natural circulation with only 1 of 2 steam generator atmospheric dump valves in operation, even if there is a 4-hour hold time at hot standby. The analyst estimated that the failure probability of operators to cool down and initiate shutdown cooling is 1.0E-2. Therefore, assuming a zero base case, the estimated delta- core damage frequency of the finding is 2.7E-5/yr. (1.0E-2) =2.7E-7/yr. The inspectors also determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee did not ensure that equipment was available and adequate to assure nuclear safety by minimization of long-standing equipment issues in that the valve was not being maintained through a preventive maintenance program.
05000361/FIN-2010006-022010Q2San OnofreFailure to Translate Design Basis Information for Turbine-Driven Auxiliary Feedwater Pump Steam Admission ValvesThe inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion Ill, \"Design Control,\" involving the failure to translate nonconservative errors in calculations and procedures identified during review of external operating experiences. The first example involved the sizing calculation for the condensate storage tank failing to account for effects of auxiliary feedwater pump heat during recirculation. The second example involved the failure to update procedural guidance concerning the adverse effects of placing the low pressure safety injection system into operation following use of the residual heat removal system in the shutdown cooling mode of operation above 200F. This issue was entered into the licensee\'s corrective action program as Nuclear Notification 200886265. The licensee initiated actions to correct its procedure and calculation for each instance. The finding is of more than minor significance because it adversely affects the design control attribute of the mitigating systems cornerstone objective. Using Inspection Manual Chapter 0609.04, Phase 1, \"Initial Screening and Characterization of Findings,\" the finding was determined to have a very low safety significance (Green) because the finding did not result in a loss of system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the operating experience component because the licensee failed to implement and institutionalize operating experience information, including vendor recommendations, through changes to plant processes, procedures, equipment, and training programs.
05000361/FIN-2010006-012010Q2San OnofreInadequate Operability Determination for Turbine-Driven Auxiliary Feedwater Pump Steam Admission ValvesThe inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, \\\"Instructions, Procedures, and Drawings,\\\" involving the failure to follow procedural requirements for performing operability determinations. Specifically, the licensee\\\'s operability evaluation for a degraded turbine-driven auxiliary feedwater pump steam admission valve failed to address all the specified safety functions of the affected component as described in the Final Safety Analysis Report and design basis documents. For exampie, the operabiiity determination incorrectly stated that manual closure of the valves was not a credited safety function and incorrectly assumed nonsafety-related instrument air would always be available to close the valves. This finding was entered into the licensee\\\'s corrective action program as Nuclear Notifications 200869281 and 200887620. The licensee\\\'s corrective actions included re-performing the evaluation and emphasizing with licensee staff the importance of ensuring ali design basis information is considered in operability evaluations. The finding was more than minor because it impacted the Mitigating Systems Cornerstones and its objective to ensure the availability and reliability of equipment that responds to initiating events. Using Inspection Manual Chapter 0609 the issue screened to a Phase 3 analysis because it represented a loss of safety function for greater than the allowed technical specification allowed outage time and it screened to greater than Green using the Phase 2 pre-solved worksheet. The senior reactor analyst determined that this finding was of very low safety significance (Green) based on a bounding calculation which assumed inoperability of the component for a year. The senior reactor analyst determined that the combined significance of these scenarios was a delta-core damage frequency of 1.3E-7/yr and a delta-large early release frequency of 4.2E-8/yr. Therefore the violation was determined to be of very low safety significance (Green). The analyst determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with decision making. Specifically, the licensee utilized unsupportable assumptions in its evaluation that were not consistent with the Final Safety Analysis Report or the valve vendor manual.
05000361/FIN-2010006-072010Q2San OnofreFailure to Establish Component Cooling Water Radiation Monitoring ProceduresThe inspectors identified a noncited violation of Technical Specification 5.5.1.1.a, \"Scope,\" involving the failure to establish procedures for component cooling water system alignments such that leakage of radionuclides to the environment would be monitored during all operational alignments of component cooling water. Specifically, radiation monitors could be aligned to only one train of component cooling water at a time and the licensee\'s procedures had no provision for monitoring the second train when both trains were in-service. This finding was entered into the licensee\'s corrective action program as Nuclear It Notification 200871387, and actions were implemented to require periodic grab sampling of the train which was not being monitored. The inspectors determined that this finding was more than minor because this issue impacted the Public Radiation Protection Cornerstone and its objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, the radiation monitors for component cooling water were not sufficient to ensure adequate release measurements. The inspectors evaluated the significance of this finding using Phase 1 of Inspection Manual Chapter 0609.04 and determined that the finding screened to Inspection Manual Chapter 0609, Appendix D, \"Public Radiation Safety Significance Determination Process.\" The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609, Appendix D, and determined that the finding was of very low safety significance (Green) because dose did not exceed Appendix I criteria. This finding was determined to have a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program in that the plant operators did not have a low threshold for identifying deficiencies in procedures.
05000361/FIN-2010006-042010Q2San OnofreFailure to Report Conditions That Could of Prevented Fulfillment of Safety FunctionThe inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73, \"Licensee Event Report System,\" in which the licensee failed to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. On January 26, 2010, the valve which isolates nonseismic piping from condensate storage tank T-120 failed its in-service test when the hand wheel stem snapped after a leveraging device was used in an attempt to close the valve. This isolation valve, 2HV5715, must be closed within 90 minutes of an operating basis earthquake in order to prevent the loss of condensate storage tank T-120 water inventory from a line break in the nonseismic portion of the condensate system. The failure of this valve resulted in a condition prohibited by Technical Specification 3.7.6 and therefore was reportable. This finding was entered into the licensee\'s corrective action program as Nuclear Notification 200888616, and the licensee was taking actions to send a licensee event report to the NRC for this event. The inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was affected. Specifically, the NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated in accordance with the NRC Enforcement Policy. The finding was reviewed by NRC management, and because the violation was determined to be of very low safety significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation consistent with the NRC Enforcement Policy. This finding was determined to have a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program in that the licensee failed to appropriately evaluate corrective maintenance as a basis for past operability.
05000361/FIN-2010006-112010Q2San OnofreFailure to Secure Loose Items in the Electrical SwitchyardThe inspectors identified a noncitied violation of Technical Specification 5.5.1.1.a involving the failure to follow procedural guidance of S0123-XX-11, \"Switchyard Work Performance.\" Specifically, the inspectors identified temporary equipment stored in the switchyard that was not tethered or otherwise secured in accordance with the procedure. The licensee entered a notification in its corrective action program as Nuclear Notification 200870138, and removed or secured the items. This finding is more than minor because it impacts the protection against the external factors attribute of the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the Inspection Manual Chapter 0609 \"Significance Determination Process,\" Phase 1 Worksheet, the inspectors determined that the finding was of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. This finding also has a human performance crosscutting aspect associated with the work control component in that personnel failed to appropriately plan work activities involving job site conditions which may impact plant structures, systems and components.
05000361/FIN-2010006-132010Q2San OnofreFailure to Meet Action Plan for Substantive Crosscutting IssuesThe inspectors identified a Green finding associated with the licensee\\\'s failure to meet the actions described to the NRC in letters dated April 21,2009, and October 29 and 30, 2009, addressing corrective actions to improve site performance in the areas of human performance and problem identification and resolution. Specifically, 16 actions were not implemented on time and a number of actions were modified from what was previously described, all prior to informing the NRC. These findings were documented in Nuclear Notification 200848923. The inspectors determined that the licensee\\\'s failure to perform actions as documented in its plan to the NRC was more than minor because if left uncorrected could result in a more significant safety concern. Using Inspection Manual Chapter 0609, Appendix M, this finding was reviewed by NRC management and was determined to be of very low safety significance (Green). This finding has a crosscutting aspect in the areas of human performance.
05000261/FIN-2010002-012010Q1RobinsonInaccurate Drawings Result I Loss of RWST Level Indication Due to FreezingA self-revealing non-cited violation of Technical Specification 5.4.1, Procedures, was identified in that the licensee used inaccurate drawings to hang clearances on freeze protection circuits which resulted in the Refueling Water Storage Tank (RWST) level instrument lines freezing. The licensee failed to properly translate the design of the freeze protection circuits to the drawings used in the clearances, causing the RWST level sensing line freeze protection to be unavailable. The licensee removed the clearance, re-energized the freeze protection and level indications were restored. The licensee entered the drawing discrepancy issue into the corrective action program as AR 374561 The disabling of the RWST level instrument freeze protection during the RHR pump work is a performance deficiency. The finding is more than minor because it affected the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events. Specifically, the RWST level instrument line freezing caused the required post accident instrumentation of the RWST to be inoperable. Using Appendix A of the Significance Process (SDP) described in IMC 0609, Mitigating System Cornerstone, this finding was determined to have very low safety significance (Green) because no loss of operability or functionality of the RWST resulted from the level sensing line freezing. There is no cross-cutting aspect of this NCV since the incorrect drawing that resulted in the inaccurate clearance was last revised in 1986 and is not indicative of current licensee performance.
05000261/FIN-2010002-022010Q1RobinsonA Emergency Diesel Generator Fuel Oil Transfer Pump Power Supply Cable Subjected to Continuous Submersion in Water Design DeficiencyThe inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that the licensee failed to maintain a safety-related cable in an environment for which it was designed. Specifically, the A Emergency Diesel (EDG) Fuel Oil Transfer Pump power supply cable was exposed to continuous submersion in water. The licensee removed the accumulated water from the hand hole, resealed, and reinstalled the hand hole cover. The licensee entered the issue into the corrective action program as AR 370343. Failure to maintain a safety related cable in an environment for which it was designed is a performance deficiency. The finding is more than minor in accordance with IMC 0612, Appendix B (Block 9, Figure 2), Issue Screening, because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Specifically, subjecting the A EDG fuel oil transfer pump cable to continuous submersion could, over time degrade the cable and result in failure. In accordance with IMC 0609 (Table 4a), Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency which resulted in a loss of operability or functionality. The cause of the finding was directly related to the problem evaluation cross-cutting aspect in the corrective action program component of the Problem Identification and Resolution area because the licensee did not thoroughly evaluate the condition described in NRC Generic Letter 2007-01 Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients (P.1 (c)
05000261/FIN-2009005-022009Q4RobinsonFailure to Identify Oil Leakage on a Operating Charging PumpThe inspectors identified a Green finding for the licensees failure to identify an oil leak on the A charging pump. This failure was determined to be a performance deficiency with respect to licensee procedure OMM-001-11, Logkeeping, which requires oil leakage be identified and abnormal conditions reported to shift management. The licensee responded by stopping the A charging pump to verify proper oil level. An addition of 6.5 quarts was required to restore the oil level to normal. Additionally, to maintain operability, the licensee established a compensatory action to stop the A charging pump every three days to verify oil level until the oil leak was repaired. The licensee entered the issue into the corrective action program as AR 360876. The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Given the history of continuous operation of the charging pumps for up to 37 days, if the identified oil leak remained uncorrected, a loss of lubrication failure of the A charging pump would occur. The charging pumps are technical specification required equipment and are used in the emergency operating procedures to mitigate the consequences of an event. This finding was determined to be green because no loss of operability or functionality of the A charging pump resulted from the identified oil leakage. The apparent cause of this finding was a failure to implement a procedural requirement to identify and communicate an oil leak to shift management. The inspectors determined no cross-cutting aspect was associated with this performance deficiency
05000280/FIN-2009006-012009Q4SurryFailure to Demonstrate Effective Preventive Maintenance of Safety Injection Check Valves Nor Set Goals and Monitor Under 10CFR50.65(A)(1)The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants, for failure to demonstrate effective preventive maintenance of Unit 1 low head safety injection (LHSI) cold leg check valves in accordance with 10CFR50.65(a)(2) and not establish goals and monitor against those goals in accordance with 10CFR50.65(a)(1). The finding is more than minor because it affected the Barrier Integrity cornerstone objective of providing reasonable assurance that physical design barriers (e.g., reactor coolant system (RCS)) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding affected the LHSI cold leg check valves, which provide an isolation barrier from the high pressure RCS when the SI System is in standby to ensure that the integrity of the reactor RCS boundary is maintained. The finding is also associated with the cornerstone attribute of reactor coolant system equipment and barrier performance. The inspectors determined that this performance deficiency was a separate consequence of the degraded performance associated with the LHSI cold leg check valves. Because of this characterization, the inspectors determined that this issue should not be processed through the Significance Determination Process. Therefore, in accordance with the guidance in NRC Inspection Procedure 71111.12, Appendix D, this issue was determined to be a maintenance rule Category II finding and is of very low safety significance (Green). Based on the assessment performed by the team on the current licensees implementation of 10CFR50.65, the results of the licensees extent of condition review for this finding, and because this finding occurred on November 18, 2007, the team determined that this finding was not indicative of current licensee performance and, therefore, no Cross Cutting Aspect was assigned to this issue. This issue was entered in the licensees CAP as CR02560. The licensee restored compliance by establishing goals and monitoring the system performance against those goals in accordance with 10CFR50.65(a)(1)
05000261/FIN-2009005-042009Q4RobinsonEmergency Diesel Generator Inoperable in Excess of Technical Specifications Allowed Completion TimeA violation of TS 3.8.1. B was identified when the B Emergency Diesel Generator (EDG) was inoperable in excess of the TS allowed outage time. Enforcement discretion was exercised for this violation. No performance deficiency was identified. On April 20, 2009, the output breaker for the B EDG failed to close during the performance of planned surveillance testing. The licensee determined the cause of the breaker failure was due to the rotation of a cotter pin, used to retain a control relay lift linkage, during the previous breaker opening which prevented the lift linkage from returning to the normal position. The licensee entered the issue into the corrective action program as AR 331663 and initiated a root cause and extent of condition review. Based on the failure mechanism, the licensee, using engineering judgment concluded the B EDG had been inoperable for greater than the 7 days allowed by TS 3.8.1.B.4 and Condition C. The last successful breaker closure was March 28, 2009. As discussed in the licensees root cause report, the vendor had previously modified the breakers lifting link assembly. The drive screw/rolled pin that was originally used to retain the relays mechanical lift linkage was substituted with a cotter pin retaining component. This substitution created a design flaw because the cotter pin was susceptible to an unrecognized failure mechanism. The design flaw was reported by Westinghouse in accordance with 10 CFR 21 Reporting of Defects and Noncompliance, on May 28, 2009 (EN 45100). Because the cotter pin substitution was a vendor performed design change, the cause was not reasonably within the licensees ability to foresee and correct, therefore, no performance deficiency was identified. The inspector determined a violation of TS 3.8.1.B occurred since the B EDG was inoperable in excess of the TS allowed outage time (7 days). The inspectors determined that this violation was more than minor because it affected the equipment performance attribute of the Mitigating System cornerstone and because it affects the cornerstone objective of ensuring mitigating system availability. The inspectors determined that the breaker failure was not a performance deficiency because the cause of the failure was not reasonably within the licensees ability to foresee and correct to prevent the failure. Because a performance deficiency was not associated with this issue, it was not subject to evaluation under the formal Significance Determination Process (SDP) using Inspection Manual Chapter 0609. However, an assessment of the significance of the event was performed by the inspectors. This review resulted in the matter being assigned a risk assessment of low to moderate significance. In addition, the licensees risk evaluation found an increase in core damage probability of 2.72 E-6 (also low to moderate significance). The event was mitigated by the redundant A EDG and Dedicated Shutdown Diesel Generator being available to respond to an event. Additionally, the licensee concluded that several response actions to recover the B EDG, such as the discovery of the misaligned relay lift linkage or replacing the affected breaker with a spare could be accomplished in an estimated time frame which ranged from one to four hours. The inspectors reviewed the licensees assessment and corrective actions for the event, and determined they were appropriate to the circumstances. All similar breakers at the Robinson Plant which are susceptible to this failure are scheduled to be modified by May 15, 2010. Prior to implementation of this modification, satisfactory compensatory actions have been implemented which will ensure successful operation of the breaker.
05000261/FIN-2009005-012009Q4Robinsona EDG Fuel Transfer Pump Power Supply Cable Subberged in WaterThe inspectors identified an unresolved item (URI) associated with the submergence of a safety-related cable. The inspectors identified approximately 3 inches of standing water in the manhole which contained the A EDG fuel oil transfer pump cabling. This item is unresolved pending further review and evaluation of the licensees environmental qualifications of the submerged 600V cable. During an inspection of the underground cable manhole/bunkers, the A EDG fuel oil transfer pump power supply cable was identified as being submerged in 3 inches of water. Additional inspection activities are needed to determine if the A EDG fuel oil transfer pump power supply cable is suitable for exposure to submersion in water. Pending the results of this additional inspection an Unresolved Item will be opened and designated as URI 05000261/2009005-01, A EDG Fuel Transfer Pump Power Supply Cable Submerged in Water
05000261/FIN-2009005-032009Q4RobinsonLicensee-Identified ViolationTS 3.3.2 required that PC-953A containment pressure switch channel be placed and maintained in the tripped condition. Contrary to this on June 29, 2009, during repair activities the channel was inadvertently removed from the tripped condition. The cause of the error was inadequate work instructions. The channel was restored to the tripped condition in approximately two minutes. This condition was documented in Condition Report 342793. This violation is of very low safety significance because the condition was promptly corrected in approximately 2 minutes and redundant channels were operable
05000400/FIN-2008003-022008Q2HarrisLicensee-Identified ViolationTS 3.6.2.2 requires that the containment spray additive system be operable with two spray additive eductors each capable of adding sodium hydroxide solution from the chemical additive tank to a containment spray system pump flow. Contrary to this, between October 21, 2007 and May 18, 2008 the licensee was unable to maintain proper sodium hydroxide flow in both eductors of the spray additive system. Additional details are located in section 4OA3 of this report. This was identified in the licensees CAP as AR 00254402
05000400/FIN-2008003-012008Q2HarrisFailure to Properly Categorize Maintenance Rule Functional FailuresThe inspectors identified a non-cited violation (NCV) of 10 CFR 50.65 (a)(2) for the licensees failure to categorize two failures of the condenser vacuum pump effluent radiation monitor (REM-3534) as maintenance rule functional failures and accordingly, failed to monitor the component as required by 10 CFR 50.65 (a)(1). The licensee entered this issue into the Corrective Action Program (CAP) as Condition Report 283579. The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems which responds to initiating events to prevent undesirable consequences. In addition, Example 7.b provided in Manual Chapter 0612, Appendix E, states that violations of Paragraph 10 CFR 50.65 (a)(2), failure to demonstrate effective control of performance or condition and not putting the affected Systems, Structures, and Components (SSCs) in (a)(1), are not minor because they necessarily involve degraded SSC performance or condition. The inspectors determined this finding is of very low safety significance because the REM-3534 is not a risk-significant component and a back-up means of detecting a primary to secondary leak, the steam generator blowdown radiation monitor, was functional during the time periods when REM-3534 was not functional. The finding occurred because of the two missed failures in 2005. All of the failures of REM-3534 since 2005 have been properly counted. Therefore, the cause of this finding was not associated with a cross-cutting area because it is not reflective of current licensee performance. (Section 1R12
05000338/FIN-2006501-012006Q4North AnnaLicensee-Identified ViolationThe following violation of very low concern was identified (Severity Level IV) by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV. 10 CFR 50.54(q) states in part that the licensee may make changes to these plans without Commission approval only if the changes do not decrease the effectiveness of the plans. Proposed changes that decrease the effectiveness of the approved emergency plans may not be implemented without application to and approval by the Commission. Contrary to the above, on October 06, 2000, the licensee made changes to their Emergency Plan with Revision 23, which resulted a decrease in effectiveness (DIE) of the emergency plan. The change replaced the licensees previous standard minimum Protective Action Recommendation (PAR) based on NRC and FEMA guidance, which used the keyhole approach (i.e., evacuate all sectors to 2 miles and downwind sectors 2- 5 miles) with an evacuation to 5 miles in all directions. The change may be overly conservative in such a way as to place members of the public at unnecessary risk during evacuation of an area unaffected by a radiological release which would be more appropriately recommended for sheltering. These changes were not submitted to the NRC for approval prior to implementation. This finding is documented in the licensee's corrective action program as N2006-0568.
05000331/FIN-2005011-022005Q2Duane ArnoldFailure to Comply with the Requirements of 10 CFR 50.59 for a Change to the Procedures for NON-NUCLEAR Heat Class 1 System Leakage Pressure Tests

The inspectors identified that a change to the procedures for the Non- Nuclear Heat Class 1 System Leakage Pressure Tests required prior NRC approval. The procedural changes lengthened the amount of time that the reactor coolant system would exceed 212EF, an unplanned mode change, and added control rod drive scram time testing. These procedural changes resulted in a need for a change in the Technical Specifications (TS) 3.10.1, System Leakage and Hydrostatic Testing Operation.

The finding involved a violation of 10 CFR 50.59, an activity that may impact the regulatory process. Therefore, the finding was evaluated in accordance with the traditional enforcement process. Because the finding was not entered into your corrective action program and you did not restore compliance within a reasonable period of time, a Notice of Violation is being issued. The finding was determined to be of very low safety significance since the System Leakage Test was performed at the end of the refueling outage when the decay heat rate was very low and multiple trains of emergency core cooling systems were available for accident purposes. This finding was determined to be a Severity Level IV violation of 10 CFR 50.59.

05000454/FIN-2005004-052005Q2ByronReview of Missed Ventilation And Filtration System Technical Specification Surveillance RequirementsOn January 13, 2005, during a Nuclear Oversight Audit, the licensee identified that 15 Technical Specifications required ventilation surveillance tests were not performed. The licensee's subsequent root cause evaluation and investigation determined that the missed surveillance tests were due to willful falsification of documents by a non-licensed employee. The licensee's associated extent of condition review identified 12 additional TS required ventilation surveillance tests that were also falsified. Upon performing the 27 falsified surveillance requirements, six failed. The NRC determined that this issue was a violation of Byron Station Technical Specifications. By providing false information regarding the surveillances, the non-licensed employee also caused the licensee to be in violation of 10 CFR 50.9, " Completeness and Accuracy of Information. In addition, the activities of the employee also placed himself in violation of 10 CFR 50.5, " Deliberate Misconduct." The enforcement aspects of this issue were described in the Notice of Violation EA-05-159, "Byron Station - Notice of Violation (NRC Office of Investigations Report No. 3-2005-008," from James L. Caldwell to Christropher M. Crane, dated October 27, 2005. This URI is closed.
05000373/FIN-2004002-062004Q1LaSalleUnauthorized Entry Into Unit 1 694' Reactor Building Raceway HRA by Contract PersonnelOn January 25, 2004, a contract foreman and three contract workers were assigned to conduct outage work associated with a valve located in the reactor building. In preparation for the work, the foreman signed in on a radiation work permit (RWP) associated with entry into HRAs located in the turbine and auxiliary buildings but not for entry into the Unit 1 reactor building raceway described below. The three contract workers signed in on a pre-outage RWP associated with minor maintenance activities which did not permit entry into HRAs. During a walk down of the work, the contract workers could not locate the valve, and the foreman took the contract workers into a posted HRA in the Unit 1 reactor building raceway to locate the valve. Prior to entering the HRA, at least one of the contract workers told the foreman that the contract workers were not signed in on an RWP that permitted entry into HRAs. Before entering the HRA, two of the contract workers were aware that they had not received a briefing by radiation protection personnel for the HRA, a prerequisite for entry into HRAs. Therefore, the actions of the foreman and two contractor workers are considered a willful violation, representing careless disregard of requirements, and the violation has been categorized in accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600 at Severity Level III.
05000373/FIN-2001016-012001Q4LaSalleFailure to Follow Procedure Adequately Comply with Procedureal Requirements to Further Evaluate Common Cause Analysis Outcomes

During this inspection, several examples of procedural non-compliance were identified that were associated with the station corrective action program procedure. An adverse performance trend in procedural compliance appeared to be developing in several cornerstone elements. The specific procedural adherence issues were associated with AD-AA-106 Corrective Action Process Program Procedure, Section 4.6.2.2, Class B Evaluations where the licensee had not implemented the requirement to initiate new condition reports following a class B Common Cause Analysis when a potential adverse trend was validated. One non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified 

The issue was of very low safety significance based on the inspector risk significance screening of this finding in accordance with NRC Inspection Manual Chapter 0610*, Power Reactor Inspection Reports, Appendix B, Thresholds for Documentation. Because the failure to initiate Condition Reports (CRs) when common causes or trends were identified did not have an actual or credible impact on safety, the issue was not evaluated using NRC Manual Chapter 0609, Significance Determination Process. However, the finding was more than minor based on extenuating circumstances (Group 3 Questions). The finding was considered to be a substantive cross-cutting issue because the issue was captured in a number of examples noted in the different functional areas examined during the inspection and across plant departments which indicated an adverse performance pattern.
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