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05000325/FIN-2017009-012017Q2BrunswickInoperability of EDG1 due to Cyclic Fatigue Failure of Hydraulic Fuel Rack ControlGreen . A self -revealing Green non- cited violation ( NCV ) of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, was identified on February 19, 2017, when emergency diesel generator ( EDG ) number one was determined to be inoperable due to an oil leak o n the linkshaft hydraulic control assembly. This violation of regulatory requirement existed from October 27, 2015 u ntil February 20, 2017. The licensee entered this issue in their corrective action program as nuclear condition report ( NCR) 02101084. The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failu re to correct a condition adverse to quality led to the inoperability of EDG1. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At -Power, dated June 19, 2012, Based on Exhibit 2, Q uestion A3, the inspectors determined that a detailed risk evaluation was necessary given the uncertainty over how long EDG1 would have operated while leaking oil. A regional senior reactor analyst (SRA) conducted the risk assessment and screened the issu e to Green based on an increase in risk of less than 1E -6. The inspectors determined that this finding did not have an associated cross cutting aspect because this finding was not reflective of current licensee performance due to enhancements of site procedures guiding creation of work orders.
05000395/FIN-2016007-022016Q4SummerFailure to Correct a Condition Adverse to Quality Associated with a Previously Issued NCVThe inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct a condition adverse to quality associated with a previously issued NCV, 05000395/2012004-02, Inadequate Installation of Unit 1 Service Water Piping and Related Pipe Support. The licensee entered the issue in the correction action program as condition report (CR)-16-04621. The PD is more than minor because if left uncorrected, the reduction in design margin of the pipe support could affect the Unit 1 SW systems ability to mitigate a seismic event. Specifically, Unit 1 service water (SW) piping and support had been impacted by the reduction in design margin and without formally updating the associated drawings and calculations or restoring to the original design, future modifications to the system could impact the systems ability to mitigate a seismic event. Using Manual Chapter 0609 Attachment 04, Initial Characterization of Findings, Table 2, dated October 07, 2016, the finding was determined to adversely affect the External Event Mitigating Systems. The inspectors screened the finding using Inspection Manual Chapter (IMC) 0609, Appendix A, Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012, and determined that it screened as Green (very low safety significance) because the service water system maintained its functionality to mitigate a seismic event. The inspectors determined that the finding had a cross-cutting aspect in the area of PI&R because the licensee did not take effective corrective actions to address this issue in a timely manner (P.3).
05000395/FIN-2016007-012016Q4SummerFailure to Implement Corrective Actions and Restore Compliance for Previous NRCIdentified SLIV NCVThe inspectors identified a cited Severity Level (SL) IV violation of Operating Licensee Condition 2.C.(18) for failure to ensure that conditions adverse to fire protection as noted in a previous NRC-identified SLIV NCV, 05000395/2016001-01, Failure to Implement Adequate Administrative Controls Following a Departure from National Fire Protection Association (NFPA) 80-1973 and Provide NRC Staff Complete and Accurate Information, were promptly corrected. Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for (1) the noncompliance with 10 CFR 50.9 to provide staff complete and accurate information and (2) fire doors DRIB/105A&B currently do not meet self-closing requirements in accordance with the current NFPA 805 licensing basis and no actions were specified in licensees corrective action program to restore compliance. The licensee entered the issue in their corrective action program as condition report (CR)-16-04701. The inspectors determined that the performance deficiency was more than minor because it impacted the ability of the NRC to perform its regulatory oversight function and was dispositioned using traditional enforcement. Because the licensee failed to implement corrective actions and restore compliance in a timely manner, this violation is being treated as a cited violation, consistent with Section 2.3.2. a of the NRC Enforcement Policy. This violation involved traditional enforcement and a cross-cutting aspect was not assigned to this violation.
05000390/FIN-2015405-012014Q4Watts BarSecurity
05000390/FIN-2014007-012014Q4Watts BarSecurity
05000390/FIN-2014004-012014Q3Watts BarFailure to follow scaffold procedure impacts Appendix R operator manual actionsThe inspectors identified a Green non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, for the licensees failure to follow procedure MMTP-108, Erection of Scaffolds/Temporary Work Platforms and Ladders, Revision 8. Specifically, on August 18, 2014, a scaffold was erected in the 1B-B charging pump room and Operations personnel failed to adequately evaluate the scaffold for plant equipment access impairments as required by the procedure. The inspectors determined that the licensees failure to adequately evaluate the completed scaffold for plant equipment access/operability/impairments as required by MMTP-108, Erection of Scaffolds/Temporary Work Platforms and Ladders, Revision 8, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding had the potential to affect the feasibility of performing operator manual actions (OMAs) required for fire safe shutdown in the event of a fire. The finding was evaluated using IMC 0609 Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, and was determined to require a detailed risk analysis because evaluation was beyond the scope of IMC 0609 Attachment 1, Fire Protection Significance Determination Process Worksheet, Phase 2 Quantitative Screening Approach. A bounding analysis was performed by a regional senior reactor analyst using the guidance of IMC 0609, Appendix F because the finding affected the ability to reach and maintain safe-shutdown conditions in case of fire. The analysis determined that the risk associated with the performance deficiency represented an increase in core damage frequency of less than 1E-6/year, a finding of very low safety significance (Green). The cause of the finding was directly related to the aspect of Conservative Bias in the Human Performance crosscutting area because the licensee failed to use decision making practices that emphasize prudent choices over those that are simply allowable when performing the scaffold evaluation.
05000390/FIN-2014004-032014Q3Watts BarLicensee-Identified Violation10 CFR Appendix R, Section III, G, Fire Protection and Safe Shutdown Capability, paragraph 3, states in part that Postulated Alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room, zone under consideration should be provided. Contrary to this requirement, since original construction until refueling outage RF12, in the spring of 2014, the licensee failed to meet the separation requirements of paragraph 2 of this part. Specifically, a postulated fire occurring in either of two fire areas could have resulted in the complete loss of reactor coolant pump seal cooling, resulting in a loss of coolant accident. Inspectors determined that this performance condition was more than minor because it adversely impacted the Initiating Events Cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and performed a significance determination in accordance with IMC 609 Appendix F, phase I and II risk evaluation. The condition was determined to be of very low safety significance (Green) consistent with Task 2.7.5: Screening Check, because the CDF was less than or equal to 1E-6. This condition was captured in the licensees corrective action program as PER 809167.
05000390/FIN-2014004-022014Q3Watts BarFailure to Perform an Adequate Post Maintenance Test Results in Draining of the Unit 1 RWST to a Level Below Technical Specification LimitA Green self-revealing finding was documented by the inspectors for the licensees failure to adequately perform a post maintenance test for Design Change Notice (DCN) 60683, Stage 8, resulting in draining approximately 3300 gallons of radioactive contaminated water from the Unit 1 refueling water storage tank into the auxiliary building. The inspectors determined that the licensees failure to implement an adequate post maintenance test for DCN 60683, install new connections for Fukushima modifications, as required by NPG-SPP-06.9.3, Revision 5, Plant Modification Testing, was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the Design Control attribute of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to implement an adequate post maintenance test resulted in the inoperability of the Unit 1 RWST. Using the screening worksheet of IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the resulting loss of Unit 1 RWST inventory was restored within the Technical Specification allowable time. The cause of the finding was directly related to the aspect of work management in the Human Performance crosscutting area because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority.
05000261/FIN-2014008-012014Q2RobinsonFailure to Take Adequate Corrective Action to Preclude Repetition of a Significant Condition Adverse to Quality Associated with the Steam Generator Tube LeakThe team identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take adequate corrective action to prevent repetition of a significant condition adverse to quality regarding steam generator tube leakage due to poor maintenance practices. Specifically, on February 27, 2014, the C steam generator showed indications of a primary to secondary tube leak due to foreign material that was introduced during the fall 2013 refueling outage. As immediate corrective actions, on March 7, 2014, the licensee shutdown the plant and repaired the leak. This violation was entered into the licensees CAP as nuclear condition reports (NCRs) 683695, 683593, and 683591. The licensees failure to implement appropriate corrective actions to address poor worker practices to prevent recurrence of a steam generator tube leak was a performance deficiency. The finding was more than minor because it was associated with the initiating events cornerstone equipment performance attribute and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, foreign material entered the steam generator and damaged a steam generator tube, which increased the likelihood of a steam generator tube rupture. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At-Power, dated June 19, 2012. The finding screened as Green per Section D of Exhibit 1, Initiating Events Screening Questions, because testing showed that the affected steam generator tube could sustain three times the differential pressure across the tube during normal full power and that the steam generator did not violate the accident leakage performance criterion. The performance deficiency does not have a cross cutting aspect because the last revision of the root cause evaluation was completed in 2011 and it is not indicative of current licensee performance.
05000259/FIN-2014007-012014Q1Browns FerryFailure to Identify the Root Cause of the Failure of the 1B Standby Liquid Control Pump BreakerAn NRC identified non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to adequately identify the root cause for a significant condition adverse to quality as defined in NPG-SPP-22.302 Revision 1, Corrective Action Program Screening and Oversight. Specifically, the licensee initially failed to identify the root cause of the failure of the 1B Standby Liquid Control (SLC) Pump breaker that resulted in the equipment exceeding the Technical Specification Limiting Condition for Operation. The issue was documented in the licensees corrective action program as Service Request (SR) 851718. This performance deficiency was more than minor since it adversely affected the Reactor Safety Mitigating Systems cornerstone objective of availability and reliability of affected equipment. Specifically, the failure to determine the cause of a crack in the breakers phase arc chute that fatigued over time impacted the ability to assign effective corrective actions to prevent recurrence and challenges the reliability of the safety-related equipment to provide required reactivity control capability when required for accident mitigation. The inspectors evaluated the risk of this finding using Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at Power. This determination was based on the evaluation that the inoperable equipment did not concurrently affect a single reactor protection system (RPS) trip signal to initiate a reactor scram, nor did it involve control manipulations that unintentionally added positive reactivity or result in a mismanagement of reactivity by operators. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, in the component of Evaluation, since the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000259/FIN-2014007-022014Q1Browns FerryLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and constituted a violation of NRC requirements which met the criteria of Section 2.3.2 of the NRC Enforcement Policy for being dispositioned as Non-Cited Violations. 10 CFR 50, Appendix B, Criterion III, Design Control, states, in part that, measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that deviations from such standards are controlled. Engineering Document Change 69623 modified plant drawings to add a Furminite injection fitting to 2-FCV-73-81, HPCI Steam Line Warm-up Valve. Contrary to the above, on May 15, 2009, following maintenance performed on 2-FCV-73-81, the licensee failed to reinstall the Furminite injection fitting to the valve resulting in a steam leak determined to exceed the allowable leakage to maintain operability per Technical Specification 3.6.1.3. Using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, the inspectors determined the violation was of very low safety significance (Green) because the penetration was considered a small line (1 to 2 inches) and not expected to contribute to the Large Early Release Frequency. This violation was documented in the licensees corrective action program as PER 56687.
05000269/FIN-2013007-012013Q3OconeeModifications to Fire Doors did not Receive Engineering Equivalency EvaluationsAn NRC-identified Green non-cited violation (NCV) of Oconee Nuclear Station Units 1, 2, and 3 Renewed Facility Operating License Condition 3.D and NFPA 805 was identified for the licensees modification of five fire doors from their tested configurations without performing engineering equivalency evaluations. The licensee entered this issue into the corrective action program as Problem Investigation Program O-13-06900, and declared the door nonfunctional and implemented fire watches in accordance with Selected License Commitment 16.9.5 Fire Barriers. The licensees modification of fire doors from their tested configuration without performing engineering equivalency evaluations was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events (i.e., fire) and it adversely affected the cornerstone objective in that the modifications performed on the five fire doors adversely affected the capability of the doors to provide the required level of fire resistance. The finding was determined to be of very low safety significance (Green) because the fire doors would have either provided a two-hour or greater fire endurance rating, or would have provided a minimum of 20 minutes fire endurance protection; and the fixed fire ignition sources, and combustible or flammable materials, were positioned such that the degraded fire doors would not have been subjected to direct flame impingement. A cross-cutting aspect was not assigned because the performance deficiency did not reflect current licensee performance.
05000269/FIN-2013007-022013Q3OconeeFailure to Identify Ignition Sources and Targets During Initial Fire Scenario DevelopmentAn NRC-identified Apparent Violation (AV) was identified for the licensees failure to comply with the requirements of 10 CFR 50.48(c) and National Fire Protection Association Standard 805 (NFPA 805). The Oconee fire probabilistic risk assessment (Fire PRA) failed to address the risk contributions associated with all potentially risk significant fire scenarios. This finding does not represent an immediate safety concern because the licensee entered the issue in the corrective action program as Problem Investigation Program (PIP) O-13-08059 and PIP O-13-08061 and implemented fire watches as compensatory measures. Failure to comply with the requirements of 10 CFR 50.48(c) and NFPA 805 to address the risk contributions associated with all potentially risk significant fire scenarios was a performance deficiency. This performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone objective of protection against external events (i.e., fire), and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be potentially greater than Green. Therefore, further analysis is required to assess the significance of the finding. The cause of this finding was determined to have a crosscutting aspect of H.4(c) in the Work Practices component of the Human Performance area because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported.
05000269/FIN-2013007-032013Q3OconeeFire Protection Program Change did not Meet Oconee License Condition Requirements for NFPA 805 Chapter ThreeAn NRC-identified Apparent Violation (AV) and associated traditional enforcement violation of Oconee Nuclear Station Renewed Facility Operating License Condition 3.D for Units 1, 2, and 3 was identified for the licensees failure to implement and maintain in effect all provisions of the approved fire protection program (FPP) that comply with 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805. The licensee made a change to the approved FPP involving control of combustible materials when the definition of transient fire loads was revised to exclude fire retardant scaffolding materials as transient fire loads, which would not require the licensee to track these items as combustible fire loads. The licensee also failed to submit the FPP change to the NRC for review and approval prior to implementation which impacted the ability of the NRC to perform its regulatory oversight function. The licensee entered this issue into the corrective action program as Problem Investigation Program O-13-08584. This finding did not represent an immediate safety concern because the licensee implemented compensatory measures in the form of combustible tracking impairments and fire watches in the high safety significant fire zones which contained the scaffolding. Failure to comply with Oconee Operating License Condition 3.D for a change to the approved FPP involving control of fire retardant scaffolding materials was a performance deficiency. This performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events (i.e. fire), and it adversely affected the cornerstone objective in that the change to the FPP had the potential to adversely affect the ability to achieve and maintain safe and stable plant conditions due to the increased transient fire load in the affected fire zones. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), Attachment 4, Initial Characterization of Findings, which determined that an IMC 0609 Appendix F, Fire Protection Significance Determination Process, review was required as the finding affected fire prevention and administrative controls. The performance deficiency applied to most fire zones within the plant because the licensee stopped tracking the use of the fire retardant scaffolding materials. The team determined that a systematic plant-wide assessment effort was beyond the intended scope of the fire protection SDP. Therefore additional analysis is required to assess the significance of this finding. The cause of this finding was determined to have a cross-cutting aspect of H.1(b) in the Decision- Making component of the Human Performance area because the licensee used nonconservative assumptions in the decision making associated with this FPP change. Additionally, the licensees failure to submit the FPP change to the NRC was a traditional enforcement violation. The severity level of the traditional enforcement violation will be assigned based on the significance determination of the associated finding.
05000269/FIN-2013007-042013Q3OconeeFailure to Evaluate Unapproved Combustibles in Accordance With ProceduresAn NRC-identified Green non-cited violation (NCV) of Oconee Nuclear Station Units 1, 2, and 3 Renewed Facility Operating License Condition 3.D was identified for the licensees failure to follow procedures for the control of transient combustible materials. The team identified five examples where the licensee failed to follow procedure Nuclear System Directive (NSD) 313, Control of Transient Fire Loads, in that unapproved combustible materials were stored in fire areas/fire zones without proper evaluation and without appropriate compensatory actions being implemented. The licensee entered these issues into the corrective action program as Problem Investigation Program documents O-13-07896, O-13-07897, O-13-07989, O-13-08051, and O-13-08459; and initiated immediate corrective actions to remove the unapproved combustibles from the identified fire areas/fire zones. The licensees failure to follow procedure NSD 313 for storage of transient combustibles in fire areas/fire zones was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external events (i.e. fire), and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because it only affected the ability to reach and maintain cold shutdown conditions. The cause of this finding was determined to have a cross-cutting aspect of H.4(b) in the Work Practices component of the Human Performance area, because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel did not follow procedures.
05000269/FIN-2013501-012013Q3OconeeLicensee-Identified ViolationTechnical Specification 5.4.1(a), Procedures, required in part that written procedures be established, implemented, and maintained covering the applicable procedures in Regulatory Guide 1.33, Rev. 2, Appendix A, February 1978. Procedure OP/1/A/1102/008, Enclosure 4.35, On Line Valve Lineup for MOV Maintenance, Step 2.5, stated, in part, for the operator to cycle 1LP-22 (1B LPI BWST suction). Contrary to the above, on June 26, 2013, the licensee operator failed to follow written procedure when he closed 1LP-21 (1A LPI BWST suction) which isolated the operable LPI train from the BWST rendered Unit 1 LPI inoperable. The licensee restored the LPI A train to its proper alignment within thirteen minutes. The finding was determined not to be greater than Green because the loss of function of at least a single train did not exceed its TS allowed outage time. The licensee entered the issue into their CAP as PIP O-13-06879.
05000269/FIN-2013501-022013Q3OconeeLicensee-Identified ViolationTechnical Specification 5.4.1(a), Procedures, required in part that written procedures be established, implemented, and maintained covering the applicable procedures in Regulatory Guide 1.33, Rev. 2, Appendix A, February 1978. Procedure OP/0/A/1107/016, Enclosure 4.4, Removal and Restoration of 230KV Switchyard PCB, Step 2.2.4, stated, in part, Ensure locked closed PCB (27) Yellow (Red) Bus Side Disconnect. Contrary to the above, on October 22, 2012, the licensee failed to ensure PCB27 was locked closed. The licensee discovered and corrected this condition on April 24, 2013. The finding was determined to represent a loss of system and/or function which required a risk evaluation by a Senior Reactor Analyst (SRA). The SRA estimated the likelihood of faults that could lead to damage of the disconnect and multiplied these by the change in conditional core damage probability due to a loss of the transformer impacted. Dominant cutsets involved failure of one Keowee hydro unit in conjunction with LOOP sequences, operators failure to recover offsite power, or the Keowee faults within 4 hours, and failure of EFW. The risk impact was less than 1E-7 for the exposure period. In addition, the risk impact of seismic events was estimated not to be a major contributor to the change in risk. Because the risk impact was less than 1E-7, the finding was determined not to be greater than Green. Licensee personnel entered the issue into their corrective action program as PIP O-13-04503.
05000269/FIN-2013007-052013Q3OconeeNon-Compliance to License Condition Requiring Modifications to LPG Tank was not Identified During Transition to NFPA 805The team identified an unsecured 500 gallon (water capacity) LPG storage tank in the Transformer Yard adjacent to the CT1 transformer and the Unit 1 Turbine Building. The LPG storage tank sat on four concrete blocks and did not appear to have an excess flow control valve installed to prevent a large release of propane gas from a ruptured supply line to the Auxiliary Boiler. The licensee had previously identified these problems in PIPs O-06-01385, O-08-02163, O-11-10119 and O-13-03819. Initially, a work request was written to secure the tank but it was later closed without the work being performed. Subsequently, an engineering change request was initiated to address the issue but was never approved. In ONS License Amendment 64 for Unit Nos. 1 and 2 and License Amendment 61 for Unit 3, each units license was amended to state, in part, that The licensee is authorized to proceed and is required to complete modifications identified in Table 3.1 of the NRCs Fire Protection SER dated August 11, 1978. The modifications shall be completed on the schedule specified in Table 3.1. SER Section 3.1.7, states, Propane tanks located outside of the turbine building will be anchored and provided with excess flow valves. Table 3.1 states in part, The modification will be completed by the end of the first refuel outage for any unit which occurs after 6 months from the date of issuance of this Safety Evaluation. A letter from Duke Power Company to the NRC dated June 29, 1979 stated, in part that, The required modification had been completed. When questioned about the current configuration of the tank, the licensee stated that since the transition to their approved NPFA 805 FPP all prior FPP SERs and commitments have been superseded in their entirety by the revised license condition and that the tank was in compliance with the requirements of NFPA 805, Section 3.3.7.1 for the storage of flammable gases located outdoors. Offset distances from the tank to structures, systems or components were judged by the licensee to be sufficient to prevent adverse impact from fires or explosions. The team did not agree with this position and stated that the ONS April 14, 2010 License Amendment Request (LAR) did not address the tank or the non-compliance with the license amendment requirement of 1978. The NRC has requested additional information from the licensee to determine if a prior change to the license, made before the transition to NFPA 805, allowed the tank to remain in its current location without the originally required modifications; and, to determine if the tank had, at one time, been in compliance, but had been improperly relocated under a work order performed in 1986. This issue is unresolved pending NRC review of additional information requested to determine if the issue of concern constitutes a violation of NRC requirements. This issue is identified as URI 05000269, 270, 287/2013007-05, Non-Compliance to License Condition Requiring Modifications to LPG Tank was not Identified During Transition to NFPA 805.
05000324/FIN-2013009-012013Q2BrunswickExtent of Condition Review for NCR 490292In May of 2011 inspectors identified holes, gaps, and other degradations in the four day emergency diesel generator fuel oil tank enclosure. The issue was dispositioned as a white finding in inspection report 05000324, 325/2011014. A root cause was later performed under NCR 490292. The root cause evaluation resulted in NCR 492979 being documented to perform flood protection vulnerability walkdowns as an extent of condition for the fuel oil day tank enclosure degradation. The walkdowns were performed in October of 2011 and documented in NCR 492979 assignment 2. Inspectors noted the following regarding these walkdowns: Engineers identified 23 deficiencies in the Unit 1 Reactor Building and 22 flooding deficiencies in the Unit 2 Reactor Building in their walkdown reports. These deficiencies included corroded and leaking pipe and conduit link seals, water stains and evidence of past leakage below penetrations, cracks in the concrete wall, and an active leak coming from a crack in the wall. FnEngineers identified 26 deficiencies in the common Service Water Building in their walkdown report. These deficiencies included minor leakage from penetration seals, stains on the wall from past leakage, and rust/scaling on the walls. FnEngineers identified 14 deficiencies in the common Emergency Diesel Generator Building in their walkdown report. These deficiencies included stains and evidence of past leakage from junction boxes, links seals, and conduit seals. Several corroded link seals were also identified. NCR 490292 assignment 53 was created to correct the deficiencies documented in the walkdown reports. However, inspectors were not able to verify that all of the documented deficiencies were entered into the CAP in the form of an NCR or WR. Some WRs that documented some of the deficiencies were found, but for example, only 10 of the 26 deficiencies documented for the Service Water Building could be found in the CAP. During the course of this inspection the licensee was unable to provide the inspectors with documentation which would enable the inspectors to verify that all issues identified during the October 2011 and the August 2012 walkdowns had been entered into the CAP. Inspectors need to perform additional inspection, including (1) verify deficiencies identified in the walkdown reports were identified in NCR 490292 assignment 53; (2) verification that adequate corrective action was taken for issues from NCR 490292 assignment 53 that were appropriately documented in the CAP, if any; and (3) review of the results of (1) and (2) above to determine if a performance deficiency exists and if it is more than minor. This issue is identified as URI 05000324, 325/2013009-01, Extent of Condition Review for NCR 490292.
05000324/FIN-2013009-022013Q2BrunswickFailure to Correct a Leaking Service Water Building Pipe Penetration SealWR 00091804 was written on April 8, 2003 for water leaking past a link seal on 30 service water pipe 1-SW-100-30-157 in the Service Water Building. The WR directed repair or replacement of the faulty link seal. The associated work order was not worked until the penetration was injected with a temporary sealant under work order 02034121-19 on May 1, 2012. The work order was then left open to completely replace the degraded link seal, which is currently planned for 2014. The licensee did not have a justification for the timeliness of the associated corrective actions to address the identified concern. The inspectors determined that the leaking seal was a condition adverse to quality and that the failure to correct this condition for 9 years was a performance deficiency and a violation of 10 CFR 50, Appendix B, Criterion XVI which requires conditions adverse to quality to be promptly identified and corrected. In accordance with the part 9900 technical guidance on degraded and nonconforming conditions: If the licensee does not resolve the degraded or nonconforming condition at the first available opportunity or does not appropriately justify a longer completion schedule, the staff would conclude that corrective action has not been timely and would consider taking enforcement action. Pending completion of additional inspection and review of information to determine if the performance deficiency is more than minor, this is identified as URI 05000324, 325/2013009-02, Failure to Correct a Leaking Service Water Building Pipe Penetration Seal.
05000324/FIN-2012005-032012Q4BrunswickInadequate Maintenance Procedure for the EDG Jacket Water Pump Wear Ring TolerancesA self-revealing Green NCV of Technical Specification (TS) 5.4.1a, Procedures, was identified because the licensee did not have an adequate maintenance procedure to perform work on the emergency diesel generator (EDG) 3 engine-driven jacket water pump (JWP). Specifically, between July 25, 1992 and November 15, 2012, Procedure 0CM-ENG528, Gould Engine Driven Jacket Water Pump Model 3736, did not provide the correct tolerances for the EDG JWP wear rings, resulting in the JWP seizure. The licensees corrective actions included replacing the casing wear rings with wear rings with the correct tolerance and revising Procedure 0CM-ENG528. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 572546. The performance deficiency associated with this finding was the failure of the licensee to have an adequate procedure for maintenance on the EDG 3 engine-driven JWP. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate procedure resulted in reduced availability of EDG 3 to repair the engine-driven JWP and reduced reliability of the jacket water system during operation. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding was of very low safety significance because the finding did not affect the design or qualification of a mitigating structure, system and component (SSC), the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding does not have a cross-cutting aspect since the performance deficiency is not indicative of current plant performance. Procedure 0CM-ENG528 included the incorrect tolerances since July 25, 1992.
05000324/FIN-2012005-042012Q4BrunswickInadequate Design of EDG 2 ASSD Switch A1The inspectors identified a Green NCV of 10 CFR 50 Appendix B, Criterion III, Design Control, for failure to assure that the design basis for EDG 2 Alternate Safe Shutdown (ASSD) Switch A1 was correctly translated into specifications and drawings. Specifically, between original EDG 2 installation and September 1, 2012, a wiring discrepancy existed associated with EDG 2 ASSD Switch A1 which resulted in an induced fault that could have impacted the ability to locally control EDG 2 during certain fire scenarios. The licensees corrective actions included correcting the EDG 2 control circuit wiring to ensure it was in accordance with the existing approved design and returning EDG 2 to operable status. The licensee entered this issue into the CAP as NCR 557897. The performance deficiency associated with this finding was the failure to assure that the design basis for EDG 2 ASSD Switch A1 was correctly translated into specifications and drawings. The finding was more than minor because it was associated with the protection against external factors (i.e. fire) attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, an induced fault could have impacted the ability to locally control EDG 2 during certain fire scenarios. Using IMC 0609, Attachment 4, issued June 19, 2012, Initial Characterization of Findings, and IMC 0609, Appendix F, Attachment 1, Part 1: Application of Fire Protection SDP Phase 1 Worksheet, the results of this evaluation required further significance evaluation. A phase 3 analysis was performed by a regional SRA in accordance with NRC IMC 0609 Appendix F. The finding affected the capability to achieve alternate safe shutdown for Unit 1. The result of the analysis was an increase in core damage frequency of <1E-6/year a GREEN finding of very low safety significance. The finding does not have a cross-cutting aspect since the performance deficiency is not indicative of current plant performance. The EDG 2 ASSD Switch A1 wiring discrepancy has existed since original EDG installation.
05000324/FIN-2012005-052012Q4BrunswickInadequate Maintenance Procedure for Fluorescent Lights over Safety-related EquipmentThe inspectors identified a Green finding for the licensee not having an adequate procedure for maintenance on fluorescent lights over safety-related equipment. Specifically, between plant startup and August 29, 2012, the licensee did not have instructions for closing S-hooks on fluorescent lights over safety-related equipment during maintenance on the fluorescent lights. This resulted in over 40 S-hooks open in safety-related buildings which could result in fluorescent lights falling and impacting safety-related equipment during a seismic event. The licensees corrective actions included closing the open S-hooks and adding instructions for closing S-hooks to work order (WO) 431558. The licensee entered this issue into the CAP as NCR 551646. The performance deficiency associated with this finding was the failure of the licensee to have an adequate procedure for maintenance on fluorescent lights over safety-related equipment. The finding was more than minor because if left uncorrected, the deficiencies could lead to a more significant safety concern. If left uncorrected, the failure to provide procedural guidance to close the S-hooks on fluorescent lights over safety-related equipment could lead to fluorescent lights falling on safety-related instruments during a seismic event resulting in a reactor trip. This finding is also associated with the design control attribute of the Initiating Events Systems Cornerstone. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding was of very low safety significance because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the CAP attribute because the licensee did not identify the open S-hook issue completely, accurately, and in a timely manner commensurate with their safety significance during the Fukushima walkdowns.
05000261/FIN-2012008-012012Q4RobinsonFailure to Maintain Fire Load Limits in Fire Area aThe inspectors identified a Green Non-Cited Violation (NCV) of Robinsons License Condition 3. E, Fire Protection Program, for increasing the amounts of in-situ combustibles in several fire zones within fire area A above required limits in NRC approved licensing exemptions. The licensee has entered the finding into the corrective action program as Nuclear Condition Report (NCR) 567284 and established continuous fire watches as well as restricted transient combustibles permits for the fire areas of concern. The failure to properly evaluate increases in the quantities of in-situ combustibles above those stated as upper limits for anticipated fire severity and duration in HBR2 exemption requests and an associated SER was a performance deficiency. This performance deficiency was determined to be more than minor because the finding was associated with the protection against external events (i.e. fire) attribute of the mitigating systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to external events that prevent undesirable consequences. In seven fire zones upper limits for fire severity were exceeded. Using the guidance contained in the IMC 0609 Appendix F, the inspectors concluded that a Phase 3 analysis was necessary because the noncompliance involved fires leading to main control room abandonment. The Phase 3 analysis concluded that this item would be associated with a finding of very low safety significance (Green) because the fire detection systems, fire brigade, fire barriers, and the Dedicated Shutdown Systems were not affected by increase in fire loading in the affected fire areas. The inspectors identified a cross cutting aspect in the decision making component of the human performance area because the licensee did not use conservative assumptions in decision making to demonstrate the proposed action did not affect the validity of the technical basis for granted exemptions.
05000261/FIN-2012008-022012Q4RobinsonFailure to Identify and Correct Deficiencies in the Emergency Lighting System Preventive Maintenance ProgramThe inspectors identified a Green NCV of 10 CFR Part 50.65, Maintenance Rule, for the licensees failure to identify and correct deficiencies in the emergency lighting system (ELS) preventive maintenance program. The licensee entered the issues into their corrective action program as NCRs 567517 and 567632. The deficiency will be mitigated by the operators use of flashlights until the deficiencies are corrected. The licensees failure to identify and correct deficiencies in the ELS preventive maintenance program was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the high failure rate of battery load test resulted in a lack of reasonable assurance that the ELS would perform its design function of providing illumination for 8 hours during fire events. Utilizing IMC 0609, Appendix F, Fire Protection Significance Determination Process, the team assigned the performance deficiency to the Post-fire Safe Shutdown category since it affected systems or functions relied upon for post-fire safe shutdown. The finding was then assigned a low degradation rating since the finding minimally impacted the performance and reliability of the fire protection program element. Specifically, the team noted that operators were required to obtain and carry flashlights. Therefore, the finding screened as having very low safety significance (Green). The team identified a cross-cutting aspect in the corrective action program component of the problem identification and resolution area.
05000324/FIN-2012005-062012Q4BrunswickFire Related Unanalyzed Condition that could Impact Equipment Credited in Safe Shutdown AnalysisTitle 10 of the Code of Federal Regulations, 50.48(b)(1) requires, in part, that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of Appendix R, Section III.G. Appendix R, Section III.G.3 states, in part, that alternative or dedicated shutdown capability be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of 10 CFR 50, Appendix R, Section III.G.2. Contrary to the above, from original plant startup to October 13, 2011, the licensee failed to provide an alternative or dedicated shutdown capability when the requirements of 10 CFR 50, Appendix R, Section III.G.2 were not met. Specifically, the licensees alternative/dedicated post-fire SSD strategy for five FAs failed to ensure alternative shutdown capability because the licensee had not considered the possibility of certain fire-induced spurious actuations of critical components that would potentially result in the loss of equipment required for safe shutdown. Because the licensee committed to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c), and this commitment was documented prior to December 31, 2005, the NRC is exercising enforcement and reactor oversight process discretion for this issue in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. This issue was identified and addressed during the licensees transition to NFPA 805, it was entered into the licensees CAP as NCR 493784, immediate corrective action and compensatory measures were taken, it was not likely to have been previously identified by routine licensee efforts, it was not willful, and it was not associated with a finding of high safety significance.
05000369/FIN-2012008-012012Q3Mcguire
McGuire		Failure to Perform Required Extent of Condition Assessments for Quick Cause Evaluations in accordance with McGuires Quality Assurance ProgramA finding of very low safety significance and associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the licensees failure to perform required extent of condition assessments for Quick Cause Evaluations (QCE) in accordance with McGuires Quality Assurance Program. Specifically, Nuclear System Directive (NSD) 212, Cause Analysis, requires in part that an Extent of Condition review shall be conducted as soon as possible when a QCE is performed. One example included the licensees failure to perform an extent of condition assessment for a QCE of the safety-related NSW system. To address this issue, the license entered PIP M-12-6309 into their CAP. The failure to perform the required extent of condition assessments for QCE in accordance with NSD 212 was considered a performance deficiency. The finding was determined to be more than minor because it adversely affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to evaluate events for extent of condition applicability for the Nuclear Service Water issue (PIP M-12-0106) was not only a failure to follow a procedure requirement, but allowed the station to be susceptible to the existence of similar discrepancies in other systems, units, organizations, programs, processes, components, or trains. The finding was determined to be of very low safety significance (Green) because the finding did not result in a loss of system safety function or a loss of safety function of a single train for greater than allowed technical specification allowed outage time. The team identified a cross-cutting aspect in the work practices component of the Human Performance area, because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel did not follow procedures.
05000424/FIN-2012007-012012Q3VogtleFailure to Identify and Repair an Inoperable Fire Penetration SealAn NRC-identified non-cited violation of Vogtle Unit 2, Operating License Condition 2.G, was identified for the licensees failure to identify and repair an inoperable fire penetration seal. The NCV was associated with the licensees failure to identify and repair Auxiliary Feedwater Pumphouse penetration seal 2-59-031-1 that was missing half of 1 damming board material on one side of the seal. The inoperable fire penetration seal is in a 3-hour fire rated wall of the Auxiliary Feedwater Pumphouse. The licensee took immediate corrective actions to declare the penetration seal inoperable, entered the issue in their corrective action program as CR 467932, established a continuous fire watch, and repaired the penetration seal to its design condition. Additionally, the licensee performed an extent of condition inspection of the Auxiliary Feedwater Pumphouse to verify that there were not any other penetration seals in the same degraded condition. The inoperable fire penetration seal represented a performance deficiency, since the partial missing damming board would be expected to be identified and corrected by the licensee during performance of Procedure 29144-C, Fire Boundaries and Fire Rated Penetration Seals-18 Month Visual Inspection. The finding adversely affected the fire containment capability defense-in-depth element. The finding was determined to be more than minor because it was associated with the protection against external events attribute, (i.e., fire), and degraded the Mitigating Systems cornerstone objective to ensure the availability of systems that respond to initiating events. Using NRC IMC 0609, Appendix F, Fire Protection SDP Phase 1 Worksheet, the inspectors conducted a screening and determined the finding to be of very low safety significance (Green) because the remaining penetration seal depth and damming material provided at least 2- hours of fire resistance. The team identified a cross-cutting aspect in the resources component of the human performance area because the licensee did not ensure that personnel and procedures were available, and adequate to assure nuclear safety. Specifically, because the licensee did not identify any work activities that may have damaged the seal since the completion of the most recent inspection, it was reasonable to assume that the deficiency was missed during the surveillance performed on May 9, 2012.
05000424/FIN-2012007-022012Q3VogtleFailure to Install Freeze Protection for Exposed Fire Protection PipingAn NRC-identified non-cited violation of Vogtle Unit 1, Operating License Condition 2.G, was identified for the licensees failure to provide proper freeze protection for an 8 diameter above ground fire water line located alongside the Unit 1 Main Steam Valve Room. The NCV was associated with exposed fire protection lines and piping that was not provided electrical freeze protection or insulated. Specifically, Design Basis Document DC-2301, Fire Protection Water System Section 3.3.6 stated, in part, that exposed lines shall be electrically freeze-protected and insulated. Vogtles NFPA codes of record, NFPA-14 (1983 Edition) and NFPA-24(1984 Edition), required proper safeguards to be provided to prevent freezing for areas that were unheated and that exposed lines and equipment shall be electrically freeze protected and insulated. The licensee documented the deficiency in their corrective action program as CR482524. No immediate compensatory measures were needed because the temperature at the time of the discovery was well above freezing. The licensees failure to provide freeze protection for the Unit 1 fire protection piping, as required by the design bases document and applicable NFPA codes, was a performance deficiency. The performance deficiency was more than minor because it adversely affected the Mitigating Systems cornerstone attribute of Protection Against External Events and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to provide freeze protection for exposed sections of fire protection piping could result in the unavailability of fire suppression capability during a fire event. In accordance with NRC IMC 0609, Appendix F, Fire Protection (SDP) Phase 1 Worksheet, the inspectors conducted a screening and determined the finding to be of very low safety significance (Green) because temperatures at Vogtle are normally well above freezing and there was a low likelihood of complete loss of suppression capability; therefore, the deficiency was determined to be low degradation. No cross cutting aspect was assigned to this finding because the NRC concluded the finding did not reflect current licensee performance.
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.
05000348/FIN-2013009-012012Q1FarleyDeliberate Failure to Conduct Fire WatchesThe violations occurred between September and December 2011, when four contract employees willfully failed to complete fire watch rounds required to ensure that Farley remained in compliance with 10 CFR 50.48, Fire Protection. In addition, these same employees falsified fire watch logs by annotating that hourly fire watches were completed when in fact they had not been performed. These actions caused FNP to be in violation of 10 CFR 50.48 together with a site implementing procedure requiring roving fire watch patrols, and 10 CFR 50.9(a), requiring the accurate and complete documentation of such fire watches. In this case, the violations did not result in any actual consequences because there was no fire at the facility during the time period. In addition, the potential consequences of missed fire watches were low, due to the different forms of fire mitigation available such as fire detection systems, fire barriers, sprinklers, and fire extinguishers. Furthermore, FNP maintains a staffed fire brigade, and security and operations personnel conduct daily rounds throughout areas of the plant that overlap numerous fire watch areas. However, the actions of multiple employees deliberately failing to complete fire watches and falsifying associated documentation is a concern. As discussed in the NRC Enforcement Policy, deliberate violations are of particular concern to the NRC because our regulatory program is based on licensees, contractors, and their employees acting with integrity. Based on the above, and in light of the interrelationship of the two violations, the NRC has concluded that the violations are appropriately characterized as a Severity Level III problem in accordance with the NRC Enforcement Policy. ...... As discussed in FNPs letter of February 8, 2013, corrective actions included but were not limited to: (1) the prompt initiation of an investigation into the matter, (2) an extent of condition review to determine whether the incident was isolated to one individual, (3) conduct of stand down meetings with the contractor to ensure performance expectations were clearly understood, (4) activities to improve active contractor oversight and control to verify the consistent performance of required fire watches, and (5) other SNC fleet activities to strengthen oversight of supplemental personnel. Based on the above, credit is warranted for the factor of Corrective Action. Therefore, in recognition of your prompt and comprehensive corrective actions to preclude recurrence of similar future violations, I have been authorized, after consultation with the Director, Office of Enforcement, to propose that a civil penalty not be assessed in this case. The NRC has concluded that information regarding the reason for the violations, the corrective actions taken and planned to correct the violations and prevent recurrence, and the date when full compliance was achieved is already adequately addressed on the docket in FNPs letter of February 8, 2013.
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-2012007-022012Q1Browns 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-2012007-032012Q1Browns 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-2012007-042012Q1Browns 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-2012007-062012Q1Browns 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.
05000261/FIN-2011012-012011Q3RobinsonLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Robinson Fire Protection License Condition 2.E states that Carolina Power & Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Fire Protection Safety Evaluation Report dated February 28, 1978, and supplements thereto. The supplemental SER dated August 8, 1984, states that alternate shutdown, pursuant to paragraph III.G.3 of Appendix R to 10 CFR 50, is required for a fire in the emergency switchgear room. Paragraph III.G.3 requires, in part, that alternative shutdown capability be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph III.G.2. Contrary to this, on January 21, 2008, the licensee identified that a credited Appendix R safe shutdown pathway was not available to support safe shutdown, due to an improperly locked gate (see Section 4OA3 of this report). This issue was determined to be of very low safety significance (Green) based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase III Quantitative Screening Approach. This violation was documented in the licensees corrective action program as NCR 280422. Although H.B. Robinson committed, prior to December 31, 2005, to adopt NFPA 805 and change their fire protection licensing basis to comply with 10 CFR 50.48(c), the NRC is not exercising enforcement discretion for this issue, because it does not meet all of the criteria of NRC Enforcement Policy, Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48). Specifically, the inspectors determined that it was not likely that the licensee would have identified this issue as part of its NFPA 805 transition activities.
05000400/FIN-2011011-012011Q3HarrisInadequate Procedure AOP-036.04 for Fire Area 1-A-BAL-C Post-Fire Safe ShutdownThe team identified a non-cited violation of Harris Nuclear Plant Technical Specification 6.8.1.a. for inadequate guidance in fire response abnormal operating procedure AOP-036.04, Fire Areas: 1-A-BAL-C, 1-A-BAL-D, 1-A-BAL-F, 1-G, FPYARD, Revision 17. Specifically, the procedure could not have been performed as written in that, AOP-036.04, Section 3.1, directed operators to implement a step in the procedure that did not exist. The licensee initiated Nuclear Condition Report 489092 to address this issue in the Corrective Action Program and subsequently revised the procedure. The team determined that inadequate fire response procedure guidance was a performance deficiency. This finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and it affected the cornerstone objective of protection against external events (i.e., fire). The team assessed this finding using IMC 0609, Appendix F, Fire Protection Significance Determination Process. The team assigned a low degradation rating to this finding because the abnormal operating procedure deficiency was compensated by available emergency operating procedure guidance, operator experience/familiarity, and training. It was likely that plant operators would have been able to assess plant parameters and would have taken the appropriate actions required to ensure post-fire safe and stable plant conditions. Therefore, this finding was of very low safety significance (Green). The cause of this finding was determined to have a cross cutting aspect in the Human Performance Area, Resources Component, because the licensees validation and verification process did not ensure that the procedure was adequate and accurate.
05000400/FIN-2011011-022011Q3HarrisNoncompliance for Providing Inadequate Procedural Guidance for Post-Fire Safe ShutdownShearon Harris License Condition 2.F, Fire Protection Program states, in part, that Carolina Power & Light Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report (FSAR) for the facility as amended. Section 9.5.1, Fire Protection System, of the FSAR incorporates, by reference, Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1, Revision 3, dated May 7, 1986. o Section C.5.b (2) of Fire Protection Evaluation and Comparison to NUREG- 0800, BTP CMEB 9.5-1 requires one train of systems necessary to achieve and maintain hot standby conditions from either the control room or emergency control station(s) be free of fire damage by providing one of the means described in Section C.5.b (2) (i.e., use of spatial separation, passive fire barriers, and fire detection and an automatic fire suppression system). o Section C.5.b (3), of Fire Protection Evaluation and Comparison to NUREG- 0800, BTP CMEB 9.5-1 requires that alternative or dedicated shutdown capability be provided where the guidelines of Section C.5.b (1) and C.5.b (2) cannot be met. Section 9.5.1.5.4, Quality Assurance Program, of the FSAR states that the fire protection quality assurance program elements are included in Section 17.3 of the FSAR. Section 17.3.1.1, Methodology, of the FSAR states, in part, that the HNP quality assurance program prescribes measures for the control and accomplishment of activities for the operation of safety related and fire protection SSCs. Section 17.3.1.1 also commits to the requirements of 10 CFR 50, Appendix B. 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances. Contrary to the above, the licensee failed to meet the requirements of its documented fire protection program, in that: The licensee failed to protect redundant systems necessary to achieve and/or maintain hot shutdown conditions from the MCR or emergency control station(s) from fire damage by one of the means described in Section C.5.b(2) of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1. The licensee failed to ensure alternative shutdown capability was available for two fire areas where the guidelines for ensuring one redundant train for safe shutdown remain free of fire damage, detailed in Section C.5.b (1) and C.5.b (2) of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5- 1 could not be met. The licensee failed to provide adequate procedural guidance, in that the licensees fire safe shutdown procedure failed to incorporate instructions to alert operators concerning time constraints for restoring cooling to the RCP seals. Additionally, the licensees fire safe shutdown procedure included steps that were not appropriate to the circumstances in that a required procedural step may not have been feasible due to the presence of postulated smoke, under certain conditions. Because this issue relates to fire protection, and the associated noncompliances were resolved by compliance with 10 CFR 50.48(c), the NRC is exercising enforcement and reactor oversight process discretion for this issue in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. Specifically, the licensee entered the noncompliances into their corrective action program and implemented appropriate compensatory measures. The noncompliances were not associated with a finding of high safety significance (Red), the noncompliances were not willful, and the licensee submitted a letter of intent stating its intention to transition to 10 CFR 50.48(c) by December 31, 2005. LER 05000400/2002-004-09, Unanalyzed Condition Due to Inadequate Separation of Associated Circuits; LER 05000400/2004-004-00, Unanalyzed Condition Due to Inadequate Separation of Associated Circuits; and URI 05000400/2005007-01, Fire Response Procedures May Not Be Adequate To Prevent RCP Seal Failure and Subsequent Seal Loss of Coolant Accident For a Fire in Certain Fire Areas, are closed.
05000280/FIN-2011003-082011Q2SurryLicensee-Identified ViolationNUHOMS Certificate of Compliance 1030, Amendment 0, Technical Specifications 2.1.c, Functional and Operating Limits, requires, in part, that the spent nuclear fuel stored in each 32PTH DSC/HSM-H at the Independent Spent Fuel Storage Installation (ISFSI) is to be qualified for four (4) heat load zones designated as Zones 1a, 1b, 2 and 3. Contrary to this requirement, the licensee identified that it failed to properly load fuel assemblies into four NUHOMS Dry Shielded Canisters (DSCs) resulting in the fuel assemblies exceeding the decay heat limit for the loading zones in two of the four center zones. Specifically, the Zone 1a and Zone 1b locations were reversed, resulting in the DSC Zone 1b heat load limits being slightly exceeded (less than one per cent in the worst case) at the time of loading. An evaluation performed by the licensee showed that all of the affected DSCs are currently in a safe condition as loaded in the HSMs. This issue is in the licensees CAP as CR419237, NUHOMS DSCs Loaded to Incorrect Heat Load Limits for Specific Orientation. This Severity Level IV violation is being treated as a non-cited violation (NCV), consistent with Section 2.3.2.b of the NRC Enforcement Policy; specifically, the violation was identified by the licensee, the issue was placed into the licensees CAP, the violation was not repetitive as a result of inadequate corrective action, and the violation was not willful.
05000280/FIN-2011003-012011Q2SurryUnplanned Dilution of Unit 2 RCSOn May 28, 2011, while Unit 2 was operating in Intermediate Shutdown (>200 F, 310 psi), a control room operator noticed a decreasing level trend in the primary grade water tank over the past 2.5 hours. Additionally, it was noted that volume control tank and pressurizer level trends were increasing and charging seal injection flow was 101 gpm with letdown flow of 85 gpm. The licensee entered their abnormal procedure for emergency boration and conducted two emergency borations of the RCS while sampling RCS boron concentration and monitoring shutdown margin. Subsequently, it was identified that the cation demineralizer primary grade header isolation valve, 2-CH-19, indicated closed but was allowing primary grade water to leak by. This caused reverse flow through the cation demineralizer and introduced primary grade water into the RCS via the VCT. The licensee estimated that up to 30,000 gallons of PG water could have entered the RCS. Just prior to this event maintenance was conducted on 2-CH-19 and the valve was returned to service in a condition that allowed the primary grade water leakage flow path described above. The licensee entered this issue into their CAP as CR428947, and initiated Root Cause Evaluation (RCE) 001054. The inspectors require additional information, including the licensees completed investigation in RCE001054, to determine if there is a performance deficiency which is more than minor. This issue is identified as URI 05000281/2011003-01, Unplanned Dilution of Unit 2 RCS.
05000280/FIN-2011003-022011Q2SurryFailure to Classify and Declare a Notification of Unusual EventA Green non-cited violation was identified by the inspectors for the licensees failure to classify and declare a Notification of Unusual Event when conditions warranted as required by 10 CFR 50.54(q) and 10 CFR 50.47(b)(4). The inspectors reviewed IMC0612, Appendix B, and determined that the finding was more than minor because it adversely affected the Emergency Response Organization performance attribute of the Emergency Preparedness cornerstone objective to ensure 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. Since the finding involved a failure to comply with regulatory requirements during an actual event, the inspectors reviewed IMC0609, Appendix B, Sheet 2, and determined that this was a finding of very low safety significance (Green) because it involved the failure to declare a Notification of Unusual Event. The cause of this finding involved the cross-cutting area of human performance, the component of decision making, and the aspect of conservative assumptions and safe actions, H.1(b), because the licensee failed to use conservative assumptions in the decision to not classify and declare the event as an Unusual Event.
05000280/FIN-2011003-032011Q2SurryInadequate Qualification Testing of Fire Barrier Penetration SealsA Green non-cited violation of Surry Units 1 and 2 Operating License Condition 3.I, Fire Protection, was identified by the inspectors for failure to have adequate qualification testing results, as directed by Appendix A to Branch Technical Position APCSB 9.5-1. Specifically, the licensee did not have sufficient testing results to qualify certain aluminum conduit configurations that penetrate 3-hour fire rated barriers separating fire areas containing redundant equipment required for safe shutdown. As part of the corrective actions, the licensee performed testing to determine the qualification of aluminum conduit penetrations, and performed modifications, as appropriate, to restore compliance. 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. Specifically, not having qualification testing results for aluminum conduits that penetrate fire rated barriers adversely affected the fire confinement capability defense-in-depth element because subsequent testing revealed some conduit configurations that did not meet the penetration seal criteria established in Branch Technical Position APCSB 9.5-1. The inspectors used the guidance of NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and determined that the performance deficiency represented a finding of very low safety significance (Green). Specifically, the fire areas in question either contained a non degraded automatic gaseous or water-based fire suppression system, or the exposed fire areas did not contain potential damage targets that are unique from those in the exposing fire areas. Inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance.
05000280/FIN-2011003-042011Q2SurryLicensee-Identified ViolationLicensee Technical Specification, 3.1.B.3, requires, in part, that the spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 deg F. Contrary to this, the licensee identified that this specification was exceeded on Unit 1 on April 17, 2011. The licensee created Engineering Technical Evaluation ETE-SU-2011-0042 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR422769.
05000280/FIN-2011003-052011Q2SurryLicensee-Identified ViolationLicensee Technical Specification, 3.1.B.3, requires, in part, that the spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 deg F. Contrary to this, the licensee identified that this specification was exceeded on Unit 2 on April 17, 2011. The licensee created Engineering Technical Evaluation ETE-SU-2011-0058 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR422778.
05000280/FIN-2011003-062011Q2SurryLicensee-Identified ViolationLicensee Technical Specification, 3.1.B.3, requires, in part, that the pressurizer heatup rate shall not exceed 100 degF per hour. Contrary to this, the licensee identified that this specification was exceeded on Unit 1 on April 20, 2011. The licensee created Engineering Technical Evaluation ETE-CEM-2011-0005 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR423197.
05000280/FIN-2011003-072011Q2SurryLicensee-Identified ViolationLicensee Technical Specification, 3.1.B.3, requires, in part, that the pressurizer heatup rate shall not exceed 100 degF per hour. Contrary to this, the licensee identified that this specification was exceeded on Unit 2 on May 26, 2011. The licensee created Engineering Technical Evaluation ETE-SU-2011-0073 to evaluate the acceptability of the Pressurizer for continued operation. The evaluation concluded that the structural integrity impact of the transient was within design fatigue analysis margin and therefore did not affect Pressurizer operability. The inspectors determined the finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding was a low safety of significance (Green) because there was no actual degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The inspectors determined that licensee correctly evaluated the finding and developed appropriate corrective action as documented in the licensees CAP as CR428788.
05000250/FIN-2011003-012011Q2Turkey PointFailure to properly perform a procedure results in damage to an RHR pumpA self-revealing, non-cited violation (NCV) of Technical Specifications 6.8.1.a, Procedures, was identified when operators did not properly align the RHR system from shutdown cooling mode to injection mode. As a result, the 4A RHR pump was left running with no suction source causing a failure of the pump mechanical seal and minor flooding in the Unit 4, A RHR pump room. The pump was not available for either injection or shutdown cooling operations until the seal was replaced. The issue was documented in the corrective action program as AR 1644427 and a root cause investigation was initiated. Failure to properly align the RHR system to the injection lineup was contrary to plant procedures and was a performance deficiency. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating System Cornerstone and resulted in damage to an RHR pump. The finding was screened using IMC 0609, Appendix A, Phase 1, and because there was no loss of safety function with the alternate RHR pump remaining operable, the finding was determined to be of very low safety significance. The finding affected the cross-cutting area of Human Performance, Work Practices because personnel did not adequately implement error prevention techniques, such as pre-job briefings, self and peer checks, and proper documentation of activities
05000327/FIN-2011002-012011Q1SequoyahReactor Trip due to Unplugged Steam Dump Load Reject ControllerA self-revealing finding was identified for the licensees failure to perform adequate post-maintenance testing, as specified by procedures SPP-8.3, Post- Modification Testing, revision 10, and NPG-SPP-06.3, Pre-/Post-Maintenance Testing, revision 0, in conjunction with a work order which implemented a plant modification on Unit 1 and included the relocation of the steam dump load reject controller. This resulted in a manual trip of Unit 1 following a turbine trip from 26 percent rated thermal power due to the steam dump load reject controller power supply not being properly connected. The licensee entered this issue into their corrective action program as PERs 285349. The licensee implemented corrective actions to include a revision to post-modification testing procedures to require an additional post maintenance testing (PMT) review for large/complex modifications, as well as revision to applicable maintenance procedures to require verification for plugin type connections. The finding was determined to be greater than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, this finding resulted in a reactor trip. Using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because although it did contribute to the likelihood of a reactor trip, it did not contribute to the likelihood that mitigating systems will not be available. The cause of this finding was determined to have a cross-cutting aspect of Work Planning, in the area of Human Performance associated with the Work Control component. The work planning processes failed to identify the need to include steps to verify the operational status of the controller following completion of the activity, considering the physical conditions and requirements associated with relocating the device. (H.3(a)).
05000327/FIN-2011002-022011Q1Sequoyah\\\"Failure to Adequately Qualify Molded-Case Circuit Breakers to Safety-Related Application Through Commercial Grade Dedication\\\"The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that appropriate quality standards were specified and included in design documents and that deviations from such standards were controlled. Specifically, the licensee failed to ensure that the molded case circuit breakers utilized in the station 120VAC vital instrument power boards were properly seismically qualified for their application. The licensee entered this issue into their corrective action program as PERs 264271, 266599, 286156, and 319161. Corrective actions included revision of applicable procedures to perform re-alignment of breakers in the vital instrument power boards. The finding was determined to be greater than minor because it was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that the 120VAC vital instrumentation board components had proper seismic qualification had the potential to affect the ability of safety-related equipment to perform its required function under design basis conditions. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because it did not represent an actual loss of safety function. No cross-cutting aspect was identified, since the issue was determined to not reflect current licensee performance.
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