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05000390/FIN-2016013-042016Q4Watts BarLicensee-Identified ViolationTechnical Specification 3.5.2 Emergency Core Cooling Systems (ECCS) Operating Condition A required, in part, that while in Mode 1 that if one train becomes inoperable that it be restored to an operable status in 72 hours. Condition B required action to place the unit in Mode 3 in 6 hours and Mode 4 in 12 hours if that train is not restored in 72 hours. Contrary to the above, the Unit 1 1B-B CCP was inoperable from July 24, 2016, until August 5, 2016, in excess of the allowed outage time of Condition A without the unit being placed in Mode 3 in 6 hours and Mode 4 in 12 hours as required by Condition B. This issue was documented in the licensees corrective action program as CR 1199024. The finding was screened using IMC 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The finding required a detailed risk evaluation because a single train of CCP was inoperable for greater than its allowed outage time. The regional Senior Reactor Analyst reviewed the inspector provided detailed risk evaluation that was performed using the Saphire SDP module. The finding was determined to be Green.
05000390/FIN-2016013-012016Q4Watts BarFailure to Implement the Program Requirement to Enter Issues into the CAPGreen. The NRC identified a Finding for the licensees failure to consistently implement the program requirements of the CAP. Specifically, the licensee failed to implement NPG-SPP-22.301, section 3.2.2 which required the licensees staff to initiate a Condition Report (CR) to enter various items into their CAP. The licensee placed this issue into their corrective action program. The performance deficiency was more than minor because, if left uncorrected, issues would remain unanalyzed that could represent a more significant safety concern. The performance deficiency was screened using IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Cornerstone dated June 19, 2012. The finding screened to Green because none of the examples were related to any structure, system, component, (SSC) 3 exceeding its technical specification allowed outage time. A cross cutting aspect of Identification was assigned because the licensees threshold for identifying and entering issues into their CAP was not low enough as defined by their procedures. (P.1)
05000391/FIN-2016013-032016Q4Watts BarFailure to Implement Confirmatory Order Requirement for Adverse Employment ActionTBD. The NRC identified an Apparent Violation of Confirmatory Order Modifying License, (EA-09-009,203) Dated December 22, 2009 (ML093510993) for the licensees failure to; (1) implement a process to review proposed licensee adverse employment actions at Watts Bar Nuclear plant before actions were taken to determine whether the proposed action comports with employee protection regulations, and whether the proposed actions could negatively impact the SCWE; and (2) implement a process to review proposed significant adverse employment actions by contractors performing services at TVAs nuclear plant sites before the actions were taken to determine whether the proposed action comports with employee protection regulations, and whether the proposed action could negatively impact the SCWE. The NRC determined this violation constituted a more than minor traditional enforcement violation associated with failure to implement actions required by Confirmatory Order Modifying License, (EA-09-009,203). The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address violations which impede the NRCs ability to regulate using traditional enforcement. The inspector determined that the licensees failure to implement the requirements of the Confirmatory Order had the potential to impede or impact the regulatory process, and therefore subject to traditional enforcement as described in the NRC Enforcement Policy, dated November 1, 2016. The NRC has not made an enforcement decision on this matter.
05000390/FIN-2016013-022016Q4Watts BarFailure to Provide Accurate InformationSL-IV. The NRC identified a Non-cited Violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information for the licensees failure to provide accurate information in all material respects to the Commission. The team determined on April 22, 2016, the licensee provided inaccurate information in a letter to the NRC titled, RESPONSE TO NRC LETTER CONCERNING A CHILLED WORK ENVIRONMENT FOR RAISING AND ADDRESSING SAFETY CONCERNS AT THE WATTS BAR NUCLEAR PLANT (ML16113A228). This information was material because the NRC relied on this information to conclude that TVA was in compliance with CO-EA-09-009/203 requirements. The licensee placed this issue into their corrective action program. The NRC determined this violation constituted a more than minor traditional enforcement violation associated with failure to provide accurate information. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address violations which impede the NRCs ability to regulate using traditional enforcement. The inspector determined that the licensees failure to provide accurate information was a violation of 10CFR50.9 which had the potential to impede or impact the regulatory process, and therefore subject to traditional enforcement as described in the NRC Enforcement Policy, dated November 1, 2016. This violation is characterized as a Severity Level IV violation because it was similar to Example Section 6.9.d.1 of the NRC Enforcement Policy.
05000390/FIN-2015004-062015Q4Watts BarFailure to Identify a Condition Adverse to Quality for Unacceptable Preconditioning of the 1A-A Charging Pump Discharge Check ValveThe NRC identified a NCV of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify a condition adverse to quality. Specifically, the licensee unacceptably preconditioned the 1A-A charging pump discharge check valve 1-CKV-62-525 and failed to identify this as a condition adverse to quality or take appropriate corrective action. The inspectors determined that the performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, unacceptable preconditioning could mask the actual as-found conditions and result in the loss of degradation trending information of component performance. The inspectors determined the finding to be of very low safety significance (Green) because the finding did not result in the loss of operability of 1-CKV-62-525. This finding had a cross-cutting aspect in the area of Human Performance, work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees work management process was not able to prevent the unacceptable preconditioning of the 1A-A discharge check valve even after it was identified as a possibility prior to the planned maintenance.
05000390/FIN-2015004-042015Q4Watts BarShield Building Operability RequirementsThe inspectors identified an unresolved item (URI) associated with the requirements of Watts Bar Unit 1 technical specification (TS) 3.6.15, Shield Building. Additional inspection is required to determine if the requirements of 3.6.15.B applied during a specific testing alignment. On September 10, 2015, the licensee conducted 0-SI-65-6-A, Emergency Gas Treatment System (EGTS) Train A 10-Hour Operation. During the 10-hour time period of the test when the EGTS was in service, the auxiliary gas building treatment system was also in service for a Unit 2 construction test. This unique ventilation combination is not normally experienced during the 0-SI-65-6-A surveillance. As a result, shield building annulus differential pressure fell below the limit established by TS surveillance requirement (TSSR) 3.6.15.1 limits for the entire duration of the 10-hr EGTS surveillance. TS limiting condition for operation (LCO) 3.6.15.B requires annulus pressure be restored when it is outside of limits with a required completion time of 8-hrs. The licensee considered the note associated with TS LCO 3.6.15.B, which states that the annulus pressure requirement is not applicable during ventilating operations, required annulus entries, or auxiliary building isolations not exceeding one hour in duration. The licensee considered the alignment they were in at the time to be ventilating operations and thus the requirements of TS LCO 3.6.15.B did not apply. The licensee further considered that the note, as written, allowed grace from the annulus pressure requirement for ventilating operations for an unlimited amount of time. The inspectors were concerned about a possible allowance in the TS to have grace from annulus pressure requirements for longer than the allowed LCO required action completion time. Furthermore, a basis for the note and what can be considered ventilating operations was not immediately apparent. Because more information is necessary to evaluate the proper applicability of TS LCO 3.6.15.B and the associated note, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to determine if a TS compliance issue exists. This is identified as URI 0500390/2015004-04, Shield Building Operability Requirements.
05000390/FIN-2015004-022015Q4Watts BarAFWST Permanent Plant ModificationThe inspectors identified an unresolved item (URI) associated with the 50.59 screening performed for the installation of the auxiliary feedwater storage tank (AFWST). Additional inspection is required to determine if the plant modification which installed the tank would have required NRC permission in the form of a license amendment prior to the change. The AFWST is a 500,000 gallon source of clean water for the auxiliary feedwater (AFW) pumps. It was installed as part of the licensees post-Fukushima (FLEX) modifications to meet the mitigating strategies order (EA-12-049). The new tank was needed because the licensee determined they could not credit their existing condensate storage tanks (CSTs) for FLEX strategies due to seismic requirements necessary to survive the extended loss of AC power (ELAP) event. The AFWST was connected to the existing condensate system in the AFW supply piping upstream from the AFW pumps and downstream from the CSTs. The modification was evaluated in two separate DCNs, each with its own 50.59 applicability screening. DCN 60060 evaluated the installation of the tank and DCN 61422 evaluated the piping connections to the condensate system. The piping connections included new check valves in the CST piping to prevent AFWST inventory loss in the event the CSTs are damaged in the ELAP event. There were also two air-operated supply valves on AFWST outlet piping which automatically open on low pressure in the downstream condensate piping and also fail open on a loss of power or air. Inspectors noted a number of deficiencies in the 50.59 screening for DCN 61422. Inspectors determined that several potentially adverse impacts were introduced by the modification and were not adequately considered in the 50.59 screening. The licensee re-performed the screening and concluded that the modification would require a 50.59 evaluation due to adverse impacts brought up by the inspectors. Because more information is necessary to properly evaluate the 50.59 evaluation that was completed late in the quarter, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to determine if prior NRC approval was required for the installation of the AFWST. This is identified as URI 05000390/2015004-02, AFWST Permanent Plant Modification.
05000390/FIN-2015004-012015Q4Watts BarFailure to Perform ISI General Visual Examination of Containment Moisture Barrier Associated with Containment Liner Leak-chase Test Connection Threaded Pipe PlugsThe inspectors identified a Green NCV of Title 10 of the10 CFR Part 50.55a, Codes and Standards, involving the licensees failure to properly apply Subsection IWE of American Society of Mechanical Engineers, Section XI, for conducting general visual examinations of the metal-to-metal pipe plugs of the leak-chase channel test connections, installed inside the access box, that provide a moisture barrier to the basemat containment liner seam welds. Following the inspectors identification of this issue, the licensee initiated actions to conduct the required inservice inspection (ISI) general visual examinations. Inspection of the access boxes and leak-chase channels revealed the presence of standing water as well as general corrosion in both locations. The licensee took actions to remove the water and evaluate the condition of the applicable structure, system, and components to verify that containment integrity had been maintained, and would continue to be maintained through the expected life of the plant. The licensee updated the ISI plan such that the required inspections will be performed in the future. The inspectors determined that the licensee had taken adequate immediate corrective actions to address the deficiencies identified, and to ensure the leak-tight integrity of the containment. The issue was entered into the licensees corrective action program (CAP) as Condition Report 1092415. This performance deficiency was of more than minor significance because the failure to conduct required visual examinations and identify the degraded moisture barriers which allowed the intrusion of water into the liner leak-chase channel, if left uncorrected, would have resulted in more significant corrosion degradation of the containment liner or associated liner welds. The finding was associated with the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, visual examinations of the containment metal liner provide assurance that the liner remains capable of performing its intended safety function. The inspectors used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment.
05000390/FIN-2015004-052015Q4Watts BarCore Barrel Lift Error Resulted in Unintended high Dose RatesA self-revealing NCV of TS 5.7.1, Procedures, Programs and Manuals, was identified when the unit one core barrel (CB) was raised above the height limit specified in licensee procedure1-MI-68.003, Removal and Replacement of the Unit 1 Reactor Vessel Lower Internals, Revision 0003. Specifically, step 6.11(20) states in part, ...slowly raise the lower internals package UNTIL the lower internals is at or above EL. 75910 as indicated by the break of the laser indicator on the wall target. On October 5, 2015, while moving the CB from the storage stand to the reactor vessel, the CB was inadvertently lifted approximately three feet higher than the 75910 elevation and required radiation protection (RP) intervention to stop the lift when dose rates in and around containment exceeded anticipated levels. The licensee entered this issue into the CAP as CR 1090220. Corrective actions included stand-downs with each crew to review expectations for critical steps, increased field oversight, and revision of the lift procedure to clarify the steps regarding use of the laser indicator. This finding was determined to be greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance, Program and Processes (procedures for monitoring and RP controls) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The finding was not related to As Low As Reasonably Achievable planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Human Performance, Work Management (H.5) because distractions at the work location contributed to the failure to recognize that the CB had been raised above the procedural limit.
05000390/FIN-2015004-032015Q4Watts BarFailure to Comply with Source Range Neutron Flux Channel Technical Specification RequiprementsA self-revealing non-cited violation (NCV) of Technical Specification (TS) 3.3.1, was identified for the licensees failure to take the actions of Table 3.3.1-1, Function 5, action J.1 to immediately open the reactor trip breakers (RTBs) when two source range neutron flux channels were inoperable with the RTB closed and the rod control system capable of rod withdrawal. Specifically, the licensee failed to identify both required channels of the source range trip function were bypassed and proceeded to withdraw control rods for testing and reactor startup. The performance deficiency was more than minor because it affected the configuration control attribute of the mitigating 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 source range level trip switches were left in bypass, outside of their required configuration, thereby removing a trip function that is required by TS during rod withdrawal. The inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a mismanagement of reactivity by the operators. This finding had a cross-cutting aspect in the area of Human Performance, avoid complacency, because the licensee failed to recognize and plan for the possibility of mistakes and latent issues or use appropriate error reduction tools.
05000390/FIN-2015004-072015Q4Watts BarLicensee-Identified ViolationWatts Bar Nuclear Plant TS 3.6.12 states that the ice condenser inlet doors, intermediate deck doors, and top deck doors shall be operable and closed. TS 3.6.12 Condition B requires that maximum ice bed temperature is verified to be less than 27 degrees F once per four hours (Action B1) when one or more doors is inoperable. Contrary to the above, four intermediate deck doors were inoperable from September 8, 2015 until September 17, 2015 and required action B1 of TS 3.6.12 Condition B was not performed. WBN maintenance personnel erected scaffolding on September 8, 2015 which blocked four intermediate deck doors in the Unit 1 upper ice condenser, which made the doors inoperable since the scaffolding would have prevented them from opening. The TS implications of the scaffold were not immediately recognized and therefore the required TS action B1 was not performed. The licensee identified this condition on September 16, 2015 and took immediate actions to enter TS LCO 3.6.12, Condition B, requiring that maximum ice bed temperature is verified to be less than 27 degrees F once per four hours (Action B1) and to restore the doors to operable status in 14 days (Action B2). The scaffold was removed on September 17, 2015; therefore, the 14-day completion time of TS 3.6.12 was not exceeded. A review of ice bed temperatures between September 8, 2015 and September 17, 2015 showed that ice bed temperatures never exceeded 27 degrees F as required by TS 3.6.12 Action B1. Using IMC 0609, Appendix A, Exhibit 2 (Mitigating Systems); this finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of function of at least a single train of equipment for greater than its technical specification allowed outage time. This violation was entered into the WBN CAP under CR 1082469.
05000390/FIN-2015002-052015Q2Watts BarReview of 10 CFR 50.59 Evaluation for the Emergency Diesel Generator Heat ExchangerThe inspectors identified an unresolved item (URI) regarding the licensees 10 CFR 50.59 evaluation for a modification to the operational configuration of the inlet motor operated valves (MOVs) for the EDG Heat Exchanger. Additional inspection would be required to determine if the licensees 10 CFR 50.59 evaluation properly addressed whether the modification resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structures, systems, or components (SSCs) important to safety previously evaluated in the UFSAR. Watts Bar has four EDGs that are each cooled by two heat exchangers supplied by the ERCW system. Prior to the modification, flow through the heat exchangers was continuous due to the inlet MOVs (1-FCV-067-0066-A, 2-FCV-067- 0066-A, 1-FCV-067-0067-B and 22FCV-067-0067-B) being locked open. In order to ensure sufficient flow is available to components served by ERCW during dual-unit operations, the licensee modified the position of these MOVs from normally open with power removed, to normally closed with breakers closed. This resulted in the EDG heat exchangers being isolated during normal operation from the ERCW system. Flow, however, would be restored by the MOVs active function to open upon receipt of a signal from the EDG speed switch, should the EDGs startup. The inspectors reviewed the results of the licensees 10 CFR 50.59 evaluation related to the impact of the modification on the failure probability of the EDG. The inspectors concluded that additional information and review was necessary to determine whether the modification resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component important to safety previously evaluated in the UFSAR. Particularly, the inspectors needed additional information on the specific inputs, assumptions, and evaluation methodology used to determine the increase in EDG failure probability. This issue was identified as URI 05000390/2015002-05, Review of 10 CFR 50.59 Evaluation for the EDG Heat Exchanger.
05000259/FIN-2013011-052013Q2Browns FerryMaintenance Personnel Not Following Clearance Procedure ViolationThe team identified a Green non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures. The team determined that the maintenance Primary Authorized Employee (PAE) did not verify that all blocking points were danger tagged to ensure worker personal safety and equipment protection for the A2 RHRSW pump planned maintenance. The PAEs decision to only verify two of nine clearance components was a violation of TVA Corporate Procedure NPG-SPP-10.2, Rev. 5, Clearance Procedure to Safely Control Energy . The maintenance PAE did not ensure that the A2 RHRSW pump was isolated from an unexpected release of energy that could have resulted in personnel injury or pump damage. The PAE did not verify or recognize that the A2 RHRSW pump manual discharge valve was full open and not danger tagged closed on May, 6, 2013. This performance deficiency was reasonably within BFNs ability to foresee and correct. This Finding was more than minor because, if left uncorrected the BFN Maintenance Supervisors failure to follow the clearance and tagging procedure requirement to verify all danger tag blocking points, he only verified two of nine danger tags, for the A 2 RHRSW planned pump the performance deficiency would have the potential to lead to a more significant safety concern, such as more severe plant transients, engineered safeguard system malfunctions, and a higher probability of personnel injury. The team determined that this Finding was of very low safety significance (Green) because it did not represent an actual loss of safety function or safety systems out of service for greater than the TS allowed outage time. The team identified a cross-cutting aspect in the Work Practices component of the Human Performance area. Specifically, the licensee ensures supervisory and management oversight of work activities such that nuclear safety is supported.
05000259/FIN-2013011-022013Q2Browns FerryFailure to Follow Procedure during Implementation of Plant Modifications to the Residual Heat Removal and Core Spray SystemsThe team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for the licensees failure to maintain effective configuration control as required by Procedure NPG-SPP-09.3, Rev. 13, Plant Modifications and Engineering Change Control. Specifically, the licensee partially implemented permanent plant modifications to the Residual Heat Removal (RHR) and Core Spray (CS) systems under Design Change Notices (DCN) 69466 and 69467 and left these DCNs in partially implemented status beyond two refueling outages without approval of the Vice President of Engineering. This created the potential for a loss of configuration control of the CS and RHR systems. The licensee entered this issue of concern in their corrective action program as SR 739929 and PER 740729 that included actions to evaluate completion or cancellation of the remaining portions of the DCNs. The team determined the Finding was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The finding was of very low significance because it was not a design or qualification deficiency, and it did not result in an actual loss of one or more trains of the RHR or CS systems and/or their function. The finding had a cross-cutting aspect in the area of Human Performance, Work Control because the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of partially implemented DCNs on the plant.
05000259/FIN-2013011-032013Q2Browns FerryFailure to Perform 10 CFR 50.59 Evaluation Intergranular Stress Corrosion Cracking Examination on ASME Code Class 1 Piping WeldThe team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, for the licensees failure to perform an evaluation of a change to the facility as described in the Updated Final Safety Analysis Report (UFSAR) and an associated Green Finding for the licensees failure to perform an acceptable Ultrasonic (UT) examination in accordance with American Society of Mechanical Engineers (ASME) Code, Section XI requirements. Specifically, this change resulted in a departure from the method of evaluation used to inspect for intragranular stress corrosion cracking (IGSCC) in reactor coolant pressure boundary components at BFN as described in the UFSAR and therefore, required a 10 CFR 50.59 evaluation to determine if the change would have required a license amendment request pursuant to 10 CFR 50.90. The licensee performed the required 10 CFR 50.59 evaluation and entered this issue of concern in their corrective action program (CAP) under SR 743380 and PER 744849. The team determined the underlying PD was more than minor and a Finding, because the PD affected the Barrier Integrity cornerstone and if left uncorrected, could become a more significant safety concern. Absent NRC identification of this PD, the licensee could have continued to perform UT examinations to detect IGSCC on safety-related components without obtaining the minimum required examination volume. This could result in IGSCC susceptible welds on ASME Code Class 1 piping being only partially examined for IGSCC flaws and could lead to safety-related components with potentially unacceptable service-induced flaws not detected during UT examinations being returned to service. The team evaluated the Findings significance in accordance with IMC 0609, Appendix G, Shut-down Operations Significance Determination Process, because the PD occurred while Unit 2 was in cold shutdown. The team reviewed IMC 0609, Appendix G, Attachment 1, Checklists 5, 6, 7, and 8 and determined this Finding did not require a quantitative assessment. Therefore the Finding screened as having very low safety significance. The team determined the traditional violation was more than minor because of reasonable likelihood the departure from weld inspection methodology as described in the UFSAR would have required Commission review and approval prior to implementation. The team concluded that the violation of 10 CFR 50.59 was a Severity Level IV because the underlying PD screened Green under the SDP. The team also concluded that this Finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because the licensee did not make safety significant or risk-significant decisions using a systematic process when faced with uncertain or unexpected plant conditions, to ensure safety was maintained.
05000259/FIN-2013011-162013Q2Browns FerryFailure to Establish Qualified Ultrasonic Examination ProceduresThe team identified a NCV of 10 CFR 50, Appendix B, Criterion IX, Control of Special Processes for the licensees failure to control non-destructive examination (NDE) activities by not having qualified NDE procedures required by applicable codes, standards, specifications, criteria, and other special requirements. Specifically, four Ultrasonic (UT) examination procedures did not contain any of the required American Society of Mechanical Engineers (ASME) Code Section XI, Appendix VIII essential variables or the explicit requirement to perform the UT examinations using applicable Performance Demonstrated Initiative (PDI) procedures. The licensee initiated prompt corrective actions to revise all UT implementing procedures to become qualified in accordance with ASME Code Section XI, Appendix VIII requirements and entered the issue into their corrective action program (PERs 730250 and 721446). The Finding was more than minor, because it affected the Initiating Event cornerstone and if left uncorrected, could become a more significant safety concern. Absent NRC identification of this PD, the licensee could have continued performance of UT examinations on safety-related components using unqualified procedures. Performance of UT examination using unqualified procedures could lead to safety-related components with unacceptable service-induced flaws being returned to service without ASME codespecified evaluation or repair. The team determined the Finding was of very low significance because the Finding was not likely to result in exceeding the RCS leak rate for a small loss of coolant accident (LOCA) or cause total loss of function for a LOCA mitigating system. This Finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience (OE) because the licensee did not implement and institutionalize OE pertaining to UT examination procedure issues through changes to station processes, procedures, and training programs to support plant safety.
05000259/FIN-2013011-152013Q2Browns FerryDeficient Design Control for RHR Service Water Freeze ProtectionThe team identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to maintain adequate design control measures associated with the residual heat remove service water (RHRSW) system freeze protection. Specifically, the team identified that freeze protection was not installed on two RHRSW pump air relief valves (ARV) to maintain operability of the RHRSW system during extended periods of cold weather. BFN entered the issue into their corrective action program under SRs 731375, 727908, and 732519 and PER 732519 and concluded that an immediate operability concern was not present due to the current warm weather conditions and recent satisfactory pump testing. Additionally, BFN performed a detailed inspection of ARVs on all 12 RHRSW pumps, and identified deficiencies on ARVs for eight pumps and entered each item into the CAP. The team determined that failure to maintain adequate design control measures associated with the RHRSW system freeze protection was a performance deficiency. This Finding was more than minor because it adversely affected the design control attribute of the Mitigating Systems cornerstone and the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the Finding was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating system, structure or component (SSC), where the SSC maintained its operability. The Finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program problem identification, because BFN did not maintain a low threshold for issue identification such that this issue was identified and resolved during numerous previous focused inspections of the RHRSW system configuration.
05000259/FIN-2013011-142013Q2Browns FerryFailure to Implement an Adequate Test Program for RHRSWS and EECSThe team identified a non-cited violation of 10CFR50, Appendix B, Criterion XI, Test Control, because the licensee did not establish a test program for Residual Heat Removal Service Water (RHRSW) and Emergency Equipment Cooling Water (EECW) pumps such that the test adequately demonstrated the pumps would perform satisfactorily in service. Specifically, BFN did not perform RHRSW/EECW pump performance testing such that it adequately accounted for river water temperature impact on the pump lift, which affected pump flow and vibration performance. The test program did not account for changes to pump lift caused by river water temperature changes; as a result the test program did not adequately monitor pump and system performance and degradation. The licensee completed a prompt operability determination verifying that the pumps remained operable and documented the issue in PERs 730497 and 741036. The Finding was more than minor because at affected the Mitigating System Cornerstone and if left uncorrected, could become a more significant safety concern. The team determined the Finding was of very low safety significance because it was not a design or qualification deficiency, and it did not result in an actual loss of one or more trains of the RHRSW or EECW systems and/or their function. The Finding had a crosscutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not to thoroughly evaluate the changes in RHRSW and EECW pump performance such that the resolution addressed the causes and extent-of-condition.
05000259/FIN-2013011-132013Q2Browns FerryFailure To Translate The Design Into Procedure 3-SR-3.3.8.2.1(B)The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate seismic uncertainties into acceptance criteria and measuring and test equipment accuracy requirements into the Reactor Protection System circuit protector calibration surveillance procedure. This was determined to be a performance deficiency. Prompt corrective actions included determination that the equipment remained operable and entry of this issue into their corrective action program as problem evaluation report 723605 and 730495. The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern because it could have affected the operability of the relays. The team used Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, for mitigating systems, and Inspection Manual Chapter 0609, Appendix. A, The Significance Determination Process for Findings at Power, issued June 19, 2012, and determined the Finding to be of very low safety significance (Green) because the Finding did not result in the loss of functionality or operability of a structure, system, or component. The team did not identify a cross-cutting aspect because this performance deficiency has existed since 2006 and is not indicative of current licensee performance.
05000259/FIN-2013011-112013Q2Browns FerryInadequate Corrective Actions to Address Programmatic Procedure Quality IssueThe team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, due to BFNs failure to take corrective action to preclude repetition of a significant condition adverse to quality regarding procedure quality. Specifically, BFN self-identified corrective actions implemented to address inadequate procedures but did not identify and address a significant contributor to the inadequate procedures, resulting in several additional plant performance issues. The team identified multiple inadequate procedures across most BFN departments during the inspection document review and onsite inspection. BFN has conducted root causes, developed and implemented numerous corrective actions; however, procedural deficiencies continued to contribute to plant shutdowns, unplanned component unavailability, and rework activities. BFN documented the issue in PERs 680792 739429, and 740212. This Finding was determined to be more than minor because it associated with the human performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit this likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the Finding was of very low safety significance because it 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 (e.g. loss of condenser, loss of feedwater). The team determined that the Finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because BFN did not thoroughly evaluate the extent of condition associated with inadequate procedures such that the corrective actions resolved the issue and prevented repetition.
05000259/FIN-2013011-102013Q2Browns FerryRequirements for Concurrent Verification, Independent Verification, and Peer ChecksThe team identified a Green, non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures. The team determined that BFNs Requirements for Concurrent Verification, Independent Verification, and Peer Checks were not consistently applied to plant procedures, instructions, and work documents as required by TVA Corporate Procedure NPG-SPP-10.3, Rev.1, Verification Program, and regulatory requirement ANSI N18.7-1976/ANS-3.2, Administrative Controls and Quality Assurance for Operational Phase Nuclear Power Plants. BFN documented the issue in SRs 722559, 726755, and PERs 707531, 722859, and 727405. This finding was more than minor because, if BFN site verification procedure requirement issues and adherence are left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern, such as more severe plant transients, or engineered safeguard system actuations or malfunctions. Additionally, this issue is similar to Example 4.b in IMC 0612, Appendix E, Examples of Minor Issues, in that the recent inadequate use of human performance error prevention tools (self-checking, peer checking, and missing IVs and CVs in the Procedure NPGSPP- 10.3, Appendix A, list of 35 BFN systems that are required to have verifications for procedures, instructions, and work documents) have resulted in a reactor scrams, unplanned safety and risk significant system inoperability and unavailability, or other transients. The Finding was determined to be of very low safety significance (Green) in accordance with Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, The Significant Determination Process (SDP) for Findings At-Power, because it did not represent an actual loss of safety function or safety systems out of service for greater than the TS allowed outage time. The team identified a cross-cutting aspect in the Resources component of the Human Performance area, because the licensee did not ensure that procedures were available and adequate to assure nuclear safety. Specifically, accurate and up-to-date procedures, work packages, and correct labeling of components.
05000259/FIN-2013011-092013Q2Browns FerryFailure to control a modification to the seismically mounted control room ceiling light diffusersThe team identified a Green, NRC identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to control deviations from the as built control room envelope design for seismically mounted ceiling light diffusers in accordance with instructions that assure quality standards are controlled. Specifically, contrary to the procedure the licensee unsecured three seismically mounted control room ceiling light diffusers and slid them over the top of other light diffusers creating a seismic missile hazard that could have impacted control room ventilation damper actuators. Once the licensee understood that unfastening the ceiling light diffusers and sliding them over top of other diffusers was creating unanalyzed modifications, the licensee removed the ceiling diffusers from the overhead and placed them in a seismically safe condition. In addition, the licensee clarified the procedure step to have the ceiling light diffusers removed completely. The licensee entered this issue into their CAP as PER 730443. The failure to control a planned modification of the seismically mounted control room ceiling light diffusers was a performance deficiency (PD). The PD was more than minor because it is associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1-Initial Screening and Characterization of Findings, the team determined that the Finding had very low safety significance (Green) because the Finding only represents a degradation of the radiological barrier function for the control room. This Finding has a cross-cutting aspect in the area of human performance because the licensee did not define and effectively communicate expectations regarding procedural compliance and personnel follow procedures.
05000259/FIN-2013011-082013Q2Browns FerryFailure to Manage Emergent Risk Condition during A1 and A2 RHRSW InoperabilityThe team identified a self-revealing, Green non-cited violation (NCV) of 10 CFR 50.65 (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, due to BFNs failure to adequately manage the impact of an emergent risk condition related to the A1 residual heat removal service water (RHRSW) quarterly surveillance test. BFN recognized the online maintenance risk condition however, failed to implement appropriate risk management actions (RMAs) in accordance with Procedure BFN-ODM-4.18, Protected Equipment. The A and B emergency diesel generators were required to be protected. BFN entered this issue into their corrective action program (CAP) as SR 730356. Specifically, on May 6, 2013, with the A2 RHRSW pump inoperable for planned maintenance, the A1 RHRSW pump was declared inoperable during the A1 RHRSW pump quarterly test due to a tagging error that resulted in Assistant Unit Operators closing and danger tagging the A1 pump manual discharge valve instead of the required A2 pump discharge valve. Upon starting the A1 RHRSW pump, control room alarms provided the operators indication of a system problem, and in the course of responding to the alarm, the operators noted the danger tag. The tags were removed and the pump was declared inoperable to fill and vent the system prior to returning it to an operable status. This issue was entered in to the corrective action program as PER 722859 and 731570. The team determined that BFNs failure to adequately manage the impact of an emergent risk condition related to the A1 residual heat removal service water (RHRSW) quarterly surveillance test was a performance deficiency that was reasonably within BFNs ability to foresee and correct. The performance deficiency was determined to be more than minor and a Finding because, if the deficiency was left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to take adequate RMAs could have led to unplanned inoperability of redundant TS or risk significant mitigating systems being relied upon to respond to initiating events to prevent undesirable consequences. The performance deficiency was also determined to be more than minor since it is similar to more than minor Example 7.e of Inspection Manual Chapter (IMC) 0612, Appendix E Examples of Minor Issues. The Finding was evaluated in accordance with Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, of IMC 0609, Significance Determination Process, and was determined to be of very low safety significance (Green). This Finding has a cross-cutting aspect in the area of Human Performance, Work Control, because BFN failed to implement immediate RMAs and communicate to the station personnel the change in plant risk condition and protected equipment requirements that may affect work activities.
05000259/FIN-2013011-042013Q2Browns FerryTwo BFN Assistant Unit Operators Closed and Danger Tagged the A1 RHRSW Pump Manual Discharge Valve Instead of the Required A2 RHRSW Pump Discharge ValveThe team identified a Green, self-revealing non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures. The team determined that BFNs clearance and tagging application related to the planned A2 residual heat removal service water (RHRSW) pump maintenance was not properly applied and verified as required by TVA Corporate Procedures NPG-SPP-10.2, Rev. 5, Clearance Procedure to Safely Control Energy, and NPG-SPP-10.3, Rev.1, Verification Program. Two BFN assistant unit operators (AUOs) closed and danger tagged the A1 RHRSW pump manual discharge valve instead of the required A2 RHRSW pump discharge valve on May, 6, 2013. Upon starting the A1 RHRSW pump, control room alarms provided the operators indication of a system problem, and in the course of responding to the alarm, the operators noted the danger tag. The tags were removed and the pump was declared inoperable to fill and vent the system prior to returning it to an operable status. This issue was entered in to the corrective action program as PER 722859. The performance deficiencies were reasonably within BFNs ability to foresee and correct. This Finding was more than minor because it was associated with the human performance attribute which occurred when the AUOs closed and tagged the wrong RHRSW pump discharge valve. The AUOs errors adversely affected the Mitigating System cornerstone objective of ensuring the availability, reliability, and capability of the RHRSW and RHR systems that respond to initiating events to prevent undesirable consequences. The team determined that this Finding was of very low safety significance (Green) because it did not represent an actual loss of safety function or safety systems out of service for greater than the TS allowed outage time. The team determined that this Finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because BFN AUOs did not use self-checking and peer checking human error prevention techniques to prevent the inadvertent closure and danger tagging of the A1 RHRSW pump manual discharge valve instead of the required A2 RHRSW pump valve during the application of a tagging clearance.
05000259/FIN-2013011-062013Q2Browns FerryConduct of Operations Procedure ViolationThe team identified a Green, non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures. The team determined that assistant unit operators (AUOs) failure to comply with Procedure OPDP-1, Rev. 26, Conduct of Operations, Sections 4.2 K. and M., related to the missing A1 RHRSW pump discharge valve label plate and the AUOs inadequate walkdown of the A1 RHRSW pump prior to the planned quarterly surveillance test pump start on May 6, 2013, were performance deficiencies that were reasonably within BFNs ability to foresee and correct. Immediate corrective actions by the licensee included revising the conduct of operations procedure, and enter the issue in the corrective action program as PERs 13161, 701486, and 722859. This Finding was more than minor because, if TVAs failure to follow the Procedure OPDP-1 requirements was left uncorrected, the performance deficiencies would have the potential to lead to a more significant safety concern, such as more severe plant transients, or engineered safeguard system actuations or malfunctions. Additionally, this issue is similar to Example 4.e in IMC 0612, Appendix E, Examples of Minor Issues, in that the A1 RHRSW pump discharge valve was missing the valve label plate and AUOs did not stop the A2 RHRSW pump clearance application to correct the valve label issue prior to proceeding with the danger tag application. This action was required by TVA Corporate Procedure OPDP-1, Rev. 26, Conduct of Operations, and resulted in an improper valve manipulation due, in part, to the missing label plate. The team determined that this Finding was of very low safety significance (GREEN) because it did not represent an actual loss of safety function or safety systems out of service for greater than the TS allowed outage time. The team determined that this Finding had a cross-cutting aspect in the area of Human Performance, Work Control. Specifically, the licensee plans and coordinates work activities, consistent with nuclear safety. In addition, the licensee appropriately coordinates work activities by incorporating actions to address: the impact of changes to the work scope or activity on the plant and human performance, 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, the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities.
05000259/FIN-2013011-072013Q2Browns FerryFailure to Adequately Implement Procedure 3-SR-3.3.8.2.1(B)The team identified a non-cited violation of Technical Specification (TS) 5.4.1, which requires written procedures be established, implemented, and maintained covering activities referenced in NRC Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978, including surveillance tests. Specifically, a performance deficiency occurred, when the licensee failed to implement the procedure, which required that approved measuring and test equipment be used to measure the underfrequency relay settings during the performance of the Reactor Protection System circuit protector calibration surveillance procedure. Prompt corrective actions included determination that the equipment remained operable and entry of this issue into their corrective action program as problem evaluation report 731144. The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern, because it could have affected the operability of the relays. The team used Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, for mitigating systems, and Inspection Manual Chapter 0609, Appendix. A, The Significance Determination Process for Findings at Power, issued June 19, 2012, and determined the Finding to be of very low safety significance (Green) because the Finding did not result in the loss of functionality or operability of a structure, system, or component. The team identified a crosscutting 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.
05000259/FIN-2013011-012013Q2Browns FerryFailure to Perform Evaluation of Nonconforming Material during Commercial Grade Dedication of Safety-Related BearingsThe team identified a Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion III, Design Control in that the licensee did not adequately evaluate a commercial grade dedication (CGD) of bearings prior to installing the bearings in a safety-related low pressure coolant injection (LPCI) motor generator (MG) set. Specifically, BFN did not perform an acceptance evaluation of non-conforming materials as required by Section 3.2.6 of NPG-SPP-04.2, Material Receipt and Inspection, Rev. 2. The licensee subsequently initiated prompt corrective actions that included an evaluation of acceptance of the installed bearings, a LPCI operability determination, an extent-ofcondition review, and entered the issue in their corrective action program (PER 729646). The Finding was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally the Finding was similar to Example 5.c in Appendix E of IMC 0612. The Finding was of very low significance because the finding was a design qualification deficiency and the affected structure system component (SSC) (3EN LPCI MG set) maintained its operability. This Finding had a cross-cutting aspect in the area of Human Performance, Decision Making because the licensee did not use conservative assumptions when making the decision to accept non-conforming commercial grade bearings for safety-related use, such that nuclear safety was supported.
05000259/FIN-2013011-122013Q2Browns FerryDeficient Acceptance Criteria for Main Battery Bank 1The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to incorporate appropriate quantitative acceptance criteria into a station battery inspection Procedure. Specifically, Procedure EPI-00248-BAT005, Annual Inspection of 250V DC Main Battery Banks 1, 2, 3 and Associated Chargers, Revisions 18 and 19 did not provide the correct acceptance criteria for the battery bank connection resistance results. Prompt corrective actions included determination that main battery bank 1 remained operable and entry of the issue into the corrective action program (SR 731341 and PER 732511). The team determined that BFNs failure to establish correct quantitative acceptance criteria after main bank battery replacement and after changing the battery inspection methodology in the annual battery test inspection procedure was a performance deficiency. The performance deficiency was determined to be more than minor and a Finding because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The Finding was of very low safety significance (Green) because it was not a design or qualification deficiency and did not result in an actual loss of system and/or function. The Finding had a cross-cutting aspect in the area of Human Performance, Resources - Procedures, because BFN did not provide accurate and up-to-date procedures for the inspection of safety-related station batteries.
05000269/FIN-2010005-042010Q4OconeePotential Inoperability of the Unit 3 Standby Shutdown Facility Reactor Coolant Makeup PumpDuring the performance of a quarterly In-service Testing surveillance test of the Unit 3 SSF RCM pump on August 24, 2010, the licensee observed an increase in the normal reactor building sump level and subsequently found to have been caused by seat leakage in relief valve 3HP-404. It was identified that while leakage during the test was less than 0.5 gpm, a gradual increase in leakage through the relief valve had occurred during tests performed over the preceding 13 months. The SSF RCM Pump was used to provide makeup to the reactor coolant system (RCS) and RCP seal cooling during an SSF-design basis event. Leakage through the relief valve would reduce the amount of water that would reach the RCS and could result in RCP seal failure or the inability of operators to control RCS inventory if the relieve valve leakage continued to increase. The licensee will conduct testing to determine if the leakage rate would degrade enough over the 72-hour mission time of the SSF to prevent the SSF RCM pump from performing its safety function. The licensee has replaced the relief valve. This issue is identified as URI 05000287/2010005-04: Potential Inoperability of the Unit 3 Standby Shutdown Facility Reactor Coolant Makeup Pump.
05000269/FIN-2010005-012010Q4OconeeFailure to Adequately Protect Risk Significant And Safety-Related Systems, Structures or Components (SSCs) from Cold Weather ConditionsAn NRC-identified non-cited violation (NCV) of TS 5.4.1.a was identified for the licensees failure to implement procedures to ensure equipment associated with cold weather protection of risk significant and safety-related systems, structures or components (SSCs) was in-service and functional prior to the onset of cold weather. This issue was entered into the licensees corrective action program as PIP O-10-9308. Corrective actions taken include expediting maintenance on equipment determined to be non-functional, assigning an individual as a cold weather protection point-of-contact and revising/developing procedures to ensure similar deficiencies do not occur in the future. The licensees failure implement cold weather procedures was a performance deficiency (PD). The PD was more than minor because, if left uncorrected, it would have the potential to become a more significant safety concern in that safety-related or risk significant SSCs could be adversely affected by cold ambient temperatures. The finding was of very low safety significance (Green) because the finding did not result in the likelihood of a reactor trip at the same time that mitigation equipment or associated functions would not be available. The finding involved the cross-cutting area of Human Performance under the Management Oversight aspect of the Work Practices component in that the licensee failed to provide the appropriate management oversight to ensure the activities required to prepare the plant for cold weather conditions were completed prior to the onset of cold weather.
05000269/FIN-2010005-022010Q4OconeeFailure to Prescribe Procedures for Inspecting the East Penetration Room Floor SealsAn NRC-identified finding was identified for the licensees failure to verify the operability of the East Penetration Room (EPR) expansion joint floor seals for all three units since 2006. Selected Licensee Commitment (SLC) Surveillance Requirement (SR) 16.9.11a.7 required the licensee to verify the operability of auxiliary building (AB) floor seals every eighteen months. The licensees failure to ensure that the required EPR expansion joint floor seal inspections were performed as required by SLC SR 16.9.11a.7 was a PD. The PD was more than minor because, if left uncorrected, it would have the potential to become a more significant safety concern in that the floor seals could further degrade and affect the function of the flood outlet devices (FOD) to protect safety-related related equipment from flooding after a HELB in the EPR. The inspectors determined that the finding was of very low safety significance (Green) because the degradation the EPR floor seals did not result in the loss of operability or functionality of equipment they were designed to protect. The cause of this finding was directly related to the complete, accurate, and up-to-date design documentation, procedures and work packages aspect of the Resources component of the Human Performance cross-cutting area, in that, procedures and work packages to perform the surveillance were not updated following the FOD modification.
05000269/FIN-2010005-032010Q4OconeeFailure to Install Structural Rebar as Required by Instructions and DrawingsAn NRC-identified non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to adhere to drawings and instructions during the installation of rebar in QA-1 structures prior to concrete placement. The inspectors identified two examples where rebar installation did not meet the concrete coverage requirements specified in ACI Code 117-06. This violation has been entered into the licensees corrective action program as PIPs O-10-9091 and O-10-9351. The licensees failure to follow approved drawings and instructions for construction of QA-1 structures was a PD. The PD was more than minor because, if left uncorrected, insufficient concrete coverage on the rebar could lead to rebar corrosion and challenge the integrity of the QA-1 structures under construction. The finding was of very low safety significance (Green) because the finding did not result in the actual loss of function of the PSW duct bank, the Emergency Condensate Cooling Water pipe, or the PSW Building roof. The finding was directly related to the cross-cutting area of Human Performance under the Procedural Compliance aspect of the Work Practices component because the licensee failed to effectively ensure workers followed procedures and written guidance in the performance of their activities.
05000335/FIN-2010005-022010Q4Saint LucieFailure to Take Timely and Effective Corrective Actions for ECCS Fan Damper FailuresThe inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failing to take timely and effective corrective actions for Emergency Core Cooling System (ECCS) area exhaust fan damper louver failures resulting in TS Limiting Conditions for Operation (LCO) entries for an inoperable ECCS area exhaust air filter train. Specifically, multiple damper failures occurred over at least a two year period where the root cause of the failures was not identified and corrected to prevent recurrence. The finding was more than minor because it is similar to Example 4.f in IMC 0612, Appendix E, in that the failure to adequately correct a condition adverse to quality affected the 1-HVE-9A ECCS area exhaust fans operability. The finding was evaluated in accordance with IMC 0609.04, Significance Determination Process (SDP) Phase 1 screening worksheets and determined to be of very low safety significance because the finding did not represent a degradation of the radiological barrier function provided for the auxiliary building, or represent a degradation of the control room barrier function, or an actual open pathway of containment, or a reduction in function of containment hydrogen ignitors. The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not thoroughl evaluate the problem such that the resolution addressed causes, as necessary.
05000269/FIN-2010005-052010Q4OconeeLicensee-Identified ViolationTS 3.3.7, Engineered Safeguards Protective System Digital Automatic Actuation Logic Channels, required that eight Engineered Safeguards Protective System digital automatic actuation output logic channels shall be operable. TS SR 3.3.7.1 required that a digital automatic actuation output logic channel functional test be performed on a 92 day frequency. Contrary to the above, on September 17, 2010, the licensee identified that the channel functional test for 1) Keowee Emergency Start, 2) Load Shed and Standby Breaker Initiate, and 3) Standby Bus Feeder Breaker signal were not being performed on the 92 day frequency required by TS SR 3.3.7.1. The licensee entered TS SR 3.0.3 which required the licensee to perform a risk evaluation and manage the risk of the delayed performance of the surveillance until the surveillance could be completed. The inspectors verified the licensee performed the risk evaluation and managed the risk appropriately. The licensee developed a procedure to perform the digital automatic actuation logic channel functional test for the three functions on-line and each function was determined to be operable. The licensee entered the violation into their CAP as PIP O-10-7227.
05000269/FIN-2010005-062010Q4OconeeLicensee-Identified Violation10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings requires, in part, that activities affecting quality shall be prescribed by documented instructions or procedures appropriate to the circumstances. Contrary to the above, on September 23, 2010, during post maintenance testing following the stator repair of Keowee Hydro Unit (KHU) 1, the licensee identified that post maintenance test procedure, PT/0/A/0610/024, Keowee Emergency Start for Troubleshooting and Post Maintenance Checkouts, did not properly verify that TS SR 3.8.1.9 acceptance criteria was met. The same deficiency was also found to exist in the routine emergency start surveillance test procedure, PT/0/A/0620/016, Keowee Hydro Emergency Start Test. Based on subsequent analysis of the actual test data, the licensee concluded that operability of the KHUs had been maintained. Following identification of the procedural deficiency, the licensee developed corrective actions to revise the procedures to ensure the test acceptance criteria is appropriately defined so that future tests ensure operability was verified during their performance. The licensee entered the finding into their CAP as PIP O-10-7357.
05000335/FIN-2010005-012010Q4Saint LucieFailure to Identify and Correct a Condition Adverse to Quality that Resulted in the 1C-AFW Pump Being Out of Service for Greater Than Its Allowed Outage TimeA self-revealing Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to promptly identify and correct a condition adverse to quality (CAQ) that resulted in the 1C Auxiliary Feedwater (AFW) pump being inoperable for greater than its Technical Specifications (TS) allowed outage time (ACT). Specifically, in December 2009, the licensee identified a concern with housekeeping in both Unit 1 and Unit 2 AFW pump areas that could affect the pump motor, bearings, seals, and turbine controls and linkages. Then in June 2010, these same housekeeping issues combined with extended operation of the atmospheric dump valves (ADV5) caused failure of the 1 C AFW pump to reach rated speed during its scheduled surveillance test. The finding was determined to be more than minor because it is similar to Example 4.f in IMC 0612, Appendix E, in that the failure to adequately correct a CAQ affected the 1C-AFW pumps operability and affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capacity of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated in accordance with IMC 0609.04, Significance Determination Process (SDP) Phase 1 screening worksheets. Because it represented an actual loss of safety function of a single train for greater than its TS ACT, SDP Phase 2 worksheets were evaluated. The phase 2 notebook produced an overly conservative result for a short exposure time (less than 2 week duration), and consequently a phase 3 SDP evaluation was performed. The resultant core damage frequency (CDF) was <1E-6 Green. The inspectors determined that the cause of this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
05000335/FIN-2010005-032010Q4Saint LucieLicensee-Identified ViolationTS 3.0.3 requires that when a limiting condition of operation (LCO) is not met, except as provided in the associated action requirements, within 1 hour, action shall be initiated to place the unit in a mode which the specification does not apply. Contrary to this, from June 17-24, 2010, two auxiliary feed water pumps were not operable, and actions were not taken to place the unit in the required mode of operation. This was identified in the licensees CAP as condition report 2010-1 6485 and Unit 1 LER 02010-007-00. The analyst determined the finding was of very low safety significance (<1E-6) Green.
05000369/FIN-2010004-032010Q3Mcguire
McGuire		Failure to Determine the Cause And Take Corrective Action to Preclude Repetition For Control Room Area Chilled Water SystemAn NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to determine the cause of a significant condition adverse to quality involving both trains of Control Room Area Chilled Water System (CRACWS) being out of service at the same time. This resulted in insufficient corrective action to preclude repetition. The licensee reopened the root cause investigation to determine the cause and was resolving the high cycle fatigue issue on the hot gas bypass line. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern in that failing to identify corrective actions to preclude repetition could result in the loss of safety function of more risksignificant equipment such as emergency diesel generators. This finding was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. This finding was associated with the cross-cutting aspect of supervisory and management oversight in the Work Practices component of the Human Performance area because managements establishment of the scope and reviews of the completed root cause evaluation failed to provide adequate oversight to ensure the cause of a significant condition adverse to quality was determined and corrective actions were taken to preclude repetition. (H.4(c)) (Section 4OA3
05000413/FIN-2010004-022010Q3CatawbaFailure to Report a Permanent Change in Medical Condition for a Licensed OperatorAn Unresolved Item (URI) concerning a potential failure to report a permanent change in medical condition for a licensed operator as required by 10 CFR 55.25 was identified. During a review of a random sample of medical records and associated documents, the inspectors identified that the licensee potentially failed to report a permanent medical condition of an individual who was a licensed operator. The inspectors determined that a licensed operator had been diagnosed with a permanent medical condition on September 11, 2006. The licensed operator notified the licensee of the medical condition on the next day. A member of the licensees medical staff submitted the medical details to the corporate medical director, and questioned if NRC notification was required. On September 13, 2006, the corporate medical director determined that the requirements of ANSI/ANS 3.4 were met and no NRC notification was required. On September 22, 2008, the licensee sent a license renewal letter to the NRC, including NRC Form 396. The only restriction identified on NRC Form 396 was that the individual required corrective lenses to be worn during performance of his licensed operator duties. On November 13, 2008, based on the information provided by the licensee, the NRC renewed the individuals operator license with one restrictive condition identified, specifically that the individual shall wear corrective lenses while performing licensed operator duties. The licensee entered the issue of concern into its corrective action program as PIP C-10-04465. This issue is unresolved pending receipt of further information from the licensee, including NRC Form 396s for the above individual and any other licensed individuals identified during an extent-of-condition review. Once the additional information is received, an NRC medical review will occur, and a determination will be made if this issue constitutes a performance deficiency that is more than minor: URI 05000413,414/2010004-02: Failure to Report a Permanent Change in Medical Condition for a Licensed Operator.
05000369/FIN-2010004-012010Q3Mcguire
McGuire		Failure to Update the UFSAR For a Modification to the VG SystemAn NRC-identified SL-IV NCV was identified when the licensee did not update the Updated Final Safety Analysis Report (UFSAR) for a modification to the emergency diesel generator air start system (VG) on both units. This modification installed cross-connect piping between the two VG receivers on each emergency diesel generator to allow maintaining receiver pressure when an air compressor was out of service. Licensee corrective actions include updating the UFSAR and Design Basis Documents and processing a Technical Specification (TS) change to make the TS applicable to the crossconnected configuration. This violation is in the licensees corrective action program as PIPs M-10-5299 and M-10-5504. This performance deficiency was considered as traditional enforcement because not having an updated UFSAR hinders the licensees ability to perform adequate 10 CFR 50.59 evaluations and can impact the NRCs ability to perform its regulatory function such as license amendment reviews and inspections. This violation was determined to be a SL-IV violation using Section 6.1 of the NRCs Enforcement Policy because the inaccurate information was not used to make an unacceptable change to the facility. Cross-cutting aspects are not assigned to traditional enforcement violations. (Section 1R04
05000369/FIN-2010004-022010Q3Mcguire
McGuire		Failure to Update the UFSAR For New EDG Tripping FunctionsA NRC-identified SL-IV NCV of 10 CFR 50.71(e) was identified when the licensee failed to update the UFSAR following a modification that installed new protective functions for the emergency diesel generators (EDGs). This violation is in the licensees corrective action program as PIP M-10-05718 This performance deficiency was considered as traditional enforcement because not having an updated UFSAR hinders the licensees ability to perform adequate 10 CFR 50.59 evaluations and can impact the NRCs ability to perform its regulatory function such as license amendment reviews and inspections. This violation was determined to be a SL-IV violation using Section 6.1 of the NRCs Enforcement Policy because the inaccurate information was not used to make an unacceptable change to the facility. Cross-cutting aspects are not assigned to traditional enforcement violations. (Section 1R17
05000413/FIN-2010004-012010Q3CatawbaFailure to Adequately Control Transient Combustible Material in Accordance with the Fire Protection ProgramAn NRC-identified Green NCV of the Fire Protection Program (FPP) was identified when transient combustible materials of greater than 15 pounds and located near an ignition source were stored in the Unit 2 electrical penetration room without prior review and approval as required by NSD 313, Control of Combustible and Flammable Material. The issue was entered into the licensees corrective action program as Problem Investigation Program report (PIP) C-10-5521. The performance deficiency was more than minor because it was associated with the Initiating Events cornerstone attribute of fire and adversely affected the cornerstone objective in that the adjacent 600V pressurizer heater breaker panel could ignite the combustibles and cause damage to safety-related containment pressure transmitters. The finding was determined to be of very low safety significance (Green) because the transient combustibles did not involve low flash point liquids or self igniting material. This finding was associated with the cross-cutting aspect of the licensee defining and effectively communicating expectations regarding procedural compliance in the Work Practices component of the Human Performance area because the requirements of NSD 313 were not clearly communicated (H.4(b)).
05000390/FIN-2010006-022010Q2Watts BarWorst Case 6900 VAC Bus Voltage in Design CalculationsThe team indentified an Unresolved Item (URI) regarding calculations that supported the degraded voltage protection scheme. The calculations that analyzed the Class 1E 6900 VAC and 480 VAC motor loads take credit for administratively limiting the minimum 161kV offsite power supply bus voltage and credit performance of the nonsafety- related automatic load tap changers on the common station service transformers (CSSTs) to limit the minimum voltage on the Class 1E 6900 VAC and 480 VAC buses. The calculations did not evaluate the Class 1E 6900 VAC and 480 VAC motor loads at the worst possible case low voltages which could drop as low as the bottom end of the acceptable tolerance band of the degraded voltage relays. Offsite power is normally provided to the Class 1E 6900 VAC buses from the 161kV offsite power system through the CSSTs. The CSSTs have non-safety automatic load tap changers which are designed to maintain approximately 6900 VAC on the Class 1E buses through a dynamic range of 161kV offsite power supply voltages. The Class 1E 480 VAC buses are then powered from fixed-tap 6900/480VAC transformers powered from the respective Class 1E 6900 VAC buses. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Appendix 8-A, Branch Technical Position PSB-1: Adequacy of Station Electric Distribution System Voltages, Rev. 2 (07/1981) is part of the licensing basis for the Watts Bar Nuclear Plant. This document states, in part, that the selection of under-voltage and time-delay setpoints shall be determined from an analysis of the voltage requirements of the Class 1E loads at all onsite distribution levels. Watts Bar calculation WBNEEBMSTI060029, Degraded Voltage Analysis, Rev. 31, evaluated transient motor starting voltages at the beginning of a design basis loss of coolant accident (LOCA). This calculation was based on the voltages where the minimum 161kV offsite power supply bus voltage was limited by taking credit for administrative controls rather than assuming a worst-case 161kV offsite power supply voltage drop which would still allow voltage recovery to the degraded voltage relay reset setpoint (minus setpoint tolerance) before the expiration of the degraded voltage relay nominal 10 second time delay, and thereby leave the Class 1E 6900 VAC buses connected to the offsite power supply. In addition, calculations for motor starting during steady-state conditions credited voltage improvement based on performance of the non-safety related CSST automatic load tap changers instead of being based on worst-case conditions. Summary. This issue is unresolved pending further inspection to determine (1) the actual worst-case voltage required to be analyzed on the Class 1E 6900 VAC and 480 VAC buses for safety-related loads in accordance with the facility licensing basis; and (2) the impact of not using the worst-case bus voltage afforded by the degraded voltage protection scheme in safety-related 6900 VAC and 480 VAC motor starting studies. Additionally, this issue is very similar to a URI reported in the Sequoyah Nuclear Plants inspection report 05000327,328/2010007-01. (URI 05000390/2010006-02, Worst Case 6900 VAC Bus Voltage in Design Calculations
05000390/FIN-2010006-012010Q2Watts BarInadequate Assessment of Seismic Qualification of ERCW StrainersThe team identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to update ERCW strainer mounting (seismic/structural) calculations to reflect the as-built conditions, a failure which was allowed to exist since commercial operations began. This calculation was then used in making acceptance conclusions for a modification installed in recent months. The licensee entered this condition into their corrective action program as Problem Evaluation Reports (PERs) 221018, 220754, and 223677 and took immediate actions to determine the seismic acceptability of the current installation, utilizing calculational conclusions of a similar installation at the licensees Sequoyah Nuclear Plant. The finding was determined to be more than minor because it was associated with the design control attribute within the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, in that there was reasonable doubt as to the operability of the ERCW strainers as a result of the performance deficiency. The team evaluated the finding to be of very low safety significance (Green) utilizing IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings worksheet, as it was a calculational error subsequently determined to not result in an operability issue. No cross-cutting aspect was identified since the issue was not reflective of current licensee performance
05000348/FIN-2010002-022010Q1FarleyFailure to Re-Evaluate Significant Changes in Assumption to Prompt Operability Determination of Unit 2 TDAFW PumpThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for the failure to implement procedure NMP-AD-012, Operability Determinations and Functionality Assessments. Specifically, the licensee failed to revise the existing prompt determination of operability (PDO) as required by NMP-AD-012 for the Unit 2 Turbine Driven Auxiliary Feedwater (TDAFW) pump when significant non-conservative changes in water content of oil samples challenged assumptions used to establish pump operability. This issue was entered into the licensees CAP as CR 2010101426. The finding is more than minor because it is associated with the reactor safety mitigating systems cornerstone attribute of equipment performance and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, this finding was analogous to MC0612, Appendix E example 3.j in that a reasonable doubt about the continued operability of the pump existed prior to further evaluation. This finding was assessed using the Phase 1 screening worksheets of Appendix 4 of MC 0609, SDP and determined to be of very low safety significance because the finding did not result in the loss of safety function of a single train or screen as risk significant due to external events. This finding was assigned a cross-cutting aspect in the Resources component of the Human Performance area in that complete, accurate and up-to-date design documentation, procedures, work packages, and correct labeling of components were not provided (H.2(c)). Specifically, the oil sampling program procedures and methods lacked the detail and rigor necessary to verify assumptions in the PDO and called into question the continued operability of the TDAFW pump
05000348/FIN-2010002-032010Q1FarleyViolation of Technical Specification 5.4.1 for Failure to Maintain Procedures for Full Flow Recirculation After a Loss of Coolant AccidentThe inspectors identified a Green NCV of TS 5.4.1 for the failure to maintain emergency procedure FNP-1/2-ESP-1.3, Transfer to Cold Leg Recirculation, Rev. 19. ESP-1.3 contained a step to verify containment sump level was sufficient to adequately cover the containment sump screens prior to initiating cold leg recirculation following a loss of coolant accident (LOCA) which led to a full flow recirculation. The containment sump level specified by the procedure was not sufficient to ensure suction vortexing and air ingestion into the emergency core cooling system (ECCS) would have been prevented. This finding was entered into the licensees corrective action program as condition report (CR) 20101101103. Planned corrective actions included issuing a standing night order to ensure adequate containment sump level is verified prior to transferring to cold leg recirculation and formally changing the value in ESP-1.3. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems (containment spray and residual heat removal) that respond to initiating events (LOCAs which lead to full flow recirculation phase) to prevent undesirable consequences (i.e., core damage) and the cornerstone attribute of Procedure Quality, i.e. Operating (Post Event) Procedures (EOPs). The team assessed this finding using the SDP and determined that the finding was of very low safety significance (Green) because the inspectors determined that there was no loss of safety system function. Safety system function was determined to be maintained since the analyzed LOCAs in the accident analysis of the facility updated final safety analysis report (UFSAR) would introduce sufficient water into the containment from ECCS and the reactor coolant system (RCS) to provide sufficient containment sump level to ensure water level above the sump screens to prevent air introduction. This finding was reviewed for cross-cutting aspects and none were identified since the performance deficiency has existed since initial operation and is not indicative of current licensee performance
05000348/FIN-2010002-042010Q1FarleyFailure to Maintain Control of Combustible MaterialA self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to maintain combustible material a distance of 35 feet or greater from the hot work area as required by station procedure FNP-0-AP-38, Use of Open Flame. Unit 1 entered a Notification of Unusual Event (NOUE) emergency action level on December 12, 2009, when a fire occurred in the CCW heat exchanger/pump room. The fire occurred below and to one side of the scaffold near the service water (SW) supply to the 1A CCW heat exchanger isolation valve Q1P16V003A. The cause of the fire was combustible material left in the work area by licensee personnel performing lead abatement on piping supports for a plant modification. Welding personnel had later entered the area, performed welding/grinding activities, then placed work-related material in a concentrated area under the work area. The licensee entered this performance deficiency into their CAP (CR 2009114825) for resolution. The finding was more than minor because it adversely affected the protection against the external factors attribute of the IE cornerstone to limit the likelihood of those events upsetting plant stability and challenge critical safety functions during shutdown, as well as power operations. Specifically, this finding resulted in upsetting plant stability and potentially affected plant safety-related equipment. This finding was assessed using the Phase 1 screening worksheets of Appendix 4 and Appendix F of MC 0609 SDP, and determined a Phase 2 analysis was required. Fire Damage State (FDS) 0 was assigned to the actual fire and any postulated fires due to the performance deficiency. FDS 0 indicated that no functions failed as a consequence of these fires. In the actual fire there was no functional damage to any target. Also, the peak heat release had happened and passed when the fire was extinguished. Consistent with Inspection Manual Chapter 0609, Appendix F, a maximum heat release rate of 200 KW was selected for the postulated transient combustible fires. No targets were observed in the zone of influence where the combustible material was located. Under step 2.2 of Appendix F performance deficiencies associated with FDS 0 fires were not analyzed in the Fire Protection SDP as a risk contributor. Therefore, the finding was determined to be of very low safety significance (Green). A contributing cause of the finding is the failure of supervisory personnel to ensure the area was free of combustible material as required by FNP-0-AP-38 and the actual open flame permit. Therefore, this finding was assigned a cross-cutting contributing cause related to the Human Performance work-practices component, and its aspect of the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (H.4(c))
05000348/FIN-2010002-012010Q1FarleyFailure to Control Combustible Material in a No Intervening Combustible Allowed Area of the PlantAn NRC-identified NCV of License Condition 2.C.(4) was identified for the failure to control combustible material in the Unit 1 Component Cooling Water (CCW) Pump area as required by the licensees administrative controls program. Workers left combustible material in the area of the 1A CCW pump motor, which is identified as a 10 CFR 50, Appendix R, Section III.G.2.b area. Twenty feet of cable separation exists in the area, but because no fire barrier exists, no intervening combustibles or fire hazards are allowed. Work Order (WO) 1082262401 was generated by licensee personnel to clean the sight-glasses on the inboard and outboard motor bearings of the 1A CCW pump. Part of the preparation and planning process includes a transient fire load analysis, which is included in the maintenance work instructions. In the case of this WO, the instructions utilized the fire load analysis data for the Unit 1 CCW heat exchanger area instead of the CCW pump area, and was included in the written instructions. The inspectors determined these inadequate work instructions contributed to the performance deficiency. The licensee entered their failure to control combustible material into their CAP for resolution (CR 2009114934) for resolution. The licensees immediate corrective action was removal of the material from the location. The finding was more than minor because it adversely affected protection against the external factors attribute of the Initiating Events (IE) cornerstone, to limit the likelihood of those events upsetting plant stability and challenging critical safety functions during shutdown, as well as power operations. Specifically, this finding affected plant safety-related equipment required for the safe shutdown of the plant in the event of a plant fire. This finding was assessed using the Phase 1 screening worksheets of Appendix 4 and Appendix F of MC 0609. The inspectors determined the presence of combustible materials was a low degradation finding against the fire protection program, because the identified material had a low likelihood of causing a fire from an existing source of heat or electrical energy. The inspectors determined the finding was of very low safety significance (Green) because of the low degradation rating. This finding was assigned a cross-cutting aspect in the resources component of the Human Performance area because complete, accurate and up-to-date design documentation, procedures, work packages, and correct labeling of components were not provided (H.2(c))
05000369/FIN-2010002-042010Q1Mcguire
McGuire		Power Reduction on both operating units due to entry into TS Limiting Condition for Operation 3.0.3 caused by the inoperability of both trains of the Control Room Area Chilled Water SystemThe inspectors identified an unresolved item (URI) regarding NOED 10-2-001 granted on January 13.The inspectors reviewed NOED 10-2-001 granted on January 13 and related documents to determine the accuracy and consistency with the licensees assertions and implementation of the licensees compensatory measures and commitments, those of which included protecting the B CRACWS chiller, nuclear service water (RN), and power availability, and verifying mitigating equipment in the accordance with licensee procedure AP-39, Control Room Hi Temperature. The licensee issued LER 05000369/2010-001 on March 11, 2010. Additional inspection is required to conduct a review of the LER, root cause, and planned corrective actions. This URI is identified as: URI 05000369,370/2010002-04 Power Reduction on both operating units due to entry into TS Limiting Condition for Operation 3.0.3 caused by the inoperability of both trains of the Control Room Area Chilled Water System
05000261/FIN-2009005-032009Q4RobinsonLicensee-Identified ViolationTS 3.3.2 required that PC-953A containment pressure switch channel be placed and maintained in the tripped condition. Contrary to this on June 29, 2009, during repair activities the channel was inadvertently removed from the tripped condition. The cause of the error was inadequate work instructions. The channel was restored to the tripped condition in approximately two minutes. This condition was documented in Condition Report 342793. This violation is of very low safety significance because the condition was promptly corrected in approximately 2 minutes and redundant channels were operable
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