Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000321/FIN-2017502-012017Q4HatchLicensee-Identified ViolationThe following licensee-identified violation of NRC requirements was determined to be of very low safety significance or Severity Level IV and meet the NRC Enforcement Policy criteria for being dispositioned as a Non-Cited Violation. Because it had the potential for impacting the NRCs ability to perform its regulatory function, traditional enforcement is applicable in accordance with Inspection Manual Chapter 0612, Appendix B. This finding was also determined to be a Severity Level IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involved the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan was reduced. Title 10 of the Code of Federal Regulations, Part 50.54(q) states, in part, that a licensee may make changes to emergency plans without prior NRC approval only if the changes do not reduce the effectiveness of the plans and the plans, as changed, continue to meet the standards of 50.47(b) and the requirements of Appendix E. Proposed changes that reduce the effectiveness of the approved emergency plans may not be implemented without application to and approval by the NRC. Contrary to the above, on multiple occasions between 2008 and 2014, the licensee implemented changes to their Radiological Emergency Plan and Emergency Action Levels (EALs) which reduced the effectiveness of the Plan. Specifically, the licensee deleted and/or changed EAL threshold values, all of which would have resulted in a change that reduced the effectiveness of the approved Emergency Plan and was implemented without application to and approval by the NRC. Because the violation was entered into the licensees corrective action program as Condition Report 10421212, it is being treated as a Green non-cited licensee-identified SL IV violation consistent with Section 2.3.2 of the Enforcement Policy.
05000413/FIN-2015301-012015Q2CatawbaLicensee-Identified ViolationThe following Severity Level IV violation was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Following the facilitys administration of the initial written examination on May 28, 2015, the licensee identified that an earlier version of the examination was inadvertently provided to the RO applicants. The licensee immediately informed the NRC. The earlier version of the examination did not include the changes that were made to resolve NRC comments provided during the preexamination review of the written examination. This earlier version of the examination had not been approved by the NRC for administration to the license applicants. 10 CRF 55.49, Integrity of examinations and tests states, in part, that facility licensees shall not engage in any activity that compromises the integrity of any test or examination required by this part. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. Contrary to the above, on May 28, 2015, the licensee administered an unapproved RO written examination, an activity that compromised the integrity of the written examination. The RO applicants did not get the benefit of the question enhancements that occurred during the examination review. The SRO applicants experienced a higher quality exam than that of the RO applicants. This is neither equitable nor consistent. The unapproved version of the exam was subsequently reviewed by the NRC and was determined to be valid. See enclosure 3 to this report to review the analysis of the administered written exam for validity of the exam. A violation of 10 CFR 55.49 is a violation that potentially impacts the regulatory process, because the examination results are used by the NRC to make licensing decisions. An improperly administered examination has the potential to provide inaccurate information to the NRC regarding the competence of the applicants. There were no actual or potential safety consequences. This violation is being treated as a Severity Level IV non-cited violation consistent with Section 2.3.2.a. of the NRC Enforcement Policy. The violation was entered into the licensees corrective action program as Nuclear Condition report 01931989.
05000348/FIN-2014405-012014Q1FarleyLicensee-Identified Violation
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-2013011-162013Q2Browns 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-2013011-122013Q2Browns 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.
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-2013011-092013Q2Browns 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-2013011-072013Q2Browns 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-2013011-062013Q2Browns 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-2013011-052013Q2Browns 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-2013011-042013Q2Browns 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-2013011-032013Q2Browns 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-2013011-022013Q2Browns 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-2012005-022012Q4Browns FerryLicensee-Identified ViolationUnit 3 Technical Specification 3.4.3, Safety/Relief Valves, required that twelve of thirteen main steam safety relief valves (MSRVs) lift at a setpoint within plus or minus three percent of a specified value. Contrary to this, during TS required surveillance testing following the Unit 3 Cycle 9 refueling outage, the licensee discovered that the lift setpoints of two MSRVs exceeded the plus or minus three percent TS allowed pressure band. This TS violation was entered into the licensees CAP as PER 558488. The finding was determined to be of very low safety significance because the as-found lift setpoint conditions of the Unit 3 MSRVs were evaluated and determined to meet the design basis criteria for the most limiting reactor pressure vessel over-pressurization events.
05000259/FIN-2012005-012012Q4Browns FerryLicensee-Identified ViolationThe licensee-identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI for the licensees failure to assure that conditions adverse to quality, such as deficiencies, and nonconformances are promptly identified and corrected. Specifically, the licensee failed to take timely corrective actions to address an extensive backlog of EQ information releases which resulted in not meeting their environmental qualification program and the 10CFR 50.49 auditability requirements. Contrary to this requirement, since January of 2010, the licensee failed to take prompt and appropriate corrective actions to evaluate and correct an extensive backlog of EQIRs, which resulted in 81 of the licensees 99 required Environmental Qualification equipment files not being updated to reflect the as-installed specifications and configuration of EQ equipment. The licensee entered this issue into their corrective action program as PERs 238931 and 624137. The finding was determined to be of very low safety significance (Green) using Attachment 4 to IMC 0609, Significance Determination Process, because the incomplete corrective actions did not result in an actual loss of safety function.
05000259/FIN-2012004-052012Q3Browns FerryLicensee-Identified ViolationTechnical Specifications 3.6.1.3, Primary Containment Isolation Valves (PCIVs), required that while Unit 2 is in Modes 1, 2, and 3, each PCIV, except reactor building-to-suppression chamber vacuum breakers shall be operable. The Technical Specification (TS) action statement A.1 required in part that an affected flow path be isolated by use of at least one closed and de-activated automatic valve within four hours. Contrary to the above, between June 7, 2012, and June 13, 2012, while Unit 2 was in Mode 1, the licensee identified a leak coming from the 2-FCV-73-0081, a one-inch HPCI steam warm-up bypass PCIV and action was not taken to isolate the affected penetration flow path within four hours. The finding was screened in accordance with IMC 0609 Appendix H, Containment Integrity SDP and was characterized to be of very low safety significance (Green) because the 2-FCV-73- 0081 valve was a one-inch valve and would not generally contribute to Large Early Release Frequency (LERF) as discussed in IMC 0609, Appendix H.
05000259/FIN-2012004-042012Q3Browns FerryLicensee-Identified ViolationTechnical Specifications 5.4.1.d, Procedures, required that written procedures for the Fire Protection Program shall be established, implemented and maintained. Section 7.2.2.c, Combustible Material Control Procedures of the Fire Protection Plan established that the storage or staging of transient combustibles during modes 1, 2, or 3 would be restricted from within the twenty foot zone of separation visibly marked as red floors . Contrary to the above, since approximately the year 2000 when the Radiation Protection remote camera system was installed, the licensee failed to adequately control transient combustibles in a red floor area in the Units 1, 2 and 3 reactor buildings as required by the Fire Protection Plan. The licensee entered this issue into their CAP as PERs 529001 and 558964. The safety significance of this finding was characterized to be of very low safety significance (Green) in accordance with IMC 0609, Appendix F, because the finding was assigned a low degradation rating and reflected a fire protection program element whose performance was minimally impacted by the inspection finding.
05000259/FIN-2012004-032012Q3Browns FerryAutomatic Reactor Scram Due to Inadequate Testing of Current TransformerA self-revealing finding (FIN) was identified for the licensees failure to adequately test a Unit 3 main turbine generator current transformer (CT) as required by TVA-NQA-PLN89-A, Nuclear Quality Assurance Plan which resulted in the improper wiring of the CT. The licensee switched the CT leads to correct the input to the main transformer relay, adequately tested all other new Unit 3 relays, implemented a transition plan to incorporate the protective relay group into the nuclear organization, and planned post startup monitoring for the Unit 1 and 2 digital differential protective relays. The licensee entered this issue into their corrective action program as PER 558183. This finding was determined to be more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability. Specifically, the failure to adequately test a Unit 3 main turbine generator CT directly contributed to a reactor scram of Unit 3. The significance of the finding was evaluated using Phase 1 of the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter 0609 Attachment 4 and was determined to be of very low safety significance (Green) because it did not contribute to both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The cause of this finding was directly related to the cross-cutting aspect of Supervisory and Management Oversight in the Work Practices component of the Human Performance area, because the supervisors failed to ensure proper procedure quality, procedure usage, worker qualification, and proper work preparation associated with the protective relay groups work activities such that nuclear safety was supported.
05000259/FIN-2012004-022012Q3Browns FerryAutomatic Reactor Scram Due to Inadequate Design Review of Relay SettingA self-revealing finding (FIN) was identified for the licensees failure to provide an adequate design review of vendor calculations as required by TVA-NQAPLN89- A, Nuclear Quality Assurance Plan which resulted in the 3A Unit Station Service Transformer (USST) differential current protection relay trip settings being incorrectly set. The licensee reset and adequately tested the function of the relay. The licensee has evaluated vendor-provided modifications for similar protective relays and plans to revise the design review process to provide increased licensee accountability and specificity of reviews for vendor designs. The licensee entered this issue into their corrective action program as problem evaluation report (PER) 555573. This finding was determined to be more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability. Specifically, the failure to provide an adequate design review of vendor calculations directly contributed to a reactor scram of Unit 3. The significance of the finding was evaluated using Phase 1 of the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter 0609 Attachment 4 and Appendix A and 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 mitigating equipment or functions were not available. The cause of this finding was directly related to the cross-cutting aspect of Complete Documentation in the Resources component of the Human Performance area, because the licensee failed to ensure procedure NEDP-5, Design Document Reviews was consistent with TVANQA- PLN89-A, Nuclear Quality Assurance Plan
05000259/FIN-2012004-012012Q3Browns FerryLoss of Seismic Monitoring CapabilityThe inspectors identified a non-cited violation (NCV) of 10 CFR 50.54(q)(2) for the licensees failure to follow and maintain an emergency plan that meets the requirements of emergency planning standard 10 CFR 50.47(b)(4). Specifically, due to a plant modification, the licensee failed to maintain configuration control of seismic instrumentation necessary for the declaration of emergency events from August 17 to August 31, 2012. Completion of installation of the power and instrumentation logic signal to the control room annunciators on August 31, 2012, restored compliance with the emergency plan requirements. The licensee entered this issue into their corrective action program as PER 610625. This finding was determined to be more than minor because it was associated with the Emergency Response Organization (ERO) Performance Attribute of the Emergency Preparedness Cornerstone and affected the cornerstone objective of ensuring the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, one Alert and one Notification of Unusual Event Emergency Action Level (EAL) initiating condition would have been rendered ineffective such that a seismic event may not have been appropriately declared. The significance of this finding was evaluated in accordance with the IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, and was determined to be of very low safety significance because an ineffective or degraded EAL scheme that affects Alert declarations was categorized as a Green violation. The cause of this finding was directly related to the cross cutting aspect of Documents, Procedures and Component Labeling in the Resources component of the Human Performance area. Specifically, a lack of complete, accurate and up-to-date design documentation resulted in a loss of configuration control and degradation of information necessary to classify a seismic event.
05000259/FIN-2012004-062012Q3Browns FerryUnanalyzed Conditions Discovered During NFPA 805 Transition Review10 CFR Part 50.48(b)(1) requires that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of 10 CFR Part 50, Appendix R, Section III.G. 10 CFR 50, Appendix R, Section III.G.2, states, in part, that where cables or equipment, that could prevent operation or cause mal-operation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided: separation of cables and equipment by a fire barrier having a 3-hour rating, separation of cables and equipment by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards and with fire detectors and an automatic fire suppression system in the fire area, and enclosure of cables and equipment in a fire barrier having a 1-hour rating and with fire detectors and an automatic fire suppression system in the fire area. Contrary to the above, the licensee failed to use one of means described in Appendix R, Section III.G.2.a, b, or c to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. Specifically, on April 5, 2012, the licensee identified the failure to protect equipment that was required to mitigate fire events. The licensee determined that fire damage could prevent operation or cause mal-operation of the reactor pressure instrument loops, safety relief valve overpressure logic, automatic depressurization logic, RHR test return valves, drywell spray valves, suppression pool spray valves MSIVs, and a 4kV shutdown board due to hot shorts, open circuits, or shorts to ground. This condition has existed since initial plant startup for Units 1, 2 and 3. The licensee entered this issue into the corrective action program (PERs 229734, 259787 and 424389) and implemented compensatory actions in the form of fire watches for Units 1, 2, and 3. Because the licensee committed to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c), the NRC is exercising enforcement and reactor oversight process (ROP) discretion for these issues in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. Specifically, these issues were identified and will be addressed during the licensees transition to NFPA 805, they were entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, they were not likely to have been previously identified by routine licensee efforts, they were not willful, and they were not associated with a finding of high safety significance (Red).
05000259/FIN-2012003-052012Q2Browns FerryLicensee-Identified ViolationA violation of Technical Specification 5.4.1.a was identified by the licensee for the failure to establish adequate work instructions to ensure proper installation of the gap setting between the actuator stem and valve stem of Unit 1 HPCI, (High Pressure Coolant Injection), turbine stop valve, 1-FCV-073-0018. On April 19, 2012, during the performance of a quarterly surveillance test the turbine stop valve, 1-FCV-073- 0018, failed to close upon repeated demands. A Phase 3 analysis determined the significance of the finding was very low safety significance (Green) The regional Senior Reactor Analyst performed a Phase 3 SDP analysis on the finding. The risk was dominated by the unavailability of the HPCI during the repair time after discovery of the Stop Valve issue. The finding was determined to be GREEN in the SDP, primarily due to the short period of time it was fully non-functional. The licensee initiated PER 539040 to enter the issue into their corrective action program.
05000259/FIN-2012003-042012Q2Browns FerryFailure to Establish Preventive Maintenance for Unit 2 and 3 Main Control Room Annunciator Power SuppliesA self-revealing finding (FIN) was identified for the licensees failure to perform preventive maintenance on the Unit 3 Main Control Room (MCR) annunciator power supplies. As a result, a power supply failed which led to a fire in annunciator panel 3-X-55-5A in the Unit 3 control room. The licensee initiated actions to extinguish the fire, replace the two affected power supplies and develop a preventive maintenance program to replace the power supplies every ten years. Additional corrective actions to replace all power supplies that have been installed for more than four years are pending. This was captured in the licensees corrective action program as problem event report (PER) 496592. The performance deficiency was determined to be more than minor because it was considered sufficiently similar to example 4.f of Inspection Manual Chapter (IMC) 0612, Appendix E, for an issue that resulted in a fire hazard in a safety-related area of the plant. The finding was associated with the Initiating Events Cornerstone and required a phase 3 analysis in accordance with IMC 0609 because the finding increased the likelihood of, and actually caused, a fire in the Unit 3 control room. The phase 3 analysis determined that without an impact to additional plant equipment, or a major impact on human action failure rates, the finding was determined to be Green. The cause of this finding was related to the cross cutting aspect of Problem Identification in the Corrective Action Program component of the Problem Identification and Resolution area because the licensee should have recognized the electrolytic capacitors were installed beyond their recommended service life and scheduled replacement prior to their failure (P.1(a)).
05000259/FIN-2012003-032012Q2Browns FerryFailure to Implement DOT Type A Package Closure RequirementsA self-revealing non-cited violation (NCV) of 10 CFR 71.5, Transportation of Licensed Material, was identified by inspectors for the licensees failure to comply with Department of Transportation (DOT) regulations during shipment of radioactive materials. Specifically, the licensee failed to ensure proper closure of a DOT 7A Type A package as required by Department of Transportation (DOT) regulations in 49 CFR 173.475, Quality Control Requirements Prior To Each Shipment Of Class 7 (Radioactive) Materials. This issue has been entered into the licensees corrective action program as SR 571151. The finding was more than minor because it is associated with the Public Radiation Safety Cornerstone, Plant Facilities/Equipment and Instrumentation attribute, involving transportation packaging and adversely affected the cornerstone objective, to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, the failure to apply the correct torque to the package closure bolts could have resulted in incomplete sealing of the container or failure of the cover bolts during transportation. The finding was determined to be of very low safety significance (Green) because it did not involve radiation limits being exceeded, a package breach, a certificate of compliance issue, a low-level burial ground non-conformance, or a failure to make emergency notifications. The cause of this finding was directly related to the cross cutting aspect of Documents, Procedures and Component Labeling in the Resources component of the Human Performance area because the licensee did not effectively incorporate the vendor provided container loading and shipping instructions into their work package and transportation program to ensure correct torque values were used to close the shipping container. (H.2(c))
05000259/FIN-2012003-022012Q2Browns FerryFailure to Properly Prepare a DOT Type A Package for TransportA self-revealing non-cited violation (NCV) of 10 CFR 71.5, Transportation of Licensed Material, was identified by inspectors for the licensees failure to comply with Department of Transportation (DOT) regulations during shipment of radioactive materials. Specifically, the licensee failed to ensure proper packaging of two DOT 7A Type A packages as required by Department of Transportation (DOT) regulations in 49 CFR 173.475, Quality Control Requirements Prior To Each Shipment Of Class 7 (Radioactive) Materials. This issue has been entered into the licensees corrective action program as SR 570902. The finding was more than minor because it is associated with the Public Radiation Safety Cornerstone, Plant Facilities/Equipment and Instrumentation attribute, involving transportation packaging and adversely affected the cornerstone objective, to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, the failure to correctly secure the package contents to prevent movement could have resulted in damage or failure of the container during transportation. The finding was determined to be of very low safety significance (Green) because it did not involve radiation limits being exceeded, a package breach, a certificate of compliance issue, a low-level burial ground nonconformance, or a failure to make emergency notifications. The cause of this finding was directly related to the cross cutting aspect of Documents, Procedures and Component Labeling in the Resources component of the Human Performance area because the licensee did not effectively incorporate package design specifications into their transportation program to ensure that all internal restraining devices are correctly installed to secure the CRDM in place to prevent damage to the transport package. (H.2(c))
05000259/FIN-2012003-012012Q2Browns FerryFailure to Maintain Flood Barrier Results in Inoperable Safety Related PumpsAn NRC-identified non-cited violation (NCV) of the Technical Specifications 5.4.1.a was identified for the licensees failure to maintain an Emergency Equipment Cooling Water (EECW) pump flood barrier in accordance with written procedures which resulted in the inoperability of two other safety related pumps. The licensee immediately restored the flood protection configuration of the C Residual Heat Removal Service Water (RHRSW) pump room by properly re-installing the flood protection cover and permanently stenciled the aluminum plate with the required procedure for installation. The licensee entered this issue into their corrective action program as PER 532050. The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of RHRSW pumps to perform their intended safety function during a design basis flooding event. Specifically, the improper re-installation of an external flood protection cover resulted in the inoperability of two Residual Heat Removal Service Water (RHRSW) pumps. The significance of this finding was evaluated in accordance with the IMC 0609 Attachment 4, Phase 1- Initial Screening and Characterization of Findings, which required a Phase 3 analysis because the finding involved the degradation of equipment designed to mitigate a flooding event and it was risk significant due to external initiating event core damage sequences. The finding was determined to be Green because of the short exposure time, and the low likelihood of the flood. The cause of this finding was directly related to the cross cutting aspect of Supervisory Oversight in the Work Practices component of the Human Performance area, because of the foremans assumption that workers knew to restore the flood protection cover to meet procedural requirements without a formal pre-job brief (H.4(c)).
05000259/FIN-2012002-052012Q1Browns FerryLicensee-Identified ViolationUnit 1 Technical Specification 3.3.8.2, Reactor Protection System (RPS) Electric Power Monitoring, required that, for each in-service RPS motor generator set or alternate power supply, two RPS electric power monitoring assemblies be operable in Modes 1, 2, and 3; and in Modes 4 and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. With one electric power monitoring assembly inoperable for one or both in-service power supplies, the associated in-service power supply(s) were required to be removed from service in 72 hours or be in Mode 3 within 12 hours and in Mode 4 within 36 hours. In addition, TS 3.0.4 prohibited Mode changes with TS 3.3.8.2 not met. Contrary to this, on October 6, 2011, while performing an operability determination for the channel A RPS power monitoring system under-voltage trips, the licensee determined that the as-found under-voltage trip for the RPS 1A1 relay was less than the required TS acceptance criteria during multiple previous TS surveillances and that the RPS 1A1 relay was inoperable from April 30, 2007 to October 5, 2011. This TS violation was entered into the licensees CAP as PERs 413140 and 442914. The finding was determined to be of very low safety significance because the finding does not represent an actual loss of the RPS safety function in that the remaining operable channel A RPS electric power monitoring assembly still provided protection to the RPS bus powered components under degraded voltage conditions.
05000259/FIN-2012002-032012Q1Browns FerryFailure to Ensure ECCS Design Calculation Does Not Exceed Maximum Clad TemperatureThe NRC identified a Green non-cited violation of 10CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, for the licensees failure to ensure that the ECCS was satisfactorily designed such that the maximum fuel element cladding temperature specified in 10 CFR 50.46(b)(1) would not be exceeded. On May 29, 2011, operating limitations were implemented to account for the error in calculations. This violation has been entered into the licensees CAP as PER 372764. This performance deficiency was considered greater than minor because it was associated with the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents. The inspectors determined the finding to not be greater than green based on the remaining barriers to fission product release were unaffected. The cause of this finding was directly related to the cross-cutting aspect of Issue Identification in the Corrective Action Program component of the Problem Identification and Resolution area because the licensee failed to completely, accurately, and in a timely manner identify the errors with the ECCS evaluation model
05000259/FIN-2012002-022012Q1Browns FerryFailure to Immediately Report a Plant FireThe NRC identified a non-cited violation of Technical Specification 5.4.1.d, Fire Protection Program implementation associated with the licensees failure to report a fire in the Unit 1 Turbine Building to the main control room (MCR). Specifically, the failure to report a plant fire resulted in a failure of the MCR operators to implement Emergency Plan Implementing Procedure EPIP-17, Fire Emergency Response. Following the event, plant staff performed additional inspections of plant areas and either removed electrical extension cords or ensured each cord had a required GFCI and was not overloaded. Expectations for plant employees discovering and responding to fires were reinforced by plant management. The licensee entered this event into their corrective action program as PER 527090. The performance deficiency was determined to be more than minor because if left uncorrected, the failure to notify the MCR of plant fire events would have the potential to lead to a more significant safety concern. Specifically, emergency response procedures for plant fires would not be entered and implemented and the Fire Brigade response would be delayed. The significance of this finding was evaluated in accordance with the IMC 0609, Appendix F, Attachment 1, Part 1, Fire Protection SDP Phase 1 Worksheet. The inspectors concluded that the significance of this finding was Green due to a low degradation rating for this fire event because it was a small electrical fire with no combustible material within the vicinity of the fire. The cause of this finding was directly related to the cross cutting aspect of Procedural Compliance in the Work Practices component of the Human Performance area, because the licensee failed to recognize the requirement to immediately report a fire and enter the appropriate fire emergency response procedures
05000259/FIN-2012002-012012Q1Browns FerryFailure to Adequately Implement Impaired Fire Barrier and Detector ControlsThe NRC identified a non-cited violation of Technical Specification 5.4.1.d, Fire Protection Program, for the licensees failure to adequately implement Limiting Conditions For Operation in accordance with Fire Protection Report Volume 1, Fire Protection Plan. Specifically, the licensee failed to adequately implement impaired fire barrier and detector controls which resulted in the failure to establish a continuous fire watch for an impaired fire barrier having smoke detection identified as unavailable to protect either side of the inoperable barrier. The licensee subsequently returned the impaired fire door and smoke detection to service. The licensee entered this event into their corrective action program as PERs 529543 and 527311. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events, and adversely affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, inadequate implementation of the licensees FPIP and LCO processes resulted in the licensee missing a LCO entry condition and not establishing a continuous fire watch for an impaired fire door. The significance of this finding was evaluated in accordance with the IMC 0609 Appendix F, Attachment 01, Part 1, Fire Protection SDP Phase 1 Worksheet. The finding was determined to be of very low safety significance (Green) because the condition represented a low degradation of fire prevention and administrative controls. Specifically, a smoke detection system on one side of the impaired fire door was discovered functional. The cause of this finding was directly related to the cross cutting aspect of Procedural Compliance in the Work Practices component of the Human Performance area, because licensee expectations were ineffectively communicated and fire protection procedures inadequately implemented to maintain a site understanding of fire barrier and detector configuration
05000259/FIN-2012002-042012Q1Browns FerryRepeated Failure to Report ECCS Analyses Methodology Change or ErrorsDuring discussions between the NRC staff, the fuel vendor, and the licensee starting in April 2010, the NRC staff questioned the appropriateness of the application of credit for spray cooling in the Units 2 and 3 ECCS evaluation, and the effect non-single failure proof ADS would have on the ECCS evaluation model for the BFN units. In a letter dated April 30, 2010 the licensee acknowledged the single failure issue with ADS and indicated that the estimated effect of the change or error on peak clad temperature (PCT) was not significant (greater than 50 degrees Fahrenheit). TVA committed to modify the ADS to provide a single failure proof automatic initiation capability of 4 ADS valves. The licensee also outlined the compensatory measures intended to address the identified degraded/nonconforming condition. Subsequently, on June 30, 2011, TVA submitted the annual ECCS evaluation model report and indicated a minor change to the radiative heat transfer model which resulted in a minor change in PCT for Units 2 and 3. On October 7, 2011, TVA submitted a revised ECCS analysis in support of a Unit 1 fuel transition request. This analysis provided a methodology change to address the evaluation model error associated with spray cooling, which had been identified by the NRC staff, and for which the licensee implemented operating restrictions to ensure that the effects of the error would not cause the predicted PCTs at Units 2 and 3 to exceed 2200F.. This analysis was also applicable for current operating conditions for Units 2 and 3 and was not previously reported to the NRC. NRC review identified that the effect of the evaluation model error would have resulted in greater than a 50 degree increase in predicted PCT for Units 2 and 3. On February 29, 2012, TVA initiated Service Request 514121 which recognized that a 30-day report for a significant change in peak clad temperature consistent with 10 CFR 50.46 had not been submitted. As of March 30, 2012, TVA had not submitted the required 30-day report for a significant change in peak clad temperature consistent with 10 CFR 50.46 which was identified on February 29, 2012. Following the end of the reporting period, TVA submitted the required report per 10 CFR 50.46 on April 18, 2012. Analysis: The inspectors determined that the licensees repeated failure to report changes or errors in the ECCS analyses was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612, Appendix B, and determined the performance deficiency did not constitute a Finding, but the failure to report impacted the regulatory process and was subject to traditional enforcement consistent with the discussion for Block 7, Figure 2, Paragraph 2.a.v. The violation was determined to be more than minor per the NRC Enforcement Manual, Section 2.10.F, since the NRC has evidence that this failure to report has occurred repeatedly. This violation was determined to be a Severity Level IV violation based on section 6.9 of the NRC Enforcement Policy. Enforcement: 10 CFR 50.46 (a)(3)(ii), requires for each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46 requirements. Contrary to the above, the licensee failed to report each change or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation for Units 2 and 3. Specifically, from May 29, 2011 to April 18, 2012, the licensee failed to report a significant change in peak clad temperature associated with an error related to spray cooling to the NRC within 30 days, and include with the report a proposed schedule for providing reanalysis or taking other action as may be needed to show compliance. The licensee subsequently submitted the required report per 10 CFR 50.46. Because this violation was determined to be a Severity Level IV violation and was entered into the licensees CAP as PER 531752, this violation is being treated as an NCV consistent with the Enforcement Policy. This NCV is identified as NCV 05000260(296)/2012002-04, Repeated Failure to Report ECCS Analyses Methodology Change or Errors.
05000259/FIN-2011005-012011Q4Browns FerryFailure to Report a Valve Motor Operator Manufacturing Defect Pursuant to 10CFR21.21 in a Timely MannerThe inspectors reviewed the two specific structures, systems and components (SSC) within the scope of the Maintenance Rule (MR) (10CFR50.65) with regard to some or all of the following attributes, as applicable: (1) Appropriate work practices; (2) Identifying and addressing common cause failures; (3) Scoping in accordance with 10 CFR 50.65(b) of the MR; (4) Characterizing reliability issues for performance monitoring; (5) Tracking unavailability for performance monitoring; (6) Balancing reliability and unavailability; (7) Trending key parameters for condition monitoring; (8) System classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2); (9) Appropriateness of performance criteria in accordance with 10 CFR 50.65(a)(2); and (10) Appropriateness and adequacy of 10 CFR 50.65 (a)(1) goals, monitoring and corrective actions (i.e., Ten Point Plan). The inspectors also compared the licensees performance against site procedure NPG-SPP-3.4, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting; Technical Instruction 0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting; and NPG SPP 3.1, Corrective Action Program. The inspectors also reviewed, as applicable, work orders, surveillance records, PERs, system health reports, engineering evaluations, and MR expert panel minutes; and attended MR expert panel meetings to verify that regulatory and procedural requirements were met. FnResidual Heat Removal Service Water (RHRSW) 023-C, Vessel / Containment Flooding MR Function Reclassified as Risk Significant FnUnit 1, Loop I RHR Low Pressure Coolant Injection (LPCI) Outboard Injection Valve (1-FCV-74-52) Failure and 10CFR50.65(a)(1) corrective action plan b. Findings Introduction: The NRC inspectors identified a Severity Level (SL) IV non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 21, Reporting of Defects and Noncompliance, for the licensees failure to report a known defect as soon as practicable, but in all cases within 60 days of discovery. More specifically, the licensee did not submit an interim report or notify the NRC in a timely manner pursuant to 10CFR21.21 regarding a manufacturing defect that caused a failure of the Unit 1 RHR Loop I outboard injection valve (1-FCV-74-052) on August 2, 2011. Description: On August 2, 2011, the Unit 1 Loop I LPCI Outboard Injection Valve (1- FCV-74-52), experienced a failure of the motor operator during performance of surveillance test procedure 1-SR-3.3.5.1.6(C I), Functional Testing of RHR Loop I Valve Logic and Interlocks. During this surveillance test, 1-FCV-74-52 was cycled successfully multiple times until it suddenly failed to reopen. The licensee promptly entered the required TS Limiting Condition of Operation (LCO) 3.5.1 seven day action statement, and initiated PER 410394 to also enter this issue into the corrective action program (CAP). The valve was repaired and returned to service within its TS allowed outage time (AOT). The actual failure of 1-FCV-74-52 did not involve a past operability concern or licensee performance deficiency. The root cause evaluation for PER 410394 was presented to the Corrective Action Review Board (CARB) on September 14, 2011. As part of the root cause analysis the licensee determined the motor operator failure was a manufacturing defect due to inadequate vendor assembly procedures and manuals for the SMB-5(T) motor operator which led to incomplete lug engagement of the clutching mechanism that subsequently rendered the valve non-functional. At the conclusion of the CARB, the NRC inspectors questioned the root cause team leader regarding the lack of a Part 21 evaluation and notification. The inspectors were informed that the licensee was working with the valve motor operator vendor on further corrective actions, and any required Part 21 notification would be addressed with the vendor. On September 20, 2011, PER 435444 was initiated stating that the root cause determination did not include a Part 21 evaluation. As a result of this PER, the licensee implemented their procedure NPG-SPP-03.5, Regulatory Reporting Requirements, and recognized this issue was potentially reportable per the requirements of 10CFR21.21. The inspectors subsequently concluded that the time of discovery for a Part 21 evaluation was September 14, 2011, for which 10CFR21.21 required the licensee to complete their Part 21 evaluation within the next 60 days, and then notify the NRC within the following seven days; or submit an interim report within 60 days of discovery if the Part 21 evaluation could not be completed within the 60 days. This timeframe required the issuance of a Part 21 interim report to the NRC by November 15, 2011, or a Part 21 initial Notification by November 20, 2011. However, no interim report was issued, and the licensee did not make an initial Part 21 Notification. The valve motor operator vendor (Flowserve) did submit the required Part 21 written report on November 29, 2011. The untimely Part 21 Notification was entered into licensees CAP as PER 487357. Analysis: The inspectors determined that the licensees failure to issue an interim report within 60 days or make an initial Notification of a Part 21 reportable condition constituted a violation of 10CFR21.21. Specifically, the licensee did not ensure that the failure of the 1-FCV-74-52 motor operator due to a manufacturing defect was evaluated and reported in accordance with the timeliness requirements of Part 21. This violation was evaluated using traditional enforcement because it had the potential for impacting the regulatory process. In accordance with the guidance in Section 2.2.2 and Section 6.9.d. of the NRC Enforcement Policy, the inspectors determined this violation was a Severity Level (SL) IV violation of low safety significance because the failure to report this condition did not substantially impact the Agency\\\'s regulatory responsibilities and the Agency would not have responded in a significantly different manner had the information been properly reported. The inspectors also concluded that failing to recognize this as a Part 21 reportable issue in a timely manner was a performance deficiency under the Reactor Oversight Process (ROP). In accordance with NRC IMC 0612, Appendix B, Issue Screening, the inspectors concluded that this performance deficiency was minor. Because this performance deficiency was minor and the violation was evaluated using Traditional Enforcement, a cross-cutting aspect is not assigned in accordance with IMC 0612. Enforcement: 10CFR21.21(a) required in part that the licensee shall evaluate deviations to identify defects associated with a substantial safety hazard as soon as practicable, but in all cases within sixty (60) days of discovery. Upon completion of this evaluation, an initial Notification to the Commission was required within seven days. However, if an evaluation of an identified defect potentially associated with a substantial safety hazard could not be completed within 60 days from discovery of the deviation, an interim report was required to be submitted to the Commission within the 60 days of discovery. Contrary to the above requirements, following the discovery of a manufacturing defect associated with the motor operator for 1-FCV-74-52, Loop I LPCI Outboard Injection Valve on September 14, 2011, the licensee failed to make either an initial Notification or submit an interim report within the time requirements of 10CFR21.21. The NRC was not notified of the Part 21 defect until the vendor (Flowserve) submitted a written report on November 29, 2011. This violation was a SL IV violation of low safety significance because the failure to report this condition did not substantively impact the Agency\\\'s regulatory responsibilities and the Agency would not have responded in a substantially different manner had the information been properly reported. Because this violation was of very low safety significance and it was entered into the licensees CAP as PER 487357, this violation was treated as an NCV, consistent with the NRC Enforcement Policy. This NCV is identified as NCV 05000259, 260, 296/2011005-01, Failure to Report a Valve Motor Operator Manufacturing Defect Pursuant to 10CFR21.21 in a Timely Manner.
05000259/FIN-2011005-022011Q4Browns FerryUnit 1 TS 3.0.3 Entry Caused by the Failure of RPS M-G Set 1BA self-revealing finding (FIN) was identified for the licensees failure to adequately evaluate historically high vibrations on the Unit 1 Reactor Protection System (RPS) Motor Generator (M-G) Set 1B. Consequently, the impact of high vibrations was not considered in the determination of an adequate preventive maintenance frequency to ensure proper wire and cable terminal tightness in the RPS M-G Set control panel. The failure to ensure proper wire and cable tightness resulted in a loss of all Technical Specifications required reactor coolant system (RCS) leak detection. The licensee replaced the voltage regulator and performed all required tightness checks on RPS M-G Set 1B. In addition, the licensee verified appropriate tightness checks were scheduled in the next component outage window for the remaining M-G sets, and initiated actions to increase the preventive maintenance (PM) frequency to every two years, perform a design review to relocate the control panel, and evaluate the vibration program alert limit evaluation process. This issue was entered into the licensees corrective action program as problem evaluation report (PER) 412934. The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of RCS Equipment and Barrier Performance, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to ensure proper wire and cable tightness resulted in loose connectors and/or cable assemblies that caused the failure of the Unit 1 RPS M-G Set 1B which resulted in a loss of all Unit 1 RCS leakage detection systems and an unplanned power reduction on August 6, 2011. The significance of the finding was evaluated using Phase 1 of the SDP in accordance with the IMC 0609 Attachment 4, and was determined to be of very low safety significance (Green) because the finding did not represent a pressurized thermal shock, fuel barrier, or spent fuel pool issue. The cause of this finding was directly related to the cross-cutting aspect of Long Term Plant Safety Through Proper Maintenance Practices in the Resources component of the Human Performance area, because the vibration issues associated with Unit 1 RPS M-G Set 1B had been a long-standing equipment issue that was not adequately addressed by the vibration program alert limit evaluation process (H.2(a)).
05000259/FIN-2011005-032011Q4Browns FerryFailure to Control Temporary Equipment Resulted in a FireA self-revealing non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings was identified for the licensees failure to install and maintain adequate control of temporary lighting in the intake cable tunnel as required by the Tennessee Valley Authority (TVA) Safety Manual and NPG-SPP- 09.17, Temporary Equipment Control. Consequently, a temporary light string was left improperly installed, without ground fault circuit interrupt (GFCI) device(s), for over two years until it faulted electrically and caused a fire in the intake cable tunnel on October 12, 2011. The fire brigade extinguished the fire in approximately 10 minutes and removed the temporary light string from the cable tunnel. The licensee entered this event into their corrective action program as PER 445331. The finding was determined to be more than minor because it was considered sufficiently similar to example 4.f of Inspection Manual Chapter (IMC) 0612, Appendix E, for an issue of concern that resulted in a fire hazard in a safety-related area of the plant. The finding was associated with the Initiating Events Cornerstone and characterized according to IMC 0609, Significance Determination Process (SDP), Attachment 04, Phase 1 - Initial Screening and Characterization of Findings. The results of this analysis required an evaluation in accordance with IMC 0609, Appendix F, Attachment 01, Part 1, Fire Protection SDP Phase 1 Worksheet. For the SDP Phase 1 evaluation a high degradation rating was assigned for this fire event with a duration factor greater than 30 days. When compared against the SDP Phase 1 screening criteria, this resulted in a SDP Phase 2 evaluation. The inspectors concluded that this finding screened to Green in the Appendix F Phase 2 analysis using Appendix F Attachment 01, Part 2, Fire Protection SDP Phase 2 Worksheet. Specifically, it was determined that the fire could not reach the temperature threshold for fire-induced cable failure and would not spread to other combustible materials in the area. The cause of this finding was directly related to the cross cutting aspect of Long-Standing Equipment Issues in the Resources component of the Human Performance area, because the deficiencies with the permanently installed lighting system in the intake cable necessitated the use of the temporary light stringer for more than two years (H.2(a))
05000259/FIN-2011005-042011Q4Browns FerryLicensee-Identified ViolationUnit 1 TS 5.4.1.a, required that written procedures recommended in RG 1.33, Revision 2, Appendix A, shall be established, implemented, and maintained. Item 9.a of RG 1.33, Appendix A, stated, in part, that maintenance affecting the performance of safety-related equipment be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, on March 16, 2005, the licensee failed to establish an adequate procedure for the performance of maintenance that affected the performance of a piece of safety related equipment. Specifically, the level of detail in work order package WO 2002-013120-030 and procedure MCI-0-073-PMP002 was inadequate to ensure the proper installation of the Unit 1 HPCI booster pump outboard thrust bearings, which directly led to severe bearing damage and would have eventually resulted in failure of the HPCI pump. The licensee initiated PER 408067 to enter this issue into their CAP and performed corrective maintenance to replace the bearings. The finding was determined to be of very low safety significance in accordance with a Phase 3 SDP of IMC 0609 because the licensee identified and repaired the booster pump prior to an actual failure.
05000259/FIN-2011004-012011Q3Browns FerryFailure to Control Transient Combustible Materials in the Unit 1 Reactor BuildingA NRC-identified non-cited violation of the Technical Specifications 5.4.1.d, Fire Protection Program Implementation, was identified for the licensees failure to control transient combustible materials in a designated exclusion area between Fire Zones 1-1 and 1-2 in the Unit 1 reactor building. Specifically, on August 12, 2011, the inspectors identified transient combustible materials left unattended in the designated exclusion area between Loops I and II of the low pressure coolant injection (LPCI) system following LPCI injection valve maintenance activities. Upon notification by the inspectors, the licensee promptly removed the materials. This issue was entered into the licensees corrective action program as problem evaluation report (PER) 418101. The finding was determined to be greater than minor because it was similar to example 4.k. of Inspection Manual Chapter (IMC) 0612, Appendix E, for an issue of concern involving transient combustibles in a designated combustible free area required for separation of redundant safe shutdown trains. The safety significance of the finding was characterized using IMC 0609, Significance Determination Process (SDP), Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, and determined to be of very low safety significance because of a low degradation rating since a roving fire watch was already established in this same area for an another fire impairment while the transient combustibles were left unattended. The cause of this finding was directly related to the cross cutting aspect of effectively communicating expectations regarding procedural compliance in the Work Practices component of the Human Performance area, because the expectations for the removal of combustible materials from this area were not effectively communicated to the night shift personnel
05000259/FIN-2011011-052011Q3Browns FerryASME Code Compliance PDNRC Inspection Report 05000259/2011008 (ML111290500) documents additional reviews conducted regarding the adequacy of the IST program at BFN. Specifically, the NRC documented that TVAs failure to implement in IST program in accordance with the American Society of Mechanical Engineer (ASME), Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 1995 Edition, 1996 Addendum, Section ISTC 4.1, precluded the timely identification that the RHR Loop II subsystem was unable to fulfill its safety function due to a failure of the LPCI Outboard Injection Valve 1-FCV-74-66. The NRC concluded that TVAs IST program inadequacy was well within its purview, and represents a performance deficiency. Details of the NRCs final determination regarding the performance deficiency are discussed in Enclosure 2. In a letter dated June 8, 2011, TVA appealed the final significance determination of the Red finding. This letter indicated that other licensees understand and implement ASME Operation and Maintenance Code Section ISTC 4.1 in a similar manner to TVA. The NRC recognized the potential generic implications associated with this issue. In the August 16, 2011, Letter to TVA, the NRC states: With respect to the IST performance deficiency described in our May 9 inspection report, the NRC determined that the requirements of the ASME OM Code concerning the verification of valve obturator position warrant additional clarification due to the diversity of views among NRC staff and industry experts. As a result, the NRC staff will continue to pursue generic resolution of the OM code testing issues separate from the resolution of this finding. Until the ASME OM Code testing issues are resolved and there is clarification and guidance on the requirements of ASME OM Code Section ISTC 4.1, this issue is considered an URI. Once resolution has been determined, TVAs IST program will be re-evaluated to determine whether it met ASME OM Code Section ISTC 4.1 requirements and whether a performance deficiency exists or not. (URI 05000259/2011011; Verification of Valve Obturator as Required by ASME OM Code)
05000259/FIN-2011004-052011Q3Browns FerryLicensee-Identified Violation10 CFR 50 Appendix B, Criterion XVI, Corrective Action required in part that measures shall be established to assure that conditions adverse to quality, such as failures are promptly identified and corrected. Contrary to this requirement the licensee failed to adequately identify and correct the failure of normally energized relay 2-RLY-075-14A-K30B during performance of 2-SR-3.3.5.1.6(CS II), Core Spray System Logic Functional Test Loop II on August 8, 2011 prior to returning it to an operable status. The licensee entered this issue into their CAP as PER 415242 and replaced the failed relay on August 13. The safety significance of this finding was characterized to be of very low safety significance in accordance with the Phase 1 SDP of IMC 0609, Attachment 4, because the finding did not represent an actual loss of a safety function of a single train of EDGs for greater than the TS allowed outage time.
05000259/FIN-2011004-042011Q3Browns FerryUnit 3 Loss of Shutdown Cooling During Primary Containment Isolation System Relay ReplacementA self-revealing non-cited violation of Technical Specifications 5.4.1.a was identified for the licensees failure to establish adequate work order instructions for maintenance activities on CR120A relays associated with the Unit 3 Primary Containment Isolation System (PCIS). Consequently, on May 12, 2011, while performing maintenance on a CR120A relay, electricians inadvertently initiated a PCIS Group 2 actuation which resulted in a loss of Unit 3 shutdown cooling (SDC). The licensee immediately restored the affected relay wiring and reestablished Unit 3 SDC. Additional, corrective actions to revise CR120A relay maintenance procedures were in progress. This issue was entered into the licensees corrective action program as problem evaluation report (PER) 368764. The finding was determined to be greater than minor because it was associated with the Initiating Events Cornerstone attribute of Procedure Quality, and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. Specifically, the work package to replace the Unit 3 PCIS relays did not include specific work precautions or instructions to require that jumpers be installed to prevent an inadvertent Group 2 PCIS actuation. According to Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), Appendix G, Shutdown Operations, Table 1, Losses of Control, the finding was determined to be of very low safety significance because the change in temperature during the inadvertent loss of SDC did not exceed 20 percent of the temperature margin to boil. In addition, Checklist 8 of Appendix G, Attachment 1, Shutdown Operations, confirmed adequate mitigation capability remained available for all of the shutdown safety functions to be considered of very low safety significance. The cause of this finding was directly related to the cross-cutting aspect of complete documentation in the Resources component of the Human Performance area, because the licensee failed to provide adequate work package details concerning the replacement of PCIS relays which resulted in the loss of SDC
05000259/FIN-2012010-012011Q3Browns FerryIncomplete and inaccurate information provided in Generic Letter 89-10 response10 CFR 50.9 requires, in part, that information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commissions regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects. Contrary to the above, on January 6, 1997, and May 5, 2004, TVA provided information to the Commission that was not complete and accurate in all material respects, related to its NRC Generic Letter 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance testing program. Specifically, in a letter dated January 6, 1997, TVA responded to a prior NRC question, and stated that Closure of valves FCV-74-52 and FCV-74-66 is not required by plant procedures to operate the RHR system in the suppression pool cooling mode. Therefore, these valves have no redundant safety function and will not be included in the GL 89-10 program. This information was inaccurate because the FCV-74-52 and FCV-74-66 valves do have a safety function to shut to operate the RHR system in the suppression pool cooling mode as described in EOI Appendix-17A, RHR System Operation Suppression Pool Cooling, and should therefore have been included in Browns Ferrys GL 89-10 MOV monitoring program. Additionally, TVA also provided incomplete and inaccurate information in a letter to the NRC dated May 5, 2004. This letter referenced 18 valves, including valves FCV-74-52 and FCV-74-66, that are not in the GL 89-10 program, since the valves are normally in their safety position. This letter stated that TVAs review and documentation of the design basis for the operation of each Unit 1 MOV within the scope of the GL 89-10 program, the methods for determining and adjusting its switch settings, testing, surveillance, and maintenance are the same as with the Units 2 and 3 program. This information was material to the NRC because it was used, in part, as the basis for determining that valves FCV-74-52 and FCV-74-66 did not meet the conditions necessary that would require them to be in Browns Ferrys GL 89-10 MOV monitoring program. This is a Severity Level III violation.
05000259/FIN-2011004-032011Q3Browns FerryUnit 1 Loss of Shutdown Cooling Caused by Emergency Diesel Generator Output Breaker TripA self-revealing non-cited violation of Technical Specifications 5.4.1.a was identified for the licensees failure to establish an adequate maintenance procedure to ensure appropriate calibration and alignment of the Emergency Diesel Generator (EDG) overspeed trip limit switch (OTLS) arm. The lack of procedure guidance resulted in an improperly adjusted OTLS that caused a premature trip of the A EDG output breaker and loss of Unit 1 shutdown cooling (SDC) on May 2, 2011. The licensee replaced and properly set the OTLS on the A EDG, verified the OTLS setpoint on all other seven EDGs, and initiated revisions to applicable maintenance procedures. This issue was entered into the licensees corrective action program as problem evaluation report (PER) 362340. The finding was determined to be greater than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. Specifically, the misadjusted A EDG OTLS resulted in a premature trip of the A EDG output breaker and a loss of Unit 1 SDC. According to Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), Appendix G, Shutdown Operations, Table 1, Losses of Control, the safety significance of the finding was initially characterized to be potentially greater than very low safety significance because the inadvertent loss of SDC represented a loss of control due to a loss of thermal margin to boiling greater than 20 percent. However, a Phase 3 analysis was performed by a Senior Reactor Analyst, it was determined the loss of SDC event was of very low risk significance (i.e., Green), due in part to a low change in risk because of a high chance of recovery of offsite power before the duration of time required to cause the EDG to trip, and the likelihood of recovery of the tripped EDG. The cause of this finding was directly related to the cross-cutting aspect of appropriate self assessments in the Self and Independent Assessments component of the Problem Identification and Resolution area, because inadequate technical rigor applied by the licensee to recognize single point system vulnerabilities resulted in inadequate procedural guidance for maintenance personnel to appropriately calibrate and align the OTLS switch arm and overspeed trip lever
05000259/FIN-2011004-022011Q3Browns FerryFailure to Properly Install Unit 1 High Pressure Coolant Injection Booster Pump Outboard BearingsA licensee-identified apparent violation of Technical Specifications 5.4.1.a was identified for the licensee failing to establish an adequate maintenance instruction for properly installing the Unit 1 High Pressure Coolant Injection (HPCI) booster pump outboard bearing. On July 20, 2011, visual inspections confirmed the booster pump outboard bearing was installed incorrectly and exhibited severe damage. The licensee replaced the HPCI booster pump outboard bearing and the issue was entered into the licensees corrective action program as problem evaluation reports (PER) 405165 and 408067. The finding was determined to be greater than minor because it was associated with the Mitigating Systems Cornerstone attributes of Equipment Performance and Procedure Quality, 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). Specifically, the work package to replace the HPCI booster pump outboard bearing did not include sufficiently detailed instructions to ensure that the bearings were installed in the correct back to back arrangement. Failure to correctly install the HPCI booster pump bearing resulted in severe bearing damage that would have eventually led to a failure of the Unit 1 HPCI pump. The significance of this finding was characterized using Inspector Manual Chapter (IMC) 609, Significance Determination Process (SDP), Attachment 04, Phase 1 - Initial Screening and Characterization of Findings, which did not screen as Green for the Mitigating Systems Cornerstone because it involved a loss of system safety function. A further characterization of the safety significance was then performed using IMC 609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The Phase 2 SDP of Appendix A determined the finding to be potentially greater than very low safety significance (Green) based on the Browns Ferry Phase 2 pre-solved table. Since this finding was potentially greater than Green it will necessitate a Phase 3 SDP to characterize the safety significance. Because the safety significance of this finding has not been finalized, it will be designated as To Be Determined (TBD). No crosscutting aspect was assigned because the incorrect bearing installation did not occur within the past three years, and therefore, was not reflective of current licensee performance.
05000259/FIN-2011003-032011Q2Browns FerryUse of Inappropriately Qualified Methods to Evaluate Emergency Core Cooling During Accident MitigationThe inspectors identified an URI associated with the emergency core cooling system (ECCS) evaluation performed for the Units 2 and 3. The inspectors reviewed Calculation ANP-2908(P), Browns Ferry Units 1, 2, and 3 105% OLTP LOCA Break Spectrum Analysis. The inspectors determined that the analysis, which used the ECCS Evaluation Model described in EMF-2361(P)(A), EXEM BWR-2000 ECCS Evaluation Model was not an adequate evaluation for application at Browns Ferry. The Browns Ferry ECCS evaluation was somewhat unique for two reasons: (1) In most BWR cases, the ADS was single failure-proof; however, for Browns Ferry it was not, and (2) The most severe postulated loss of coolant accidents (LOCA) for Browns Ferry were those arising from small breaks, rather than a large break. Therefore, certain aspects of the approved evaluation model were not applicable to the unique plant configuration at Browns Ferry. Because of this, the evaluation model described in EMF-2361(P)(A) was not entirely applicable to Browns Ferry while the ADS system design was considered to be non-single-failure-proof. Dialog with the licensee and vendor have identified the staffs concerns and resolutions were being pursued through inspection efforts with the licensee and fuel vendor.
05000327/FIN-2011003-012011Q2SequoyahFailure to Perform Instrumentation Surveillance Testing within Required FrequencyThe inspectors identified a non-cited violation of Units 1 and 2 TS Surveillance Requirement (SR) 4.0.2 for the licensees failure to perform SRs specified in Units 1 and 2 TS 3/4.3.1, Reactor Trip System Instrumentation, and 3/4.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, within the required surveillance frequencies. The inspectors identified eight examples over the last three years (five examples on Unit 1 and three examples on Unit 2) where the interval between tests of the automatic actuation logic and reactor trip breaker functions required by SRs 4.3.1.1.1 and 4.3.2.1.1 exceeded the maximum surveillance interval allowed by TS. The licensee entered this issue into their corrective action program as PER 369938. Corrective actions included ensuring that work control processes correctly implement the required surveillance intervals. The finding was determined to be greater than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, extending beyond the required maximum interval between TS surveillance tests affects the ability to confirm continued availability of TS equipment, and the ability to detect potential latent operability concerns in a timely manner. Using Inspection IMC 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) since it did not represent an actual loss of safety function of a single train for greater than the associated TS allowed outage time. The inspectors did not identify that the cause of this finding was related to any of the cross-cutting aspects defined in IMC 0310, and therefore no cross-cutting aspect was assigned to this finding
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Eugene-20Guthrie-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"