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05000259/FIN-2018003-02Licensee-Identified Violation2018Q3This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Violation: 10 CFR 50.48(c)(3)(ii) required, in part, the licensee to complete its implementation of the methodology in Chapter 2 of NFPA 805 (including all required evaluations and analyses) and, upon completion, modify the fire protection plan. NFPA 805 Chapter 2, section 2.4.2.2.1, Circuits Required in Nuclear Safety Functions required, in part, that circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation or that result in the mal-operation of the equipment identified. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.
05000296/FIN-2018003-01Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints2018Q3A self-revealed SL IV NCV of Technical Specification (TS) 3.4.3, Safety Relief Valves, was identified when the licensee discovered, through as found test results, that three of the thirteen main steam relief valves (MSRVs) that were removed during the Spring 2018 Unit 3 outage had as found lift settings outside of the +/- 3 percent band required for their operability. The LER was associated with three of the thirteen MSRVs as found setpoints being outside of the +/- 3 percent setpoint band required for their operability. This was discovered on May 17, 2018, following as-found testing results conducted on all thirteen MSRVs that were removed during the refueling outage. The licensee determined that the three MSRV pilot discs had corrosion bonding to their valve seats as a result of their platinum anti-corrosion coatings flaking off. The licensee determined that these three MSRVs were inoperable for an indeterminate period of time from March 26, 2016, when the unit entered Mode 2 (beginning of operating cycle) to February 17, 2018, when the unit entered Mode 4 (beginning of refueling outage). The inspectors reviewed the licensee event report and determined that the report adequately documented the summary of the event including the cause and potential safety consequences. The inspectors also reviewed other documents that indicate that this type of failure is a known industry issue associated with this type of valve.
05000296/FIN-2018012-01Failure to correct an inoperable 250V Shutdown Board Battery Charger2018Q3A self-revealed, Green, NCV of Technical Specifications (TS) 3.8.4 was identified when the licensee failed to correct an inoperable 250V Shutdown Board (SDBD) 3EB Battery Charger on Unit 3. Specifically, in 2014 the 250V SDBD 3EB Battery Charger was entered into the Corrective Action Program (CAP) as a Condition Adverse to Quality (CAQ), but no actions were taken to correct the condition, which led to the component being in inoperable for longer than the allowed outage time defined in TS 3.8.4.
05000259/FIN-2018412-01Security2018Q2
05000259/FIN-2018010-01Licensee-Identified Violation2018Q2The Browns Ferry Nuclear Plant, Unit 3, Renewed Facility Operating License, DPR-68, License condition 2.C(7) required, in part, that TVA Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c)... Specifically, 10 CFR 50.48(c)incorporated by reference National Fire Protection Association Standard 805 (NFPA 805), and NFPA 805 section 2.4.2.2.2, Other Required Circuits, required in part, Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. Contrary to the above, since June 22, 2016, when the NFPA 805 requirements went into effect, the licensee did not implement and maintain in effect all provisions of the approved fire protection program, because the licensee did not correctly evaluate circuits that share common power supply for their impact on their ability to achieve nuclear safety performance criteria in accordance with NFPA 805.Significance: The team evaluated the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, for Mitigating Systems, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, issued May 2, 2018, and determined the finding to be of very low
05000259/FIN-2018002-04Licensee Identified Non-Cited Violation2018Q2

LER 05000259, 260, 296/2018-003-00 identified a violation of 10 CFR 50.48(c)(4)(iii). This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. 

Violation: 10 CFR 50.48(c)(4)(iii) Fire Protection required, in part, that the licensee maintain fire protection defense in depth (post-fire safe shutdown capability). Contrary to the above, from October 28, 2015 until March 10, 2018, the C3 Emergency Equipment Cooling Water (EECW) pump did not have the Fire Protection Plan required backup control panel function. Significance/Severity: Using IMC 0609 Appendix F, the violation was screened to green following a risk analysis performed by the licensee that a NRC Senior Risk Analyst reviewed and agreed was correctly performed. Corrective Action Reference(s): CR 1394604
05000259/FIN-2018002-03Failure to analyze for a Water Hammer event due to Spurious Operation of Residual Heat Removal Service Water (RHRSW) Valves during a Fire Event2018Q2An Apparent Violation (AV) of 10 CFR 50.48(c)(3)(ii) was identified for the failure to perform a required analysis using the methodology in Chapter 2 of NFPA 805 for the RHRSW piping as a result of a postulated fire scenario.
05000296/FIN-2018002-02Inoperable Residual Heat Removal (RHR) Pump Results in Condition Prohibited by Technical Specifications2018Q2A self-revealed SL IV NCV of TS 3.5.1 and 3.6.2.3 was identified when the licensee discovered that the 3A RHR pump was inoperable for longer than the allowed outage time and follow on action completion time.
05000259/FIN-2018002-01HPCI System Over Pressurization due to Failure to Maintain Procedure2018Q2A self-revealed, Green, NCV of 10 CFR 50, Appendix B, Criterion V Instructions, Procedures, and Drawings was identified for failure to maintain procedure 2-SR-3.8.4.3(MB-2) Revision 11, Main Bank 2 Battery Service Test. Specifically, the licensee failed to evaluate the impact of an emergent, Unit 2 procedure revision to a step intended to mitigate over pressurizing Unit 1 High Pressure Coolant Injection (HPCI) system
05000259/FIN-2018001-02Unauthorized Entry into a High Radiation Area(HRA2018Q1A self-revealing, Green, NCVof Technical Specification (TS)5.7.1, was identified for a worker who entered a HRA without proper authorization. Specifically, the worker entered the Unit 3 A & C Residual Heat Removal Heat Exchanger Room using an incorrect Radiation Work Permit and without being briefed on the radiological conditions.
05000296/FIN-2018001-04Inadequate Configuration Control of High Pressure Coolant Injection (HPCI)ValveDesign Issues2018Q1A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion III,was identified when the licensee failed to ensure adequate control of valve design configurations in accordance with NPG-SPP-9.3, Plant Modifications and Engineering Change Control Revision 6. Specifically, the licensee changed, over time, HPCI discharge valve yoke nut and bearing components contrary to original design without documenting or evaluating the changes
05000259/FIN-2018001-01Inadequate Post-Maintenance Testing of 4kV Breaker Stationary Switches2018Q1A self-revealing,Green, NCV of 10 CFR Part 50 Appendix B, Criterion V,was identified when the licensee failed to perform an adequate post-maintenance test in accordance with NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, the post maintenance testing on the 3C diesel generator output breaker did not ensure that all contacts on replacement stationary switches successfully changed state after installation.
05000259/FIN-2018001-03Failure to Implement Controls for Locked High Radiation Area (LHRA) Access2018Q1A self-revealing, Green, NCVof TS 5.7.2, was identified for the failure to control access to a LHRA. Specifically, a worker installed and climbed a ladder in the Unit 3 drywell without Radiological Personnel (RP) present. In doing so, the worker accessed an area with dose rates >1 rem/hr that had not been posted, locked, or surveyed prior to entry
05000260/FIN-2017004-01Inadequate Determination of Operability for the HPCI System2017Q4Two examples of an NRC-identified NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" were identified for the licensee's failure to properly implement operability evaluation requirements for degraded High Pressure Coolant Injection (HPCI) components. Specifically, from September 23 to September 28, 2017, the operability evaluations for degraded Unit 2 and 3 HPCI injection valves 2/3- FCV-73-44 did not provide reasonable assurance of operability as per the sites operability review procedures.The performance deficiency was determined to be more-than-minor because it impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. As an immediate corrective action, the licensee later performed maintenance to open and inspect these valves. Subsequently the licensee initiated condition reports and a Performance Assessment Worksheet to assess the training for such evaluations. The violation was entered into the licensee's corrective action program (CAP) as CR 1341458. The inspectors determined that the finding had a cross-cutting aspect of Evaluation in the Problem Identification and Resolution area (P.2), because the organization concluded Technical Specification operability prior to thoroughly investigating these issues commensurate with their potential safety significance.
05000259/FIN-2017004-03Licensee-Identified Violation2017Q4The following licensee-identified violation of NRC requirements was determined to be of very low safety significance and met the NRC Enforcement Policy criteria for being dispositioned as an NCV. 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, required, in part, thatmeasures shall be established to assure that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, from May 2011 to September 26, 2017, Browns Ferry staff had not identified that a Reactor Building Crane degraded, adverse condition existed. The crane, which supports the safety-related movement function for irradiated fuel assemblies, had not been meeting the required Single Failure-Proof qualification as described in the Final Safety Analysis Report section 12.2.2.5 and licensee commitments to NRC NUREG 0554 since the last wire rope replacement in May 2011. Review for cause by the licensee determined that the Main Hoist Equalizer Arm had been resting on one of its stops, causing a loss of the shock absorbing Single Failure function. The finding screens to green per IMC 0609 Appendix A, Exhibit 3 as it was a only a qualification issue that did not cause mechanical damage to fuel, did not result in a loss of spent fuel pool water inventory, and did not affect SFP component placements. The licensee entered this issue into the CAP as CR 1341964. Immediate corrective action was to reestablish the Single Failure qualification per WO 119082082.
05000296/FIN-2017004-02Failure to Perform an IDO without delay for 3A EDG after Observing Indications of a Degraded Condition2017Q4The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for failure to perform an immediate operability determination (IDO) for 3A Emergency Diesel Generator (EDG) upon discovering a degraded condition. Specifically, on December 19, 2017, the licensee failed to perform an IDO after identifying and confirming less than minimum cooling flow, thus leaving the EDG in an indeterminate state of operability.The performance deficiency is more than minor because it was associated with the equipment performance attribute and affected the associated cornerstone objective to ensure availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. As a corrective action, the licensee performed operations to restore flow within the acceptable range and performed an IDO. The violation was entered into the licensee's CAP as CR 1370601. The inspectors determined that the finding had a cross-cutting aspect in the human performance area of H.13, Consistent Process, because the performance deficiency was caused by not following a consistent, systematic approach to making a decision concerning operability of the affected DG.
05000259/FIN-2017003-02Failure to Maintain Intake Building Flood Barrier2017Q3An NRC- identified NCV of Technical Specification (TS) 5.4.1, Procedures, was identified for the failure to follow procedure MCI -0-023- PMP003, Emergency Equipment Cooling Water (EECW) and Residual Heat Removal Service Water Pump (RHRSW) Removal and Reinstallation, Revision 22. The performance deficiency is more than minor because it affected the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective. A detailed risk evaluation by a regional SRA determined the finding was Green . The licensee entered the violation into the CAP as CR 1338684. The finding had a cross cutting aspect in the Avoid Complacency component of the Human Performance area because the maintenance staff chose to not refer to a previously related condition report (CR) (PER 599190) or the maintenance procedure that were corrective actions for a previous NRC finding. (H.12).
05000259/FIN-2017003-01Degraded EDG Flood Door Seals2017Q3An NRC- identified non- cited violation (NCV of 10 CFR Part 50, Appendix B, Criterion V was identified for the licensee's failure to use appropriate procedural surveillance criteria to ensure the diesel generator buildings were protected against flood- water up to the design basis flood elevation. The annual door inspection procedure did not contain instructions with appropriate acceptance criteria to determine whether the diesel generator building doors would create a watertight seal when closed. The performance deficiency is more -than -minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective. A detailed risk evaluation by a regional Senior Risk Analyst ( SRA ) determined the finding was of very low safety significance (Green) . The licensee entered the violation into the corrective action program (CAP) as CR 1306268. The inspectors determined that the finding had a cross -cutting aspect in the Self -Assessment area of the Problem Identification and Resolution aspect (P.6), because recent self - assessments had not been self -critical of the external flood protection program and practices.
05000260/FIN-2017002-04Failure to Implement Corrective Actions to Prevent the Recurrence of a Reactor Scram Due to IRM spiking2017Q2Green . A self -revealing non- cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI , Corrective Action, was reviewed for the licensees failure to establish measures to assure that corrective action was taken to preclude repetition of a significant condition adverse to quality (SCAQ) . The licensee failed to correct electronic noise problems with the scram reset switch which led to a March 29, 2017, reactor scram. As an immediate corrective action, the licensee initiated more rigorous test s to identify noise vulnerabilities on Intermediate Range and Source Range Monitors . The licens ee entered this issue into their corrective action program as Condition Report (CR) 1278595. This performance deficiency wa s more -than- minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective in that the licensee failed to implement corrective actions to address I ntermediate Range Monitor (IRM) spiking following the May 24, 2012, reactor scram . T he finding was determined to be Green because it did not involve the loss of mitigation equipment . The inspectors determined that the finding had a cross -cutting aspect of Challenge the Unknown (H.11) with in the cross -cutting area of Human Performance because the licensee failed to res olve the unknown noise paths to ensure that scram vulnerabilities were corrected.
05000259/FIN-2017002-03Failure to Assure EECW Design Basis Capability2017Q2Green . An NRC- identified non- cited violation of 10 CFR Part 50, Appendix B, Criterion III was identified for the licensee's failure to correctly translate the design basis of the EECW system into technical instruction 0 -TI-579(EECW). The effects of instrument uncertainty and diesel frequency variations were not considered when establishing the minimum allowed inservice test low alert pump flow limits . As an immediate corrective action, the licensee evaluated the operability of the EECW pump and initiated corrective action to make changes to the test criteria and/or the system design analysis . The violation was entered into the licensee's corrective action program as CR 1288208. The performance deficiency was more- than- minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that there was a reasonable doubt on the operability of the B3 EECW pump since portions of the adjusted pump curve would be below the minimum pump curve established in the design basis calcul ation. Additionally, there was a significant reduction in available margin for the pump under design basis conditions. The finding was determined to be Green because the finding was a deficiency affecting the design of a mitigating system, but the pump maintained its operability. The inspectors determined that the finding had a cross -cutting aspect of Human Performance (H.6 ) within the cross -cutting area of De sign Margins because engineers did not demonstrate the characteristic of ensuring that design margins were guarded and changed only through a systematic and rigorous process .
05000259/FIN-2017002-02Non -conservative Assumptions in Emergency Drain Capacity Design Review2017Q2Green . An NRC- identifi ed non- cited violation of 10 CFR 50, Appendix B, Criterion III was identified for the licensee's failure to verify the adequacy of the U nit 1 and 2 diesel building emergency drain pipe to mitigate a postulated internal flood. Specifically, the licensees design review contained non- conservative assumptions. As an immediate corrective action, the licensee reevaluated the potential water accumulation and concluded the diesel generators were still protected. The violation was entered into the licensee's corrective action program as CR 1303737. The performance deficiency 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 capability of systems that respond to initiating events to prevent undesirable consequenc es. Specifically, non- conservative assumptions in calculation MDQ00004020110008 resulted in inaccurate conclusions about the capacity of the drain and the resulting water accumulation in the building. The finding was determined to be Green because it represented a deficiency affecting the design of the drain piping, but it maintained its functionality. Functionality was preserved because additional evaluation showed that the resulting water accumulation would not affect any safety related equipment . No cross -cutting aspect was assigned because it was not considered to be reflective of current licensee performance because the performance deficiency occurred more than three years ago .
05000259/FIN-2017002-01Inadequate Fire Risk Evaluation for Postulated Fires Affecting EECW Strainers2017Q2Green . An NRC- identified non- cited violation of 10 CFR 50.48(c) and NFPA 805, Section 2.4.2.4 was identified for the licensee's failure to perform an adequate engineering analysis to determine the effects of fire on the ability to achieve the nuclear safety performance criteria . Specifically, the licensees fire risk evaluation (FRE) of the effects of fire on the Emerge ncy Equipment Cooling Water (EECW ) strainers did not have an adequate basis . As an immediate corrective action, the licensee performed plant -specific analyses to determine the effects of fire on the functionality of EECW strainers and EECW system . The violation was entered into the licensee's corrective action program as CR 1263434. The performance deficiency was determined to be more -than- minor because it wa s associated with the protection against external factors attribute of t he Mitigating Systems cornerstone and adversely impacted the cornerstone objective in that failure to adequately 3 analyze the effects of fire damaged cables for the EECW strainers and backwash valves impacted the objective of ensuring the reliability of the E ECW system during a fire. This finding was determined to be Green because the finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. The inspectors determined that the finding had a cross -cutting aspect of Avoid Complacency (H.12) within the cross- cutting area of Human Performance because the licensee did not recognize that historical assumptions about long -term strainer functionality could contain mistakes and latent issues during development of the nuclear safety capability analysis.
05000259/FIN-2017002-05Licensee-Identified Violation2017Q210 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, required, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, between October 5, 2016 , and December 22, 2016, the 4kV shutdown board C degraded voltage relay timer was not installed in accordance with MAI -3.8, Installation of Electrical Components. The failure to install mounting screws of an appropriate length with suitable thread engagement for the seismic restraining strap resulted in the relay being inoperable for longer than the Technical Specification allowed outage time. The licensee entered the violation into the corrective action program as CR 1244680 and replaced t he damaged mounting screw and installed the seismic restraining strap. Using an exposure time of 78 days, the change in core damage frequency was conservatively estimated to be less than 4E -8 per year. The most dominant core damage s equences were those involving the loss of the high pressure injection systems. The significance of the finding was limited because it did not affect the 22 ability of the diesel generator to automatically start under loss of of fsite power conditions and it did not affect the ability of operators to manually start the diesel generator in response to degraded voltage conditions. The inspectors determined the finding was Green .
05000259/FIN-2017408-01Security2017Q2
05000259/FIN-2017405-01Licensee-Identified Violation2017Q1
05000259/FIN-2017001-03Failure to Perform Airborne Radioactivity Surveys2017Q1Green. An inspector-identified NCV of TS 5.4.1 was identified for the licensees failure to obtain an air sample while performing work in an area with smearable contamination levels greater than 50,000 disintegrations per minute (DPM) per 100cm2. This performance deficiency was determined to be greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors determined the finding to be of very low safety significance (Green). The licensee entered the issue into their CAP (CR 1219539) and, since the work created airborne radioactivity in the area, performed in-vivo monitoring on the affected workers to assess doses from the intake of radioactive material. This finding involved the cross-cutting aspect of Human Performance, Avoid Complacency, (H.12), because, considering the contamination levels present, RP staff underestimated the risk for potential airborne radioactive material in the area
05000260/FIN-2017001-01Failure to Take Corrective Actions to Preclude a Repeat Failure of a Containment Isolation Valve2017Q1Green. An NRC identified non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees inadequate corrective actions to preclude repetition (CAPR) of a significant condition adverse to quality (SCAQ). The licensees failure to take appropriate CAPRs for a SCAQ that resulted in an inoperable RCIC containment isolation check valve was a performance deficiency. The licensee entered the condition into their corrective action plan as condition report (CR) 1265552, performed repairs to the valve, and initiated a new root cause analysis. This performance deficiency was more than minor, because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective because the misalignment of the stem to disc for 2-CKV-71-14 resulted in a loss of reliability. The finding screened as Green because the RCIC subsystem remained operable. The finding was not assigned a cross-cutting aspect because the cause was not related to current licensee performance.
05000259/FIN-2017001-02Unauthorized Entry into a High Radiation Area2017Q1Green. A self-revealing NCV of Technical Specifications (TS) 5.7.1 was identified for a worker who entered a High Radiation Area (HRA) (Unit 1 reactor building steam tunnel) without proper authorization. This performance deficiency was determined to be greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors determined the finding to be of very low safety significance (Green). The licensee entered the issue into their Corrective Action Program (CAP) as CR 1219539 and took immediate corrective actions including restricting Radiologically Controlled Area (RCA) access for the individuals involved and performing confirmatory surveys of the area. This finding involved the cross-cutting aspect of Human Performance, Teamwork, (H.4), because a significant contributor to this event was poor communication between different work groups (workers entering the reactor building steam tunnel and RP personnel at the control point). (Section 2RS1)
05000259/FIN-2017001-05Licensee-Identified Violation2017Q110 CFR 50 Appendix B, Criterion III, Design Control required, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, between July 17, 2014, and January 8, 2017, the licensee failed to correctly translate into applicable drawings as required by their NPG-SPP-9.3 Nuclear Plant Modifications and Engineering Change Control procedure the changes associated with DCN 70491 to the EDG D output breaker. This resulted in two separate modifications using the same terminal point that caused a short circuit when the breaker was manually closed. This violation is documented in the licensees CAP as CR 1248939. This violation screened as Green because it was determined that the EDG D was operable during this entire period.
05000259/FIN-2017001-04Failure to Control the Issuance of Instructions and Drawings for Transformer Replacements2017Q1Green. An NRC identified NCV of 10 CFR Part 50, Appendix B, Criterion VI, Document Control, was identified after maintenance on safety-related 4kv to 480 volt transformers TS1A and TS1B (Unit 1) resulted in the windings tap setting being misconfigured. The licensees failure to develop work instructions to change TS1A and TS1B transformer configuration was a performance deficiency. This performance deficiency was more than minor because it impacted the Mitigating Systems cornerstone attribute of configuration control in that the loads supplied by 480 volt shutdown boards 1A and 1B were challenged by this misconfiguration. The finding screened as Green because the electrical system remained operable. The licensee entered the condition into their corrective action plan as CR 1221265 and corrected the tap setting. The finding was not assigned a cross-cutting aspect because the cause was not related to current licensee performance.
05000259/FIN-2017406-01Security2017Q1
05000296/FIN-2016004-02Inadequate Prompt Determination of Operability for the HPCI System2016Q4Green. An NRC identified NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" was identified for the licensee's failure to accomplish the Prompt Determination of Operability (PDO) for CR 1039036 in accordance with the requirements of NEDP-22, "Operability Determinations and Functional Evaluations," Sections 3.2.2.E, 3.2.2.G, and Attachment 2. As an immediate corrective action, the licensee revised the PDO to include an evaluation that supported a reasonable expectation of operability. The licensee entered the violation into the corrective action program as CR 1219620. The performance deficiency was more-than-minor because it was associated with the Human Performance 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). Specifically, after considering the inadequacies of the PDO, additional and significant evaluation was required to maintain reasonable assurance of the HPCI system operability. The doubt stemmed from uncertainty about the actual water level in the turbine, the expected transient severity, and the unanalyzed effects of the piping configuration. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Exhibit 2 "Mitigating Systems Screening Questions," dated June 19, 2012. The inspectors determined the finding was Green because it was a deficiency affecting the qualification of HPCI, but it maintained its operability. The inspectors determined that the finding had a cross-cutting aspect of Evaluation in the Problem Identification and Resolution area (P.2), because the organization did not thoroughly investigate this issue commensurate with its potential safety significance.
05000259/FIN-2016004-01Inadequate Reassembly Procedure for HPCI Steam Line Inboard Isolation Valve Actuator2016Q4Green. A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" was identified for the licensee's failure to provide sufficient detail, in this case, appropriate to the work activity in procedure, MCI-0-000-ACT004, Maintenance of SMB-0 through SMB-4T Limitorque Actuators, which impacted the design features of HPCI valve 1-FCV-73-2. As an immediate corrective action, the valve was repaired and corrective actions initiated to address the quality and details of motor operated valve procedures. The licensee entered the violation into their corrective action program as Condition Reports (CRs) 1228056 and 1229289. The performance deficiency was more than minor because it was associated with the Equipment Performance 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). Specifically, the reliability of the valve was reduced due to the impending worm gear teeth failure. While the valve was full open, the High Pressure Coolant Injection (HPCI) pump was able to fulfill its safety function of injecting water into the reactor. Since the valve was able to close upon entering outage U1R11, the HPCI system was able to isolate the HPCI steam supply line in the event of a HPCI steam line break. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Mitigating Systems Screening Questions. The inspectors determined the finding screened to Green as HPCI was not unavailable longer than its TS allowed outage time and the finding did not involve the loss or degradation of equipment designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect of Procedure Adherence in the Human Performance area (H.8), because individual staff members did not review procedures and instructions prior to work to validate they were appropriate for the scope of work.
05000259/FIN-2016011-02Failure to Adequately Identify and Evaluate All Circuit Failures for NSCA Credited Equipment2016Q3The NRC identified a violation of 10 CFR 50.48(c) for the licensees failure to properly identify circuits required for the nuclear safety function. Specifically, the licensees Nuclear Safety Capability Assessment (NSCA) failed to identify that fire-induced failure of cables associated with the undervoltage trip function of the 4KV Shutdown Board could cause the shutdown board to not shed loads upon an undervoltage condition. This could lead to overloading the emergency diesel generator (EDG) credited for powering the shutdown board. This item was entered into the CAP as CR 1199002. The affected area was already covered by an hourly roving fire watch as a compensatory measure. Additionally, the licensee submitted EN 52150 to the NRC, documenting this as an unanalyzed condition. The licensees failure to identify circuits required for the nuclear safety function, as required by Section 2.4.2.2.1 of NFPA 805 was a PD. The PD was determined to be more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e., fire) and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to analyze the effects of fire damage on the 4kV shutdown bus undervoltage circuitry could result in overloading the emergency diesel generator (EDG) credited for powering the shutdown board. Using the guidance of IMC 0609, App. F, the finding was screened as Green because the risk increase associated with the finding was an increase of core damage frequency of <1E-6/year. There was no cross cutting aspect assigned to this finding because it was not indicative of current licensee performance since the original ignition source and target walkdowns were performed more than 3 years ago. (Section 1R05.06)
05000259/FIN-2016003-04Inadequate Prompt Determination of Operability for HPCI Steam Line Inboard Isolation Valve2016Q3An NRC identified NCV of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" was identified for the licensee's failure to promptly identify conditions adverse to quality associated with the prompt determination of operability (PDO) for CR 1061051. As an immediate corrective action, the licensee entered the violation into the licensee's corrective action program as CR 1193943. The performance deficiency was more-than-minor because it was associated with the Human Performance 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). Specifically, had the deficiencies in the PDO been identified, engineers would have recognized that the resulting stresses exceeded allowable design stresses in the valve vendor's weak link analysis and approached the yield strength of the stem material. As a result, the practice was permitted to continue until the valve stem catastrophically failed. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined the finding required a detailed risk evaluation because the finding represented a loss of system function and/or function for the high pressure coolant injection (HPCI) system. Senior Reactor Analyst performed a detailed risk evaluation using the Standardized Plant Analysis Risk (SPAR) model for Browns Ferry Unit 1. The HPCI system was modeled as unavailable for a conservative exposure period of 7 days. The delta CDF estimate was less than 1E-6/yr range, which represents a finding of very low safety significance (Green). The dominant core damage sequence was an inadvertent open relief valve, failure of HPCI, and failure to depressurize. The availability of additional injection sources helped minimize the risk significance. The inspectors determined that the finding had a cross-cutting aspect in the Design Margins area of the Human Performance aspect (H.6), because engineers did not demonstrate the behavior of carefully guarding margins to ensure that safety related equipment was operated and maintained within design margins.
05000296/FIN-2016003-05Alternate Depressurization Valve Inoperable Longer than the Allowed Outage Time2016Q3A self-revealing NCV of TS 3.5.1, Emergency Core Cooling Systems, Condition E in that an inoperable Automatic Depressurization System (ADS) valve function existed longer than the allowed technical specification time. The licensee implemented corrective actions by declaring the affected component inoperable per technical specifications, identified preventative maintenance procedures as the cause, repaired the breaker stabs to restore the circuit, and re-performed the surveillance to establish operability. This issue was entered into the licensee's corrective action program as CR 1161991. The performance deficiency was more than minor because it adversely affected the Mitigating Systems cornerstone attribute of equipment performance. Specifically, one of the TS required ADS valves opening capability was not fully qualified. Using NRC IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) because the finding did not represent a loss of system safety function as the other five Main Steam Relief Valve (MSRV) ADS functions were still available. The inspectors assigned a cross cutting aspect of Identification since the licensee had not taken sufficient post maintenance actions to verify function of the alternate breaker for the ADS valve 3-PCV-001-0022. (P.1)
05000260/FIN-2016003-01Failure to Ensure Adequate Piping Clearances After MOV Modification2016Q3An NRC identified non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" was identified for the licensee's failure to ensure sufficient clearance was available following a replacement of the Core Spray minimum flow valve actuator motors. Modifications personnel failed to identify that the resulting clearances were less than permitted by TVA procedure MAI-4.10 Piping Clearance Instruction and that they required an engineering evaluation. As an immediate corrective action, the licensee cut away portions of floor grating to establish an acceptable amount of clearance for the valves. The violation was entered into the licensee's corrective action program as CRs 1161330 and 1169591. The performance deficiency 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). Specifically, the inadequate clearance resulted in an analysis showing that ASME code allowable design stresses would be exceeded under accident conditions. Exceeding design stresses created a reasonable doubt on the operability and reliability of loop 2 of the Core Spray system for Units 2 and 3. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined the finding was Green because the finding was a deficiency affecting the qualification of the Core Spray loop. Operability was maintained because an engineering evaluation demonstrated, through the use of alternative analytical methods, that the piping stress criteria in Appendix F of Section III of the ASME Boiler and Pressure Vessel Code was satisfied and that the stresses in the valve would not cause distortions of a magnitude that would prevent operation of the valve. The inspectors did not assign a crosscutting aspect because the performance deficiency was not reflective of present licensee performance since it occurred more than three years ago.
05000260/FIN-2016003-02Failure to Implement Compensatory Roving Fire Watch2016Q3An NRC identified non-cited violation (NCV) of Renewed License Number DPR-52, condition 2.C.(14) was identified for the licensees failure to implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c). Specifically, the licensee failed to establish a compensatory roving fire watch, within 1 hour of rendering the spray systems that protect the Main 500kV transformer 2B and Unit Service Station Transformer (USST) 2B nonfunctional. As an immediate corrective action, the licensee established the required fire watch and entered the violation into the licensee's corrective action program as CR 1203990. The performance deficiency was more-than-minor because it was associated with the protection against external factors (Fire) attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. This finding was evaluated in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013. The inspectors determined the finding was Green because the finding did not affect the reactors ability to reach and maintain the fuel in a safe and stable condition. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance area of Change Management (H.3) because leaders failed to clearly establish the control room's ownership of Fire Protection Requirements Manual (FPRM) usage as part of the NFPA 805 transition.
05000259/FIN-2016011-01Failure to Identify and Evaluate All Targets Within the Zone of Influence of Ignition Sources2016Q3The NRC identified a violation of 10 CFR 50.48(c) for the licensees failure to address in the Fire Probabilistic Risk Assessment (Fire PRA) the risk contribution associated with all potentially risk significant fire scenarios for a given fire compartment/fire area. The licensee did not identify and evaluate all targets that were within the zone of influence (ZOI) of ignition sources for selected fire scenarios that could potentially contribute to the risk for the fire scenarios. The licensee entered the issue in the corrective action program (CAP) as Condition Reports (CRs) 1195603 and 1197392. The affected area was already covered by an hourly roving fire watch as a compensatory measure. The licensees failure to address the risk contribution associated with all potentially risksignificant fire scenarios, as required by section 2.4.3.2 of NFPA 805, was a performance deficiency. For each example, the performance deficiency was determined to be more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e., fire) and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to analyze the full risk impact of the selected fire scenarios, and the missed targets in the ZOI for the selected fire scenarios had the potential to impact the ability to achieve safe and stable conditions. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the finding was screened as Green in step 1.6.1 Screen by Licensee PRA-Based Safety Evaluation. There was no cross cutting aspect assigned to this finding because it was not indicative of current licensee performance since the original ignition source and target walkdowns were performed more than 3 years ago. (Section 1R05.06)
05000296/FIN-2016003-06Main Steam Relief Valves Inoperable Longer than Allowed Outage Time2016Q3A self-revealing NCV of TS 3.4.3, Safety Relief Valves was identified for two required MSRVs being inoperable longer than the allowed outage time and follow on action completion time. The licensees immediate corrective action was to replace all Unit 3 MSRV pilot valves prior to the completion of the refueling outage. This issue was entered into the licensees corrective action program as CR 1157981. The performance deficiency was more than minor because it adversely affected the Mitigating Systems cornerstone attribute of equipment performance. Specifically, two required MSRVs were not able to lift within their required pressure band. This performance deficiency was screened using NRC IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. This performance deficiency screens to Green because although the system was inoperable for greater than its allowed outage time and follow on action completion time, the system maintained its safety function. The inspectors assigned a cross cutting aspect of Resolution since the licensee has not taken sufficient corrective actions to address the continued out of tolerance lift results caused by corrosion bonding of the MSRV pilot valve seats. (P.3)
05000259/FIN-2016003-07Licensee-Identified Violation2016Q3Title 10 of the Code of Federal Regulations (CFR) Part 50.54(i-1), states, in part, ...the licensee shall have in effect an operator requalification program. The operator requalification program must, as a minimum, meet the requirements of 55.59(c) of this chapter. Notwithstanding the provisions of 50.59, the licensee may not, except as specifically authorized by the Commission decrease the scope of an approved operator requalification program. Contrary to the above, the licensee reduced the scope of the requalification program for a licensed Reactor Operator (RO) which did not meet the requalification examination requirements of 10 CFR 55.59(c)(4)(i) from January 1, 2012, until the licensee requested the ROs license be withdrawn on September 30, 2016. Specifically, the operator did not complete the requalification cycle for the years 2011- 2012 and did not take an annual operating exam or biennial written exam as required by 10 CFR 55.59. In accordance with the NRC Enforcement Policy, this violation was classified as Severity Level IV Violation (Section 6.4.d) because the operator was administratively restricted from performing licensed duties during this time. This violation was entered into the licensees corrective action program under CR 1195643.
05000259/FIN-2016003-03Failure to Maintain The High Pressure Fire Protection System Piping2016Q3A self-revealing Non-cited Violation (NCV) of Technical Specification (TS) 5.4.1.d, Fire Protection Program Implementation, was identified for the licensees failure to maintain the integrity of the high pressure fire protection piping. The licensees immediate corrective action was to isolate the leak and entered this issue into their corrective action program as CR 1102016. This performance deficiency was more than minor because it adversely affected the Initiating Events cornerstone objective of protection against external factors such as fire. Specifically, the high pressure fire protection system piping was unable to maintain the required pressure during a system demand. This finding was evaluated in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013. The inspectors determined the finding was Green because the finding did not affect the reactors ability to reach and maintain the fuel in a safe and stable condition. The inspectors assigned a cross cutting aspect of Operating Experience because there was a similar occurrence of a fire protection piping break at Browns Ferry caused by heavy construction vehicle traffic in 2014 (P.5).
05000259/FIN-2016010-01Failure to Include Required Gasket Replacement in Limit Switch Surveillance Procedure2016Q2An NRC-identified non-cited violation (NCV) of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include vendor requirements for maintaining the environmental qualification of the main steam isolation valve (MSIV) limit switches in maintenance procedures. Specifically, not maintaining the MSIV limit switches in their qualified condition impacts their reliability. The licensee entered this issue into the corrective action program as CR 1160702. The licensee evaluated the impact of the incorrect guidance, and determined that all three units were affected, and that the MSIV limit switches remained operable, although they were in an unqualified condition. The licensee plans to correct the affected procedures. This performance deficiency was more than minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining the MSIV limit switches in their qualified condition impacted their reliability. The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The team determined that no cross-cutting aspect was applicable because the finding was not indicative of current licensee performance.
05000296/FIN-2016002-02Failure to Declare Notification of Unusual Event2016Q2The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50.54(q)(2), for the licensees failure to declare a Notification of Unusual Event (NOUE) within 15 minutes of entry conditions being met. Specifically, on April 6, 2016, at 3:05 pm, Browns Ferry Unit 3 main control room (MCR) operators received a high-high radiation alarm on the main steam lines (MSL) that met Emergency Action Level (EAL) 1.4-U for declaring a NOUE. The licensee initiated CR 1159943 to address the issue. This performance deficiency was more than minor because it was associated with the Emergency Preparedness cornerstone attribute of Emergency Response Organization Performance, and adversely affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, on April 6, 2016, personnel did not declare a NOUE within 15 minutes of initial indications that EAL 1.4-U had been exceeded. The performance deficiency is associated with the Emergency Classification Planning Standard, and is considered a Risk Significant Planning Standard (RSPS). The failure to declare a NOUE when directed by the EAL Matrix is considered a lost or degraded RSPS in accordance with Section 4 of Inspection Manual Chapter (IMC) 0609, Appendix B. Section 4.3.e of IMC 0609, Appendix B, provides the significance determination for a Failure to Implement, and the performance deficiency was determined to be of very low safety significance (Green). The finding was associated with a cross-cutting aspect in the Procedure Adherence component of the Human Performance area because individuals did not follow processes, procedures and work instructions that would have led them to declare in a timely manner (H.8).
05000296/FIN-2016002-01Failure to Provide Adequate Maintenance Results in Loss of Core Flow While Shutdown2016Q2A self-revealing, finding for the licensees failure to provide adequate work instructions for maintenance on the Unit 3 recirculation pump discharge valve motors which included appropriate testing as described in Procedure NPG SPP 06.9.3 Post Modification testing, was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown operations. The inspector performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix G, Attachment 3, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings and determined that the finding was of very low safety significance. This finding had a cross-cutting aspect in the area of human performance because the licensee did not ensure that design documentation was correct and that work packages provided the proper tests to ensure the Variable Frequency Drives (VFD) / Recirculation pump trip logic. (H.7).
05000260/FIN-2016002-03Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety Function2016Q2An NRC identified Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) 50.72(b)(3)(v) and 10 CFR 50.73(a)(2)(v) was identified for the licensee's failure to notify the NRC within 8 hours and submit an LER within 60 days of discovery of a condition that could have prevented the fulfillment of a safety function. Specifically, the licensee failed to notify the NRC that the High Pressure Coolant Injection (HPCI) system had been rendered inoperable due to an equipment failure. As an immediate corrective action, the licensee entered the violation into the licensee's corrective action program as CR 1185268. The licensees failure to provide the required notification constitutes a traditional enforcement violation because it impacts the NRC's ability to carry out its regulatory function. The traditional enforcement violation was determined to be Severity Level IV because it matched example 6.9.d.9 of the NRC Enforcement Policy. Because the violation is a traditional enforcement violation, no cross-cutting aspect was assigned.
05000296/FIN-2016001-09Licensee-Identified Violation2016Q1Licensee Event Report (LER) 05000296/2015-002-00 Switch Failure Rendered Automatic Startup of Some Emergency Core Cooling System Pumps Inoperable Longer than Allowed by Technical Specifications: TS 3.3.5.1 condition A required, in part, that when one or more channels of Emergency Core Cooling System (ECCS) Instrumentation were inoperable that the condition listed in table 3.3.5.1-1 be immediately entered for that channel. MJ(STA 52) switch on breaker BFN-3-BKR-211-03ED/008 failed rendering automatic start sequence timing for the 3B and 3D Core Spray pumps, the 3D RHR Pump, and the D1 RHRSW Pump sequence time to become inoperable for conditions where normal power was maintained. This resulted in the licensee not meeting the TS completion times from September 17, 2014 until January 24, 2015, for TS 3.3.5.1 condition C (Core Spray Pumps 3B and 3D), TS 3.5.1 condition B (3D RHR pump), and TS 3.7.1 condition G (D1 RHRSW pump). This licensee identified violation is documented in the licensees CAP as CR 980277. This finding was able to be screened to Green using IMC 0609 Appendix A dated June 9, 2012 because although these pumps were inoperable, their respective systems did not lose their function as emergency starts were not affected.
05000260/FIN-2016012-02Failure to Maintain Complete and Accurate Shift Logs2016Q110 CFR 50.9, Completeness and Accuracy of Information, states, in part, information required by the Commissions regulations, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. Contrary to the above, on December 21, 2014, TVA failed to maintain information required by the Commissions regulations that was complete and accurate in all material respects. Specifically, following an equipment manipulation, plant transient and subsequent realization of an operator error, TVA maintained incomplete and/or inaccurate information on the cause of the transient in the operating logs and corrective action program. Shift logs are material to the NRC, as the logs are used to provide information in the determination of chronologies, root and contributing causes, and corrective actions for post-transient safety reviews and investigation by TVA and by the NRC.
05000259/FIN-2016001-07Licensee-Identified Violation2016Q1Licensee Event Report (LER) 05000259/2015-005-00 Inboard Main Steam Isolation Valve Actuators Inoperable for Longer Than Allowed by Technical Specifications. TS 3.6.1.3 condition A required, in part, that when one or more penetration flow paths with one Primary Containment Isolation Valve (PCIV) inoperable except due to MSIV leakage not within limits that within 4 hours the affected penetration flow path be isolated by use of at least one closed and de-activated automatic valve with flow through the valve secured. TS 3.6.1.3 condition E required, in part, that when the Required Action and associated Completion Time of Condition A was not met in MODE 1, that the Unit must be placed in Mode 3 within 12 hours and Mode 4 within 36 hours. Contrary to the above, on multiple occasions between December 1, 2012 and October 29, 2015, the inboard MSIVs PCIV function was inoperable on all main steam lines on all three Units longer than the allowed outage time and the follow on action completion time. This violation is documented in the licensees CAP as CR 1098857. This finding was screened to Green using IMC 0609 Appendix H dated May 6, 2004. Table 6.2 Phase 2 Risk Significance was used to screen the finding to Green because at no point during the time period between December 1, 2012 and October 29, 2015 did any outboard MSIV leakage on any Unit exceed 10,000 scfh.
05000259/FIN-2016001-02Failure to adequately maintain emergency plan implementing procedures2016Q1The inspectors identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(2), for the licensees failure to maintain the effectiveness of its emergency plan by ensuring procedures for use by the emergency response organization are maintained and up-to-date as required by 10 CFR 50.47(b)(16). Corrective actions already taken were implementation of a revision (49) to EPIP-5, effective January 7, 2016, essentially replacing Section 3.6 and references to appropriate Appendices, and a broader scope EOC to review all site EPIPs to ensure no other inadvertent omissions were made. The inspectors determined that the performance deficiency was more than minor because it was associated with the procedure quality attribute of the Emergency Preparedness (EP) cornerstone, adversely affected the associated cornerstone objective, and may have been used had an emergency been declared. The finding was evaluated using the EP significance determination process and was identified as having very low safety significance (Green) because it was a failure to comply with NRC requirements and was not a loss of the planning standard function. The finding was associated with a cross-cutting aspect in the Evaluation component of the Problem Identification and Resolution area because the licensee failed to thoroughly evaluate a similar issue at one of its other sites to ensure extent of conditions commensurate with their safety significance are thoroughly resolved. (P.2)
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