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	Quarter	Site	Title	Description
05000296/FIN-2018012-01	2018Q3	Browns Ferry	Failure to correct an inoperable 250V Shutdown Board Battery Charger	A 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-2018003-02	2018Q3	Browns Ferry	Licensee-Identified Violation	This 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-2018003-01	2018Q3	Browns Ferry	Main Steam Relief Valves Lift Settings Outside of Technical Specifications Required Setpoints	A 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.
05000259/FIN-2018002-04	2018Q2	Browns Ferry	Licensee Identified Non-Cited Violation	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-2018002-03	2018Q2	Browns Ferry	Failure to analyze for a Water Hammer event due to Spurious Operation of Residual Heat Removal Service Water (RHRSW) Valves during a Fire Event	An 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-2018002-02	2018Q2	Browns Ferry	Inoperable Residual Heat Removal (RHR) Pump Results in Condition Prohibited by Technical Specifications	A 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-2018002-01	2018Q2	Browns Ferry	HPCI System Over Pressurization due to Failure to Maintain Procedure	A 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
05000251/FIN-2017004-01	2017Q4	Turkey Point	Failure to Perform an Adequate ASME BPVC Section XI Repair/Replacement Plan for a Code Class 1 and 2 Replacement	A NRC-identified NCV of 10 CFR 50.55a, Codes and Standards, was identified for the failure to adequately perform a Boiler and Pressure Vessel Code (BPVC) class 1 and 2 replacement activity in accordance with the Turkey Point Plant American Society of Mechanical Engineers (ASME) Section XI Repair/Replacement Program. Specifically, the licensee did not ensure a system leakage test conducted on October 19, 2017, was appropriately evaluated to meet the requirements of ASME Section XI for pre-service leakage testing of a Unit 4 high head safety injection (HHSI) cold leg injection check valve that was replaced on October 15, 2017. This issue was entered into the licensees Corrective Action Program (CAP) as ARs 2235484 and 2239149. Corrective actions included documenting a formal bases for current operability via a prompt operability determination and updating work order (WO) documentation to fully comply with ASME BPVC Section XI requirements. This performance deficiency was determined to be more than minor because an inadequate inservice inspection repair/replacement plan adversely affected the Reactor Coolant System (RCS) Equipment and Barrier Performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the issue had very low safety significance because there was no actual degradation of the RCS boundary. This finding was assigned a cross-cutting aspect in the Procedure Adherence component of the Human Performance cross-cutting area, in that the licensee did not effectively evaluate and appropriately implement the ASME BPVC requirements in the 4-873A Repair/Replacement Plan which were reiterated in licensee administrative procedure 0-ADM-532, ASME Section XI Repair/Replacement Program (H.8).
05000251/FIN-2017004-03	2017Q4	Turkey Point	Inadequate Installation of Outdoor Use Electrical Enclosures Results in Manual Reactor Trip	A self-revealing finding (FIN) was identified for failure to ensure the 4B and 4C main feedwater regulating valve (MFRV) control circuits remained free from the effects of water intrusion or condensation in electrical enclosures. Specifically, a hand selector switch (HSS) enclosure for the 4C MFRV redundant positioners was flooded during wind-driven rain and resulted in the 4C MFRV failing closed, lowering 4C steam generator water level, and a subsequent Unit 4 manual reactor trip initiated by control room operators.Engineering Change (EC) 246879 appropriately selected NEMA-4X rated enclosures for the HSSs but associated SPEC-C-065 did not provide critical configuration details for the enclosure installations. Water collected in the 4B and 4C MFRV positioner HSS enclosures because the penetrations were on top of the enclosures and not properly sealed and the bottom of the enclosure did not have a weep hole.This performance deficiency 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 cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, because the failure resulted in lowering steam generator water levels and caused control room operators to complete a fast load reduction and manually trip the reactor. In accordance with NRC IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, the inspectors determined that the issue had very low safety significance because it only caused a reactor trip and did not cause the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Since EC 246879 and associated work orders were completed in 2013, the inspectors determined the finding was not indicative of current licensee performance and was not assigned a cross-cutting aspect.
05000251/FIN-2017004-02	2017Q4	Turkey Point	Failure to Identify and Correct a Deficient CCW Penetration Seal Configuration that Exacerbates External Piping Corrosion Conditions	A NRC-identified NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to promptly identify and correct an adverse condition to quality that led to continued corrosion and significant scaling and pitting of the Unit 4 component cooling water (CCW) 18-inch headers at the penetration seals from the CCW heat exchanger room to the 10-foot pipeway. This issue was entered into the licensees CAP as ARs 2217942, 2227877, 2211843, 2236687, and 2239632. Corrective actions included removing protective boots that were inappropriately installed and not in accordance with design drawings and work order instructions, and were collecting hypersaline water that wetted carbon steel piping.The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences because corrosion and pipe wastage was ongoing and unmonitored for the Unit 4 CCW headers. In accordance with IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined the finding to be of very low safety significance because it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee failed to identify the adverse condition that allowed corrosion to continue unmonitored (P.1).
05000390/FIN-2017002-05	2017Q2	Watts Bar	Licensee-Identified Violation	Watts Bar Nuclear Plant TS 5.7.1.1 states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures in Regulatory Guide (RG) 1.33, Revision. 2, Appendix A, February 1978. Procedures for surveillance tests are applicable procedures under RG 1.33 Appendix A, 8.b. Contrary to this requirement, on April 4, 2017, surveillance procedure 0-SI-82-4, 18 Month Loss of Offsite Power with Safety Injection Test DG 1B-B, Revision 63, was not implemented as written. Specifically, Step 3.1 (3) was not followed when the 1B-B safety injection pump discharge isolation valve was closed but not tagged as directed by the procedure. As a result of not being tagged, there was no programmatic control in place to return the valve to the open position upon completion of 0-SI-82-4. Therefore, the valve was left in the closed position, causing the B train of safety injection to be inoperable from April 11, 2017, until May 10, 2017, when the valve was discovered to be closed during operator rounds. Because the 1B safety injection pump was inoperable for longer than its TS allowed outage time of 72 hours, a regional senior reactor analyst conducted a detailed risk evaluation using SAPHIRE (Version 8.1.5) and the standard model for Watts Bar (SPAR Version 8.50). The resulting change in core damage frequency was less than 1E-6; therefore, the finding was determined to be of very low safety significance (Green). The licensee entered this issue into their corrective action program as CR 1294133.
05000390/FIN-2017002-03	2017Q2	Watts Bar	Failure to Follow Procedure Results in Reactor Coolant Pump Failure to Transfer and Unit 1 Reactor Trip	Green. A self-revealed Green finding was identified for the failure to follow procedure NPG-SPP-22.207, Procedure Use and Adherence Revision 4, which requires that applicable procedures are used for all activities controlled by a written procedure. The licensee entered this into their corrective action program as CR 1291140 The failure to follow procedure NPG-SPP-22.207, Procedure Use and Adherence, Revision 4, was a performance deficiency. The performance deficiency was more than minor because it affected the Initiating Events Cornerstone attribute of Human Performance and adversely affected the cornerstone objective in that it resulted in two reactor trips. The inspectors determined that the finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment. The finding was not assigned a cross-cutting aspect since none of the CCAs described in IMC 0310 corresponded to an apparent cause or most significant causal factor of the performance deficiency. (Section 4OA3.6)
05000390/FIN-2017002-02	2017Q2	Watts Bar	Failure to Implement Clearance on Containment Isolation Valve Results in TS 3.6.3 Violation	Green. A self-revealed non-cited violation of Technical Specification (TS) 3.6.3, Containment isolation Valves, was identified for a failure to properly implement a clearance for containment isolation valve surveillance testing. Clearance 1-30-1011-WW removed fuses from a different valve than the one specified in the clearance. The licensee entered this issue into their corrective action program as CR 1245529. The failure to comply with NPG-SPP-10.2, Steps 3.1.2.B.5 and 6, was a performance deficiency. The performance deficiency was more than minor because it adversely affected the configuration control attribute of the Barrier Integrity Cornerstone because the incorrectly placed clearance resulted in the inoperability of the containment isolation valve for longer than its TS allowed outage time, reducing ensurance that the containment function assumed in the safety analyses would be maintained. The inspectors determined that this violation was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters. The finding has a cross-cutting aspect in the Avoid Complacency component of the Hum an Performance area as defined in NRC IMC 0310, because multiple personnel failed to recognize and plan for the possibility of 3 mistakes and error reduction tools, such as concurrent verification, were not appropriately implemented (H.12).
05000260/FIN-2017002-04	2017Q2	Browns Ferry	Failure to Implement Corrective Actions to Prevent the Recurrence of a Reactor Scram Due to IRM spiking	Green . 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-2017002-05	2017Q2	Browns Ferry	Licensee-Identified Violation	10 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-2017002-03	2017Q2	Browns Ferry	Failure to Assure EECW Design Basis Capability	Green . 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-2017002-01	2017Q2	Browns Ferry	Inadequate Fire Risk Evaluation for Postulated Fires Affecting EECW Strainers	Green . 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-2017002-02	2017Q2	Browns Ferry	Non -conservative Assumptions in Emergency Drain Capacity Design Review	Green . 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 .
05000390/FIN-2017002-06	2017Q2	Watts Bar	Licensee-Identified Violation	Title 10 CFR 50.72(b)(3)(v)(C) requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems or components that are needed to control the release of radioactive material. Contrary to the above, on March 9, 2017, the licensee failed to notify the NRC that reactor containment was inoperable, resulting in a condition that could have prevented fulfillment of a safety function. Specifically, an inner containment door equalizing valve was not fully shut when the outer containment door was open for entry into upper containment, thereby resulting in a direct path from containment to the auxiliary building. This failure to report was assessed using Section 2.2.4 of the NRCs Enforcement Policy using the example listed in Section 6.9.d.9, A licensee fails to make a report required by 10 CFR 50.72 or 50.73, and the issue was determined to be a SL IV violation. The licensee entered this issue into their corrective action program as CR 1273873.
05000390/FIN-2017002-04	2017Q2	Watts Bar	Failure to Report Multiple Examples of a Loss of Safety Function in accordance with 10 CFR 50.72 and 50.73	Severity Level IV. The inspectors identified a Severity Level IV non-cited violation of 10 Code of Federal Regulations (CFR) 50.72 and 50.73, with multiple examples due to the licensees failure to make the required eight-hour non-emergency notification and submit a Licensee Event Report (LER) to the NRC within 60 days for conditions that, at the time of discovery, could have prevented fulfillment of a safety function. These issues have been entered into the licensees corrective action program as condition report (CR) 1310096. The inspectors determined that the licensees failure to comply with 10 CFR 50.72(b)(3)(v) and 50.72(a)(2)(v) was a performance deficiency. This performance deficiency was dispositioned under traditional enforcement because the failure to make a non-emergency notification and submit an LER within the time requirements may impact the ability of the NRC to perform its regulatory oversight function. The violation was assessed using Sections 2.2.4 and 6.9.d.9 of the NRCs Enforcement Policy and determined to be a SL IV violation. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000391/FIN-2017002-01	2017Q2	Watts Bar	Inadequate Chemistry Procedure Results in Inoperable Containment Isolation Valves	SL IV. A self-revealed severity level (SL) IV non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when implementing an inadequate procedure resulted in rendering the steam generator chemistry sample containment isolation valves inoperable. The licensee entered this issue into their corrective action program as CR 1160910. The inspectors determined that the use of an inadequate procedure that rendered the containment isolation valves inoperable was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC-2517, Appendix C, because the use of an inadequate procedure rendered the containment isolation valves inoperable. The inspectors determined this finding to be of very low safety significance because it did not represent a breakdown of the licensees quality assurance program. This finding had a cross-cutting aspect in the work management component of the Human Performance cross-cutting area because the work process did not include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities (H.5).
05000259/FIN-2017001-05	2017Q1	Browns Ferry	Licensee-Identified Violation	10 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-2017001-04	2017Q1	Browns Ferry	Failure to Control the Issuance of Instructions and Drawings for Transformer Replacements	Green. 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-2017001-03	2017Q1	Browns Ferry	Failure to Perform Airborne Radioactivity Surveys	Green. 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
05000259/FIN-2017001-02	2017Q1	Browns Ferry	Unauthorized Entry into a High Radiation Area	Green. 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)
05000260/FIN-2017001-01	2017Q1	Browns Ferry	Failure to Take Corrective Actions to Preclude a Repeat Failure of a Containment Isolation Valve	Green. 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.
05000296/FIN-2016004-02	2016Q4	Browns Ferry	Inadequate Prompt Determination of Operability for the HPCI System	Green. 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.
05000390/FIN-2016013-02	2016Q4	Watts Bar	Failure to Provide Accurate Information	SL-IV. The NRC identified a Non-cited Violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information for the licensees failure to provide accurate information in all material respects to the Commission. The team determined on April 22, 2016, the licensee provided inaccurate information in a letter to the NRC titled, RESPONSE TO NRC LETTER CONCERNING A CHILLED WORK ENVIRONMENT FOR RAISING AND ADDRESSING SAFETY CONCERNS AT THE WATTS BAR NUCLEAR PLANT (ML16113A228). This information was material because the NRC relied on this information to conclude that TVA was in compliance with CO-EA-09-009/203 requirements. The licensee placed this issue into their corrective action program. The NRC determined this violation constituted a more than minor traditional enforcement violation associated with failure to provide accurate information. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address violations which impede the NRCs ability to regulate using traditional enforcement. The inspector determined that the licensees failure to provide accurate information was a violation of 10CFR50.9 which had the potential to impede or impact the regulatory process, and therefore subject to traditional enforcement as described in the NRC Enforcement Policy, dated November 1, 2016. This violation is characterized as a Severity Level IV violation because it was similar to Example Section 6.9.d.1 of the NRC Enforcement Policy.
05000259/FIN-2016004-01	2016Q4	Browns Ferry	Inadequate Reassembly Procedure for HPCI Steam Line Inboard Isolation Valve Actuator	Green. 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.
05000390/FIN-2016013-01	2016Q4	Watts Bar	Failure to Implement the Program Requirement to Enter Issues into the CAP	Green. The NRC identified a Finding for the licensees failure to consistently implement the program requirements of the CAP. Specifically, the licensee failed to implement NPG-SPP-22.301, section 3.2.2 which required the licensees staff to initiate a Condition Report (CR) to enter various items into their CAP. The licensee placed this issue into their corrective action program. The performance deficiency was more than minor because, if left uncorrected, issues would remain unanalyzed that could represent a more significant safety concern. The performance deficiency was screened using IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Cornerstone dated June 19, 2012. The finding screened to Green because none of the examples were related to any structure, system, component, (SSC) 3 exceeding its technical specification allowed outage time. A cross cutting aspect of Identification was assigned because the licensees threshold for identifying and entering issues into their CAP was not low enough as defined by their procedures. (P.1)
05000390/FIN-2016013-04	2016Q4	Watts Bar	Licensee-Identified Violation	Technical Specification 3.5.2 Emergency Core Cooling Systems (ECCS) Operating Condition A required, in part, that while in Mode 1 that if one train becomes inoperable that it be restored to an operable status in 72 hours. Condition B required action to place the unit in Mode 3 in 6 hours and Mode 4 in 12 hours if that train is not restored in 72 hours. Contrary to the above, the Unit 1 1B-B CCP was inoperable from July 24, 2016, until August 5, 2016, in excess of the allowed outage time of Condition A without the unit being placed in Mode 3 in 6 hours and Mode 4 in 12 hours as required by Condition B. This issue was documented in the licensees corrective action program as CR 1199024. The finding was screened using IMC 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The finding required a detailed risk evaluation because a single train of CCP was inoperable for greater than its allowed outage time. The regional Senior Reactor Analyst reviewed the inspector provided detailed risk evaluation that was performed using the Saphire SDP module. The finding was determined to be Green.
05000259/FIN-2016003-07	2016Q3	Browns Ferry	Licensee-Identified Violation	Title 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.
05000296/FIN-2016003-06	2016Q3	Browns Ferry	Main Steam Relief Valves Inoperable Longer than Allowed Outage Time	A 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)
05000296/FIN-2016003-05	2016Q3	Browns Ferry	Alternate Depressurization Valve Inoperable Longer than the Allowed Outage Time	A 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)
05000259/FIN-2016003-04	2016Q3	Browns Ferry	Inadequate Prompt Determination of Operability for HPCI Steam Line Inboard Isolation Valve	An 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.
05000259/FIN-2016003-03	2016Q3	Browns Ferry	Failure to Maintain The High Pressure Fire Protection System Piping	A 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).
05000260/FIN-2016003-02	2016Q3	Browns Ferry	Failure to Implement Compensatory Roving Fire Watch	An 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.
05000260/FIN-2016003-01	2016Q3	Browns Ferry	Failure to Ensure Adequate Piping Clearances After MOV Modification	An 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.
05000424/FIN-2016003-02	2016Q3	Vogtle	Licensee-Identified Violation	Title 10 CFR Part 50.54(q)(2) required, in part, that a licensee shall follow and maintain the effectiveness of its emergency plan that meets the planning standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(4) required, in part, that a standard emergency classification and action level scheme is in use by the nuclear facility licensee. Contrary to these requirements, since 2008, emergency action level EAL HA1 #5 was not translated from the emergency plan to implementing procedure NMP-EP-110 GL03 (formally 91001-C) during the 2008 revision of the emergency plan. The licensee entered the issue into their corrective action program as CR 10251396. The inspectors determined that the finding was of very low safety significance (Green) because the finding constituted an ineffective EAL rather than a failed risk-significant planning standard.
05000424/FIN-2016003-01	2016Q3	Vogtle	Failure to Properly Implement Fire Door Inspections	An NRC-identified Green non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.d, Procedures, was identified for the licensees failure to correctly verify fire door gaps at the strike plate area and between meeting edges of double swinging metal doors were within acceptable limits. The licensee initiated hourly roving fire watches for these fire doors and took corrective maintenance action to restore affected fire doors within limits. The licensee documented this condition in condition reports 10254221 and 10252774. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Hazards (i.e. fire) and adversely affected the cornerstone objective in that door gaps outside the required limits compromised the doors fire rating qualification. The finding was determined to be of very low safety significance (i.e. Green) because either the combustible loading on both sides of each door was representative of a fire duration of less than 1.5 hours or each door maintained at least a 1-hour fire endurance rating. The finding had a cross-cutting aspect of Training in the Human Performance area because the licensee did not ensure there was adequate training to properly inspect station fire doors (H.9).
05000296/FIN-2016002-02	2016Q2	Browns Ferry	Failure to Declare Notification of Unusual Event	The 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).
05000260/FIN-2016002-03	2016Q2	Browns Ferry	Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety Function	An 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-2016002-01	2016Q2	Browns Ferry	Failure to Provide Adequate Maintenance Results in Loss of Core Flow While Shutdown	A 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).
05000424/FIN-2016502-01	2016Q2	Vogtle	Failure to Adequately Maintain Emergency Response Facilities	The 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 that adequate emergency facilities and equipment to support emergency response are provided and maintained as required by 10 CFR 50.47(b)(8). Specifically, the effectiveness of the emergency plan was reduced by a change to the Technical Support Center (TSC) functionality requirements in Technical Requirements Manual (TRM) TR 13.13.1, Emergency Response Facilities, Revision 1. The requirement to maintain climate control was removed without an adequate basis to support removal. The procedure change had been in place since September 2013, and until a corrected revision is issued, a Standing Order has been put in place. The licensee entered this finding into the corrective action program (CAP) as condition report (CR) 10221041. 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 would have affected the emergency response organizations ability to effectively perform their duties had an emergency been declared and TSC climate control non-functional. 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 or the overall function of the TSC. The finding was associated with a cross-cutting aspect in the Change Management component of the Human Performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000425/FIN-2016002-01	2016Q2	Vogtle	Failure to properly implement a maintenance procedure caused a Reactor Trip	A self-revealing non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to properly implement procedure 24750- 2, Steam Generator Level (Narrow Range) Protection Channel II 2L-519 Channel Operational Test and Channel Calibration. During testing of Unit 2 loop 1 steam generator (S/G) narrow range channel 2L-519 the channel was not removed from scan resulting in a reactor trip. The licensees immediate corrective actions were to remove the technicians performing the calibration from maintenance duties for formal remediation. The licensee documented this condition in CR 10230073. The performance deficiency (PD) was more than minor because it adversely affected the Initiating Events cornerstone objective in that the failure to properly remove channel 2L-519 from scan resulted in a reactor trip. The finding was determined to be Green because the PD did not result in a loss of mitigation equipment used to transition the reactor to a stable shutdown condition. The finding was assigned a cross cutting aspect of Avoid Complacency because maintenance technicians failed to implement appropriate error reduction tools to verify that the correct channel was removed from scan for testing.
05000260/FIN-2016001-01	2016Q1	Browns Ferry	Unacceptable Preconditioning of RCIC Valve Prior to ASME In-Service Testing	An NRC identified finding (FIN) for failure to meet TVA procedure NETP-116.3, Inservice Testing Program Preconditioning Guidelines, because unacceptable preconditioning of the Unit 2 Reactor Core Isolation Cooling (RCIC) steam supply valve occurred prior to quarterly In-Service Test (IST). Specifically, the preconditioning was unacceptable because the testing sequence was avoidable, it masked the actual asfound condition of the valve, and it could possibly result in an inability to verify the operability of the valve. As an immediate corrective action, the licensee performed an evaluation that determined the valve remained operable. The finding was entered into the licensee's corrective action program as CR 1159463 . 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 of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Additionally, if left uncorrected, the performance deficiency could lead to a more significant safety concern. Specifically, the licensees justification of this particular preconditioning event could be applied to justify additional, avoidable, preconditioning events and possibly result in an inability to verify the operability of components. 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 not a design or qualification deficiency, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its TS allowable outage time, did not result in a loss of function of non-TS equipment, and did not involve the loss of equipment or function specifically designed to mitigate an external event. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance area of Consistent Process (H.13), because individuals did not complete the required preconditioning evaluation forms described in licensee procedure NETP-116.3, which would have challenged the validity of the licensees original determination of acceptability.
05000259/FIN-2016001-02	2016Q1	Browns Ferry	Failure to adequately maintain emergency plan implementing procedures	The 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)
05000259/FIN-2016001-10	2016Q1	Browns Ferry	Licensee-Identified Violation	Licensee Event Report (LER) 05000259/260/296/2015-004-00 Containment Atmosphere Dilution B Train Supply System Inoperable Longer Than Allowed by Technical Specifications: Technical Specification LCO 3.6.3.1, Containment Atmosphere Dilution System, Condition B required that when Two CAD subsystems are inoperable that the licensee verify by administrative means that the hydrogen control function is maintained and to restore one CAD subsystem to OPERABLE status within 7 days. Condition C required action to place the affected unit in Mode 3 within 12 hours if the Condition B completion time was not met. Contrary to Technical Specification LCO 3.6.3.1 condition C, completion times were not met to place the units in Mode 3 within 12 hours when both trains of CAD were considered unavailable. This licensee identified violation is documented in the licensees CAP as CR 1087766. This finding was screened to Green using IMC 0609 Appendix H,Table 6.1 because the finding did not affect any of the listed Systems, Structures, or Components important to LERF.
05000296/FIN-2016001-09	2016Q1	Browns Ferry	Licensee-Identified Violation	Licensee 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.
05000296/FIN-2016001-08	2016Q1	Browns Ferry	Licensee-Identified Violation	Licensee Event Report (LER) 05000296/2014-003-00 Primary Containment Isolation Valve Inoperable for Longer Than Allowed by Technical Specifications 10 CFR 50, Appendix B, Criterion 5 required, in part, that activities affecting quality be implemented in accordance with documented procedures and drawings. Contrary to the above, between March 7, 2014 and June 6, 2014, relay 3-RLY-074-10A-K98A was wired incorrectly as discussed in LER 05000296/2014-003-00. The licensee corrected the wiring and entered the issue into the licensee's corrective action program as CR 892500. Inspectors screened the violation using IMC 0609, Appendix G, Attachment 1, Exhibit 3 Mitigating Systems Screening Questions, dated May 9, 2014. Because the finding degraded a functional auto-isolation of RHR on low reactor water level, a Phase 2 screening was required. Using attachment 3, Phase 2 Significance Determination Process Template for BWR During Shutdown, dated February 28, 2005, inspectors completed Worksheet 1 for Loss of Inventory in Plant Operating State 1 (Head On) and determined the risk was approximately 1e-7/yr, which was less than the 1e-6/yr threshold for a greater than Green finding. The dominant core damage sequence was the failure to isolate a reactor coolant leak and subsequent failure by operators to open vent paths (e.g. a safety relief valve) to control RCS pressure to enable continued low pressure injection. In the evaluation, no operator recovery credit was given for leak isolation, but credit was given for the redundant isolation valve that was operable which could have satisfied the automatic isolation function. The Regional Senior Reactor Analyst performed a detailed risk review of the finding. The risk review considered both the outage related risk, and the risk associated with a trip from power that would have the plant in shutdown cooling during the recovery. A screening analysis using bounding assumptions and the risk models ISLRHR event tree was performed. The dominant cutsets involved failure of the redundant valve to operate, and operator actions to recover. Because of the short exposure time during the shutdown periods, the redundant valve with the automatic action available, and the availability of operator recovery, the Finding was determined to be Green. This violation is being treated as an NCV consistent with Section 2.3.2 of the Enforcement Policy.

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