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| Start date | Site | Title | Description | |
|---|---|---|---|---|
| 05000346/FIN-2018002-03 | 30 June 2018 23:59:59 | Davis Besse | Failure to Perform a Procedure Affecting Quality | The NRC identified a finding of Green significance and an associated non-cited violation of 10 Code of Federal Regulation(CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, due to the licensees failure to implement DBOP03006, Miscellaneous Instrument Shift Checks, Specifically, the licensee declared SFAS Channel 1 operable without performing the required channel check. |
| 05000346/FIN-2018002-02 | 30 June 2018 23:59:59 | Davis Besse | Failure to Apply Technical Specification for Safety Features Actuation SystemInstrumentation | The NRC identified a finding of Green significance and an associated Non-Cited Violation of Technical Specification 3.3.5.b, Safety Features Actuation System (SFAS) Instrumentation, for the licensees failure to place the reactor in Mode 3 within six hours of identifying that two channels of Safety Features Actuation System Borated Water Storage Tank level instrumentation were inoperable. Specifically, the licensee inappropriately exited Technical Specification 3.3.5.b, and failed to place the reactor in Mode 3 while two Borated Water Storage Tank level instruments were inoperable for more than six hours. |
| 05000346/FIN-2018002-01 | 30 June 2018 23:59:59 | Davis Besse | Failure to Follow the Makeup and Purification Procedure | A self-revealed Green finding and associated Non-Cited Violation of Technical Specification 5.4.1.a, Procedures, was identified when the licensee failed to follow station procedure DBOP06006, Makeup and Purification System. Specifically, the licensee failed to open MU177, the Make-Up Filter 1 Outlet Isolation valve, which resulted in the isolation of letdown while swapping make-up filters. |
| 05000346/FIN-2018002-04 | 30 June 2018 23:59:59 | Davis Besse | Misapplication of the Operability Determination Process | The NRC identified a finding of Green significance due to the licensees misapplication of NOPOP1009, Operability Determinations and Functionality Assessments. Specifically, the licensee failed to apply the Operability Determination process in accordance with procedures. |
| 05000346/FIN-2017004-01 | 31 December 2017 23:59:59 | Davis Besse | Failure to Maintain Procedures Associated with Ventilation Air Monitoring Assessment Program | The inspectors identified a finding of very-low safety significance and an associated NCV of Technical Specification 5.4.1 for the failure to maintain procedures for station vent releases during planned scenarios. Specifically, the inspectors identified multiple procedures that were not updated when the station vent monitors were replaced in 2014. This issue has been entered into the licensees Corrective Action Program as CR201710817. Corrective actions taken included the issuance of a Standing Order for collecting samples during accident conditions, provided Just-In-Time training for chemistry technicians, and revision of the outdated procedures. The performance deficiency was determined to be more-than-minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening. Specifically, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern in that the failure to maintain procedures to collect station vent samples under all predicted conditions could result in the inability to measure the amount of gaseous radioactivity leaving the plant and to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The finding was assessed using Inspection Manual Chapter 0609 Appendix D, Public Radiation Safety Significance Determination Process, and was determined to be of very-low safety significance because the issue involved radioactive effluent releases, but did not: (1) represent a substantial failure to implement the Radioactive Effluent Release Program; or (2) result in public exposure that exceeded the dose values in Appendix I to 10 CFR, Part 50, and/or 10 CFR, Part 20.1301(e) limits. The inspectors determined that the finding had a cross-cutting component in the area of Human Performance, in the aspect of Work Management: specifically, the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. (H.5) |
| 05000346/FIN-2017004-02 | 31 December 2017 23:59:59 | Davis Besse | Interface Between New Accident Range Ventilation Monitors and the Emergency Preparedness Dose Assessment Program | During inspection activities associated with the accident range station vent monitor, the inspectors identified an unresolved item (URI) associated with the interface between the monitor and the Dose Assessment Program used to project dose to members of the public during potential accident conditions. Description: The licensee replaced the accident range station vent monitors in 2014 using ECP 040006, Replace Kaman Radiation Monitors. The replacement monitors were manufactured by a different company than the original monitors, had different detection capabilities, different system calibration, and different computer hardware to convert detector output into usable information. The licensee could not immediately provide specifics regarding the interface between the new accident range monitors and the program used during accident conditions for providing dose projections and the resulting protective action recommendations. The inspectors focus of concern was how the new accident range monitors accounted for the potentially rapidly changing mixture of radioactive gases during the early phase of a postulated accident. Consequently, this issue remains under review by the NRC awaiting for additional information from the licensee to verify the new monitor interface to determine if it represents a performance deficiency and is categorized as a URI. (URI 05000346/201700403, Interface Between New Accident Range Ventilation Monitors and the Emergency Preparedness Dose Assessment Program) |
| 05000346/FIN-2017004-03 | 31 December 2017 23:59:59 | Davis Besse | Failure to Prescribe Appropriate Work Instructions for an Activity Affecting Quality | A self-revealed finding with an Apparent Violation (AV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and an associated violation of technical specification (TS) 3.7.5, Emergency Feedwater (EFW), was identified on September 13, 2017, due to the licensees apparent failure to prescribe appropriate work instructions for an activity affecting quality of the safety-related auxiliary feedwater (AFW) system. Specifically, the licensee apparently did not provide appropriate instructions to maintain an adequate amount of oil in the AFW turbine bearing oil sumps, resulting in the failure of AFW 1 on September 13, 2017. The licensee entered this issue into the CAP as CR201709443 and CR201709857, immediately replaced the damaged bearing, and updated the lubrication manual data sheets to include sight glass marking dimensions per vendor guidance. The apparent performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and potentially adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that respond to initiating events. Specifically, the apparent performance deficiency resulted in the failure of the AFW 1. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609 Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the mitigating systems cornerstone. The inspectors determined the finding represented an apparent actual loss of function of at least a single train for greater than its technical specification allowed outage time. Therefore, a detailed risk evaluation will be performed by a regional senior reactor analyst. Because the safety characterization of this finding is not yet finalized, it is being documented with a significance of to be determined (TBD). The inspectors determined this finding affected the cross-cutting aspect of challenge the unknown in the area of Human Performance, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, licensee personnel apparently did not stop when faced with uncertain conditions in the preventive maintenance procedure for replacing the AFPT sight glasses. Although the replacement of the AFPT 1 inboard bearing sight glass occurred in 1997, the licensee had the opportunity to challenge the lack of detail in the work instructions in late 2014 when the AFPT 2 outboard bearing sight glass was replaced. (H.11) |
| 05000346/FIN-2017004-04 | 31 December 2017 23:59:59 | Davis Besse | Failure to Document a Degraded Condition on the AFPT 1 Outboard Bearing | The inspectors identified a finding of very low safety significance for the licensees failure to document a degraded condition of a safety-related system in the corrective action program (CAP), as required by licensee procedure, NOPLP2001. Specifically, during planned maintenance on auxiliary feedwater pump turbine (AFPT) 1, the licensee identified scoring on the outboard turbine bearing and failed to generate a condition report detailing the issue. The licensee entered this issue into the CAP as condition report (CR) 201712487 for evaluation. The inspectors determined the performance deficiency was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, the failure to document a degraded condition in the CAP did not allow the organization to properly assess the issue. Therefore, the underlying cause may not have been appropriately addressed. Using IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014, the inspectors determined the finding to be of very low safety significance (Green) because the inspectors answered no to all questions in Exhibit 3 of Appendix G, Attachment 1. The inspectors determined this finding affected the cross-cutting aspect of identification in the area of Problem Identification and Resolution, where the organization implements a corrective action program with a low threshold for identifying issues and individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to completely identify the degraded condition, resulting in the failure to document the issue. (P.2) |
| 05000346/FIN-2017003-05 | 30 September 2017 23:59:59 | Davis Besse | Ultrasonic Testing Records to Support Fuel Selection Were Not Being Maintained | A Severity Level IV NCV of 10 CFR 72.174, Quality Assurance Records, was identified by the NRC inspectors for the failure of the licensee as of June 22, 2017, to maintain sufficient records to furnish evidence of activities affecting quality. Specifically, the licensee failed to maintain ultrasonic testing (UT) records which were relied upon to demonstrate that the spent fuel selected for loading in calculation CNF062.02055, Revision 0, was correctly classified as intact. The licensee documented this issue in its CAP as CR201706976 and took timely corrective actions.The inspectors determined that the violation was of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues. Example 9a is applicable to this issue in that the licensee failed to maintain UT records for many fuel assemblies classified as intact for loading, and this failure to maintain records was not an isolated incident of one or two instances. The violation screened as a Severity Level IV NCV. Cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000346/FIN-2017003-04 | 30 September 2017 23:59:59 | Davis Besse | Failure to Perform 10 CFR 50.59 Evaluation | A finding of very low safety significance and an associated NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the NRC inspectors for the licensees failure to maintain a record of a change from a method described in the USAR to another method. Specifically, the licensee failed to perform a written evaluation for the change to USAR defined load factors based on the design basis American Concrete Institute (ACI) 31863 Code to less conservative load factors based on the ACI 31871 Code. The licensee entered this issue into its CAP as CR201703025. Planned corrective action includes updating the USAR to reflect the changes to the Design Criteria Manual (DCM) for the load factors incorporated in the 1971 ACI 318 Code. 3 The inspectors determined that the licensees failure to perform a written evaluation for this change was a performance deficiency. The finding was determined to be more than minor because the inspectors could not conclude that the implemented change would not result in a departure from a method of evaluation described in the USAR) used in establishing the design bases and therefore not require a license amendment. Because the inspectors could conclude that the concrete structures designed using ACI 31871 load combinations would still have sufficient structural capacity to perform their design basis safety functions during a seismic event, the finding was determined to have very low safety significance corresponding to a Severity Level IV violation per Example 6.1.d.2 of the NRC Enforcement Policy. The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current licensee performance. |
| 05000346/FIN-2017003-03 | 30 September 2017 23:59:59 | Davis Besse | Failure to Perform Adequate Evaluation of Cask CraneComponents and Crane Support Structure | A finding of very low safety significance and an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, was identified by the NRC inspectors for the failure of the licensees design control measures to provide for verifying or checking the adequacy of design of the Auxiliary Building spent fuel cask crane and crane support structure elements. Specifically, calculations involving the Auxiliary Building structure, crane runway rails, crane rail clips, and rail clip bolts had not been verified or checked to ensure the requirements of Updated Safety Analysis Report (USAR) Section 3.8.1.2 were included. The licensee documented these issues in its Corrective Action Program (CAP) as CR201705071, CR201707084, and CR201707114, and initiated actions to restore compliance.The performance deficiency was determined to be of more-than-minor significance because it was associated with the Barrier Integrity cornerstone attribute of Design Control and adversely affected the cornerstone objective of providing reasonable assurance that the physical design barriers, i.e. the Auxiliary Building, protect the public from radionuclide releases caused by accidents or events. The inspectors screened the finding through Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process for Findings At-Power, using Exhibit 3, Barrier Integrity Screening Questions. The finding screened as of very low safety significance because the finding only represented a potential degradation of the radiological barrier function provided by the Auxiliary Building. The inspectors identified a Human Performance, Design Margin (H.6) cross-cutting aspect associated with this finding. Specifically, the licensee failed to ensure the Auxiliary Building structure, cask crane runway rails, rail clips, and rail clip bolts reflected the intended design margins established based on the design and licensing basis. |
| 05000346/FIN-2017003-02 | 30 September 2017 23:59:59 | Davis Besse | Auxiliary Feedwater Pump 1 Bearing Failure | A URI was identified by the inspectors relating to the final determination of the cause of the AFP 1 turbine inboard bearing failure. On September 13, 2017, the licensee performed the scheduled quarterlysurveillance test on AFP 1. This test requires the pump to run loaded with full flow of water, whereas the monthly test runs the pump only lightly loaded with water being 10 pumped through a minimum recirculation line. Within three minutes after the full flow adjustments were completed, the AFP 1 turbine inboard bearing high temperature alarm (>220 oF) actuated. The licensee verified the alarm was valid and manually tripped the AFP 1 turbine approximately 30 minutes after the alarm was received. Oil samples indicated bearing damage. The licensee disassembled the AFP 1 turbine bearing and observed bearing failure.Initial evaluation of the bearing by the licensee revealed that the damage was due to insufficient lubrication caused by low oil level. The oil level at the time of failure was within the indicated acceptable band of the oil sight glass, however, indicated band was significantly larger than the vendor recommended 3/8 inch and not at the correct height.The oil level in the sump was too low to sufficiently wet the oil slinger ring. This condition was determined to have existed since the previous pump quarterly test on June 21, 2017. After that test, a technician removed an oil sample, but did not replenish the oil. The oil level indicated low to mid band, but within the (incorrectly marked) acceptable range on the sight glass at the time. The licensee entered this issue into their CAP as CRs 201709443, 201709817, 201709527, and 201709857. Because the licensee had yet to complete their investigation and analysis of the event by the end of this inspection period, the issue is being treated as a URI pending the inspectors review of the licensees completed root cause evaluation. (URI 05000346/201700302, Final Cause Determination of Auxiliary Feedwater Turbine Bearing Failure) |
| 05000346/FIN-2017003-01 | 30 September 2017 23:59:59 | Davis Besse | Pinched Wiring Causing the Failure of Fuses Y210 and Y214 | An unresolved item (URI) was identified by the inspectors relating to the significance of pinched wires and licensees understanding of the condition and the extent of cause and condition. On July 6, 2017, during a planned replacement of fuse Y204 in electrical cabinet Y2, unrelated fuse Y214 blew. Both fuses were scheduled for replacement as part of the licensees project to replace Shawmut A25X style fuses that are susceptible to premature failure. The failure of fuse Y214 was unexpected, and the licensee was not able to discern a direct cause. The licensee determined that the failure was the fuse itself being so unstable that any perturbation was enough to cause failure. This failure resulted in multiple systems being declared inoperable including AFP 2, safety features actuation system channel 2, decay heat removal system interlock, and radiation element RE8447. On August 8, 2017, the same electrical cabinet, Y2, was opened for replacement of fuse Y216. Following the replacement, fuses Y210 and Y214 blew. The licensee attempted replacement of the fuses, but the replacement fuses blew again, shortly after being repowered. Initial licensee evaluation of the condition revealed thatthe wire bundle running along the hinge side of the cabinet door was unconstrained and two of the wires had become pinched between the door and cabinet frame, which damaged the wire insulation and allowed the wires to short circuit against the cabinet frame. The failure of Y210 and Y214 resulted in multiple systems being declared inoperable including AFP 2, safety features actuation system channel 2, decay heat removal system interlock, and emergency diesel generator 2. The licensee removed and replaced the damaged portion of the wires and used wire ties to constrain the wire bundle. The licensee entered this issue into their CAP as CRs 201707196 and 201708185. Because the licensee had yet to answer NRC inspector questions pertaining to the corrective actions and extent of condition by the end of this inspection period, the issue is being treated as a URI pending completion of the inspectors review. (URI 05000346/201700301, Examination of Extent of Cause and Condition of Pinched Wires in Electrical Cabinets) |
| 05000346/FIN-2017008-01 | 30 June 2017 23:59:59 | Davis Besse | Failure to Maintain Adequate Room Temperature in the Emergency Feedwater Facility | Green . A finding of very low safety significance was identified by the inspectors for failing to maintain adequate room temperature in the emergency feedwater facility (EFWF) to support equipment operation. Specifically, the inspectors identified temperatures below freezing in multiple loca tions on emergency feedwater (E FW) system piping and in the E FWF basement. In response, the licensee installed heaters to raise room temperature. This finding is not a violation of NRC requirements. The inspectors determined that failing to maintain adequate room temperature in the EFWF to support equipment was cont rary to Nuclear Energy Institute ( NEI ) 12 06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Rev. 2 and was a performance deficiency. The finding is of more than minor significance because it was associated with the cornerstone attribute of protection against external factors and adversely affected the mitigating systems cornerstone objective . A detailed risk evaluation (DTE) determined the findin g was ( Green) . This finding was assigned a cross- cutting of Challenge the Unknown . (H.11) |
| 05000346/FIN-2017001-01 | 31 March 2017 23:59:59 | Davis Besse | Failure to Establish a Test Program that Demonstrates the Emergency Core Cooling System Room Coolers Will Perform Satisfactorily in Service | Green. The inspectors identified a finding of very low safety significance (Green) and an associated Cited Violation of Title 10 of the Code of Federal Regulations, (10 CFR) Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to establish a test program that demonstrates the emergency core cooling system (ECCS) room coolers will perform satisfactorily in service. Specifically, the associated inspection procedures did not include acceptance criteria, and the inspection results were not documented and evaluated to demonstrate the ECCS room coolers thermal performance was acceptable. The licensee captured this issue in their corrective action program (CAP) as condition report (CR) 201703328 to, in part, restore compliance and assess current and past operability. The performance deficiency was determined to be more than-minor because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to demonstrate the ECCS room coolers will perform satisfactorily in service does not ensure the coolers would remain available and capable of performing their mitigating function because it has the potential to allow an unacceptable condition to go undetected. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee re-evaluated the past operability impact of the 2016 tube blockage discoveries and determined that coolers were operable by crediting actual service water temperature and flowrate conditions. The inspectors determined that the associated finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. |
| 05000346/FIN-2016409-01 | 31 December 2016 23:59:59 | Davis Besse | Security | |
| 05000346/FIN-2016410-01 | 31 December 2016 23:59:59 | Davis Besse | Security | |
| 05000346/FIN-2016009-01 | 31 December 2016 23:59:59 | Davis Besse | Fire Hazards Analysis Report Incorrectly Described Rooms 511 and 512 as being Continuously Staffed | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited Violation of License Condition 2.C.(4) for the licensees failure to implement and maintain the Fire Protection Program as described in the Updated Final Safety Analysis Report. Specifically, the current Fire Hazards Analysis Report incorrectly listed rooms 511 and 512 as not requiring a separate fire watch, for fire protection impairments, because the rooms were incorrectly assumed to be continuously staffed or visible to the continuously staffed area. The licensee entered this issue into their Corrective Action Program and updated the Fire Hazards Analysis Report to reflect the current operating practice and deleted rooms 511 and 512 from the list of rooms that were continuously staffed. The inspectors determined that the performance deficiency was more-than-minor because if left uncorrected, it could become a more significant safety concern for the failure to maintain the defense-in-depth element for the Fire Protection Program. The lack of fire watches degraded the ability to recognize conditions which could either increase the likelihood of a fire or the severity of a fire. The finding was representative of a low degradation and screened as having very low safety significance (Green) in Task 1.3.1 of IMC 0609, Appendix F. The finding did not have a cross-cutting aspect associated with it because it was not reflective of current performance. |
| 05000346/FIN-2016004-01 | 31 December 2016 23:59:59 | Davis Besse | Mispositioned Instrument Air Valves Result in Plant Transient | A self-revealed finding of very low safety significance was identified for the licensees failure to appropriately follow station procedures for aligning instrument air valves that support main feedwater (MFW) regulating valve operation. Specifically, two instrument air valves were not aligned to their normal operating position following planned maintenance. As a result, the Steam Generator 2 (SG 12) MFW Regulating Valve momentarily closed during routine steam feedwater rupture control system (SFRCS) surveillance testing and caused a plant transient. Corrective actions taken by the licensee, include but are not limited to, performance of an instrument air valve line up to validate no other valves were out of position; performance of SFRCS Actuation Channel 2 testing to verify no other half trips existed on SFRCS Actuation Channel 2 components; a configuration control stand-down with the instrument and control shop; and revisions to procedural guidance to perform additional valve position verification. The finding was of more than minor significance because it was associated with cornerstone attribute of configuration control 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. The finding was determined to be of very low safety significance because the finding did not cause a reactor scram with the loss of mitigation equipment relied upon to transition the plant from the onset of the scram to a stable shutdown condition (e.g. loss of condenser, loss of feedwater). The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. The inspectors assigned the cross-cutting aspect of Avoid Complacency to the finding because the procedural step to close valve IA1008A was marked as complete but was not performed correctly. Additionally, appropriate human performance error reduction tools were not adequately used to ensure valve manipulations were performed as intended. (H.12) |
| 05000346/FIN-2016004-02 | 31 December 2016 23:59:59 | Davis Besse | Failure to Adequately Evaluate Degraded Turbine Building Roof Vents | A finding of very low safety significance was self-revealed on September 10, 2016, when rainwater intrusion into the automatic voltage regulator caused a generator lockout and reactor trip. Specifically, station management failed to adequately assess the identified degraded condition of the turbine building roof vents in accordance with station expectations and procedures when four roof vents were left stuck open although it was identified by operators that water intrusion was possible onto the stator water cooling skid and automatic voltage regulator on August 17th, 24 days prior to the event. No violation of regulatory requirements was identified because the turbine building roof vents and automatic voltage regulator are not safety related, and the applicable maintenance procedures were not covered under Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B. The finding was of more than minor safety significance because it affected the Equipment Reliability attribute of the Initiating Events cornerstone. Specifically, the failure to fully evaluate the risk associated with the stuck open turbine building roof vents affected the availability and reliability of the automatic voltage regulator causing a reactor trip. The inspectors also reviewed the examples of minor issues in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, dated August 11, 2009, and found no similar examples. The finding was determined to be a licensee performance deficiency of very low safety significance because the performance deficiency did not cause a reactor trip with the loss of mitigating equipment. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution and the cross-cutting aspect of evaluation. The licensee did not properly evaluate the problem and assigned an incorrect priority to the work order to address the degraded roof vents. (P.2) |
| 05000346/FIN-2016010-01 | 30 September 2016 23:59:59 | Davis Besse | 10 CFR 50.59 Evaluation Failed to Consider Change to Seismic Licensing Basis | A finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50.59(b)(1), Changes, Tests, and Experiments, (effective January 1, 1991) was identified by the inspector for the licensees failure to maintain records that included a written safety evaluation which provided the bases for determining that the change to seismic licensing basis damping in calculations to support removal of snubbers under modification 90-0079 did not involve an unreviewed safety question. Specifically, licensee safety evaluation SE91-0046 did not provide a suitable basis for concluding that there was no increase in the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the safety analysis report, in that it did not address how the basis for the NRCs-approval of the seismic design of the reactor coolant system continued to be met with respect to the steam generator slider support (Lubrite plate) damping. In particular, a May 31, 1983, NRC Safety Evaluation Report approved the licensees use of 0.15g safe shutdown earthquake ground acceleration in its seismic analysis for reactor coolant system design, in part, because there is sufficient conservatism and margin in the piping systems components and supports at Davis-Besse Unit 1 to ensure safe shutdown and continued shutdown heat removal in the event of a safe shutdown earthquake having a ground acceleration of 0.20g. The licensee subsequently adopted a significantly higher damping value for the steam generator slider support while maintaining a 0.15g acceleration for the design without addressing how sufficient conservatism and margin otherwise continued to be met. The licensee entered this issue into its corrective action program. The inspector determined that the licensees failure to provide in its 10 CFR 50.59 evaluation, SE91-0046, a suitable basis for the determination that the use of damping higher than established in the seismic licensing basis for the reactor coolant system, specifically the steam generator slider support, was not an unreviewed safety question was a performance deficiency. The issue of concern was determined to be more than minor because the performance deficiency impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (reactor coolant system) protect the public from radionuclide releases caused by accidents or events and the design control attribute to maintain functionality of the reactor coolant system. The inspector evaluated the underlying technical issue using IMC 0609, The Significance Determination Process for Findings at Power, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspector answered No to all the questions in Exhibit 1. In particular, because the reactor coolant system remained operable (capable of performing its safety function during a seismic event), the finding was determined to have very-low safety significance (Green) corresponding to a Severity Level IV violation per Example 6.1.d.2 of the NRC Enforcement Policy. The inspector did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance. |
| 05000346/FIN-2016003-01 | 30 September 2016 23:59:59 | Davis Besse | Inadequate Instructions to Correctly Assemble Electrical Conductor Seal Assemblies | A self-revealed finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified for the licensees failure to provide adequate instructions to correctly assemble electrical conductor seal assemblies (ECSAs) used to provide an environmental barrier for resistance temperature detectors (RTDs). Specifically, the midlock ferrules inside two ECSAs were installed backwards during the 18th refueling outage (RFO) in 2014 which rendered multiple post accident monitoring system (PAMS) indications required by Technical Specification (TS) 3.3.17 inoperable. This issue was entered into the licensees corrective action program (CAP). Corrective actions by the licensee included, but were not limited to, replacement of the two dual element RTDs impacted and their associated ECSAs during the 2016 RFO, performance of an extent of condition review, development of enhanced procedural guidance, and implementation of additional training on ECSA components. This finding was of more than minor significance because it was associated with the cornerstone attribute of equipment performance, and adversely affected the cornerstone objective: "To ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage)." The inspectors determined the finding to be of very low safety significance because it did not represent a deficiency affecting design or qualification of a mitigating system, structure, and component (SSC); it did not represent a loss of system and/or function; it did not represent an actual loss of function for at least a single train for more than its TS allowed outage time; and it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. Specifically, the cross-cutting aspect of Training was assigned to the finding because a job task analysis was performed prior to the 2014 RFO and determined that the procedural guidance to correctly assemble the ECSAs was adequate; thus no training or procedural changes were required. But the as-found condition of the RTDs during the 2016 RFO identified that a knowledge gap and procedure deficiency existed. (H.9) |
| 05000346/FIN-2016003-02 | 30 September 2016 23:59:59 | Davis Besse | Inadequate Modification Design Control Measures Result in Reactor Protection System Inoperability | A self-revealed finding of very low safety significance and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, were identified for the licensees failure to have adequately prepared and implemented a permanent plant modification associated with steam generator (SG) replacement during the units 18th RFO in 2014. Specifically, in conjunction with SG replacement the licensee had also replaced a significant amount of reactor coolant system (RCS) piping and instrumentation, including all RCS hot leg resistance temperature detectors (RTDs). The RTD housings were improperly insulated during the modification, such that over the ensuing reactor operating cycle the RTD wiring insulation degraded to the extent that nearly all the RTDs were rendered inoperable. This issue was entered into the licensees CAP. Corrective actions by the licensee included replacement of the degraded RTDs. This finding was of more than minor safety significance because it affected the attribute of design control of the Mitigating Systems cornerstone of reactor safety, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the units RPS. Specifically, the inspectors determined that the licensees failure to have properly designed and implemented the insulation packages for the RTD housings ultimately resulted in the overheating and degradation of the RTD wiring insulation and inoperability of the RTDs associated with the RCS high temperature and RCS pressure/temperature reactor trips. The finding was determined to be of very low safety significance based on a detailed risk analysis that yielded a change in core damage frequency (CDF) of less than 1E7 events per year. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. The inspectors assigned the cross-cutting aspect of Field Presence to the finding because the licensees SG replacement project management team failed to reinforce the importance of close communication between responsible engineers with overlapping and interfacing modification packages, and did not adequately promote effective work execution through the use of clearly defined work documents that were written and structured to minimize the likelihood for human error. (H.2) |
| 05000346/FIN-2016002-01 | 30 June 2016 23:59:59 | Davis Besse | Mispositioned Instrument Air Valves Result in Plant Transient | On May 31, 2016, at approximately 10:21 a.m., planned testing of Steam and Feedwater Rupture Control System (SFRCS) Actuation Channel No. 1 was in progress. This was the first performance of this test since the unit returned to operation following RFO 19. Unexpectedly, operators in the control room received several overhead annunciator alarms coincident with a rapid swing in plant power and indications that the SG 12 MFW Regulating Valve (SP6A) had gone closed and then reopened. In accordance with established procedures for responding to such an event, control room operators took manual control of integrated control system (ICS) stations for reactor demand, SG/reactor demand, both MFW regulating valves, both MFW startup valves, and both MFW loop demands. The control room crew was then able to arrest the transient and stabilize plant power at approximately 89 percent. Initial evaluation of the transient by the licensee revealed that two instrument air (IA) valves associated with control air for SP6A (IA1008D, SVSP6A1 Bypass; and IA1008A SVSP6A1 Maintenance Isolation) were out of their normal positions. The mispositioned valves had the effect of placing P6A in a half trip condition, such that when SFRCS Actuation Channel No. 1 was being tested SP6A unintentionally responded to the test signal. The licensee entered this issue into their CAP as CRs 201607282, 201607286, 201607337, and 201608363. Because the licensee had yet to complete their investigation and analysis of the event and the IA valve mispositioning by the end of this inspection period, the issue is being treated as an unresolved item (URI) pending the inspectors review of the licensees completed cause evaluation and proposed corrective actions. (URI 05000346/201600201) |
| 05000346/FIN-2016002-03 | 30 June 2016 23:59:59 | Davis Besse | Inadequate Evaluation of Trend Related to A25X Fuse Failures | A self-revealed finding of very low safety significance and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, were identified for the licensees failure to have adequately addressed an identified adverse trend involving 10 and 15 ampere Gould Shawmut A25X series fuses. Specifically, the licensee had identified adverse trends related to failures of A25X series fuses in 2005, and again in 2015, and had entered these adverse trends into their corrective action program (CAP) as Condition Reports (CRs) 200505314 and 201503516. However, the evaluation performed under CR 201503516 did not recognize that the fuse failures were occurring much more frequently than originally anticipated and that the previously created Preventative Maintenance (PMs) were not adequate to prevent failures. Additionally, the evaluation did not adequately incorporate industry experience that also identified a trend of failures with the A25X series fuses. Corrective actions by the licensee included replacement of the existing stock of uninstalled A25X series fuses with equivalent fuses of a different style and from a different manufacturer and identification of a plan to replace in-plant installed fuses. This finding was of more than minor safety significance because it affected the attribute of equipment performance of the Initiating Events cornerstone of reactor safety, and adversely impacted the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance because it did not represent a deficiency that caused a reactor trip as well as the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feedwater, etc.) The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution. The inspectors assigned the cross cutting aspect of Resolution to the finding because the licensee failed to take action to resolve the identified adverse trend associated with premature failures of A25X series fuses in a timely manner. (P.3) |
| 05000346/FIN-2016002-02 | 30 June 2016 23:59:59 | Davis Besse | Failure to Use the Corrective Action Program to Evaluate and Document Degraded Condition with Auxiliary Feedwater Train 2 | An NRC-identified finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified for the licensees failure to have entered a degraded condition associated with Auxiliary Feedwater (AFW) Train No. 2 into their CAP until challenged by the inspectors. Specifically, a flow transient that occurred on May 7, 2016, and that caused damage to components in the AFW recirculation line during AFW Train No. 2 testing was not entered into the licensees CAP until May 8, 2016, following challenges from the inspectors. This omission on the part of the licensees staff had the effect of bypassing certain features of the licensees CAP associated with evaluating and documenting the operability of safety-related equipment. The physical event and equipment issues were entered into the licensees CAP as CR 201606515 on May 8, 2016, following prompting by the inspectors. Corrective actions taken by the licensee included repairs to all damaged equipment, detailed inspections of AFW Train No. 2, and an engineering analysis into why the event occurred. The matter of the licensees failure to enter the event into their CAP in a timely manner was documented as CR 201606516, with corrective actions including the coaching and counseling of personnel involved regarding the proper use of the CAP. This finding was of more than minor safety significance because it affected the equipment performance attribute of the Mitigating Systems cornerstone of reactor safety and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the units AFW system. The finding was determined to be of very low safety significance because it did not represent a deficiency affecting the design or qualification of a mitigating system, structure, or component (SSC); it did not, in and of itself, represent a loss of system and/or function; it did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, or two separate safety systems being out-of-service for greater than their TS allowed outage times; and it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution. The inspectors assigned the cross-cutting aspect of Identification to the finding because the licensees staff failed to identify the issue with AFW Train No. 2 within their CAP completely, accurately, and in a timely manner in accordance with program requirements. (P.1) |
| 05000346/FIN-2016002-04 | 30 June 2016 23:59:59 | Davis Besse | Inadequate Post-Manufacture Quality Control Inspections Performed for ASME Section III, Class 2 Reactor Coolant Pump Seal Cavity Vent Line Flexible Hose | A self-revealed finding of very low safety significance and an associated NCV of Technical Specification (TS) 3.4.13, Reactor Coolant System (RCS) Operational Leakage, were identified for the failure of a licensee vendor to have ensured that a replacement reactor coolant pump (RCP) seal cavity vent line flexible hose was subjected to adequate quality control testing following manufacture. Specifically, a manufacturing weld defect on the flexible hose assembly for the RCP No. 11 first stage seal cavity vent line was not detected by post-manufacture testing, such that the hose developed a very minor leak during power operations for the reactor operating cycle occurring before the licensees spring 2016 refueling outage. Corrective actions by the licensee included replacement of the failed flexible hose assembly and revising the procurement requirements for subsequently ordered flexible hose assemblies to include enhanced helium tracer probe leak testing. This finding was of more than minor safety significance because it affected the attribute of equipment performance of the Initiating Events cornerstone of reactor safety, and adversely impacted the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance because it was determined that the finding could not have resulted in exceeding the RCS leak rate for a small loss of coolant accident or affected other systems used to mitigate a loss of coolant accident. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. The inspectors assigned the cross cutting aspect of Field Presence to the finding because the licensee failed to ensure that their vendor performed adequate quality control testing following manufacture of a safety-related flexible hose assembly. (H.2) |
| 05000346/FIN-2016404-01 | 30 June 2016 23:59:59 | Davis Besse | Security | |
| 05000346/FIN-2016001-01 | 31 March 2016 23:59:59 | Davis Besse | Operation of Safety Related Butterfly Valves in a Manner Beyond Design | A self-revealed finding of very low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified for the licensees failure to incorporate applicable manufacturers limits into the operating procedures and instructions for the service water (SW) outlet isolation/throttle valves for Component Cooling Water (CCW) Heat Exchanger (HX) Nos. 1, 2, and 3 (SW36, SW38, and SW37). Specifically, the licensees procedural guidance for the operation of these valves allowed them to be throttled beyond the manufacturers recommended limits, and repeated operation of the SW37 valve in this manner beyond its design contributed to its failure. This issue was entered into the licensees corrective action program (CAP). Corrective actions by the licensee included repair of the SW37 valve. This finding was of more than minor safety significance because it affected the attributes of design control and procedure quality of the Mitigating Systems cornerstone of reactor safety, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the units CCW system. Specifically, the inspectors determined that the licensees failure to have incorporated the applicable design limits for SW37 throttle position and differential pressure across the valve into applicable operating procedures contributed to the degradation and ultimate inoperability of the valve. The finding was determined to be of very low safety significance since the finding did not result in a loss of operability of any system or component. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. The inspectors assigned the cross cutting aspect of Design Margins to the finding because the licensee had failed to ensure that the safety related SW37 butterfly valve was operated and maintained well within the manufacturers design limits. |
| 05000346/FIN-2016001-03 | 31 March 2016 23:59:59 | Davis Besse | Shield Building Emergency Ventilation System Operability with Watertight Door No. 108 Inadvertently Left Open | The shield building EVS functions to collect and process potential leakage from the containment vessel to minimize environmental activity levels resulting from all sources of containment leakage following a design-basis accident. The EVS is required to maintain a negative pressure (a minimum of 14 inch water gauge), with respect to outside atmosphere, within the annular space between the shield building and the containment vessel and in the penetration rooms following an accident. In addition, it is required to provide a filtered exhaust path from the shield building annulus and the penetration and pump rooms following an accident. The EVS consists of two independent and redundant trains. Each train consists of a prefilter, a high efficiency particulate air filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system. The EVS boundary, consisting of various walls and doors within the plants auxiliary building, must be intact and functional to ensure EVS operability. Door No. 108, Emergency Core Cooling System Pump Room No. 115 to Detergent Waste Drain Tank to Clean Waste Receiver Tank, is one such plant door. At approximately 7:53 p.m. on March 21, 2016, with the unit in Mode 1 and operating at power, operations personnel discovered a plant watertight door, Door No. 108, open and unattended. The operations personnel immediately secured the door and informed operations on-watch management of the issue. The on-watch operations shift manager determined that because the door was fully functional and closed when he was informed of the issue that neither the door nor the shield building EVS was inoperable. He then contacted the licensees on-duty management team to discuss the issue. Collectively, the licensees personnel concurred with the shift managers operability decision and determined that the issue was not immediately reportable under 10 CFR 50.72(b)(3)(v) as an Event or Condition that Could Have Prevented Fulfillment of a Safety Function, since no SSCs had ever been declared inoperable. Subsequently, licensee engineering personnel reviewing the issue determined that based on exiting plant calculations and the area of the door that it was highly improbable that the EVS would be able to have met its specified safety function with Door No. 108 open and unattended. The licensee entered this issue into their CAP as CR 201603694. An investigation by the licensee into the issue identified that the door had been inadvertently left open by contractor workforce personnel approximately five minutes before it was discovered open by operations personnel. During the next few days while conducting their routine review of the licensees CAP entries, the inspectors took note of this issue and questioned the licensee regarding their decision not to report the matter under 10 CFR 50.72(b)(3)(v). Licensee management subsequently decided to perform a special test of the EVS with Door No. 108 in the open position (under the administrative control of a designated individual) to empirically determine the capability of the EVS in this condition. The test was performed during the afternoon/evening hours on March 25, 2016. Preliminary results indicated that the EVS passed, albeit by only 0.08 seconds. Because the licensee had not yet completed their analysis of the issue following the March 25, 2016, special EVS test at the end of the inspection period, the issue is being treated as a URI pending the inspectors receipt and review of the licensees completed CAP documents and evaluation. (URI 05000346/201600103) |
| 05000346/FIN-2016001-02 | 31 March 2016 23:59:59 | Davis Besse | Service Water Header Operability While Using a Degraded No. 3 CCW HX SW Outlet Isolation Valve (SW37) for SW Header Pressure Control | As discussed in Section 1R12.1 of this report, during the colder months of the year the demand on the SW system is reduced. During these winter months, the licensee operates the system in a mode specifically intended to reduce header pressure to avoid any challenges to the SW header relief valves, as the reduced SW flow requirements would otherwise tend to cause SW header pressure to rise. In this header pressure control mode, the SW side of a spare HX is placed in service to allow flow to pass without cooling any loads, and the increased SW system flow subsequently reduces SW header pressure back down to a more nominal value. Licensee operating crews frequently utilize the swing CCW HX No. 3 to perform this function, and its associated outlet valve (SW37) is throttled by procedure to accomplish this. Again, as discussed in Section 1R12.1 of this report, SW37 has experienced a number of leakage issues, at least in part, as a result of this practice. Most recently, excessive through leakage on SW37 was identified in March of 2015 (CR 201503283). Initially, the licensees evaluation of the condition only evaluated the impact of the through leakage on the valves isolation function. The evaluation concluded that the valve could be considered operable, but degraded, since an alternate means of isolation was available. The evaluation did not, however, assess the impact of the valve degradation on operation of the SW system if CCW HX No. 3 were to be placed in service or credited to be aligned to one of the CCW and SW headers in standby. As licensee engineering and technical personnel were preparing for an upcoming SW system flow test, their analyses of the condition began to suggest that small changes in the resistance of SW37 as a result of continued valve degradation could impact SW flow and possibly challenge minimum SW design basis flow assumptions for certain accident scenarios. As a result, in January of 2016 the licensee prohibited use of CCW HX No. 3 as an in-service or standby HX (CR 201600438). However, the licensees evaluation of the condition continued to permit CCW HX No. 3 and SW37 to be used for SW header pressure control. In reviewing the issue, the inspectors noted that the licensees evaluation, as documented in CR 201600438 and entered into their CAP, did not contain any technical justification for the continued use of SW37 in header pressure control mode. Field observations by the inspectors revealed that the licensee operations staff had attached a plant information tag to the SW37 valve hand wheel warning personnel of the degraded condition of the valve and the potential for rendering SW Header No. 1 inoperable if the valve position were to be altered. Given the unknown condition of the SW37 valve internals, the unknown extent of the degradation of the valves liner/seat, and the unknown nature of the mechanism causing the degradation, the inspectors questioned how it could be possible for the licensee to conclude that use of SW37 in header pressure control mode would be acceptable. On February 18, 2016, the inspectors raised their concerns on this matter to the licensees operations supervisory staff, and asked to be provided with the licensees technical basis for continued operability of the SW system with the degraded SW37 being utilized for header pressure control. After several days had passed without receiving an answer, the inspectors elevated the question to the Site Vice President on February 23, 2016. On February 24, 2016, licensee engineering and operations management informed the inspectors that CCW HX No. 3 and SW37 had been removed from SW header pressure control and would be precluded from further use in that manner pending additional licensee analysis. The licensee entered this issue into their CAP as CR 201602667. Because the inspectors had not yet received the results of the licensees additional analysis concerning the use of CCW HX No. 3 and SW37 for SW header pressure control at the end of this inspection period, the issue is being treated as an unresolved item (URI) pending the inspectors review of the licensees completed evaluation. (URI 05000346/201600102) |
| 05000346/FIN-2016001-05 | 31 March 2016 23:59:59 | Davis Besse | Lack of Software Change Controls and Inadequate Corrective Action for an Operator Workaround Contributes to Complications Experienced During a Reactor Trip | A self-revealed finding of very low safety significance (Green) was identified for the licensees failure to implement a technically correct software change associated with the SG / Reactor Demand ICS control station. Specifically, a known logic error within the plants ICS would cause the SG / Reactor Demand control station to trip to manual from automatic coincident with a reactor trip. The licensee had instituted compensatory operator actions for this condition, but removed these actions in December 2015 when they implemented a software change to rectify the problem. However, the corrective actions were inadequate and the SG / Reactor Demand ICS control station unexpectedly tripped to manual from automatic when the unit tripped on January 29, 2016. The unexpected control station mode of operation change, combined with the absence of any compensatory operator actions, contributed to the SG No. 1 high level condition and the resultant SFRCS actuation. This issue was entered into the licensees CAP. Corrective actions taken by the licensee included initiating work on a new software change to rectify the issue of the SG / Reactor Demand ICS control station tripping from automatic to manual coincident with a reactor trip; reestablishing the operator workaround and associated compensatory actions for control room operators; and revising applicable procedures to incorporate current industry standards for controlling software life cycle changes to certain categories of software that interface with plant systems. This finding was of more than minor safety significance because it affected the design control and procedure quality attributes of the Mitigating Systems cornerstone of reactor safety and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the units MFW system and main condenser for decay heat removal. The finding was determined to be of very low safety significance because it did not represent a deficiency affecting the design or qualification of a mitigating SSC; it did not, in and of itself, represent a loss of system and/or function; it did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, or two separate safety systems being out-of-service for greater than their TS allowed outage times; and it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution. The inspectors assigned the cross-cutting aspect of Evaluation to the finding because the licensee had failed to thoroughly evaluate the issue of the SG / Reactor Demand ICS control station unexpectedly tripping from automatic to manual to ensure that the software change intended to resolve the issue actually addressed its cause. |
| 05000346/FIN-2016001-04 | 31 March 2016 23:59:59 | Davis Besse | Less than Sufficient Work Package Documentation and Instructions Resulted in an Inadequate Part Being Installed into the Plants Integrated Control System | A self-revealed finding of very low safety significance (Green) was identified for the licensees failure to include an adequate bench check for a replacement integrated control system (ICS) module that was installed into the system during the plants 2014 refueling outage (RFO) into the work package instructions for that activity. Specifically, a defeat switch on the replacement Module 528 for the ICS rapid feedwater reduction (RFR) circuit installed as preventative maintenance during the plants 18th RFO was incorrectly wired and not detected during pre-installation checks. The incorrectly wired module prevented the ICS RFR function from occurring during the unit trip on January 29, 2016, which contributed to the Steam Generator (SG) No. 1 high level condition and the resultant steam and feedwater rupture control system (SFRCS) actuation. This issue was entered into the licensees CAP. Corrective actions taken by the licensee included replacement of ICS Module 528 with a spare properly configured for the RFR defeat switch function. Additionally, a proper data package to enable bench checking ICS Module 528 to verify the capability of the module to perform its intended function was created. The licensee also created training and lessons learned from this event. This finding was of more than minor safety significance because it affected the design control and procedure quality attributes of the Mitigating Systems cornerstone of reactor safety, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the units main feedwater (MFW) system and main condenser for decay heat removal. The finding was determined to be of very low safety significance because it did not represent a deficiency affecting the design or qualification of a mitigating system, structure, or component (SSC); it did not, in and of itself, represent a loss of system and/or function; it did not represent an actual loss of function of at least a single train for greater than its Technical Specification (TS) allowed outage time, or two separate safety systems being out-of-service for greater than their TS allowed outage times; and it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. The inspectors assigned the cross-cutting aspect of Documentation to the finding because the licensee had failed to ensure that the instructions and other work package guidance available to maintenance personnel performing the ICS Module 528 replacement had contained provisions for an adequate bench check of the module prior to its installation. |
| 05000346/FIN-2016001-06 | 31 March 2016 23:59:59 | Davis Besse | Less than Adequate Procedural Instructions for Restoring Main Feedwater Following a Reactor Trip | A self-revealed finding of very low safety significance (Green), and an associated NCV of TS 5.4.1(a) were identified for the licensees failure to establish and implement adequate procedural guidance for restoring MFW following a reactor trip. Specifically, the guidance in licensee procedure DBOP06910, Trip Recovery Procedure, for restoring MFW to the SGs using the motor-driven feedwater pump (MDFP) did not ensure that the MFW piping had been sufficiently re-pressurized prior to opening the MFW to SG isolation valves. This lack of satisfactory procedural guidance allowed control room operators to prematurely open the MFW to SG No. 1 isolation valve, which resulted in a SFRCS actuation on the reverse delta pressure (P) function. This issue was entered into the licensees CAP. Corrective actions planned by the licensee included changes to licensee procedure DBOP06910, Trip Recovery Procedure, to ensure that MFW header pressure is greater that SG pressure prior to opening the MFW to SG isolation valves. This finding was of more than minor safety significance because it affected the design control and procedure quality attributes of the Mitigating Systems cornerstone of reactor safety, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the units MFW system and main condenser for decay heat removal. The finding was determined to be of very low safety significance based on the results of a detailed risk evaluation conducted by the NRC Region III Senior Reactor Analyst (SRA). The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. The inspectors assigned the cross-cutting aspect of Resources to the finding because the licensee had failed to ensure that the procedural instructions and guidance available to plant operators restoring MFW during reactor trip recovery actions took into account all relevant technical details (e.g., the differences between MFW piping runs, the amount of time needed to re-pressurize MFW piping, etc.) |
| 05000346/FIN-2015004-01 | 31 December 2015 23:59:59 | Davis Besse | Failure to Use Worst Case 4160 Vac Bus Voltage In Design Calculations | The inspectors identified a finding of very low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to have adequate analysis related to the degraded voltage relay (DVR) setpoints as specified in Technical Specifications. Specifically, the licensees analysis failed to demonstrate that the DVR setpoints would provide adequate starting and running voltage to safety-related equipment during the most limiting case accident loading. This issue was entered into the licensees corrective action program (CAP). Corrective actions planned and completed by the licensee included analysis to determine the appropriate DVR setpoints and interim compensatory measures to maintain minimum voltage on 4160 volts alternating current (Vac) essential buses above 4070 Vac to ensure adequate voltage for safety-related components. This finding was of more than minor safety significance because it affected the Design Control attribute of the Mitigating Systems Cornerstone of reactor safety, and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the sites 4160 Vac safety-related electrical buses. Specifically, the licensee failed to perform and maintain an analysis demonstrating that all safety-related loads had adequate starting voltage at the DVR setpoint. The finding was determined to be of very low safety significance since the finding did not result in a loss of operability of any system or component. The inspectors determined that there was no cross-cutting aspect associated with this finding because the finding represented a legacy issue that was not representative of current licensee performance. |
| 05000346/FIN-2015408-01 | 31 December 2015 23:59:59 | Davis Besse | Security | |
| 05000346/FIN-2015407-01 | 30 September 2015 23:59:59 | Davis Besse | Security | |
| 05000346/FIN-2015003-01 | 30 September 2015 23:59:59 | Davis Besse | Flow Accelerated Corrosion Model Not Maintained In Accordance with Industry Standards and Guidance | A self-revealed finding of very low safety significance was identified for the licensees failure to maintain an adequate flow accelerated corrosion (FAC) program in accordance with station procedures and applicable industry guidance. Specifically, an incorrect restriction orifice size entered into the FAC program software in the late 1980s significantly underestimated the wear rate of a section of moisture separator reheater (MSR) piping that ultimately failed causing control room operators to conduct a rapid power reduction and manual reactor trip and declare an unusual event in accordance with the station's emergency plan. The failed section of piping had not been previously inspected in accordance with industry guidance and station procedures, and the incorrect FAC program software inputs had never been validated. This finding was associated with the Initiating Events Cornerstone of reactor safety and was of more than minor significance because it directly impacted 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. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. Using Exhibit 1, which contains the screening questions for the Initiating Events Cornerstone of Reactor Safety, the inspectors determined a detailed risk evaluation was required because the finding was a transient initiator that resulted in both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (i.e., the loss of the main condenser as a heat sink and the loss of main feedwater). The inspectors contacted the NRC Region III Senior Reactor Analyst (SRA) to perform a detailed risk evaluation. The assumed core damage sequence used by the SRA was that the MSR pipe break occurs, followed by either main steam isolation valve (MSIV) failing to close, followed by any of four in-series main turbine stop valves (SVs) and control valves (CVs) failing to close. Mathematically, the change in core damage frequency (CDF) was estimated at: CDF = 1 (event occurs) x (9.51E-4 + 9.51E-4) x 4 x 1.5E-3 x 1.5E-3 = 1.71E-8/yr The SRA concluded the risk associated with this performance deficiency was, therefore, of very low safety significance (Green). Because the causes for the finding stemmed from deficiencies going back several years or more, the inspectors concluded that the finding represented a latent issue not necessarily indicative of present licensee performance. As a result, no cross cutting aspect was assigned to this finding. |
| 05000346/FIN-2014008-01 | 30 June 2015 23:59:59 | Davis Besse | Departure from Method of Evaluation Required Prior NRC Approval Under 10 CFR 50.59 (c)(2) | The inspectors identified a Severity Level IV NCV of Title 10, Code of Federal Regulations (CFR) Part 50.59(c)(2), and an associated finding of very-low safety significance for the licensees failure to request and obtain a license amendment pursuant to 10 CFR 50.90. Specifically, the licensees method of evaluation that accepted shield building laminar cracking represented a departure from the method of evaluation described in the Final Safety Analysis Report (as updated), and required prior NRC approval with respect to the design and licensing basis. The licensee entered this finding into its Corrective Action Program; the licensees immediate corrective action determined that shield building remained operable and capable to perform its design safety functions; the licensees planned corrective actions included revising 10 CFR 50.59 Evaluation 13-00918, and preparation of additional documents for inclusion in a license amendment request. The finding was determined to be more than minor because the finding was associated with the Barrier Integrity cornerstone attribute of Design Control, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, The SDP for Findings At-Power. Using Exhibit 3, the inspectors determined that the finding screened as very-low safety significance because all the Reactor Containment screening questions for the Barrier Integrity Cornerstone were answered No. Specifically, the inspectors concluded that the shield building remained capable of performing its design safety functions despite the identified laminar cracking. The associated violation was categorized as Severity Level IV because the issue was determined to be of very-low safety significance under the SDP. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee did not take a conservative approach to decision making for evaluation of shield building laminar cracking, particularly when information is incomplete or conditions are unusual. |
| 05000346/FIN-2015008-04 | 31 March 2015 23:59:59 | Davis Besse | Failure to Verify Several CCW System Manual Valves Were in the Correct Position | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of Technical Specification (TS) Surveillance Requirement 3.7.7.1, for the licensees failure to verify several component cooling water (CCW) system manual valves in the flow path servicing safety-related equipment that were not locked, sealed, or otherwise secured, were in the correct position every 31 days. Specifically, the unsecured CCW pump seal water flush isolation valves (two valves per pump) for the two required operable CCW pumps were not verified open every 31 days. The licensee entered this finding into their CAP, verified the correct position of the valves, and planned to revise the Locked Valve Program to include the requirement to have the valves in the locked open position. The finding was determined to be more than minor because it was similar to IMC 0612, Appendix E, Example 3.c, because more than one valve was in the required position, bu not locked, sealed, or otherwise secured in the correct position, and because it was associated with the Mitigating Systems cornerstones attribute of Configuration Control, and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because the deficiency was confirmed not to result in a loss of safety function. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. |
| 05000346/FIN-2014007-02 | 31 March 2015 23:59:59 | Davis Besse | Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations | The inspectors identified an unresolved item (URI) for the licensees failure to perform an analysis demonstrating that at the degraded voltage relay (DVR) set point specified in Technical Specifications (TSs) adequate voltage would be available to safety-related equipment to start and run during a design basis accident. On September 5, 2014, the licensee was informed by Schulz Electric that High Pressure Injection (HPI) Pump 1-2 Motor will not meet all Purchase Order Requirements. Specifically, the supplied motor may not be capable of accelerating the driven load to normal operating speed within 6.5 seconds at a minimum starting voltage of 70 percent. The licensee issued a condition report and initiated Prompt Operability Determination (POD) 2014-13985, HPI Pump 1-2 Motor to evaluate the impact of the condition on the ability of the HPI Pump 1-2 Motor to perform its design function. The licensee concluded that HPI 1-2 Pump was operable but nonconforming and placed a note to operators that in certain electrical plant alignments the HPI 1-2 Pump should be declared inoperable. On October 8, 2014, the licensee initiated CR- 2014-15452, Inconsistency in the Treatment of the Plant Restrictions in Operability Determinations. The corrective actions documented the review of POD 2014-13985 to determine if the recommended actions/instructions to operators should have been considered compensatory actions and therefore screened in the licensees 50.59 process. The licensee concluded that the instructions to operators only provided information and therefore not considered a compensatory action. During the review of POD 2014-13985, the inspectors were concerned that the licensee did not adequately address the operability of HPI Pump 1-2 Motor with respect to the DVR set point TS value of 3712 VAC. The inspectors reviewed NRC Regulatory Issue Summary (RIS) 2011-12, Revision 1, Adequacy of Station Electrical Distribution System Voltages; dated December 29, 2011, which was issued to clarify the NRC staffs technical position on existing regulatory requirements. Specifically, the RIS clarifies voltage studies necessary for DVR setting bases. The RIS states, in part, the limiting voltage at the bus monitored by the DVR can be calculated in terms of the voltage at the terminals of the most limiting safety-related component in the plant in all required operating conditions (such as starting and running). In addition, the RIS discusses that calculations of voltages at the terminals of all safety-related equipment with the voltage at the DVR monitored bus at the DVR dropout setting would ensure adequate voltage (starting and running) to all safety-related equipment. The inspectors requested the licensees DVR set point analysis as described in RIS 2011-12. The licensee had an analysis that shows all of the safety-related loads would be able to run at steady-state at the TS DVR set point during design basis accidents. However, they do not have an analysis which shows all safety-related loads would start and run at the TS DVR set point. The licensees position is that if a design basis accident were to occur and the 4160 VAC Bus (Bus C1/D1) was at the Degraded Voltage set point of 3712 VAC, because the time delay is set at 7 seconds, they would divorce from the offsite power source and power would be supplied by the Emergency Diesel Generators (EDGs). Therefore, the licensee believes that they are not required to have an analysis which demonstrates that all required safety-related loads can start and run at the DVR set point, as described in RIS 2011-12, Revision 1. Based on this information, the inspectors were concerned that the licensee does not have an analysis which demonstrates the safety-related loads could start and run at the TS DVR set point. The licensee captured the inspectors concerns in their CAP as CR 2014-17296, 2014 50.59 Inspection: Davis-Besse does not have an analysis to satisfy item 1 of RIS 2011-12, dated November 19, 2014. During the evaluation of CR 2014-17296 the licensee stated, An analysis has been performed and demonstrates that all loads receive adequate voltage to start and perform their intended function with the exception of some Motor Operated Valves (MOVs). In order to ensure the MOVs have adequate voltage to perform their function, a minimum voltage of 4070 VAC should be maintained on either 4160 VAC Bus (C1/D1). Since the minimum voltage of 4070 VAC is higher than the previous voltage of 3900 VAC, the licensee has added a compensatory action to monitor Bus C1/ D1 at the higher voltage. This issue is unresolved pending consultation with Nuclear Reactor Regulation to determine if the licensee is required to demonstrate that safety-related loads can start and run at the DVR TS set point. |
| 05000346/FIN-2015008-03 | 31 March 2015 23:59:59 | Davis Besse | Failure to Incorporate the Design Analysis Required Acceptance Limit into Surveillance Procedure | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensee failing to incorporate the design requirements and acceptance limits into test procedures. Specifically, the design limit for the minimum voltage on essential safety-related inverter YV1 bus during Modes 5 and 6 was not correctly incorporated into surveillance test procedures. The licensee entered this finding into their CAP and re-analyzed for the 116 volts as-found value at panel Y1, and determined the loads that did not have adequate rated voltage at the surveillance procedure minimum acceptance criteria voltage, 114 volts, would have had sufficient voltage to operate at the as-found measured voltage to perform their intended safety function. The finding was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because the deficiency was confirmed not to result in a loss of safety function. The inspectors did not identify cross-cutting aspect associated with this finding because the finding was not representative of current performance. |
| 05000346/FIN-2015008-05 | 31 March 2015 23:59:59 | Davis Besse | Failure to Comply with Technical Specifications for the Borated Water Storage Tank | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of TS 3.5.4, Borated Water Storage Tank (BWST), for the failure to comply with the limiting condition for operation (LCO) while the BWST was aligned to the non-seismic spent fuel pool purification system, causing the BWST to be inoperable based on no longer meeting the tanks seismic requirement. The licensee entered this finding into their CAP, and initiated an LCO Tracking Log entry to not place the BWST on spent fuel pool purification in Modes 1 through 4. The finding was determined to be more than minor because it was associated with the Mitigating Systems attribute of Protection Against External Factors, and affected th cornerstones objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. The finding resulted in the loss of system safety function (i.e., the BWST). The finding screened as of very-low safety significance (Green) based on a SDP Phase II analysis that determined the ?CDF for the finding was 1.17E-7/yr. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution because recent operating experience was not adequately evaluated to prevent the identified finding. |
| 05000346/FIN-2015008-02 | 31 March 2015 23:59:59 | Davis Besse | Failure to Comply with IEEE 308-1971 for the Required Independence of Safety-Related Essential Inverter Distribution Systems | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to demonstrate compliance with Institute of Electrical and Electronics Engineers (IEEE) 308-1971, IEEE Standard Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations, for the required independence of essential safety-related inverter distribution system channels. Specifically, a common mode failure due to inadequate fault protection on several outside distribution panels could cause the loss o redundant safety-related inverters. The licensee entered this finding into their CAP, and initiated a Standing Order that would identify the affected circuit breakers and require they be opened based on a tornado warning that could potentially affect the site. The finding was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) based on a SDP Phase II analysis that determined a conservative delta core damage frequency (?CDF) of 1.29E-7/yr. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. |
| 05000346/FIN-2015008-01 | 31 March 2015 23:59:59 | Davis Besse | Vulnerability of EDG Crosstie to a Non-Essential Bus | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to assure that applicable regulatory requirements, and the design basis were correctly translated into specifications, drawings, procedures, and instructions, and verifying the adequacy of design. Specifically, the licensee failed to verify the adequacy of procedures controlling alignment of non-essential busses to the emergency diesel generators during a design basis event. The procedures guidance could put the plant in an unanalyzed alignment with the potential to result in the failure of safety-related equipment. The licensee entered this finding into their Corrective Action Program (CAP), and initiated a Standing Order to preclude the unanalyzed alignment. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low-safety significance (Green) because the inspectors were able t answer Yes to screening question A1 because the finding represents a design deficiency confirmed not to result in loss of operability. The inspectors did not identify a cross-cutting aspect associated with this finding as it did not reflect current performance. |
| 05000346/FIN-2014007-01 | 31 March 2015 23:59:59 | Davis Besse | Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis | The inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control for the licensees failure to install and control the Reactor Coolant Pump (RCP) seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress Analysis: Reactor Coolant Pump 1-1-1 Seal Cavity Vent, into specifications, drawings, procedures, an instructions. The licensee entered this finding into their Corrective Action Program (CAP) to review the lack of controls over the installation of the flexible hoses, but determined that the flexible hoses remained operable. The performance deficiency was determined to be more than minor because, if lef uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits. The finding screened as of very-low safety significance (Green) because the finding could not result in exceeding the Reactor Coolant System (RCS) leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation. The inspectors determined this finding had an associated cross-cutting aspect, Design Margins, in the Human Performance cross-cutting area. Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. (H.6) |
| 05000346/FIN-2014005-01 | 31 December 2014 23:59:59 | Davis Besse | Additional Review of Medical Records Needed | The inspectors identified an URI concerning the auditable condition of the medical records and completeness of information that may affect the conditions required as part of a license operators license. The inspectors reviewed a sample of the licensed operator medical records during the inspection. The inspectors noted that the medical records were difficult to review because of the lack of a succinct filing within the records themselves. The inspectors found that two of the license operator medical files contained a change in information that was relevant to the license condition. One of the changes was not known by the licensee or the NRC during a period of time that the licensed operator was standing watch in a licensed position. The involved operators licenses are currently on administrative hold and the operators are not allowed to stan watch in license positions. As part of the biennial review inspection conducted during the week of December 8 through December 12, 2014, the inspectors reviewed a sample of the licensed operato medical records for compliance with 10 CFR Part 55. The inspectors reviewed seven medical records for accuracy and compliance with the license conditions. During the review, the inspectors noted that one licensed operator had failed to meet the conditions required on the license, had delayed in notifying the facility licensee, and had continued to stand watch. The inspectors also noted that another operator had failed to meet a condition of his license; but the license had previously been placed on administrative hold, and the operator had not stood watch during the failure to meet the license condition. The inspectors expanded the inspection sample and selected five additional records for review. None of the additional records were identified as containing a potential noncompliance. In total, 12 of the 58 operator license medical records at the site wer reviewed. The involved operators licenses are currently on administrative hold, and the operators are not allowed to stand watch in license positions. In reviewing the requirements of 10 CFR Part 55, the inspectors focused on the following: A person must be authorized by a license issued by the Commission to perform the function of an operator or a senior operator as defined in this part (10 CFR 55.3) The licensee shall comply with any other conditions that the Commission ma impose to protect health or to minimize danger to life or property. (10 CFR 55.53(l)) The inspectors determined that 2 of the 12 records reviewed (approximately 17 percent) had discrepancies which involved potential violations of 10 CFR 55. The records reviewed were found to have issues involving the auditable condition of the information contained in the records, and further inspection of the medical records should be performed to determine if additional discrepancies exist. (URI 05000346/201400501) |
| 05000346/FIN-2014005-02 | 31 December 2014 23:59:59 | Davis Besse | Inadequate Procedural Guidance During Restoration From Valve Maintenance Results in Feedwater Heater System and Plant Power Transient | A self-revealed finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1(a) were identified when the licensee failed to provide proper procedural guidance for the restoration from valve maintenance on HD291G, a manual isolation valve for the level controller for HD291A, the emergency drain valve for High Pressure (HP) Feedwater Heater No. 14, on November 13, 2014. Specifically, the licensee's restoration instructions did not isolate HD291A prior to restoring its associated level controller. As a result, when a perturbation in the level controller during restoration caused HD291A to rapidly reposition to the fully open position, the resulting HP Feedwater Train 1 transient caused HP Feedwater Heaters 14, 15, and 16 to trip The change in plant efficiency that resulted momentarily drove plant power slightly above 100 percent This finding was associated with the Initiating Events Cornerstone of reactor safety and was of more than minor significance because it directly impacted 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. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. Using Exhibit 1, the inspectors determined that the finding screened as very low safety significance because all screening questions for the Initiating Events Cornerstone of reactor safety were answered No. This finding als was determined to have a cross-cutting component in the area of human performance, work management aspect, because during the work planning process for this maintenance activity the licensee failed to identify the risk associated with not isolating the HP Feedwater Heater No. 14 Emergency Drain Valve, HD291A, prior to restoring its associated level controller to service. |
| 05000346/FIN-2014201-01 | 30 September 2014 23:59:59 | Davis Besse | Security | |
| 05000346/FIN-2014004-01 | 30 September 2014 23:59:59 | Davis Besse | Failure to Properly Perform Required Fire Watch | An NRC-identified finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.4.1(d) were identified when the licensee failed to properly implement station procedures for fire protection impairments and fire watches. Specifically, a required compensatory fire watch on numerous occasions did not enter a room for which fire impairments had existed because of a door problem. Upon identification the licensee entered the issue in the corrective action program and implemented corrective actions including modification of fire protection software to track administrative impairments and placing a camera in the room until the door was repaired. This finding was determined to be of more than minor safety significance because it was associated with the Initiating Events cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during plant operations. Specifically, required fire watches established as compensatory measures should have been maintained for the duration of the impairments so that the sites ability to promptly detect and suppress a fire would be maintained. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1Initial Screening and Characterization of Findings. Because the finding involved fire protection, the inspectors transitioned to IMC 0609, Appendix F, Fire Protection Significant Determination Process. The finding was characterized according to IMC 0609, SDP, Appendix F, Attachment 1, "Fire Protection SDP Phase 1 Worksheet," dated September 20, 2013. This issue screened as low safety significance per Attachment 1, Question 1.3.1.A, because it did not affect the ability of the reactor to reach and maintain safe shutdown. This finding had a cross-cutting aspect in the area of human performance associated with conservative bias such that individuals use decision making practices that emphasize prudent choices over those that are simply allowable. In particular, the shift manager made an inaccurate assessment of existing fire impairments by only checking the fire protection software and not the fire watch log, which was readily available. The shift manager also made the decision to not document the approval for modifying how the compensatory fire watch was being performed such that on-coming personnel would be aware of the change. |