Source: https://kanterella.com/wiki/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BSite::Davis-20Besse-5D-5D/-3FStart-20date/-3FSite/-3FIdentified-20by/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3DStart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BDavis-20Besse-20findings-5D
Timestamp: 2020-07-13 10:41:17
Document Index: 294710017

Matched Legal Cases: ['art 50', 'art 50', 'art 20', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art-20', 'art-20']

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05000346/FIN-2018002-03 30 June 2018 23:59:59 Davis Besse NRC identified 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-2018002-04 30 June 2018 23:59:59 Davis Besse NRC identified 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-2018002-01 30 June 2018 23:59:59 Davis Besse Self-revealing 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-2018002-02 30 June 2018 23:59:59 Davis Besse NRC identified 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-2017004-02 31 December 2017 23:59:59 Davis Besse NRC identified 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-2017004-01 31 December 2017 23:59:59 Davis Besse NRC identified 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-2017004-04 31 December 2017 23:59:59 Davis Besse NRC identified 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-2017004-03 31 December 2017 23:59:59 Davis Besse Self-revealing 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-2017003-03 30 September 2017 23:59:59 Davis Besse NRC identified 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-2017003-04 30 September 2017 23:59:59 Davis Besse NRC identified 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-2017003-05 30 September 2017 23:59:59 Davis Besse NRC identified 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-2017003-01 30 September 2017 23:59:59 Davis Besse NRC identified 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-2017003-02 30 September 2017 23:59:59 Davis Besse NRC identified 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-2017002-01 30 June 2017 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation On February 6, 2017, the licensee identified during an engineering review that a vendor recommendation for containment air cooler (CAC) motors was not incorporated into plant procedures. The CAC fan motor vendor manual (M40000002) states that the motors were designed and manufactured to meet the requirements of National Electrical Manufacturers Association (NEMA) standard MG1 for motors and generators which recommends no more than two cold starts and one hot start per hour. The CAC monthly surveillance test procedures (DBSP03294 (CAC 1 Monthly Test), DBSP03295 (CAC 2 Monthly Test), and DBSP03296 (CAC 3 Monthly Test)) did not specify 31 limitations on the number of allowable hot and cold starts per hour. As a result, the motors were routinely operated with more than one hot start per hour, and the inspectors concluded it contributed to the failure of the CAC 1 fan motor in May 2014 as discussed in section 4OA2.3. 10 CFR Part 50, Appendix B, Criterion V Instructions, Procedures, and Drawings states: Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to these requirements, the licensee failed to incorporate appropriate vendor recommendations on the number of hot and cold starts allowed per hour for the CAC fan motors into the CAC monthly surveillance procedures and was at least a contributor to the failure of CAC 1 in May 2014. The licensee had operated these motors in this manner for several years prior to the failure of CAC 1 motor. The objective of the Mitigating System Cornerstone of Reactor Safety is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A key attribute of this objective involves maintaining procedure quality of maintenance and testing procedures. In accordance with NRC Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. Specifically, the failure to have incorporated into station procedures the limit and precaution that CAC motors should be limited to two cold starts and one hot start per hour resulted in routinely cycling the containment air coolers with more than one hot start per hour, and ultimately was a contributor to the failure of CAC 1 motor in May 2014. Using NRC IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating System Screening Questions, the inspectors determined that the violation was of very low safety significance (Green), since the inspectors answered no to all of the screening questions. The licensee had entered this issue into their CAP as CR 201701306. Licensee corrective actions included, but were not limited to, updating the CAC monthly surveillance procedures to add a new limit and precaution on allowable CAC motor starts per hour.
05000346/FIN-2017008-01 30 June 2017 23:59:59 Davis Besse NRC identified 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 NRC identified 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-2016004-01 31 December 2016 23:59:59 Davis Besse Self-revealing 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 Self-revealing 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-2016409-01 31 December 2016 23:59:59 Davis Besse NRC identified Security
05000346/FIN-2016410-01 31 December 2016 23:59:59 Davis Besse NRC identified Security
05000346/FIN-2016009-01 31 December 2016 23:59:59 Davis Besse NRC identified 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-2016403-01 31 December 2016 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation
05000346/FIN-2016003-02 30 September 2016 23:59:59 Davis Besse Self-revealing 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-2016003-03 30 September 2016 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation
Plant TS 3.3.16, Anticipatory Reactor Trip System (ARTS) Instrumentation, requires that three ARTS channels for the main turbine trip function be maintained operable with the unit operating in Mode 1 above 45 percent power, and three ARTS channels for the SFRCS / main feed pump trip function be maintained operable with the unit operating in Mode 1 at any power. While this TS provides actions and allowed outage time for a single inoperable ARTS channel, there are no provisions for more than a single ARTS channel being simultaneously inoperable. The provisions of TS Limiting Condition for Operation 3.0.3, therefore, apply when more than one ARTS channel is inoperable at the same time, and require that actions be initiated within 1 hour from the onset of the condition to
Be in Mode 3 within 7 hours
Be in Mode 4 within 13 hours; an
Be in Mode 5 within 37 hours
As discussed in Section 4OA3.5 of this report, contrary to the requirements of TS 3.3.16, all four ARTS channels were bypassed and inoperable for both the main turbine and SFRCS functions for a period of approximately 15 hours on May 910, 2016. The objective of the Mitigating Systems Cornerstone of Reactor Safety is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Key attribute associated with this objective are human performance and configuration control. In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. Specifically, plant operators in failing to adequately implement applicable operating procedures allowed the unit to enter into a mode of operation with less that the required three channels of ARTS operable and available. Using Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that a detailed risk analysis by the NRC Region III SRA was required since the issue involved the inoperability of more than one channel of ARTS, a condition for which there is no allowed outage time specified in TS 3.3.16. The SRA used the Davis-Besse SPAR Model, Version 8.19, and SAPHIRE, Version 8.1.4, for the calculation of the change in CDF for the issue. The following assumptions were made in the analysis:
The exposure time for the issue was conservatively assumed to be 15 hours
from 3:24 p.m. on May 9, 2016, when the unit entered Mode 1 and the TS 3.3.1 for the ARTS became applicable to 5:52 a.m. on May 10, 2016, when the ART bypass switches were returned to the normal/enabled state; an
With the ARTS SFRCS function bypassed, the SFRCS input to the ARTS t
provide a reactor trip signal was bypassed. Since the ARTS is not modeled i the SPAR model, it was very conservatively assumed that the RPS automati trips were bypassed during the 15hour exposure time, and only a manua reactor trip was available The result was a change in CDF of 7.6E7 events per year. The dominant core damage sequence was a transient initiating event with a failure of plant operators to manually trip the reactor, along with a failure of plant operators to initiate emergency RCS boration. Based on the detailed risk evaluation, the inspectors determined that the violation was of very low safety-significance (Green). As discussed in Section 4OA3.5 of this report, the licensee had entered this issue into their CAP as CR 201606563. In addition to the commissioning of a formal root cause evaluation, licensee corrective actions included the issuance of an operations standing order to require periodic walk downs of all control room panels by on-watch control room operators in pairs to ensure a comprehensive understanding of plant status awarenes and enhancements to applicable operating procedures.
05000346/FIN-2016010-01 30 September 2016 23:59:59 Davis Besse NRC identified 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 Self-revealing 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-2016002-03 30 June 2016 23:59:59 Davis Besse Self-revealing 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-2016002-05 30 June 2016 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation
Failure to Properly Assess and Implement Compensatory Measures for Fire Protection System Impairment Plant TS 5.4.1(d), requires, in part, the licensee to establish, implement, and maintain applicable written procedures covering fire protection program implementation. The fire protection program was implemented, in part, by Davis-Besse Procedure DBFP00009, Fire Protection Impairment and Fire Watch, Revision 21. Procedure DBFP00009, Step 6.1.2, states: Upon request, notification, or plant condition that indicates a fire protection system/component impairment exists or will exist, the Shift Manager shall ensure an Impairment Initiation Work Sheet Sections 1, 2A, 2B as needed, 2C and 3 is completed for each impairment to assess the need for compensatory measures. Contrary to this requirement, from 7:30 p.m. on June 26, 2016, to 2:30 p.m. on June 28, 2016, the licensee failed to implement and adequately assess the need for compensatory measures in multiple locations of the turbine building when an applicable fire impairment existed based on plant conditions. On June 26th, a section of underground fire protection piping was isolated for planned maintenance to repair FP355, South Underground Loop Sectionalizing Valve, due to the valve operator not functioning properly. A fire impairment initiation worksheet was completed for the pre-planned maintenance isolation boundary using the stations fire risk software program prior to the maintenance activity commencing. The assessment concluded that compensatory measures in the form of eight-hour roving fire watches were required for the SBODG building due to the sprinkler systems in those locations being isolated to support the FP355 maintenance activity. These compensatory measures were implemented at approximately 7:30 p.m. On June 28, while preparing for future fire protection maintenance on FP40, East Underground Loop Sectionalizing Valve, the licensee recognized that FP40 had a pre-existing maintenance condition since November 2015, such that the valve was stuck in the closed position (normally open). This condition, when combined with the already in-progress FP355 valve maintenance activity, was not previously assessed and extended the pre-planned FP355 isolation boundary. As a result, the sprinkler systems in the following locations in the turbine building were isolated with no compensatory measures established
Lube Oil Storage Room (Room 249  onehour fire watch required)
Oil Drum Storage Room (Room 337  onehour fire watch required)
Turbine Generator Lube Oil Room (Room 432  eight-hour fire watch required)
Janitor Closet (Room 346  eight-hour fire watch required)
Lube Oil Filter Room (Room 347  eight-hour fire watch required); an
Main Tool Room (Room 341  eight-hour fire watch required)
Upon identification and reassessment, the appropriate compensatory measures were immediately implemented around 2:30 p.m. on June 28. The inspectors reviewed this violation using the guidance contained in Appendix B, Issue Screening, of IMC 0612, Power Reactor Inspection Reports. The inspectors determined that the licensees failure to properly implement plant procedures for assessing and establishing compensatory fire watches was a performance deficiency that was reasonably within the licensees ability to foresee and correct and should have been prevented. This violation was associated with the Initiating Events cornerstone of reactor safety and was of more than minor 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, the licensee failed to implement and adequately assess the need for compensatory measures in multiple locations of the turbine building when an applicable fire impairment existed based on plant conditions. Required fire watch patrols established as compensatory measures should have been performed for the duration of the impairment so that the sites ability to promptly detect and suppress a fire would be maintained. The inspectors evaluated the violation using IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. Because it involved fire protection, the inspectors transitioned to IMC 0609, Appendix F, Fire Protection Significant Determination Process. The violation was characterized according to IMC 0609, SDP, Appendix F, Attachment 1, "Fire Protection SDP Phase 1 Worksheet," dated September 20, 2013. The violation screened as of very low safety significance (Green), per Attachment 1, Question 1.3.1.A, because it did not affect the ability of the reactor to reach and maintain safe shutdown. The licensee had entered this issue into their CAP as CR 201608266. Corrective actions include but are not limited to immediately establishing required compensatory measures upon identification of the issue and the performance of an apparent cause evaluation.
05000346/FIN-2016404-02 30 June 2016 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation
05000346/FIN-2016404-01 30 June 2016 23:59:59 Davis Besse Self-revealing Security
05000346/FIN-2016404-03 30 June 2016 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation
05000346/FIN-2016002-01 30 June 2016 23:59:59 Davis Besse NRC identified 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-2016002-04 30 June 2016 23:59:59 Davis Besse Self-revealing 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-2016002-02 30 June 2016 23:59:59 Davis Besse NRC identified 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-2016001-06 31 March 2016 23:59:59 Davis Besse NRC identified 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-2016001-07 31 March 2016 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation Licensee engineering and operations personnel performed surveillance test DBNE03214, Moderator Temperature Coefficient Measurement by Rod Swap, on October 31, 2015, to meet the requirements of TS Surveillance Requirement 3.1.3.2. Following completion of the test and analysis of the test data, licensee engineering personnel initiated CR 2015-14893 to document that the extrapolated moderator temperature coefficient was more negative than the limit specified in the plants Core Operating Limits Report (COLR). While licensee personnel correctly evaluated that operation of the unit could continue for the time being since the current moderator temperature coefficient value was within specifications, they failed to correctly interpret the entire Note associated with TS Surveillance Requirement 3.1.3.2. This Note required, in part, that the licensee calculate the minimum boron concentration at which the moderator temperature coefficient was projected to exceed its lower limit, and shutdown the unit prior to reaching this boron value. On January 27, 2016, licensee engineering and operations personnel identified that they had misinterpreted the Note associated with TS Surveillance Requirement 3.1.3.2, and a minimum RCS boron concentration value should have been established. With measurement uncertainties, a minimum RCS boron value of approximately 9.8 ppm (parts per million) was calculated by licensee engineering personnel and provided to plant operators as the minimum RCS boron limit. At that time, RCS boron had been reduced to just 16 ppm as the unit approached the normal end of the current operating cycle. Technical Specification 5.4.1(a) requires the licensee to establish, implement, and maintain applicable written procedures for the safety-related systems and activities recommended in RG 1.33, Revision 2, Appendix A. Section 2(g) of RG 1.33, Revision 2, Appendix A, requires procedures for operation of the reactor at power and process monitoring. Contrary to these requirements, the licensee failed to properly prepare and implement technically adequate written procedures and instructions for the management of RCS boron concentration. Specifically, from October 31, 2015, through January 27, 2016, operational guidance provided to the on-watch operating crews contained no minimum RCS boron value, and during this time crews were effectively attempting to reduce RCS boron concentration to zero ppm, if possible, in preparation for the units 2016 RFO. The objective of the Barrier Integrity Cornerstone of Reactor Safety is to provide reasonable assurance that physical design barriers (fuel cladding, RCS, and containment) protect the public from radionuclide releases caused by accidents or events. A key attribute of this objective involves maintaining design control parameters to protect the integrity of the plants nuclear fuel (e.g., core design analysis parameters associated with the COLR and Cycle 19 Reload Analysis, etc.) In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. Specifically, the failure to have established a minimum RCS boron concentration as directed by the Note associated with TS Surveillance Requirement 3.1.3.2 could have resulted in operations personnel reducing boron concentration to the point where the plant was operating in an unanalyzed condition, possibly outside of established accident and safety analyses. Using NRC IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that consultation with the NRC Region III SRA was necessary to establish the violations safety significance. Following discussions with the SRA, the inspectors determined that the violation was of very low safety significance (Green), since the RCS boron concentration never was decreased below the 9.8 ppm limit. The licensee had entered this issue into their CAP as CR 201601245. Licensee corrective actions included the immediate cessation of all RCS boron dilution/removal activities, the establishment of a minimum RCS boron concentration as an operational limit, and the performance of a formal causal evaluation.
05000346/FIN-2016001-02 31 March 2016 23:59:59 Davis Besse NRC identified 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-2016001-03 31 March 2016 23:59:59 Davis Besse NRC identified 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-2016001-01 31 March 2016 23:59:59 Davis Besse Self-revealing 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-2016001-04 31 March 2016 23:59:59 Davis Besse Self-revealing 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-2016001-05 31 March 2016 23:59:59 Davis Besse Self-revealing 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-2015408-01 31 December 2015 23:59:59 Davis Besse NRC identified Security
05000346/FIN-2015004-01 31 December 2015 23:59:59 Davis Besse NRC identified 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-2015004-02 31 December 2015 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation Technical Specification 3.7.11, Control Room Emergency Air Temperature Control System (CREATCS), requires that two CREATCS trains be maintained operable with the unit operating in Modes 1 through 4. While this TS provides actions and allowed outage time for a single inoperable CREATCS train, there are no provisions for both CREATCS trains being simultaneously inoperable. The provisions of TS LCO 3.0.3, therefore, apply when both CREATCS trains are inoperable at the same time, and require that actions be initiated within 1 hour from the onset of the condition to: Be in Mode 3 within 7 hours; Be in Mode 4 within 13 hours; and Be in Mode 5 within 37 hours. As discussed in Section 4OA3.2 of this report, contrary to the requirements of TS 3.7.11, both trains of CREATCS were inoperable for a period of approximately 89 hours on June 26, 2015, with the licensee taking no actions to place the unit into a Mode required by TS LCO 3.0.3. A licensee causal evaluation concluded that with CREATCS Train No. 1 already inoperable for planned maintenance, CREATCS Train No. 2 was rendered inoperable by the inadvertent opening/bumping of the supply breaker for MOV SW1395, the SW Loop No. 2 Nonessential Isolation Valve, by personnel working on planned maintenance for SW Pump No. 1. The objective of the Barrier Integrity Cornerstone of Reactor Safety is to provide reasonable assurance that physical design barriers (fuel cladding, RCS, and containment) protect the public from radionuclide releases caused by accidents or events. A key attribute of this objective involves maintaining the radiological barrier functionality of the plants control room, which is supported by CREATCS. In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. Specifically, two independent and redundant trains of the CREATCS are required to be operable to ensure that at least one is available, assuming a single failure disables the other train. The loss of both CREATCS trains could result in equipment within the control room exceeding operational temperature limits in the event of an accident. Using NRC IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the violation was of very low safety significance (Green). Specifically, while the issue was not exclusively limited to the degradation of the radiological barrier function provided for the control room, it did not also simultaneously represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. As discussed in Section 4OA3.2 of this report, the licensee had entered this issue into their CAP as CR 201508774. Licensee corrective actions included the planned installation of protective covers over critical circuit breakers located on motor-control centers in the SW pump room.
05000346/FIN-2015407-01 30 September 2015 23:59:59 Davis Besse NRC identified Security
05000346/FIN-2015003-01 30 September 2015 23:59:59 Davis Besse Self-revealing 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-2015002-01 30 June 2015 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation
Error in Procedure Use and Execution Results in Steam Feed Rupture Control System Logic Channel 1 Unplanned Inoperability During Testing Appendix B of 10 CFR Part 50, Criterion V, Instructions, Procedures, Drawings requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to this requirement, on May 29, 2015, two maintenance technicians failed to adequately perform procedure DB-MI-03205, "Channel Functional Test/Calibration and Response Time of RCP Monitor (RC3601) to SFRCS Logic Channel 1 and RPS Channel 1." Specifically, the technicians did not perform a portion of a step during the safety system testing procedure even though the step required concurrent verification and was signed off by both technicians as being completed. During the performance of Step 8.1.4.a.2, the technicians were required to: "Remove two (2) of the knurled screws stored on TB5R (terminal block) and screw them into the shorting bars for TB6L terminals 3 and 4." Although the technicians appropriately removed two knurled screws from TB5R as specified, they did not insert the screws into the shorting bars for TB6L terminals 3 and 4 as was also required by the procedure step. Later, during the restoration steps for DB-MI-03205, the technicians recognized that the knurled screws had not been inserted into the shorting bars for TB6L terminals 3 and 4 as required; they stopped work after placing equipment in a safe condition and reported the error to supervisory personnel. While RPS Channel 1 was already inoperable and in "bypass" for the testing that was in progress, because of the procedure compliance error plant operators were forced to declare the affected SFRCS Logic Channel 1 inoperable and place it in a tripped condition to comply with the requirements of TS 3.3.11, Condition A. The maintenance technicians performing the testing were removed from duty pending an investigation. To recover from the error and restore the operability of RPS Channel 1 and SFRCS Logic Channel 1, plant management directed that the test procedure be re-briefed and performed over using different maintenance technicians. The DB-MI-03205 procedure was successfully completed under enhanced supervisory oversight later that same day and both RPS Channel 1 and SFRCS Logic Channel 1 were restored to an operable status. The objective of the Mitigating Systems Cornerstone of Reactor Safety is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A key attribute of this objective is human performance, and specifically, configuration control. In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. The licensees failure to complete DB-MI-03205 as written resulted in the unplanned inoperability of SFRCS Logic Channel 1, and needlessly extended the time RPS Channel 1 was inoperable and in a bypassed condition. The licensee had entered this issue into their CAP as CR 2015-07685. An apparent cause evaluation was commissioned and corrective actions taken and planned included, but were not limited to:
A lessons learned communication was provided to each station workgroup during a stand down conducted on June 1, 2015
An enhancement is planned to DB-MI-03205 to split the applicable procedure step into two distinct actions, with the concurrent verification being required for specifically installing the knurled screws; an
An interim action was established to ensure a consistent standard of concurrent verification / independent verification performance within the station's maintenance organization. Specifically, this action required all concurrent verification / independent verification maintenance steps have direct oversight by a qualified maintenance supervisor.
05000346/FIN-2014008-01 30 June 2015 23:59:59 Davis Besse NRC identified 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-2015502-01 30 June 2015 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation The licensee-identified a finding of very low safety significance (Green) and an associated violation of 10 CFR 50.54 (q)(2) and 10 CFR Part 50.47(b)(14). Title 10 CFR 50.54(q)(2), requires, in part, that a holder of a license under this part, shall follow and maintain the effectiveness of an emergency plan that meets the requirements of Appendix E, of Part 50, and for nuclear power reactor licensees, the planning standards of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(14) requires, in part, that Periodic Exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. Section 8.1.2.c.6 of the Davis-Besse Nuclear Power Station Emergency Plan, Revision 30, states, Semiannual Health Physics drills will be conducted which involve response to, and analysis of, simulated elevated airborne and liquid samples and direct radiation measurements in the environment. Contrary to the above, from mid-2013 to the end of 2014, the licensee failed to comply with the established drill and exercise program. Specifically, the health physics drill objectives were only being partially met during this time period. The drill scenarios were limited and did not provide an opportunity for the participants to complete sampling/analysis of liquid samples. As part of the corrective actions after the discovery of this issue, the licensees Emergency Response Staff conducted a drill on December 11, 2014, to ensure that all aspects of the Health Physics drill objectives were met. The performance deficiency was more than minor because the issue was associated with the Emergency Preparedness cornerstone and adversely affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the drill scenarios were limited in scope and health physics objectives specified in the emergency plan, were not carried out fully during exercises and drills, that were conducted in the middle of 2013 through the end of 2014. The NRC determined that this was a failure to comply with the licensees emergency plan and a degradation of a planning standard function in accordance with 10 CFR, Part 50.47(b)(14), and was a very low safety significance issue (Green) as indicated in Inspection Manuel Chapter 0609, Emergency Preparedness SDP, Appendix B, Attachment 2, Failure to Comply Significance Logic. Because the finding is of very low safety significance (Green) and it was entered into the licensees Corrective Action Program as Condition Report, CR-2014-16715, this violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000346/FIN-2015001-01 31 March 2015 23:59:59 Davis Besse Licensee-identified Licensee-Identified Violation Plant TS 5.4.1(d), requires, in part, the licensee to establish, implement, and maintain applicable written procedures covering fire protection program implementation. The fire protection program was implemented, in part, by Davis-Besse Procedure DBFP00009, Fire Protection Impairment and Fire Watch, Revision 20. Procedure DBFP00009, Step 6.3.4, states: Roving Fire Watches shall observe the assigned patrol area(s)/room(s)/panel(s) which is (are) to be observed as related to the impairment, for example, room associated with impaired door, damper, penetration seal, detector, etc. Contrary to this requirement, for several hours on February 16, 2015, the licensee failed to observe the interior of Room 425, the Radiation Protection (RP) Instrument Calibration Room, when an applicable fire impairment had existed. Late in the morning on February 16th, plant maintenance technicians began a planned work activity to replace 27 smoke detectors in Fire Detection Zone 412A. To compensate for this planned loss of fire detection capability, the licensee's fire protection program required an hourly fire watch patrol to be performed for each of the 17 rooms covered by Fire Detection Zone 412A. This hourly fire watch patrol was instituted by the on-watch Operations crew at 12:00 p.m., and assigned to a member of the site's Security Department, as was the standard station practice. Because the door to Room 425 was locked, however, the individual performing the hourly fire watch patrol did not enter Room 425, as required, and only checked the door to the room. This practice was repeated again at 2:00 p.m., and continued hourly by each individual who performed the fire watch patrol until 8:00 p.m. At that time, the individual performing the hourly fire watch patrol raised a question about the locked door to Room 425 to the Operations supervisors who were on watch, and it was confirmed that the room needed to be entered to adequately perform the fire watch patrol. A key for Room 425 was obtained from RP, and the fire watch patrol performed correctly from that point on. The inspectors reviewed this violation using the guidance contained in Appendix B, Issue Screening, of IMC 0612, Power Reactor Inspection Reports. The inspectors determined that the licensees failure to properly implement plant procedures for performing compensatory fire watches was a performance deficiency that was reasonably within the licensees ability to foresee and correct and should have been prevented. This violation was associated with the Initiating Events cornerstone of reactor safety and was of more than minor 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 watch patrols established as compensatory measures should have been properly performed for the duration of the impairment so that the sites ability to promptly detect and suppress a fire would be maintained. The inspectors evaluated the violation using IMC 0609, Attachment 4, Phase 1Initial Screening and Characterization of Findings. Because it involved fire protection, the inspectors transitioned to IMC 0609, Appendix F, Fire Protection Significant Determination Process (SDP). The violation was characterized according to IMC 0609, SDP, Appendix F, Attachment 1, "Fire Protection SDP Phase 1 Worksheet," dated September 20, 2013. The violation screened as of very low safety significance (Green), per Attachment 1, Question 1.3.1.A, because it did not affect the ability of the reactor to reach and maintain safe shutdown. The licensee had entered this issue into their CAP as CRs 201502119, 201502126, 201504246, and 201504248. A limited apparent cause evaluation was performed and corrective actions included, but were not limited to: Creation of a formal pre-job brief check list for fire watch patrols to be used and maintained at the Work Support Center desk for ready reference, and maps for multiple room fire watch patrols; and Communication of lessons learned, identification and discussion of knowledge gaps and reinforcement of the site's human performance tools with respect to fire watch patrol expectations.
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