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05000259/FIN-2016011-012016Q3Browns FerryFailure to Identify and Evaluate All Targets Within the Zone of Influence of Ignition SourcesThe NRC identified a violation of 10 CFR 50.48(c) for the licensees failure to address in the Fire Probabilistic Risk Assessment (Fire PRA) the risk contribution associated with all potentially risk significant fire scenarios for a given fire compartment/fire area. The licensee did not identify and evaluate all targets that were within the zone of influence (ZOI) of ignition sources for selected fire scenarios that could potentially contribute to the risk for the fire scenarios. The licensee entered the issue in the corrective action program (CAP) as Condition Reports (CRs) 1195603 and 1197392. The affected area was already covered by an hourly roving fire watch as a compensatory measure. The licensees failure to address the risk contribution associated with all potentially risksignificant fire scenarios, as required by section 2.4.3.2 of NFPA 805, was a performance deficiency. For each example, the performance deficiency was determined to be more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e., fire) and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to analyze the full risk impact of the selected fire scenarios, and the missed targets in the ZOI for the selected fire scenarios had the potential to impact the ability to achieve safe and stable conditions. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the finding was screened as Green in step 1.6.1 Screen by Licensee PRA-Based Safety Evaluation. There was no cross cutting aspect assigned to this finding because it was not indicative of current licensee performance since the original ignition source and target walkdowns were performed more than 3 years ago. (Section 1R05.06)
05000259/FIN-2016011-022016Q3Browns FerryFailure to Adequately Identify and Evaluate All Circuit Failures for NSCA Credited EquipmentThe NRC identified a violation of 10 CFR 50.48(c) for the licensees failure to properly identify circuits required for the nuclear safety function. Specifically, the licensees Nuclear Safety Capability Assessment (NSCA) failed to identify that fire-induced failure of cables associated with the undervoltage trip function of the 4KV Shutdown Board could cause the shutdown board to not shed loads upon an undervoltage condition. This could lead to overloading the emergency diesel generator (EDG) credited for powering the shutdown board. This item was entered into the CAP as CR 1199002. The affected area was already covered by an hourly roving fire watch as a compensatory measure. Additionally, the licensee submitted EN 52150 to the NRC, documenting this as an unanalyzed condition. The licensees failure to identify circuits required for the nuclear safety function, as required by Section 2.4.2.2.1 of NFPA 805 was a PD. The PD was determined to be more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e., fire) and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to analyze the effects of fire damage on the 4kV shutdown bus undervoltage circuitry could result in overloading the emergency diesel generator (EDG) credited for powering the shutdown board. Using the guidance of IMC 0609, App. F, the finding was screened as Green because the risk increase associated with the finding was an increase of core damage frequency of <1E-6/year. There was no cross cutting aspect assigned to this finding because it was not indicative of current licensee performance since the original ignition source and target walkdowns were performed more than 3 years ago. (Section 1R05.06)
05000335/FIN-2016011-022016Q2Saint LucieFailure to modify the Diesel Oil Storage Tank Overflow Line as Required by a Fire Protection License RequirementInspectors identified a Severity Level IV violation of 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, for the licensees failure to modify the Unit 2A and 2B diesel oil storage tank (DOST) overflow lines as required by a fire protection license requirement. The issue was entered into the sites corrective action program as AR 2140024. The licensees failure to notify the NRC of changes to a licensed activity that was stipulated in the fire protection license condition (Table S-1) was a performance deficiency. The inspectors determined the PD was more than minor because the licensee failed to notify the NRC that the Unit 2 DOSTs overflow lines would not be modified; and, subsequently failed to request an exemption from the requirements of NFPA 30. Traditional enforcement was applied because the PD impacted the ability of the NRC to perform its regulatory oversight function. In accordance with the NRC Enforcement Manual, Part II, Section 2.2, Actions Involving Fire Protection, the inspectors evaluated this finding to be a Severity Level IV violation. The inspectors determined that a cross-cutting aspect was not applicable because the issue was associated with a traditional enforcement violation.
05000335/FIN-2016011-012016Q2Saint LucieFailure to Meet the Quality Requirements Specified By NFPA 805Inspectors identified a Severity Level IV violation of 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, for failing to maintain adequate documentation and quality of analyses. Specifically, the NRC identified multiple examples when the licensee failed to comply with site quality assurance procedures. The issue was entered into the sites corrective action program as ARs 2139768, 2139986, and 2139993. The licensees failure to maintain adequate documentation and quality of analyses to maintain configuration control, such that they could be checked for adequacy and accuracy, was a performance deficiency (PD). The inspectors determined that the issue was more than minor because the ability of the NRC to verify aspects of the licensees NFPA 805 program was impacted. The inspectors determined that the Fire Protection Significance Determination Process (IMC 0609, Appendix F) was not suitable for screening this issue. Traditional enforcement was applied because the PD impacted a regulatory oversight function. In accordance with the NRC Enforcement Manual, Part II, Section 2.2, Actions Involving Fire Protection, the inspectors evaluated this finding to be a Severity Level IV violation. A cross-cutting aspect was not applicable because the issue was associated with a traditional enforcement violation.
05000390/FIN-2016002-102016Q2Watts BarUntimely 10 CFR 50.73 Notification of an Inoperable Rod Position IndicationThe NRC identified a SL IV NCV of 10 CFR 50.73(a)(2)(i)(B) for the licensee's failure notify the NRC that the TS LCO 3.1.8 required action and completion time were not met when the analog rod position indication (ARPI) and the demand position indication system were not operable. Subsequently, the licensee submitted LER 2016-007-00 for this issue on June 20, 2016. This violation was placed in the licensees corrective action program as CR 1163150. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects.
05000390/FIN-2016002-092016Q2Watts BarUntimely 10 CFR 50.73 Notification of Failure to Meet Technical Specification Surveillance Requirement 3.5.2.3 for the Emergency Core Cooling SystemThe NRC identified a SL IV NCV of 10 CFR 50.73(a)(2)(i)(B) for the licensee's failure to report, within 60 days of discovery, a condition which was prohibited by the plants TS associated with recent performances of TS surveillance requirement (SR) 3.5.2.3 for verification that emergency core cooling system (ECCS) piping is full of water. Subsequently, the licensee submitted LER 2016-003-00 for this issue on May 10, 2016. This violation was placed in the licensees corrective action program as CR 1166564. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects.
05000391/FIN-2016002-082016Q2Watts BarFailure to Follow Maintenance Procedure Results in overspeed trip of the 2C-S Turbine Driven Auxiliary Feedwater PumpA self-revealed Severity Level (SL) IV non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified at Watts Bar Unit 2 for the licensees failure to follow procedure 0-MI-1.003, Disassembly, Inspection, and Reassembly of Auxiliary Feedwater Pump Turbine. Specifically, the valve stem spring coil gap was not set in accordance with procedure, causing the turbine-driven auxiliary feedwater (TDAFW) pump to trip on electrical overspeed when the level control valves (LCVs) were closed. This issue was corrected on May 30, 2016, when the proper spring coil gap was set and verified and the post maintenance test was performed satisfactorily. The issue was entered into the licensees corrective action program as CR 1175968. The performance deficiency was more than minor because it represented an improper or uncontrolled work practice that could impact quality or safety involving safety-related structures, systems, and components (SSCs). The finding was a SL IV violation because it represented a failure to meet a regulatory requirement, specifically a quality assurance (QA) criteria to follow quality-related procedures, which had more than minor safety significance. The finding was assigned a crosscutting aspect of resources in the Human Performance area because the licensee failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the procedure that set the coil spring gap lacked sufficient detail and rigor to ensure that the coil gap would be set appropriately by the technicians.
05000390/FIN-2016002-072016Q2Watts BarUntimely 10 CFR 50.73 Notification of Inoperable Containment PenetrationsThe NRC identified a SL IV NCV of 10 CFR 50.73(a)(2)(i)(B) for the licensee's failure notify the NRC that the TS LCO 3.6.3 required action and completion time were not met for an inoperable emergency raw cooling water (ERCW) containment isolation valve. Subsequently, the licensee submitted LER 2016-009-00 for this issue on July 15, 2016. This issue was placed in the licensees corrective action program as CR 1174000. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects.
05000391/FIN-2016002-052016Q2Watts BarFailure to Perform A TDAFW Surveillance In Accordance With ProceduresThe NRC identified a SL IV NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 2 for the licensees failure to follow the surveillance test program procedure by making adjustments to the turbine-driven auxiliary feedwater (TDAFW) pump control system during the performance of a surveillance instruction. The licensee reperformed the surveillance instruction with satisfactory results. The issue was entered into the licensees corrective action program as CR 1167102. The performance deficiency was more than minor because making adjustments to the TDAFW pump control system during the performance of a surveillance instruction could invalidate the test and result in the TDAFW pump being inappropriately declared operable. As described in IMC 2517, the significance of this issue was determined using traditional enforcement, because the cornerstone associated with this finding was not being assessed by the reactor oversight process (ROP). The inspectors determined this finding to be of very low safety significance, SL IV, because it represented a failure to meet a regulatory requirement, specifically a QA criteria to follow quality-related procedures, which had more than minor safety significance. The finding was assigned a cross-cutting aspect of Conservative Bias in the Human Performance area because numerous individuals were aware the speed adjustment had been made while completing the surveillance instruction but did not question the appropriateness of that adjustment until prompted by NRC inspectors.
05000391/FIN-2016002-042016Q2Watts BarFailure to Follow Operability Procedure Results in Potential Inoperability of the 2A-A Auxiliary Feedwater PumpThe NRC identified a SL IV NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 2 for the licensees failure to follow procedure OPDP-8, Operability Determination Process and Limiting Condition for Operation Tracking, Revision 22. Specifically, the 2A-A motor-driven auxiliary feedwater pump (MDAFW) was potentially inoperable in mode 3 due to inadequate compensatory measures that were being controlled outside of the operability process. The issue was corrected and the pump returned to operable status on April 19, 2016. The issue was entered into the licensees corrective action program as CR 1163431. The performance deficiency was more than minor because it represented an improper or uncontrolled work practice that could impact quality or safety, involving safety-related SSCs. Specifically, failure to appropriately use the operability process when measures must be established to compensate for degraded or nonconforming conditions can lead to SSC inoperability. As described in IMC 2517, the significance of this issue was determined using traditional enforcement, because the cornerstone associated with this finding was not being assessed by the reactor oversight process (ROP). The inspectors determined this finding to be of very low safety significance, SL IV because it represented a failure to meet a regulatory requirement, specifically a quality assurance (QA) criteria to follow quality-related procedures, which had more than minor safety significance. The finding was assigned a cross-cutting aspect of Work Management in the Human Performance area because the minor maintenance work order created to compensate for the oil loss from the 2A-A MDAFW pump was never reviewed by operations, which could have identified the out of process error. (H.5).
05000390/FIN-2016002-032016Q2Watts BarUntimely 10 CFR 50.73 Notification of an Inoperable Charging PumpThe NRC identified a Severity Level (SL) IV non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.73(a)(2)(i)(B) for the licensee's failure to notify the NRC that the technical specification (TS) limiting condition for operation (LCO) 3.5.2 required action and completion time were not met when the 1B-B centrifugal charging pump (CCP) was inoperable due to an inoperable room cooler. Subsequently, the licensee submitted LER 2016-006-00 for this event on June 30, 2016. This issue was placed in the licensees corrective action program (CAP) as CR 1165380. Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this performance deficiency was dispositioned under traditional enforcement and the violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy. Using the example listed in Section 6.9.d.9, A licensee fails to make report required by 10 CFR 50.73, the issue was determined to be a SL IV violation. In accordance with IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, traditional enforcement violations are not assessed for cross-cutting aspects.
05000390/FIN-2016002-012016Q2Watts BarFailure to Ensure that a Train of Source Range Detection was Available to Monitor Neutron Population During a Fire EventThe NRC identified a Green NCV of Operating License Condition 2.F for the licensees failure to ensure that a train of source range detection was available to monitor neutron population during the initial stages of a fire event on Unit 1. This issue was entered into the licensees corrective action program as CR 1098240. The licensees failure to ensure a train of source range detection was free from fire damage was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to maintain the capability to monitor neutron population during the early stage of a fire event. In accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, the finding was determined to be of very low safety significance (Green) because the reactor would have been able to reach and maintain a stable plant condition. No cross-cutting aspect was identified for this issue.
05000390/FIN-2016002-022016Q2Watts BarFailure to Translate Design Requirements into a Maintenance Procedure for the 1B-B Charging Pump Room CoolerThe NRC identified a Green NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control for the licensees failure to specify nominal shaft size along with specific acceptance criteria for shaft tolerance measurements for the 1B-B centrifugal charging pump (CCP) room cooler fan shaft. The licensee repaired the room cooler by replacing the fan shaft and the finding was entered into the licensees corrective action program as CR 1146474. The performance deficiency was more than minor because it affected the equipment performance attribute of the mitigating system cornerstone 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 that this finding required a detailed risk analysis since it represented an actual loss of function of a single train for greater than its TS-allowed outage time. The finding does not present an immediate safety concern because the licensee has verified current operability. A Senior Reactor Analyst evaluated the increase in core damage frequency due to the pump being non-functional over the exposure period and determined it was 3.6E-7/year (Green). The dominant scenario was a loss of component cooling water, which combined with a loss of RCP seal injection causes a loss of coolant accident and leads to core damage. The risk increase was very low because of the limited exposure time, the availability of the opposite train pump, and the time dependent nature of the pump failing due to lack of room cooling. The inspectors determined that the finding had a cross-cutting aspect of design margin in the area of Human Performance because the licensee failed to carefully guard margins through a systematic and rigorous process. Specifically, the translation of shaft diameter from design documents into 0-MI-0.16 lacked rigor and allowed an undersized shaft to go undetected, leading to cooler failure.
05000390/FIN-2016002-062016Q2Watts BarFailure to Satisfy TS LCO 3.6.3The NRC identified a Green NCV of TS for the failure to recognize and take the required actions in TS 3.6.3 for inoperable containment penetration flow paths. Specifically, the required actions of TS 3.6.3 applied on November 21, 2015, and were not taken until January 30, 2016. Upon discovery, on January 30, 2016, the affected containment penetrations were isolated by placement of a clearance, thereby satisfying the TS required actions. The licensee entered the violation into the CAP as CR 1172114. The performance deficiency was more than minor because the ERCW supply and discharge containment penetrations for the 1D upper containment cooler were inoperable for longer than the TS allowed outage time. Because the 1D upper containment cooler ERCW containment penetrations were inoperable and resulted in the failure to satisfy TS LCO 3.6.3, reasonable assurance of the integrity of the containment design barrier was adversely affected. The inspectors determined the finding was of low safety significance (Green) because the upper containment cooler ERCW penetrations are small lines (<1-2 inches in diameter) and IMC 0609, Appendix H Containment Integrity Significance Determination Process dated May 6, 2004, Table 4.1 states that small lines (<1-2 inches in diameter) would not generally contribute to LERF. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee failed to make the prudent choice to fully evaluate the unsuccessful surveillance test on November 15, 2015, and instead simply documented the issue in the corrective action program and deferred the solution, resulting in the TS violation six days later.
05000321/FIN-2016007-012016Q2HatchFailure to provide reasonable assurance that Appendix R time critical operator actions (TCOAs) can be completed in a timely mannerThe NRC identified a Green non-cited violation (NCV) violation of Hatch Technical Specifications 5.4.1.d, Procedures, for Units 1 and 2, for not ensuring manual action feasibility for actions in fire area (FA) 0024. Specifically, the licensee failed to provide reasonable assurance that a credited manual action to ensure emergency power was both feasible and reliable in response to a fire event. The licensee plans to assess the issue and entered this violation into their Corrective Action Program (CAP) based upon CR10209664, CR10213119, & CR10212821. The licensees failure to provide reasonable assurance that Appendix R time critical operator actions (TCOAs) associated with fire events can be completed in a timely manner was a performance deficiency (PD). The PD was more than minor because if left uncorrected, it could to lead to a more significant safety concern. Specifically, the exclusion of TCOAs from a validation process could lead to plant or program changes that prohibit the completion of actions required to meet the licensing basis. Using the guidance of IMC 0609, App. F, the finding was screened as Green because the finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. The deficiency was screened with IMC 0310, Aspects Within Cross Cutting Areas, to determine if any cross-cutting areas were applicable. The team concluded cross-cutting was applicable to the problem identification and resolution (PI&R) area, evaluation attribute due the licensees failure to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000324/FIN-2016002-012016Q2BrunswickFailure to Identity Broken Auto Start Control Relay on Emergency Diesel Generator 1An NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified because the licensee failed to promptly identify and correct a condition adverse to quality (CAQ) on emergency diesel generator (EDG) 1. Specifically, from February 7, 2016, until March 5, 2016, the licensee failed to promptly identify and correct a broken auto start control relay (ASCR) which resulted in reduced capacity of EDG 1 due to load oscillations and inoperability of EDG 1 due to oscillating between droop and isochronous mode. The oscillations could cause the EDG to not meet Technical Specification (TS) frequency and load requirements. The licensee replaced the ASCR and entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 2007720. The licensees failure to promptly identify and correct the broken ASCR, which resulted in reduced capacity and inoperability of EDG 1 due to load oscillations, was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct the malfunctioning ASCR resulted in reduced capacity of EDG 1 due to load oscillations, and could cause EDG 1 to not meet TS frequency and load requirements. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding screened to a more detailed risk evaluation because it represented a loss of system and/or function, and the finding represented an actual loss of a function of a single train for greater than the TS allowed outage time. The regional Senior Reactor Analyst evaluated the finding and determined it to be Green. The risk was low because of the diverse sources of AC power available, and the long duration of some of the sequences allowed a greater potential for recovery of a failed AC power source. The dominant risk sequences contained common cause failure of the diesel generators, with the supplemental EDG aligned to the other unit, and non-recovery of offsite power or of an EDG. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the identification attribute because the licensee failed to implement a CAP with a low threshold for identifying issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to write a timely NCR and identify the load oscillations as a CAQ. (P.1)
05000324/FIN-2016002-022016Q2BrunswickFailure to Verify or Check the Adequacy of Design of the EDG 3 Auto-Start CircuitryA self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to verify or check the adequacy of design of the EDG 3 emergency auto-start circuitry. Specifically, on October 24, 2011, the licensee failed to verify or check the adequacy of design of the fuse block holder modification to the EDG auto-start circuitry. This resulted in the fuse block holder connection becoming loose, a loss of continuity through the circuit, and the inoperability of EDG 3. The licensee replaced the fuse block holder, performed a continuity check, and plans to implement a design change to install continuity indication for continuous verification of continuity. The licensee entered this issue into the CAP as NCR 2007449. The licensees failure to verify or check the adequacy of design of the EDG 3 emergency auto-start circuitry fuse block holder modification was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This resulted in the fuse block holder connection becoming loose, a loss of continuity through the circuit, and the inoperability of EDG 3. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding screened to a more detailed risk evaluation because it represented a loss of system and/or function, and the finding represented an actual loss of a function of a single train for greater than the TS allowed outage time. The regional SRA performed a detailed risk review for the finding. The finding was determined to be Green. The limited duration of the EDGs failure of the auto start, the ability to manually recover the EDG, and the availability of the other EDGs and of the supplemental EDG contributed to the low risk value. The dominant risk sequences were of low value, and were Station Blackout with failure to recover offsite power or the EDGs. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the identification attribute because the licensee failed to implement a CAP with a low threshold for identifying issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to identify EDG 3 was inoperable on February 7, 2016, when the indications were apparent. (P.1)
05000335/FIN-2016011-032016Q2Saint LucieFailure to Meet the Combustible Control Requirements Specified By NFPA 805 for Work Platforms Located in the Intake Cooling Water Pump HouseInspectors identified a Green, non-cited violation (NCV) of 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, for the licensees failure to comply with the combustible control requirements for work platforms that were located in the Intake Cooling Water (ICW) Pump House. The issue was entered into the sites corrective action program as AR 2137088. The licensees failure to adequately implement combustible material control requirements in procedures ADM-27.11 and Procedure 0010434 was a performance deficiency (PD). The (PD) adversely impacted the Initiating Events cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during plant operations. Additionally, if left uncorrected, the deficiencies in the combustibles control program could result in wood platforms being staged in other areas of the plant. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, dated June 19, 2012, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, which determined that, an IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, review was required because it was a fire prevention finding. The finding was determined to be of very low safety significance (Green), at Step 1.4.1.B because the impact of a fire would be limited to no more than one train of equipment important to safety. The inspector identified a cross-cutting aspect in work management because the licensee failed to ensure that the sites combustible control requirements were met during the installation and use of wood platforms in the ICW pump house (H.5).
05000321/FIN-2016007-022016Q2HatchPassive Fire ProtectionThe NRC identified a Green NCV of Hatch Renewed Operating License Conditions (OLCs) 2.C.(3) and 2.C.(3)(a), for Units 1 and 2 respectively, because the licensee failed to adhere to branch technical position (BTP) Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1. Specifically, the licensee failed to implement the NFPA 80, Fire Doors and Windows, requirements to ensure fire confinement, thus affecting the defense in depth (DID) aspects. Description: During walkdowns of the chosen fire areas, the inspectors assessed whether the passive fire protection features adhered to the NFPA code commitments specified in the current licensing basis. Based upon the walkdown of the West DC Switchgear Room 2A (FZ 2018) and the adjacent access corridor (FZ 2014), the inspectors observed what they determined to be inadequate fire protection program implementation which would result in a degraded fire confinement ability between two fire zones. The code states, in part, that when doors are installed on only one face of a fire wall, heat responsive units shall be located on each side of the wall and interconnected so that the actuation of any one of them will permit the door to close. In this case, the heat responsive units were fusible metallic links designed to melt at a specific temperature and initiate door closure. The fusible links were not installed as required on both sides of the credited fire door between FZ 2018 and FZ 2014. Specifically, a link was only on one side of door 2L482C10. No link was installed on the side of the door in FZ 2014. The door in question was considered to be a Class A fire door and was designed to provide at least 3-hours of fire resistance between adjacent fire zones for a postulated fire event. Neither of these fire zones were protected with automatic suppression capability. In addition, there were existing exemptions in place for not meeting the 10 CFR Appendix R, III.G.2 requirements as referenced by the FHA. This further supported the need for ensuring the DID measures were adequate for fire confinement. In the second example, on October 15, 2015, the NRC resident inspector observed that an issue existed with the installation of an electro-thermal link designed to close the 2C EDG rolling fire door which separated FA 2407 from FA 0401. Specifically, it was noted that the device was installed in an improper configuration and that the electro-thermal link was mounted directly to the wall and secured using a nut and washer. In this configuration, the washer overlapped the seam of the electro-thermal link and hampered the links ability to separate and automatically close the door. The licensee declared the electro-thermal link non-functional and performed a functionality assessment. The assessment concluded that additional links designed to release the door in the event of a fire would have eventually fused, releasing the door. In addition, the gaseous carbon dioxide (CO2) fire suppression system protecting the 2C EDG would have retained the required CO2 gas concentration using the air control louvers, which were installed in series with the rolling fire door. The degraded electro-thermal link was corrected on November 10, 2015. Analysis: The licensees failure to ensure the DID aspects of the FPP were implemented consistent with the NFPA 80 requirement as specified by the current fire protection licensing basis was a PD. The PD was more than minor because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. Specifically, the lack of a required link above the fire door between the West DC Switchgear Room 2A and the adjacent access corridor fire zone and the improperly installed link between EDG Room 2C and the adjacent access corridor would have negatively impacted the expected response time of each of the fire doors to close. In addition, review of historical work orders and condition reports indicated problems with the air balance louvers coincident with the degraded ETL controlling the closure of the fire door would have impacted the likelihood of confining the CO2 gas at the required design concentration. In both these instances, the finding had a negative impact on the program DID aspects for the fire confinement category. In accordance with NRC IMC 0609, Significance Determination Process, Appendix F, the inspectors performed a Phase 1 analysis and determined the finding resulted in very low significance, Green, based on question 1.4.3-A since, in each case, the combustible loading on both sides of the barrier wall represented a fire duration less than 1.5 hours (i.e., less than 120,000 Btu/ft2). The team determined that no cross-cutting attributes were applicable based upon the issue being associated with meeting the original NFPA 80 design criteria at licensing. Enforcement: Hatch Operating License Condition (OLC) 2.C.(3) and 2.C.(3)(a), for Units 1 and 2 respectively, stated, in part, that Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report (UFSAR) for the facility. The E. I. Hatch UFSAR, Unit 2, Section 9.5, stated in part the plant fire protection system is described in the Edwin I. Hatch Nuclear Plant Units 1 and 2 Fire Hazards Analysis and Fire Protection Program (incorporated by reference into the FSAR). FHA Section 9.0, Appendix A Compliance Matrix, stated the licensee complied with the applicable sections of BTP APCSB 9.5-1. The General Guideline for Plant Protection section stated that the NFPA 80, Fire Doors and Windows was applicable for fire doors. Contrary to the above, the team identified two instances that were not consistent with the stated commitments. The licensee documented this issue with condition reports CR10085883, CR 10135493, CR 10144100, and CR10022283. The team reviewed the DID and the fire confinement provisions of NFPA 805 as referenced by Sections A.4.4.6.4 and 7.3.7. In addition, Section 5.11.3.1, which states in part that passive fire protection devices such as doors and dampers shall conform to the following NFPA standards, as applicable unless otherwise permitted by 5.11.3.2. Because Hatch committed to adopt NFPA 805 and the aforementioned issues meet the criteria as stated in the NRC Enforcement Policy (Policy), Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48), the NRC will disposition the violations in accordance with the Policy and grant enforcement discretion. The NRC will also disposition the associated findings in accordance with Inspection Manual Chapter 0305, Section 11.05, Treatment of Items Associated with Enforcement Discretion.
05000390/FIN-2016001-012016Q1Watts BarFailure to Use a Procedure Appropriate to the Circumstances for the Auxiliary Control Air System Train AA self-revealing non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Procedures was identified for the licensees failure to use a procedure appropriate to the circumstances for work associated with the A-A auxiliary control air system (ACAS) compressor. Specifically, the licensee used a section of procedure 0-SOI-32.02, Auxiliary Air System, Revision 2, that placed the air compressor in OFF when it was intended to place it in A-Auto. The licensee restored the compressor to A-Auto and entered this issue into their corrective action program as condition report (CR) 1131261. The performance deficiency was more than minor because it affected the equipment performance attribute of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the ACAS train A was nonfunctional for approximately 19.5 hours on January 29, 2016 and as a supported system, the auxiliary feedwater system was inoperable during this time. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of a single train for greater than its TS allowed outage time. The finding has a cross cutting aspect in the Work Management component of the Human Performance area because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the planning and execution of work on the A-A ACAS compressor on January 29, 2016 lacked sufficient rigor to ensure the activity was performed as intended.
05000390/FIN-2016001-072016Q1Watts BarFailure to Maintain Operating LogsThe NRC identified a NCV of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, for the licensees failure to maintain sufficient records to furnish evidence of activities affecting quality. The licensee entered this issue into their corrective action program as CR 1127691. The inspectors determined that the licensees failure to document plant operations in the operating logs in accordance with OPDP-1 was a violation of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records. This violation constitutes a traditional enforcement violation because it impacts the NRC's ability to carry out its regulatory function. The failure to maintain accurate logs was more than minor because it would have likely caused the NRC to undertake further inquiry and was consistent with Enforcement Policy section 6.9.d.1 for a SL-IV violation. Crosscutting aspects are not assigned to traditional enforcement violations.
05000390/FIN-2016001-092016Q1Watts BarAppropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant ResponseThe inspectors identified an URI associated with the timely and effective corrective action associated with an adverse trend in safety related pump performance, including mechanical seal degradation and failure. This item is unresolved pending review and evaluation of the licensees response to the CRs generated to determine if a performance deficiency exists. During Unit 1, 2015 fall outage, the 1A Safety Injection (SI) pump mechanical seal was replaced. The mechanical seal had degraded to a point at which the leakage was greater than the Technical Specification limit for ECCS leakage outside of containment. The inspectors identified several issues during a review of the Prompt Determination of Operability for CR 1125623 and WO 116050574 to replace the seal. Specifically, inspectors found that non-QA1 parts were being used for seal replacement, the seal was the original equipment manufacturer part from startup, the failure mechanism was not clearly understood, and an extent of condition review was not performed. The inspectors reviewed other safety related pump mechanical seal performance and corrective action program entries. The inspectors are awaiting the completion of the licensees evaluation to determine the licensees compliance with applicable procedures and TS relative to pump operability and ECCS leakage limits outside containment. Additional inspection activities are needed to determine the extent of condition and compliance with the procedures and TS. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000390/2016001-09, Appropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant Response.
05000390/FIN-2016001-082016Q1Watts BarCharging Pump 1B-B Room Cooler Fan Bearing FailureInspectors identified an unresolved item (URI) associated with the failure of the 1B-B charging pump room cooler. This item is unresolved pending review of an equipment apparent cause evaluation that was performed after deficiencies were identified by inspectors in the past operability evaluation. On September 27, 2015, the licensee installed a new bearings on the 1B-B CCP room cooler fan shaft as part of planned maintenance (PM) under WO 115790759. The WO noted the room cooler had a broken lubrication line close to the point where it is attached to the outboard fan shaft bearing, but the new bearing on the fan shaft, including the outboard shaft bearing, were installed without an immediate repair of the lubrication line. The bearing replacements for WO 115790759 were accomplished in accordance with maintenance procedure 0-MI-0.16, Maintenance Guidelines for Belt Driven Equipment, Rev. 7. Appendix D, Bearing Installation, Step 14 requires, All remote lubrication lines, remote vibration attachments, etc. shall be verified as attached prior to return to service. The work order noted at this step that the lubrication line to the outboard fan shaft bearing was broken in half and will need to be replaced prior to return to service and the step was left blank. The licensee did not initiate a CR for this degraded condition. Due to the broken lubrication line, the outboard fan shaft bearing was the only fan shaft bearing that was not greased during installation. October 15, 2015, the licensee completed the PMT for the room cooler and noted it to be satisfactory. The broken lubrication line documented in the PM WO was identified and CR 1093983 was initiated to document the condition. This CR stated that the broken lubrication line did not affect the functionality of the fan and could be repaired at the next scheduled PM. This assessment was not questioned during the review of the CR for operability. The fan was returned to service and declared operable. On December 4, 2015, the room cooler failed in service. The licensee declared the 1BB charging pump inoperable and entered the applicable TS LCO. Investigation revealed that the outboard fan shaft bearing had failed. At this point, the inappropriate treatment of the degraded lubrication line under 0-MI-0.16 and the associated PMT was identified. This issue was documented in the licensees CAP in CR 1111791. The licensee performed a past operability evaluation (POE) for CR 1111791 which concluded the fan was operable until several hours before the time of the failure. The POE was based largely on statements from the bearing vendor indicating that the new bearing was pre-lubricated at the factory and should have performed under load for a long period of time without needing to be pre-greased at installation. The POE was hampered by the fact that the licensee did not retain the damaged bearing for failure analysis. The inspectors reviewed the POE and determined that it failed to adequately document sufficient information to either discount the broken lubrication line as a cause of the bearing failure or to identify another cause. In response, the licensee opened an investigation of the cause of the bearing failure under an equipment apparent cause evaluation. Because more information is necessary to evaluate the cause of the 1B-B CCP room cooler fan shaft bearing failure, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to review the equipment apparent cause evaluation, which was not completed by the end of the inspection period. This is identified as URI 05000390/2016001-08, Charging Pump 1B-B Room Cooler Fan Bearing Failure.
05000269/FIN-2016007-012016Q1OconeePressure Boundary of Motor Operated Valves Could be Breached Due to Fire- Induced Hot ShortAn unresolved item was identified regarding the licensees evaluation of certain motor operated valves (MOVs) in the NSCA. Specifically, based on the conclusions in the licensees NSCA, as well as subsequent evaluations, MOVs that are subject to a hot short that bypasses the torque or limit switch could result in damage to the valve that causes an unmitigated loss of reactor coolant system (RCS) inventory due to leakage through the damaged valves pressure boundary or the valves associated sealing components. Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, stated that fire damage could cause an electrical hot short that bypasses thermal overload protection for MOVs, and that this hot short could result in damage to the valve. As a part of the licensees transition to NFPA 805, the licensee identified a number of MOVs that could be susceptible to IN 92-18 type damage. These valves were classified as non-compliant components or variances from deterministic requirements (VFDRs). The subsequent evaluation of these valves by the licensees Fire PRA group determined that these VFDRs met the acceptance criteria of the Fire Risk Evaluation, as documented in OSC-9314, as being acceptable "as-is" and that no further action was required. These VFDRs and their FPRA dispositions were communicated to the NRC in the April 2010 Oconee NFPA 805 license amendment request (LAR). Subsequent to NRCs issuance of the SER, Oconee Valve Engineering determined that, due to the size of the installed motor/gearbox, 10 MOVs could potentially suffer IN 92-18 damage to the extent that the integrity of the valve body or bonnet could be compromised. Loss of valve integrity of the valve pressure boundary was not an assumption used in the FPRA evaluation. The licensee documented this condition in AR 01906086. After further evaluation, the licensee documented in AR 01999309 that 9 of the original 10 valves identified could potentially suffer IN 92-18 damage to the extent that the integrity of the valve body or bonnet could be compromised. For the 9 affected valves, the licensee has undertaken additional evaluations to determine whether some portion of the valve would fail before the valves pressure boundary is compromised, or that any possible leakage that may result can be bounded by the credited RCS make-up sourcein this case, the reactor coolant make-up pump. Inspectors determined that no immediate safety concern existed with this item because the licensee had implemented compensatory measures in accordance with the sites approved FPP. This item is unresolved pending inspector receipt and review of the licensees evaluation.
05000269/FIN-2016007-022016Q1OconeePostulated Fire Affecting High Pressure Injection Pump Did Not Receive a VFDR EvaluationThe NRC identified a Green NCV of 10 CFR 50.48(c) and National Fire Protection Association Standard (NFPA) 805, Section 2.4.2.4 for the licensees failure to perform an adequate engineering analysis to determine the effects of fire on the ability to achieve the nuclear safety performance criteria, and consequently, did not add an associated variation from deterministic requirements (VFDR) into the Fire probabilistic risk assessment (PRA). Specifically, the licensees Nuclear Safety Capability Assessment (NSCA) failed to identify cables in the turbine building (TB) that could prevent the operation of the High Pressure Injection (HPI) Pumps. This item was entered into the corrective action program (CAP) as action request (AR) 02011673, and the licensee implemented compensatory measures in the form of hourly fire watches. The performance deficiency (PD) was more than minor because it was associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e. fire), and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to analyze the effects of fire damage on the HPI cables in the TB could result in fire damage adversely affecting the ability to achieve and maintain safe and stable conditions. Using the guidance of IMC 0609, App. F, the finding was screened as Green because the finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event (Task 1.4.5-B). A cross cutting aspect in the area of Human Performance, Consistent Process because the licensee did not use a consistent, systematic approach to make decisions, and did not incorporate appropriate risk insights (H.13).
05000390/FIN-2016001-122016Q1Watts BarLicensee-Identified ViolationTechnical Specification 5.7.1, Procedures, requires, in part, that written procedures shall be established, implemented, and maintained covering activities described in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978; Appendix A, Section 6.v, requires procedures for Combating Emergencies and other Significant Events such as Plant Fires. Contrary to the above, the licensee provided operators inadequate procedural instructions to support fire safe shutdown. Specifically, since 2012, for certain fire scenarios, fire SSD procedures did not contain necessary steps to secure all reactor coolant pumps to prevent inadvertent RCS depressurization due to spurious opening of a pressurizer spray valve. Additionally, since initial plant licensing, for certain fire scenarios, fire SSD procedures did not contain necessary steps to isolate the normal charging line to prevent inadvertent RCS depressurization due to spurious opening of an auxiliary pressurizer spray valve. This violation is of very low safety significance (Green). This issue was determined to be of very low safety significance based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase II Quantitative Screening Approach. A bounding risk assessment performed by a regional SRA reviewed the licensee and inspector risk evaluations and confirmed the CDF risk increase due to this condition was less than 1E-6, and therefore Green. This violation was documented in the licensees corrective action program as CRs 954895 and 954957.
05000390/FIN-2016001-112016Q1Watts BarLicensee-Identified ViolationWatts Bar Operating License Condition 2.F requires that the licensee shall implement and maintain in effect all provisions of the approved fire protection program, as described in the Fire Protection Report for Watts Bar Unit 1, as approved in Supplements 18 and 19 of the SER (NUREG-0847). Fire Protection Report, Part V, Section 2.1, Safe Shutdown Procedures states, in part, the fire safe shutdown procedures contained in AOI-30.2 were developed based on calculations WBN-OSG4-031, Equipment Required for Safe Shutdown per 10 CFR 50 Appendix R, and WBN-OSG4-165, Manual Actions Required for Safe Shutdown Following a Fire. Calculation WBN-OSG4-165 is contained within drawing 1-45A897-1, Manual Actions Required for Safe Shutdown Following a Fire to 10 CFR 50 Appendix R. Contrary to the above, since initial plant licensing, the licensee failed to perform an adequate calculation to support fire safe shutdown procedure AOI-30.2. Specifically, for certain fire scenarios, the licensee failed to identify all equipment required to ensure availability of the TDAFW pump; and, for certain fire scenarios, the licensee established a non-conservative time requirement to mitigate spurious opening of a pressurizer PORV to prevent an undesired safety injection. This violation is of very low safety significance (Green). This issue was determined to be of very low safety significance based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase II Quantitative Screening Approach. A bounding risk assessment performed by a regional SRA reviewed the licensee and inspector risk evaluations and confirmed the CDF risk increase due to this condition was less than 1E-6, and therefore Green. This violation was documented in the licensees corrective action program as CRs 946764 and 999926.
05000390/FIN-2016001-102016Q1Watts BarFailure to Maintain an Adequate Surveillance Procedure for Emergency Core Cooling System VentingThe inspectors identified an apparent violation of TS 5.7.1.1.a, Procedures, for the licensees failure to maintain procedure 1-SI-63-10.1-A, ECCS Discharge Pipes Venting Train A Inside Containment, Revisions 11-16, in accordance with the requirements of Regulatory Guide 1.33. Specifically, the procedure did not have provisions for quantifying accumulated gases during venting which allowed emergency core cooling system (ECCS) piping to be vented without being evaluated for potential adverse impacts on system operability. The licensee implemented manual ultrasonic testing (UT) of gas accumulation and entered this issue into their corrective action program as CR 1136359. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the potential existed for an unacceptable void affecting ECCS operability to develop prior to the next scheduled surveillance. The inspectors determined the finding could not be screened to GREEN and may require a detailed risk evaluation following a determination of whether the finding represents a loss of system and/or function. 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 that the finding had a cross-cutting aspect of Change Management in the area of Human Performance because the licensee failed to use a systematic process to implement changes to the ECCS venting procedure to ensure that Generic Letter 2008-01 commitments would continue to be met.
05000390/FIN-2016001-062016Q1Watts BarFailure to Track Applicable Technical Specification Action Statement for Charging Pump InoperabilityThe NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees failure to implement OPDP-8, Operability Determinations and LCO tracking. Specifically, the licensee failed to track the applicability of action statement B of TS LCO 3.5.3, ECCS- Shutdown, during planned testing. The licensee entered this issue into their corrective action program as CR 1134949. The licensees failure to track applicable TS LCOs, as required by Section 3.5.1 of OPDP-8 was a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, it would have had the potential to lead to a more significant safety concern in that, the failure to track an applicable TS action statement could lead to plant operations outside of TS analyzed conditions. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of a single train for greater than its TS allowed outage time nor did it represent an actual loss of function of one or more non-TS equipment for greater than 24 hours. The performance deficiency had a cross-cutting aspect of Challenge the Unknown in the area of Human Performance because licensee personnel did not appropriately stop, question, and evaluate the risks before proceeding when the 1A-A CCP oil cooler low flow alarm came in during flow testing.
05000390/FIN-2016001-052016Q1Watts BarFailure to Use Approved Procedures to Place RHR Letdown In ServiceThe NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees failure to use any approved procedures to place RHR Letdown in service. The licensee entered this issue into their corrective action program as CR 1127691. The performance deficiency was determined to be more than minor because if left uncorrected a failure to use procedures to place systems or portions of systems in service could result in equipment being operated incorrectly and that system could then become inoperable or degraded. The inspectors determined that this finding was of very low safety significance (Green) because the way that the system was placed in service did not cause any safety-related components to become inoperable nor did it 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 for greater than 24 hours. The performance deficiency had a cross-cutting aspect of safety conscious work environment (SCWE) policy in the area of Safety Conscious Work Environment because the licensee organization failed to effectively implement a policy that supports individuals rights and responsibilities to raise safety concerns, and does not tolerate harassment, intimidation, retaliation, or discrimination for doing so.
05000390/FIN-2016001-042016Q1Watts BarFailure to Place the RHR System in ECCSStandby Mode Prior to Exceeding an RCS Temperature of 212 oFThe NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees failure to place the residual heat removal (RHR) system into ECCS-Standby Mode prior to the reactor coolant system (RCS) temperature exceeding 212 oF as required by procedure 1-GO-1, Unit Startup from Cold Shutdown to Hot Standby, Revision 4. The licensee entered this issue into their corrective action program as CR 1127691. The performance deficiency was determined to be more than minor because, if left uncorrected, a failure to align a safety system under the proper plant conditions could lead to that system being inoperable or degraded. The inspectors determined that this finding was of very low safety significance (Green) because the system temperatures never rose high enough to allow the RHR pump suction header to form steam voids. The performance deficiency had a cross-cutting aspect of Avoid Complacency in the area of Human Performance because licensee personnel were complacent and failed to question the long held idea that the particular step just needed to be started prior to exceeding an RCS temperature of 212 oF.
05000390/FIN-2016001-032016Q1Watts BarFailure to Adequately Implement the Administration of Site Technical Procedures for TDAFW Pump Governor CalibrationThe NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees inadequate implementation of procedure NPG-SPP-01.2, Administration of Site Technical Procedures, Revision 8. Specifically, the licensee determined applicable acceptance criteria steps in technical procedures were not applicable (N/A) in lieu of performing a procedure change. This resulted in challenging the operability of safety-related plant equipment. The licensee entered this issue into their corrective action program as CR 1125256. The performance deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern with the use of N/A and implementation of site technical procedures. Specifically, if further adjustments outside of the acceptance criteria or additional acceptance criteria were not met, it could have resulted in the turbine-driven auxiliary feedwater pump becoming inoperable. The inspectors determined this finding to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of equipment and operability was maintained. The finding had a cross-cutting aspect of Procedure Adherence, as described in the Human Performance cross-cutting area because the licensee failed to comply with NPG-SPP-01.2.
05000390/FIN-2016001-022016Q1Watts BarInadequate Immediate Determination of Operability for the Auxiliary Control Air System Train AThe NRC identified an NCV of 10 CFR 50, Appendix B, Criterion V, Procedures, for the licensees failure to follow TVA procedure OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking, Revision 21. Specifically, the licensee failed to base an immediate determination of operability (IDO) for the auxiliary control air system on information sufficient to conclude that a reasonable expectation of operability/functionality existed. The licensee subsequently implemented compensatory measures and entered this issue into their corrective action program as CR 1129322. The performance deficiency was more than minor because it affected the equipment performance attribute of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, reasonable assurance of operability/functionality did not exist for the A train of auxiliary control air from January 13, 2016, until January 14, 2016, and it therefore should have been declared inoperable/nonfunctional. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of a single train for greater than its TS allowed outage time. This finding had a cross-cutting aspect in the area of Human Performance, conservative bias, because the licensee failed to make the conservative decisions. Specifically, the licensee reinstalled a degraded valve in the auxiliary control air system without fully understanding the failure mechanism or its impact on system operability/functionality.
05000390/FIN-2015004-042015Q4Watts BarShield Building Operability RequirementsThe inspectors identified an unresolved item (URI) associated with the requirements of Watts Bar Unit 1 technical specification (TS) 3.6.15, Shield Building. Additional inspection is required to determine if the requirements of 3.6.15.B applied during a specific testing alignment. On September 10, 2015, the licensee conducted 0-SI-65-6-A, Emergency Gas Treatment System (EGTS) Train A 10-Hour Operation. During the 10-hour time period of the test when the EGTS was in service, the auxiliary gas building treatment system was also in service for a Unit 2 construction test. This unique ventilation combination is not normally experienced during the 0-SI-65-6-A surveillance. As a result, shield building annulus differential pressure fell below the limit established by TS surveillance requirement (TSSR) 3.6.15.1 limits for the entire duration of the 10-hr EGTS surveillance. TS limiting condition for operation (LCO) 3.6.15.B requires annulus pressure be restored when it is outside of limits with a required completion time of 8-hrs. The licensee considered the note associated with TS LCO 3.6.15.B, which states that the annulus pressure requirement is not applicable during ventilating operations, required annulus entries, or auxiliary building isolations not exceeding one hour in duration. The licensee considered the alignment they were in at the time to be ventilating operations and thus the requirements of TS LCO 3.6.15.B did not apply. The licensee further considered that the note, as written, allowed grace from the annulus pressure requirement for ventilating operations for an unlimited amount of time. The inspectors were concerned about a possible allowance in the TS to have grace from annulus pressure requirements for longer than the allowed LCO required action completion time. Furthermore, a basis for the note and what can be considered ventilating operations was not immediately apparent. Because more information is necessary to evaluate the proper applicability of TS LCO 3.6.15.B and the associated note, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to determine if a TS compliance issue exists. This is identified as URI 0500390/2015004-04, Shield Building Operability Requirements.
05000338/FIN-2015008-012015Q4North AnnaInadequate Procedural Guidance for Implementing Alternative Shutdown for a Fire in the Unit 2 Quench Spray Pump HouseThe inspectors identified a Green non-cited violation (NCV) of Technical Specification 5.4.1.a, for the licensees failure to provide adequate procedural guidance for implementation of the alternative shutdown capability in the event of a fire in the quench spray pump house. In particular, the fire safe shutdown procedure did not include actions to locally fail open the Unit 2 turbine-driven auxiliary feedwater (TDAFW) pump steam admission valves to allow operation of the TDAFW pump in the event the motor driven auxiliary feedwater pumps (MDAFW) were adversely affected by fire damage. The licensee entered this issue in their corrective action program as CR 1017083 and established compensatory actions until the Unit 1 and 2 procedures were revised. The sites failure to maintain adequate procedural guidance to operate the Unit 2 TDAFW pump for a fire in the quench spray pump house was determined to be a performance deficiency. This performance deficiency was more than minor because it was associated with the procedure quality attribute of the reactor safety mitigating systems cornerstone and it affected the cornerstone objective of protection against external events (i.e., fire). The inadequate procedural guidance affected the fire protection defense-in-depth element involving safe shutdown of the reactor. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) at Task 1.3.1, Question A, based upon observations that there were no credible fire scenarios which would likely result in simultaneous fire damage to the cables for the Unit 2 TDAFW pump and both Unit 2 MDAFW pumps. No cross-cutting aspect was identified because the issue was determined to not reflect current licensee performance.
05000338/FIN-2015008-032015Q4North AnnaFailure to Ensure that the Turbine-driven Auxiliary Feed Water Pump had the Capability to Provide Sufficient Flow Such that Residual Heat Removal Entry Conditions Could Be Achieved during Fire EventThe inspectors identified a Green non-cited violation (NCV) of North Anna Power Station, Units No.1 and No. 2, Renewed Facility Operating License, Conditions 2.D, Fire Protection, for the licensees failure to ensure that the turbinedriven auxiliary feed water (AFW) pump had the capability to provide sufficient flow such that residual heat removal (RHR) entry conditions could be achieved during fire events. The licensee entered this issue in their corrective action program as CR 1017291 with an action to re-evaluate the capability of the TDAFW pumps to achieve RHR entry conditions. The sites failure to provide reasonable assurance that the turbine-driven AFW pump had the capability to provide sufficient flow such that RHR entry conditions could be met was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control attribute of the reactor safety mitigating systems cornerstone and it affected the cornerstone objective of protection against external events (i.e., fire). The performance deficiency adversely affected the sites capability to achieve cold shutdown conditions in 72 hours for a fire event. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) at Task 1.3.1, Question A because the issue was associated with achieving cold shutdown conditions. The inspectors determined that the performance deficiency had a cross-cutting aspect of Teamwork in the Human Performance area (H.4).
05000324/FIN-2015004-012015Q4BrunswickInadequate Procedure for the 2C RHRSW Booster Pump Motor BearingsA self-revealing Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure of the licensee to have an adequate procedure for the 2C residual heat removal service water (RHRSW) pump motor bearing maintenance. Specifically, licensee procedure 0CM-M503, Maintenance Instructions for the RHRSW Booster Pump Motors, did not contain information to ensure proper sealing of the 2C RHRSW motor bearings. This finding resulted in a violation of technical specification (TS) 3.0.4, Limiting Condition for Operation (LCO) Applicability, and TS 3.7.1, RHRSW System. As immediate corrective actions, the licensee applied sealant to the motor bearings. Additionally, the licensee revised procedure 0CM-M503 and added a detailed location for applying the sealant to the RHRSW pump motors. The licensee entered this issue into the Corrective Action Program (CAP) as nuclear condition report (NCR) 742643. The inspectors determined the licensees failure to have an adequate procedure for the 2C RHRSW pump motor bearing maintenance was a performance deficiency. The finding was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate procedure resulted in the inoperability of the Loop A RHRSW subsystem, and the loss of safety function while the Loop B RHRSW subsystem was out for maintenance. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding screened to a more detailed risk evaluation, since the finding represented a loss of system and/or function. The regional Senior Reactor Analyst performed a detail risk review of the finding. The at-power model was conservatively used to bound the risk that would happen at the proposed time of failure, which was many days after shutdown due to the time it takes for the oil leak to cause potential bearing failure. Since the licensee had procedures for running the service water (SW) system without the RHRSW pumps energized, and the decay heat loads at the time of failure would be low, a failure rate of only 0.1 for the loss of function was assumed. This was also conservative, since the adverse conditions that would have prevented refill of the oil were LOCA assumptions, and LOCA sequences did not contribute greatly to the risk in the model. The at-power models solution was more than an order of magnitude below the Green/White threshold for the SDP. Therefore, the finding was determined to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance associated with the challenge the unknown attribute because the licensee did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding. Specifically, the licensee continued through the 2010 and 2013 2C RHRSW pump maintenance outages, even when the bearings were found without sealant. Additionally, the licensee did not question the procedurally required location for the sealant.
05000390/FIN-2015004-022015Q4Watts BarAFWST Permanent Plant ModificationThe inspectors identified an unresolved item (URI) associated with the 50.59 screening performed for the installation of the auxiliary feedwater storage tank (AFWST). Additional inspection is required to determine if the plant modification which installed the tank would have required NRC permission in the form of a license amendment prior to the change. The AFWST is a 500,000 gallon source of clean water for the auxiliary feedwater (AFW) pumps. It was installed as part of the licensees post-Fukushima (FLEX) modifications to meet the mitigating strategies order (EA-12-049). The new tank was needed because the licensee determined they could not credit their existing condensate storage tanks (CSTs) for FLEX strategies due to seismic requirements necessary to survive the extended loss of AC power (ELAP) event. The AFWST was connected to the existing condensate system in the AFW supply piping upstream from the AFW pumps and downstream from the CSTs. The modification was evaluated in two separate DCNs, each with its own 50.59 applicability screening. DCN 60060 evaluated the installation of the tank and DCN 61422 evaluated the piping connections to the condensate system. The piping connections included new check valves in the CST piping to prevent AFWST inventory loss in the event the CSTs are damaged in the ELAP event. There were also two air-operated supply valves on AFWST outlet piping which automatically open on low pressure in the downstream condensate piping and also fail open on a loss of power or air. Inspectors noted a number of deficiencies in the 50.59 screening for DCN 61422. Inspectors determined that several potentially adverse impacts were introduced by the modification and were not adequately considered in the 50.59 screening. The licensee re-performed the screening and concluded that the modification would require a 50.59 evaluation due to adverse impacts brought up by the inspectors. Because more information is necessary to properly evaluate the 50.59 evaluation that was completed late in the quarter, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to determine if prior NRC approval was required for the installation of the AFWST. This is identified as URI 05000390/2015004-02, AFWST Permanent Plant Modification.
05000338/FIN-2015008-022015Q4North AnnaECST Level Indication/Setpoints and Associated Operator Action that Ensures the Auxiliary Feedwater Pumps have an Adequate Suction SourceThe inspectors identified an Unresolved Item (URI) associated with the emergency condensate storage tank (ECST) level indication/setpoints and associated operator actions that ensures the auxiliary feedwater (AFW) pumps have an adequate suction source. UFSAR, Section 7.4-2, states that the emergency condensate storage tank (ECST) was designed to supply the initial eight hours of water to the auxiliary feedwater (AFW) pumps during licensing bases events. The inspectors noted that the licensee utilized operator actions to reduce AFW flow during the initial stages of events which is typically accomplished in order to prevent over cooling of the primary RCS during events where maximum AFW is not required. For events where maximum AFW may be required, the licensee developed calculations to ensure that an adequate water supply was maintained. The licensees Calculation ME-0584, Maximum AFW Pump Flow and NPSH Analysis, (dated 11/04/1999) determined that AFW flow reduction was required within the initial 30 minutes of an event to ensure that the pumps had sufficient net positive suction head. The inspectors determined, in some cases, that operator actions would be required prior to the receipt of the ECST tank level alarm that was described UFSAR Section 10.4.3.3, which stated that the ECST had redundant ECST safety-level alarms (1/2-CN-LI-200A and -200B) to alert operators that sufficient inventory remained for 20 minutes of pump operation at the highest-volume flow rates. Additionally, the inspectors noted that a Virginia Electric Power Company letter, dated December 22, 1999, stated that Technical Specifications ensure that the level maintained in the ECST is adequate to mitigate the accident without operator action during a design basis accident. Therefore, the indication of ECST level is not required as a Type A variable. Indication of ECST level remains a Type D, Category 1 variable... This issue is unresolved pending the NRCs review of applicable licensing requirements, calculations, and operating procedures to assess the adequacy of the ECST level indication/setpoints and associated operator actions to ensure that the AFW pumps have an adequate suction source as described by their licensing design basis. This issue is identified as URI 05000338 & 05000339/2015008-02, ECST Level Indication/Setpoints and Associated Operator Action that Ensures the Auxiliary Feedwater Pumps have an Adequate Suction Source.
05000390/FIN-2015010-012015Q3Watts Bar420 Minute Operator Manual Action to Provide Source Range Monitoring CapabilityThe inspectors identified an unresolved item associated with a fire protection safe shutdown OMA that established a time requirement of 420 minutes to provide a functional source range monitor. The inspection team noted that procedure 1-AOI-30.2 C36, Fire Safe Shutdown Room 737-A1A, Rev. 0005 included a 420 minute operator manual action (OMA) to establish a functional source range monitor. The OMA was listed as OMA 649 in Calculation EDQ00099920090016, Appendix R Unit 1 & 2 Manual Action, Rev. 4. The inspectors also noted the following: - Westinghouse Owners Group letter, WOG-05-36 (dated 01/28/2005), Section 6.2, Long Term Cold Shutdown Capability, stated that typical instrumentation to achieve a shutdown condition during Appendix R event included the source range monitors. - Technical Specification 3.3.1.L required an operable source range neutron flux channel in Modes 3, 4, and 5; and stipulated that positive reactivity additions (such as plant cooldown) be suspended when the instrument was inoperable. - Procedure 1-AOI-30.2, Fire Safe Shutdown, Rev. 0005, Step 5.3.15, stated that at least one channel of nuclear instrumentation indication must be available to monitor shutdown neutron population. - Procedure 1-AOI-30.2 C36 included a note that stated that RCS cooldown should not be initiated until source range monitoring capability can be assured. - Procedure 1-AOI-30.2 C36 directed operators to depressurize and cooldown an action that was typically required at 60 75 minutes. The 420 minute OMA would allow shutdown and subsequent cooldown of the reactor plant without operators having the ability to monitor neutron population. The licensee contended that OMA 649 was part of the sites licensing bases and thus the capability to monitor source range was not required until 420 minutes. The inspection team determined that this issue required additional inspection because the licensee did not provide an alternative method to monitor neuron population and did not provide adequate restrictions to prevent cooldown activities until monitoring capability was restored. Additionally, the OMA conflicted with the technical specification requirements for source range availability. The issue is unresolved pending additional review to determine if a performance deficiency exists. Required actions to resolve this issue include a detailed review of applicable docketed licensing bases correspondence; consultation with NRRs fire protection and technical specification branches; and an assessment to determine the applicable fire areas if the issue is to be determined to be a more-than-minor performance deficiency. This issue will be tracked as URI 05000290/2015010-01, 420 Minute Operator Manual Action to Provide Source Range Monitoring Capability.
05000280/FIN-2015008-042015Q2SurryMultiple Design Deficiencies in the Fire Protection ProgramThe inspectors identified a Green NCV of Surrys Operating License, Condition 3.I, Fire Protection, for design control deficiencies in the fire protection program. The licensee entered this issue into their corrective action program as condition report CRs 581390. The licensees failure to adequately implement the design control requirements in the fire protection program as required by Topical Report, DOM-QA-1, Dominion Nuclear Facility Quality Assurance Program Description, Section 3.2, Design Control Program was a performance deficiency. The finding was more than minor because it was associated with the design control attribute and affected the Mitigating Systems cornerstone. Specifically, design control deficiencies resulted in a lack of assurance that the design control requirements were being adequately implemented within the fire protection program. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green) because it finding did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. No cross cutting aspect was assigned because the performance deficiency did not occur within the last three years.
05000324/FIN-2015002-042015Q2Brunswick2C Residual Heat Removal Service Water (RHRSW) Pump Oil LeakThe inspectors opened a URI to review the licensees evaluation of the motor oil leak on the 2C RHRSW pump and determine if there is a performance deficiency. On April 8, 2015, the licensee identified an oil leak on the motor for the 2C RHRSW pump in excess of the amount that would be acceptable for the pump to meet the 30-day mission time, and the pump was declared inoperable. The licensees immediate corrective actions were to apply sealant to the mechanical joints of the bearing housings. The licensee entered this issue in the CAP as NCR 742643. This issue is being tracked as a URI: URI 05000324/2015002-04, 2C Residual Heat Removal Service Water Pump Oil Leak.
05000280/FIN-2015008-012015Q2SurryFailure to Ensure a Functional Alternate Shutdown System Alignment during Appendix R Fire Events EventsThe inspectors identified a Green non-cited violation (NCV) of Surrys Operating License, Condition 3.I, Fire Protection, for the licensees failure to ensure a functional alternate safe shutdown flow path during an Appendix R fire. The licensee entered this issue into their corrective action program as condition report (CR) 580928. The licensees failure to ensure a functional alternate shutdown system alignment during an Appendix R fire event was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone. Specifically, Surry failed to implement appropriate corrective actions to mitigate the spurious closure and subsequent damage of more than one motor operated valve as identified in an engineering evaluation. The failure to re-open credited Appendix R MOV(s) would result in the loss of secondary heat removal and/or RCS make-up capability during Appendix R fire events. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green). A Region II senior risk analyst performed a bounding phase 3 analysis that determined the finding represented an increase in core damage frequency of < 1 E-6 /year. No cross cutting aspect was assigned because the performance deficiency did not occur within the last three years.
05000413/FIN-2015012-022015Q2CatawbaFire Protection Program Change did not meet CNS License Condition Requirement 2.C.5 for Units 1 and 2The NRC identified a non-cited Severity Level IV violation of the Unit 1 and 2 CNS Facility Operating License, Condition 2.C.5, for the failure to implement and maintain in effect all provisions of the approved fire protection program (FPP). Specifically, the licensee made a change to the approved FPP which involved the de-rating of a credited three hour fire barrier between the control room and the cable spreading room(s) to only a pressure and smoke barrier. The licensee entered the issue in its corrective action program as AR 01932211 and it was added to existing fire watches for the area. The failure to comply with the CNS Operating License Condition 2.C.5 for a change to the approved FPP involving the de-rating of a credited three hour fire barrier between the control room and the cable spreading room(s) was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external events (i.e. Fire.) The performance deficiency negatively affected the cornerstone objective in that the change to the FPP had the potential to adversely affect the availability of the control room to achieve and maintain stable plant conditions due to the increased likelihood of control room abandonment in the event of a fire in the cable spreading rooms. The licensees failure to submit the FPP change to the NRC was determined to impede the regulatory process because the FPP change required NRC review and approval prior to implementation. The finding was screened as Green because based upon inspection of the affected barriers, the inspectors determined that the barriers would provide a 1-hour or greater fire endurance rating. This violation was determined to be a Severity Level IV violation because the associated finding was evaluated by the SDP as having very low safety significance (i.e., Green finding). No cross cutting aspect was assigned because the finding was not indicative of current licensee performance.
05000413/FIN-2015012-012015Q2CatawbaFailure to Analyze the Spurious Operation of Control Room Area Ventilation Valves and the Adverse Impact on Control Room HabitabilityThe NRC identified an NCV of the Unit 1 and 2 Catawba Nuclear Station (CNS) Facility Operating License, Condition 2.C.5, for the failure to analyze the spurious operation of two motor operated valves (MOVs) in the control room area ventilation system (CRAVS) and the adverse impact on control room habitability. The licensee entered the issue in its correction action program as action request (AR) 01930126 and a continuous fire watch was already in place due to deficiencies identified during the sites ongoing NFPA 805 licensing activities. The failure to analyze the spurious operation of two MOVs in the CRAVS and the adverse impact on control room habitability was a performance deficiency (PD). The performance deficiency was more than minor because it was associated with the protection against external events (i.e. Fire) attribute of the Initiating Events Cornerstone and it 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. Specifically, the finding could be reasonably viewed as a precursor to a significant event based on smoke migration into the control room that could challenge control room habitability and lead to an evacuation of the control room. This PD was the result of degraded defense-in-depth features that limit the effects of a fire to one fire area. The finding was screened as Green because the reactors would be able to reach and maintain safe shutdown utilizing the standby shutdown facility. No cross cutting aspect was assigned because the finding was not indicative of current licensee performance.
05000280/FIN-2015008-032015Q2SurryFailure to Perform Required 50.59 Evaluations and Failure to Update the UFSAR for Plant Changes Associated with RCP Seal Cooling During Fire EventsThe inspectors identified a Green NCV of 10 CFR 50.59 and 10 CFR 50.71(e) for the licensees failure to perform 50.59 evaluations; and failure to update the UFSAR for plant changes associated with reactor coolant pump (RCP) seal cooling during fire events. The licensee entered this issue into their corrective action program as condition report CRs 5813388. The licensees revision of fire safe shut down procedures; and the installation of a different reactor coolant pump seal package without completing the required 50.59 evaluations was a performance deficiency. Additionally, the licensees failure to update the UFSAR as required by 10 CFR 50.71(e) was a performance deficiency. The UFSAR did not adequately describe the charging cross-tie function; and did not adequately describe the fire protection programs procedural isolation of the RCP seals for the entire duration of an Appendix R event. In accordance with the Reactor Oversight Process, the performance deficiencies were more than minor because they were associated with the design control attribute of the Mitigating Systems Cornerstone. The performance deficiencies were also assessed using traditional enforcement because the NRCs ability to perform its regulatory function such as, license amendment reviews and inspections was affected. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green) because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. No cross cutting aspect was assigned because these performance deficiencies did not occur within the last three years.
05000280/FIN-2015008-022015Q2SurryFailure to Implement In-service Testing and Inservice Inspections for Charging Cross-tie ComponentsThe inspectors identified a Green NCV of 10 CFR 50.55(a) for the licensees failure to implement in-service testing (IST) and in-service inspections (ISI) for charging cross-tie components. The licensee entered this issue into their corrective action program as CRs 581385 and 581386. The licensee failed to scope the charging cross-tie manual isolation valves and piping into the ISI and IST programs. This was a performance deficiency that resulted in the subsequent failure to perform ISI and IST activities required by the ASME OM Code-2004 and 10 CFR 50.55a(f) and (g). The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone. Specifically, the sites failure to perform required inspections and testing for charging cross-tie components, since 1989, resulted in a lack of reasonable assurance that the charging cross-tie function could perform its required function. The finding was screened in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, and determined to be of low safety significance (Green) because it did not affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event. No cross cutting aspect was assigned because the performance deficiency did not occur within the last three years.
05000400/FIN-2014008-012014Q2HarrisFailure to Identify and Evaluate All Targets Within the Zone of Influence of Ignition SourcesAn NRC-identified non-cited violation of 10 CFR 50.48 (c) and National Fire Protection Association Standard (NFPA) 805 Section 2.4.3.2 was identified for the licensees failure to address in the Fire Probabilistic Risk Assessment (Fire PRA) the risk contribution associated with all potentially risk significant fire scenarios for a given fire compartment/fire area. The licensee did not identify and evaluate all targets that were within the zone of influence (ZOI) of ignition sources for selected fire scenarios which could potentially contribute to the risk for the fire scenarios. The licensee entered the issue in the corrective action program as Nuclear Condition Reports 682633 and 685355 and established an hourly roving fire watch as compensatory measures. The licensees failure to comply with the requirements of 10 CFR 50.48(c) and NFPA 805 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute of protection against external factors (i.e., fire) and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The missed targets in the ZOI for the selected fire scenarios had the potential to impact the ability to achieve safe and stable conditions. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, which determined that an IMC 0609, Appendix F, Fire Protection Significance Determination Process, review was required as the finding affected post-fire SSD. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the finding was screened as Green in step 1.6.1 Screen by Licensee PRA-Based Safety Evaluation. An SDP Phase 3 analysis was performed to document the review of the risk determination of the missed ignition source-target interactions using the licensees Fire PRA model. A senior reactor analyst performed the Phase 3 SDP analysis in accordance with the guidance in IMC 0609 Appendix F and NUREG/CR-6850 Revisions 0 and 1. The evaluation determined that the missed ignition source-target interactions resulted in a CDF increase of 5.91E- 8/year, a Green finding of very low safety significance. There was no cross cutting aspect assigned to this finding because it was not indicative of current licensee performance since the original ignition source and target walkdowns were performed in 2006 and 2007.
05000327/FIN-2014007-042014Q1SequoyahFailure to Maintain Necessary Materials and Procedures for Cold Shutdown RepairsAn NRC-identified Green non-cited violation of Sequoyah Operating License Conditions 2.C.(16) and 2.C.(13), for Units 1 and 2 respectively, was identified for the licensee's failure to maintain necessary materials and procedures for cold shutdown repairs, as required by the approved fire protection program. The licensee entered this issue into the corrective action program as Problem Evaluation Reports 845931, 847420, 847428, 847449, and 847462. The licensees failure to provide adequate guidance for all repairs listed in the Appendix R casualty procedure and failure to maintain the required repair parts for the same procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events (fire) to prevent undesirable consequences. Inadequate procedural guidance and the lack of required materials could adversely affect the licensees capability to achieve and maintain cold shutdown conditions. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, which determined that an IMC 0609, Appendix F, Fire Protection Significance Determination Process, review was required as the finding affected fire protection defense-in-depth strategies involving post-fire safe shutdown. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) at Task 1.3.1, because it was determined that the reactor was able to reach and maintain a hot safe shutdown condition. The cause of this finding was determined to have a cross-cutting aspect of Teamwork (H4) in the Human Performance cross-cutting area because the licensee failed to assure that individuals and work groups communicated and coordinated their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, the coordination between operations department procedure writers, maintenance department procedure writers, and fire operations department personnel was inadequate to ensure the adequacy of cold shutdown repair procedures and the availability of required materials.
05000327/FIN-2014007-052014Q1SequoyahFailure to Perform the Required Reviews when Adding Fire Watches to the Fire Protection ProgramAn NRC-identified Green non-cited violation of Sequoyah Operating License Conditions 2.C.(16) and 2.C.(13), for Units 1 and 2 respectively, was identified for the licensee's failure to perform the required reviews when adding fire watches to the fire protection program. The licensee entered the issue into their corrective action program as Problem Evaluation Report 845593. The licensees failure to perform the required evaluation and review prior to revising the fire hazards analysis was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events (fire) to prevent undesirable consequences. Specifically, the sole use of fire watches as a mitigation measure for the unavailability of the credited pressurizer power operated relief valve would adversely affect the capability to achieve and maintain safe shutdown during a fire event. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, which determined that an IMC 0609, Appendix F, Fire Protection Significance Determination Process, review was required as the finding affected fire protection defense-in-depth strategies involving post-fire SSD. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the issue screened as having very low safety significance (Green) at Task 1.5.3 because the change in core damage frequency (delta CDF) was less than 1E-6 (i.e., delta CDF calculated to be 6.6E-7). The cause of this finding was determined to have a crosscutting aspect of Evaluation (P.2) in the Problem Identification and Resolution crosscutting area, because the licensee did not thoroughly evaluate the issue to ensure that resolutions addressed causes commensurate with their safety significance. Specifically, the establishment of effective corrective actions was adversely affected by the failure to perform an evaluation prior to revising the fire hazards analysis.
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