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05000390/FIN-2015004-012015Q4Watts BarFailure to Perform ISI General Visual Examination of Containment Moisture Barrier Associated with Containment Liner Leak-chase Test Connection Threaded Pipe PlugsThe inspectors identified a Green NCV of Title 10 of the10 CFR Part 50.55a, Codes and Standards, involving the licensees failure to properly apply Subsection IWE of American Society of Mechanical Engineers, Section XI, for conducting general visual examinations of the metal-to-metal pipe plugs of the leak-chase channel test connections, installed inside the access box, that provide a moisture barrier to the basemat containment liner seam welds. Following the inspectors identification of this issue, the licensee initiated actions to conduct the required inservice inspection (ISI) general visual examinations. Inspection of the access boxes and leak-chase channels revealed the presence of standing water as well as general corrosion in both locations. The licensee took actions to remove the water and evaluate the condition of the applicable structure, system, and components to verify that containment integrity had been maintained, and would continue to be maintained through the expected life of the plant. The licensee updated the ISI plan such that the required inspections will be performed in the future. The inspectors determined that the licensee had taken adequate immediate corrective actions to address the deficiencies identified, and to ensure the leak-tight integrity of the containment. The issue was entered into the licensees corrective action program (CAP) as Condition Report 1092415. This performance deficiency was of more than minor significance because the failure to conduct required visual examinations and identify the degraded moisture barriers which allowed the intrusion of water into the liner leak-chase channel, if left uncorrected, would have resulted in more significant corrosion degradation of the containment liner or associated liner welds. The finding was associated with the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, visual examinations of the containment metal liner provide assurance that the liner remains capable of performing its intended safety function. The inspectors used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment.
05000390/FIN-2015004-032015Q4Watts BarFailure to Comply with Source Range Neutron Flux Channel Technical Specification RequiprementsA self-revealing non-cited violation (NCV) of Technical Specification (TS) 3.3.1, was identified for the licensees failure to take the actions of Table 3.3.1-1, Function 5, action J.1 to immediately open the reactor trip breakers (RTBs) when two source range neutron flux channels were inoperable with the RTB closed and the rod control system capable of rod withdrawal. Specifically, the licensee failed to identify both required channels of the source range trip function were bypassed and proceeded to withdraw control rods for testing and reactor startup. The performance deficiency was more than minor because it affected the configuration control attribute of the mitigating 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 source range level trip switches were left in bypass, outside of their required configuration, thereby removing a trip function that is required by TS during rod withdrawal. The inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a mismanagement of reactivity by the operators. This finding had a cross-cutting aspect in the area of Human Performance, avoid complacency, because the licensee failed to recognize and plan for the possibility of mistakes and latent issues or use appropriate error reduction tools.
05000390/FIN-2015004-062015Q4Watts BarFailure to Identify a Condition Adverse to Quality for Unacceptable Preconditioning of the 1A-A Charging Pump Discharge Check ValveThe NRC identified a NCV of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify a condition adverse to quality. Specifically, the licensee unacceptably preconditioned the 1A-A charging pump discharge check valve 1-CKV-62-525 and failed to identify this as a condition adverse to quality or take appropriate corrective action. The inspectors determined that the performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, unacceptable preconditioning could mask the actual as-found conditions and result in the loss of degradation trending information of component performance. The inspectors determined the finding to be of very low safety significance (Green) because the finding did not result in the loss of operability of 1-CKV-62-525. This finding had a cross-cutting aspect in the area of Human Performance, work management, 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 licensees work management process was not able to prevent the unacceptable preconditioning of the 1A-A discharge check valve even after it was identified as a possibility prior to the planned maintenance.
05000390/FIN-2015004-072015Q4Watts BarLicensee-Identified ViolationWatts Bar Nuclear Plant TS 3.6.12 states that the ice condenser inlet doors, intermediate deck doors, and top deck doors shall be operable and closed. TS 3.6.12 Condition B requires that maximum ice bed temperature is verified to be less than 27 degrees F once per four hours (Action B1) when one or more doors is inoperable. Contrary to the above, four intermediate deck doors were inoperable from September 8, 2015 until September 17, 2015 and required action B1 of TS 3.6.12 Condition B was not performed. WBN maintenance personnel erected scaffolding on September 8, 2015 which blocked four intermediate deck doors in the Unit 1 upper ice condenser, which made the doors inoperable since the scaffolding would have prevented them from opening. The TS implications of the scaffold were not immediately recognized and therefore the required TS action B1 was not performed. The licensee identified this condition on September 16, 2015 and took immediate actions to enter TS LCO 3.6.12, Condition B, requiring that maximum ice bed temperature is verified to be less than 27 degrees F once per four hours (Action B1) and to restore the doors to operable status in 14 days (Action B2). The scaffold was removed on September 17, 2015; therefore, the 14-day completion time of TS 3.6.12 was not exceeded. A review of ice bed temperatures between September 8, 2015 and September 17, 2015 showed that ice bed temperatures never exceeded 27 degrees F as required by TS 3.6.12 Action B1. Using IMC 0609, Appendix A, Exhibit 2 (Mitigating Systems); this finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of function of at least a single train of equipment for greater than its technical specification allowed outage time. This violation was entered into the WBN CAP under CR 1082469.
05000390/FIN-2015004-052015Q4Watts BarCore Barrel Lift Error Resulted in Unintended high Dose RatesA self-revealing NCV of TS 5.7.1, Procedures, Programs and Manuals, was identified when the unit one core barrel (CB) was raised above the height limit specified in licensee procedure1-MI-68.003, Removal and Replacement of the Unit 1 Reactor Vessel Lower Internals, Revision 0003. Specifically, step 6.11(20) states in part, ...slowly raise the lower internals package UNTIL the lower internals is at or above EL. 75910 as indicated by the break of the laser indicator on the wall target. On October 5, 2015, while moving the CB from the storage stand to the reactor vessel, the CB was inadvertently lifted approximately three feet higher than the 75910 elevation and required radiation protection (RP) intervention to stop the lift when dose rates in and around containment exceeded anticipated levels. The licensee entered this issue into the CAP as CR 1090220. Corrective actions included stand-downs with each crew to review expectations for critical steps, increased field oversight, and revision of the lift procedure to clarify the steps regarding use of the laser indicator. This finding was determined to be greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance, Program and Processes (procedures for monitoring and RP controls) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The finding was not related to As Low As Reasonably Achievable planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Human Performance, Work Management (H.5) because distractions at the work location contributed to the failure to recognize that the CB had been raised above the procedural limit.
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.
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.
05000390/FIN-2015002-052015Q2Watts BarReview of 10 CFR 50.59 Evaluation for the Emergency Diesel Generator Heat ExchangerThe inspectors identified an unresolved item (URI) regarding the licensees 10 CFR 50.59 evaluation for a modification to the operational configuration of the inlet motor operated valves (MOVs) for the EDG Heat Exchanger. Additional inspection would be required to determine if the licensees 10 CFR 50.59 evaluation properly addressed whether the modification resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structures, systems, or components (SSCs) important to safety previously evaluated in the UFSAR. Watts Bar has four EDGs that are each cooled by two heat exchangers supplied by the ERCW system. Prior to the modification, flow through the heat exchangers was continuous due to the inlet MOVs (1-FCV-067-0066-A, 2-FCV-067- 0066-A, 1-FCV-067-0067-B and 22FCV-067-0067-B) being locked open. In order to ensure sufficient flow is available to components served by ERCW during dual-unit operations, the licensee modified the position of these MOVs from normally open with power removed, to normally closed with breakers closed. This resulted in the EDG heat exchangers being isolated during normal operation from the ERCW system. Flow, however, would be restored by the MOVs active function to open upon receipt of a signal from the EDG speed switch, should the EDGs startup. The inspectors reviewed the results of the licensees 10 CFR 50.59 evaluation related to the impact of the modification on the failure probability of the EDG. The inspectors concluded that additional information and review was necessary to determine whether the modification resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component important to safety previously evaluated in the UFSAR. Particularly, the inspectors needed additional information on the specific inputs, assumptions, and evaluation methodology used to determine the increase in EDG failure probability. This issue was identified as URI 05000390/2015002-05, Review of 10 CFR 50.59 Evaluation for the EDG Heat Exchanger.
05000390/FIN-2014003-032014Q2Watts BarFailure to Comply with Design Drawing Results in a Reactor TripAn NRC-identified finding was documented by the inspectors for the licensees failure to comply with a design drawing during a modification resulting in a trip of Unit 1 reactor. The inspectors determined that the licensees failure to properly implement Design Change Notice (DCN) 52295, complete bus differential wiring for main bus 2, as required by NPG-SPP-09.3, Revision 17, Plant Modifications and Engineering Change Control, was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the objective of the Initiating Events cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correctly translate design drawings to implementing work order 08-816022-006 resulted in Unit 1 experiencing a 100 percent load rejection and reactor trip. Using the screening worksheet of IMC 0609, Appendix A, Exhibit 1 - Initiating Events Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the resulting transient was within the design basis for Unit 1 and all plant systems functioned as required to place the unit in a stable, hot standby condition. The cause of the finding was directly related to the aspect of work management in the Human Performance cross-cutting area because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority.
05000390/FIN-2014003-012014Q2Watts BarFailure to Identify a Condition Adverse to QualityAn NRC-identified NCV of 10 Code of Federal Regulations (CFR) 50 Appendix B, Criterion XVI, Corrective Action, was documented for the licensees failure to adequately identify a condition adverse to quality associated with the installation of 480 volt breaker 0-BKR-548-0021-S with non-conforming parts which was in service in safety-related 480 volt shutdown board 1B1. Immediate corrective action was to replace the non-conforming breaker. The inspectors determined that the licensees failure to adequately identify a condition adverse to quality associated with the installation of non-conforming parts as required by 10 CFR 50 Appendix B, Criterion XVI, was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify the condition adverse to quality led to an additional six months that this non-conforming condition existed thus reducing the licensees ability to ensure the reliability and capability of plant safety systems. Using the screening worksheet of IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the deficiency only affected the qualification of the breaker. The cause of the finding was directly related to the aspect of identification in the Problem Identification and Resolution cross-cutting area because the licensee did not identify this issue completely, accurately, and in a timely manner in accordance with the program.
05000390/FIN-2014003-022014Q2Watts BarFailure to Identify a Condition Adverse to QualityA NRC-identified non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was documented for the licensees failure to adequately identify a condition adverse to quality associated with the installation of relief valve 1-RFV-67- 1026D, Upper Containment Cooler 1D, an ASME Class III component. The licensee entered the issue into the corrective action program and performed an operability determination which concluded the cooler was operable. The inspectors determined that the licensees failure to adequately identify a condition adverse to quality associated with the non-conformance of relief valve 1-RFV-67-1026D, as required by 10 CFR 50 Appendix B, Criterion XVI, was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the objective of the Mitigating Systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct the condition adverse to quality in a timely manner led to an additional 4 years that this non-conforming condition existed prior to evaluation thus reducing the licensees ability to ensure the reliability and capability of plant safety systems. Using the screening worksheet of IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because there existed an additional relief valve in the IST program that could protect the piping and cooler from over pressurization with appropriate compensatory measures. The cause of the finding was directly related to the aspect of evaluation in the Problem Identification and Resolution cross-cutting area because the licensee did not adequately evaluate this issue to ensure that an adequate resolution addressed the condition commensurate with its safety significance.
05000390/FIN-2014003-042014Q2Watts BarLicensee-Identified Violation10 CFR 50 Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, states, in part, that measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. Contrary to this requirement, the licensee did not ensure that a non-conforming component was properly identified and segregated as non-conforming to prevent possible use in a safety-related application. On March 22, 2014, during receipt inspection of a refurbished 6.9 kV circuit breaker per WO 115282290, breaker maintenance personnel determined that it was not in serviceable condition and initiated service requests 862089 and 862145 to document the non-conforming conditions. On March 31, 2014, the non-conforming circuit breaker was returned to inventory with two serviceable circuit breakers and made available for issue, rather than being identified and segregated as non-conforming, as required by procedure NPG-SPP-6.11. On April 4, 2014, discussions between warehouse personnel and maintenance personnel revealed that one of three 6.9 kV circuit breakers recently placed in inventory was unusable. All three circuit breakers were placed on hold and tagged as non-conforming. Once further investigation determined which circuit breaker was non-conforming, it was removed from inventory. This low safety significant adverse condition was captured in the licensees corrective action program as PER 872906.
05000390/FIN-2014003-052014Q2Watts BarLicensee-Identified Violation10 CFR 50 Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, states, in part, that measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. Contrary to this requirement, the licensee did not ensure that a non-conforming part was properly controlled to prevent its use in a safety-related application. On January 22, 2013, non-conforming main contactors were installed in a safetyrelated 480 volt breaker which was then installed in the plant on January 25, 2013. The breaker remained in service until May 20, 2014, when it was replaced. This low safety significant adverse condition was captured in the licensees corrective action program as PER 810826. (See Section 4OA2 for additional details)
05000390/FIN-2013004-012013Q3Watts BarContribution of Potential Current Transformer Imbalance to Reactor TripThe inspectors monitored the licensee performing a trouble shooting plan following the reactor trip. The licensee attributed the cause of the reactor trip to a loose phase A connection on a digital fault recorder. The inspectors continued to monitor the root cause development following reactor startup to determine the validity of the cause and review the associated LER for closure. On June 28, 2013, an A-phase high impedance ground fault occurred on the Roane 500kV transmission line approximately 22 miles from Watts Bar Nuclear Unit 1. Concurrently, the licensee experienced a reactor trip due to the actuation of the 1A Main Bank Transformer Feeder Differential Relay 187TF. The 500kV transmission line fault was caused by a tree that fell onto the A phase of the transmission line. The tree was cut by a local land owner. Operations personnel stabilized the plant using the Auxiliary Feedwater (AFW) System and the main steam dump valves. The secondaryside steam generator (SG) atmospheric relief valves, SG power operated relief valves (PORVs) and SG safety valves were not challenged during the transient. The Reactor Coolant System (RCS) responded to the initial plant transient as expected without actuating Pressurizer PORVs or initiating Safety Injection signals. Per design, a differential relay, such as the 187TF relay, should not trip due to an event occurring outside of the relays zone of protection because the input amperage subtracts from the output amperage equaling zero so no amperage is available to trip the relay. Specifically, the 187TF relay zone of protection covers the bus network between the two main generator output breakers and the 1A main bank transformer, which is within the plants switch yard, whereas the fault was 22 miles from the plant site. Initially, the licensee tried to verify that the differential circuits current transformers (CTs) were electrically balanced over their range of operation by injecting increasing levels of test amperage into the circuit. The CTs measure the amperage of the 500KV power and feed that measurement to the 187TF relay so verifying that the input CT amperage properly subtracts from the output CT amperage would validate the CTs characteristics. Because the technicians injecting the amperage did not disable the 86 relay, all of the remaining circuit breakers connected to the X bus opened. The 86 relay detects if any breaker on the bus fails to open when called upon. The 86 relay operation had no significant effects on the shutdown of the plant, but because the technicians did not secure the circuit properly, TVA management decided to stop the verification of the CTs. Alternatively, the technicians tried to verify circuit connections by physically moving the wiring. While handling an A phase wire connected to one of the Digital Fault Recorders, a technician stated that he noticed about 1/8 inch of movement and heard a click. The click is the locking mechanism that ensures the connection remains secure; however the design of the connector ensures electrical connection over one inch of movement. Before the 86 relay tripped the remaining bus relays, the amperage injection had passed approximately 2 amps through this portion of the circuit, which would have detected a loose connection. Inspectors observed that TVA stopped assessing other electrically significant reasons that could have tripped the 187TF relay such as CT imbalances. The root cause team determined that the click heard by the technician was the cause of the relay trip even though subsequent bench testing could not support it. In response to inspector questions, the licensee hired a 3rd party consultant, which also discounted the connector as an obvious cause of the 187TF relay trip. Both the inspectors and the 3rd party consultant believe that neither of the licensees troubleshooting techniques nor their root cause analysis has adequately addressed the cause of the 187TF relay trip, which can continue to challenge the reactor protection system on subsequent high impedance ground faults outside of the plant. The licensee plans to disable the 187TF relay during the next shutdown for refueling, in the spring of 2014, in order to measure the current sensed by the relay as the main generator load is decreased for shut down. This item is identified as unresolved item (URI) 050000390/2013004-01, Contribution of Potential Current Transformer Imbalance to Reactor Trip.
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