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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000390/FIN-2018050-01 | 2018Q2 | Watts Bar | Licensee-Identified Violation | This violation of very low safety significance (Green)was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a Non-CitedViolation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Title 10 of the Code of Federal Regulations(10 CFR) Part 50 (10 CFR 50), Appendix B, Criterion III, Design Control, requires the licensee to effectively implement design control measures for piping analysis calculations* associated with the Unit 1 and Unit 2 emergency core cooling systems (ECCS).Contrary to the above, since initial operation of Unit 1 in 1996 and Unit 2 in 2016, Tennessee Valley Authority failed to ensure the proper hydraulic time history was utilized in TVAs TPIPE special purpose computer program used to determine static and dynamic linear elastic analyses for the ECCS including the effects of pipe voiding. This resulted in non-conservative voiding design acceptance criteria for the RHR and SI systems of both units. This performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to utilize proper hydraulic time history in the licensees TPIPE computer model resulted in developing non-conservative voiding acceptance criteria that was used during operation that could challenge ECCS functionality. The finding was determined to be of very low safety significance since additional analysis determined with reasonable assurance that the systems remained operable but non-conforming and would have performed their safety function.Significance/Severity Level: Green. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because the finding affected the design or qualification of mitigating systems; however, the mitigating systems maintained their operability. Corrective Action Reference:CR 1407257 |
| 05000424/FIN-2016007-01 | 2016Q4 | Vogtle | Failure to Verify Capability of EDGs under Maximum Voltage and Frequency | The NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for failure to correctly translate the appropriate permissible limits for frequency and voltage from technical specifications into the emergency diesel generators design loading calculations as required by the licensing and design bases. The violation and related issues were entered into the licensees corrective action program as condition reports 10288732 and 10293810. The licensee was evaluating corrective actions, which included determining acceptable loads at the more limiting power demands and developing procedural guidance. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the emergency diesel generators to respond to initiating events to prevent undesirable consequences. Specifically, failing to evaluate the impact from the frequency and voltage limits allowed by technical specification could result in overloading the diesel generator if operators manually loaded additional plant protection systems during an event. The team determined the finding was of very low safety significance (Green) because it was a design deficiency that did not result in a loss of emergency diesel generators operability. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance. |
| 05000424/FIN-2016007-02 | 2016Q4 | Vogtle | Failure To Ensure Adequate Unit 1 Emergency Diesel Generator Surveillance Acceptance Criteria | The NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have adequate instructions and acceptance criteria to confirm the emergency diesel generators capability to reject the largest single load without exceeding predetermined frequency and voltage while maintaining a specified margin to the overspeed trip. The violation was entered into the licensees corrective action program as condition report 10294395. An immediate determination of operability was performed and concluded that the Emergency Diesel Generators were operable but degraded nonconforming. The licensee was evaluating corrective actions, which may include a final determination of the most severe single largest load and re-performing the surveillance tests. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, without adequate acceptance criteria in surveillance procedure SR 3.8.1.8, the procedure could not ensure availability, reliability, and capability of the EDG under the most severe power demand characteristics for electric power used by components. The team determined the finding to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of technical specification or non-technical specification equipment. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance. |
| 05000424/FIN-2016007-03 | 2016Q4 | Vogtle | Failure to Meet Isolation Requirements When Incorporating Non- Class 1E Components into Class 1E electrical Circuits | The NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III Design Control, for installing non-safety related Individual Cell Equalizer devices into the Class 1E battery charging circuits without isolation as specified by Institute of Electrical and Electronics Engineers standard 384 as amended by RG 1.75. The violation was entered into the licensees corrective action program as condition report 10294321. The licensee was evaluating corrective actions, which included the removal of the non-Class 1E components. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to conform to Class 1E design requirements for independence affected the reliability of the Class 1E battery systems. The team determined the finding to be of very low safety significance (Green), because it was a deficiency affecting the design or qualification of a SSC, and the SSC maintained its operability or functionality. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance. |
| 05000425/FIN-2016007-04 | 2016Q4 | Vogtle | Failure to Perform Required In-Service Testing of Unit 2 CST Swap over Valves | The NRC identified a Green non-cited violation of Technical Specification 5.5.8, Inservice Testing Program, for Vogtle Unit 2 failure to perform the required testing in accordance with the American Society of Mechanical Engineers Operation and Maintenance Code for nine valves that had active safety functions. Specifically, these valves were required to operate when aligning the AFW pumps from Condensate Storage Tank (CST) 1 to CST 2. The violation was entered into the licensees corrective action program as condition report 10293900. The licensee performed an immediate determination of operability and determined that the CST valves were operable but degraded nonconforming. The licensee planned to register the CST valves into the IST program and exercise those valves that that have never been exercised at the first available opportunity. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, degraded valve performance could go undetected without periodic testing and trending. The team determined the finding to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of TS or Non-TS equipment. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance. |
| 05000424/FIN-2016007-05 | 2016Q4 | Vogtle | Failure to Perform Periodic Testing Of Safety-Related Valve Interlocks | The NRC identified a Green, non-cited violation of Title 10 Code of Federal Regulations Part 50.55a(h)(2) Protection Systems, because the licensee failed to perform periodic testing of safety-related valve interlocks to ensure an adequate single failure analysis by identifying detectable failures in accordance with Institute of Electrical and Electronics Engineers standard (IEEE) 379-1972, IEEE Trial-Use Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection Systems. The violation was entered into the licensees corrective action program as condition report 10293749. The licensee performed an immediate determination of operability and determined that the affected systems were operable but degraded nonconforming. The licensee was in the process of determining and developing adequate corrective actions to conform with Institute of IEEE Standard 379-1972. The performance deficiency was determined to be 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to periodically test safety-related valve interlocks affected the adequacy of the licensees single failure analysis. The team determined the finding to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of technical specification or nontechnical specification equipment. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance. |
| 05000424/FIN-2016007-06 | 2016Q4 | Vogtle | Turbine Driven Auxiliary Feedwater (TDAFW) Pumps 1/2-1302- P4-001 and Motor Driven Auxiliary Feedwater (MDAFW) Pumps 1/2-1302-P4-002/003 | The NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control for the licensee's failure to translate the Auxiliary Feedwater (AFW) pumps design bases into adequate acceptance criteria for technical specifications SR 3.5.7.2 and for the failure to verify the adequacy of the design of the same AFW pumps. The licensee entered the violation into the corrective action program as condition reports 10293456 and 10294168. As an immediate corrective action, the licensee evaluated the operability of the Unit 1 and 2 AFW pumps, modify the allowed diesel frequency acceptance criteria, and initiated corrective action to develop new acceptance criteria and monitor pump performance for degradation. The performance deficiencies were more-than-minor because they were associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, when the quality of the established surveillance criteria was considered, there was a reasonable doubt on the operability of the Unit 1 and 2 turbine driven AFW and 2A and 1B motor driven AFW pumps. The team determined the finding to be of very low safety significance (Green) because it did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time. The team determined that the finding had a crosscutting aspect in the Human Performance area of Design Margins (H.6), because engineers did not demonstrate the characteristic of ensuring that design margins were guarded and changed only through a systematic and rigorous process. |
| 05000424/FIN-2016007-07 | 2016Q4 | Vogtle | Failure to Update the UFSAR with the Complete and Accurate Information | The NRC identified a severity level IV non-cited violation of Title 10 Code of Federal Regulations Part 50.71(e)(4) for the failure to reflect all changes made in the facility or procedures as described in the Updated Final Safety Analysis Report (UFSAR). The licensee failed to update UFSAR with the design basis of a new digital emergency diesel generator sequencers installed in 2007. This violation was entered into the licensees corrective action program as condition reports 10288350, 10293456, 10291633. The licensee planned to update the UFSAR with the applicable design basis. The failure to update the UFSAR was a performance deficiency that was determined to be a minor reactor oversight program violation because it did not meet the more than minor screening criteria. Because the issue impacted the NRCs ability to perform its regulatory process, the inspectors evaluated the violation using the traditional enforcement process. The inspectors determined the issue was a severity level IV violation because it met violation example 6.1.d.3 of the NRC Enforcement Policy. The violation represented a failure to update the UFSAR as required by Title 10 Code of Federal Regulations Part 50.71(e), but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. Cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000424/FIN-2016007-08 | 2016Q4 | Vogtle | Failure to Promptly Identify Nonconformances with Tornado Missile Protection | Enforcement Guidance Memorandum (EGM) 15-002 dated 6/10/2015, (ADAMS Accession No. ML15111A269) provided guidance to exercise enforcement discretion when an operating power reactor licensee does not comply with the plants current site-specific licensing basis for tornado-generated missile protection. Specifically, discretion would apply to the TS limiting conditions for operation (LCO) which would require a reactor shutdown or mode change, if a licensee could not meet TS LCO required action(s) within the TS completion time. The EGM background discussed Regulatory Issue Summary (RIS) 2015-06, Tornado Missile Protection, dated 6/10/2015, (ADAMS Accession No. ML15020A419) to remind licensees of the need to conform their facility to the current, site-specific licensing basis for tornado-generated missile protection. In addition the EGM stated, that upon reviewing the above-noted RIS, some licensees may discover that a TS-controlled SSC at their facility does not comply with the plants current licensing basis (CLB) and that an operability determination (or functional assessment) will be necessary. The EGM actions section specified that the NRC would exercise this enforcement discretion only when a licensee implements initial compensatory measures prior to the expiration of the time allowed by the LCO that provide additional protection such that the likelihood of tornado missile effects are lessened. The licensee initiated CR10087558 on 06/23/2015, to evaluate the RIS and conducted at least two walk-downs to identify tornado missile nonconformances. The licensee discovered potential nonconformances during these walk-downs and itemized them in a list. However, the licensee failed to identify all of these items as conditions adverse to quality (CAQs), in accordance with Appendix B, Criterion XVI. The team determined that the CAP required the evaluation of these items, CRs to document the nonconformances, and operability determinations for items affecting TS. Procedure NMP-GM-002, Corrective Action Program, Section 2, defined a condition adverse to quality in part, as an all-inclusive term used in reference to any of the following: ..., deficiencies, ..., and nonconformances potentially impacting Nuclear Safety. Nonconformances are deficiencies in characteristic, documentation, or procedure that renders the quality of an item or activity unacceptable or indeterminate. The team determined that, at the time of discovery, the itemized tornado missile vulnerabilities rendered the quality of SSCs indeterminate and thus a nonconformance in accordance with the definition in the procedure. Procedure NMP-GM-002-001, Corrective Action Program Instructions Section 4 specified that personnel should initiate a CR to identify an event, condition, problem, or process that needs correcting. (This included) nonconforming items. In addition, Section 4 specified to immediately contact the Shift Support Supervisor or Work Week Coordinator (Dispatcher) when a condition is discovered that has the potential to impact plant operation or reportability. (This included) equipment or process issues related to Technical Specifications (tech specs). The team noted that the licensee did not create any additional CRs for the itemized potential vulnerabilities as required by their corrective action instructions procedure. On October 4, 2016, the inspectors conducted plant walk downs of the SSCs selected in the CDBI inspection plan and identified potential tornado missile issues. These issues were previously highlighted as potential nonconformances by the licensee, but not identified as CAQs. As a result of these observations, the licensee initiated CRs: CR10291142, Unit 1 TDAFW Exhaust nonconformance CR10291143, Unit 2 TDAFW Exhaust nonconformance CR10291144, Unit 1 Condensate Storage Tanks nonconformance CR10291145, Unit 2 Condensate Storage Tanks nonconformance CR10291146, Unit 1 Main Steam Safety Valve Exhaust nonconformance CR10291148, Unit 2 Main Steam Safety Valves Exhaust nonconformance The licensee determined that the TDAFW Exhaust and Condensate Storage Tanks were not operable because of nonconformances with these components tornado missile protection design bases. Additionally, the licensee submitted a 10 CFR 50.72 notification report (52319) to the NRC in accordance with plant procedures and NRC requirements. |
| 05000390/FIN-2016011-05 | 2016Q3 | Watts Bar | Common Service Station Transformers A and B General Design Criteria 17 Analyses | The team identified an unresolved item (URI) related to the licensees analyses done to evaluate the use of common service station transformers (CSST) A and B as qualified offsite circuits that satisfy general design criteria (GDC) 17. This URI is to determine if a performance deficiency exists. The Class 1E power system is normally supplied from offsite power through CSST C and D. Watts Bar applied for and was issued a license amendment to, in part, add an allowance to use CSST A or B as qualified offsite circuits that satisfy GDC 17. During the review of the license amendment, the NRC requested additional information about when CSST A or B is being used as a GDC 17 source and the auxiliary systems for both units are powered from the main generator. By letter dated January 29, 2015, the licensee replied by stating, in part, that their analysis evaluated: a.) A dual unit trip as a result of abnormal operational occurrence; and c.) Accident in one unit and spurious ESF actuation in the other unit. The team has requested the licensees analyses for scenarios a.) and c.) from the licensees response and has questions about how the licensee accounted for the voltage drop due to fast transfer of loads from the main generator to CSST A or B. This issue will remain open pending the licensees response to the teams questions and subsequent review in order to determine if a PD exits. This URI is identified as 05000390, 391/2016011-05 Common Service Station Transformers A and B General Design Criteria 17 Analyses. |
| 05000391/FIN-2016011-03 | 2016Q3 | Watts Bar | Failure To Ensure Adequate Unit 2 Emergency Diesel Generator Surveillance Instructions | The NRC identified a SL IV NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have adequate instructions and acceptance criteria in the emergency diesel generators surveillance procedures to ensure that the largest load rejection test bounded the power demand of the largest load. These issues were entered into the licensees corrective action program as condition reports 1201749 and 1199001. The licensee confirmed current operability and determined that likely corrective actions will include revisions to the surveillance instructions. The performance deficiency was determined to be more than minor because it represented an inadequate procedure that, if left uncorrected, could adversely affect the quality of the testing of a safety-related SSC. Specifically, the licensees procedures to implement TS SR 3.8.1.9 failed to ensure that the tested kW level of the rejected load bounded the largest predicted post-accident load. The team determined this finding to be of very low safety significance, SL IV, because it represented a failure to meet a regulatory requirement, including one or more Quality Assurance criteria that had more than minor safety significance. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Documentation in the area of Human Performance. |
| 05000391/FIN-2016011-04 | 2016Q3 | Watts Bar | Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 2 AFW Pumps | The NRC identified a SL IV NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate the available net positive suction head to the Unit 2 auxiliary feedwater pumps. These issues were entered into the licensees corrective action program as condition report 1196925. The licensee confirmed current operability and had determined that likely corrective actions will include revisions to the net positive suction head calculation. The performance deficiency was determined to be more than minor because it represented an inadequate quality oversight function that, if left uncorrected, could adversely affect the quality of the analysis of a safety related SSC. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 57%. The team determined this finding to be of very low safety significance, SL IV, because it represented a failure to meet a regulatory requirement, including one or more Quality Assurance criteria that had more than minor safety significance. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance. |
| 05000390/FIN-2016011-02 | 2016Q3 | Watts Bar | Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 1 AFW Pumps | The NRC identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate the available net positive suction head to the Unit 1 auxiliary feedwater pumps. These issues were entered into the licensees corrective action program as condition reports 1196925 and 1201623. The licensee confirmed current operability and had determined that likely corrective actions will include revisions to the net positive suction head calculation. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 74%. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating SSC that maintained its operability. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance. |
| 05000390/FIN-2016011-01 | 2016Q3 | Watts Bar | Failure To Ensure Adequate Unit 1 Emergency Diesel Generator Surveillance Instructions | The NRC identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have adequate instructions and acceptance criteria in the emergency diesel generator surveillance instructions to ensure that the largest load rejection test bounds the power demand of the largest load. These issues were entered into the licensees corrective action program as condition reports 1201749 and 1199001. The licensee confirmed current operability and determined that likely corrective actions will include revisions to the surveillance instructions. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the licensees SIs to implement TS SR 3.8.1.9 failed to ensure that the tested kW level of the rejected load bounded the largest predicted post-accident load. The team determined the finding to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of Technical Specification or Non-Technical Specification equipment. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Documentation in the area of Human Performance. |
| 05000391/FIN-2016002-04 | 2016Q2 | Watts Bar | Failure to Follow Operability Procedure Results in Potential Inoperability of the 2A-A Auxiliary Feedwater Pump | The 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). |
| 05000391/FIN-2016002-05 | 2016Q2 | Watts Bar | Failure to Perform A TDAFW Surveillance In Accordance With Procedures | The 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. |
| 05000390/FIN-2016002-07 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of Inoperable Containment Penetrations | The 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-2016002-08 | 2016Q2 | Watts Bar | Failure to Follow Maintenance Procedure Results in overspeed trip of the 2C-S Turbine Driven Auxiliary Feedwater Pump | A 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-2016002-09 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of Failure to Meet Technical Specification Surveillance Requirement 3.5.2.3 for the Emergency Core Cooling System | The 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. |
| 05000390/FIN-2016002-10 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of an Inoperable Rod Position Indication | The 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-2016002-03 | 2016Q2 | Watts Bar | Untimely 10 CFR 50.73 Notification of an Inoperable Charging Pump | The 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-2016002-01 | 2016Q2 | Watts Bar | Failure to Ensure that a Train of Source Range Detection was Available to Monitor Neutron Population During a Fire Event | The 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-2016002-02 | 2016Q2 | Watts Bar | Failure to Translate Design Requirements into a Maintenance Procedure for the 1B-B Charging Pump Room Cooler | The 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-2016002-06 | 2016Q2 | Watts Bar | Failure to Satisfy TS LCO 3.6.3 | The 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. |
| 05000390/FIN-2016008-01 | 2016Q1 | Watts Bar | Failure to Perform 50.59 Screenings For Procedures For Unit 1 | The NRC identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 1 for the licensees failure to perform 10 CFR 50.59 screening reviews for new technical procedures and changes to technical procedures, as directed by procedure NPG-SPP- 01.2.1, Interim Administration of Site Technical Programs and Procedures for Watts Bar 1 and 2, Rev. 2. The licensee entered this issue into their corrective action program as condition report 1145320 and performed the procedurally directed screening reviews which determined that no 50.59 Evaluations were required. The licensees failure to perform 10 CFR 50.59 screening reviews for new technical procedures and changes to technical procedures as directed by procedure NPG-SPP- 01.2.1 was determined to be a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. The finding was determined to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of Technical Specification or Non-Technical Specification equipment. The finding was assigned a cross-cutting aspect of Change Management in the Human Performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remained the overriding priority (H.3). |
| 05000390/FIN-2016008-02 | 2016Q1 | Watts Bar | Failure to Perform Verification and Validation For Abnormal Operating Instructions For Unit 1 | The NRC identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 1 for the licensees failure to perform verification and validation for abnormal operating instructions as directed by technical instruction 0-TI-12.11, Emergency Operating Instruction (EOI). The licensee entered this issue into their corrective action program as condition reports 1151954 and 1153507, and performed the procedurally directed verifications and validations which determined that all of the abnormal operating instructions in question were adequate. The licensees failure to perform verification and validation for abnormal operating instructions as directed by technical instruction 0-TI-12.11 was determined to be a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. The finding was determined to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of Technical Specification or Non-Technical Specification equipment. The finding was assigned a cross-cutting aspect of Change Management in the Human Performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remained the overriding priority (H.3). |
| 05000390/FIN-2016001-12 | 2016Q1 | Watts Bar | Licensee-Identified Violation | Technical 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-2016001-07 | 2016Q1 | Watts Bar | Failure to Maintain Operating Logs | The 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-2016001-11 | 2016Q1 | Watts Bar | Licensee-Identified Violation | Watts 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-2016001-10 | 2016Q1 | Watts Bar | Failure to Maintain an Adequate Surveillance Procedure for Emergency Core Cooling System Venting | The 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. |
| 05000391/FIN-2016008-04 | 2016Q1 | Watts Bar | Failure to Perform Verification and Validation For Abnormal Operating Instructions For Unit 2 | The NRC identified a SL IV non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 2 for the licensees failure to perform verification and validation for abnormal operating instructions as directed by technical instruction 0-TI-12.11, Emergency Operating Instruction (EOI). The licensee entered this issue into their corrective action program as condition reports 1151954 and 1153507, and performed the procedurally directed verification and validations which determined that all of the abnormal operating instructions in question were adequate. The licensees failure to perform verification and validation for abnormal operating instructions as directed by technical instruction 0-TI-12.11 was determined to be a performance deficiency. 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. The inspectors determined this finding to be of very low safety significance (SL IV) in accordance with Section 6.5 of the Enforcement Policy. The finding was assigned a cross-cutting aspect of Change Management in the Human Performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remained the overriding priority (H.3). |
| 05000391/FIN-2016008-03 | 2016Q1 | Watts Bar | Failure to Perform 50.59 Screenings For Procedures For Unit 2 | The NRC identified a SL IV non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, at Watts Bar Unit 2 for the licensees failure to perform 10 CFR 50.59 screening reviews for new technical procedures and changes to technical procedures, as directed by procedure NPG-SPP- 01.2.1, Interim Administration of Site Technical Programs and Procedures for Watts Bar 1 and 2, Rev. 2. The licensee entered this issue into their corrective action program as condition report 1145320 and performed the procedurally directed screening reviews which determined that no 50.59 Evaluations were required. The licensees failure to perform 10 CFR 50.59 screening reviews for new technical procedures and changes to technical procedures as directed by procedure NPG-SPP- 01.2.1 was determined to be a performance deficiency. 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. The inspectors determined this finding to be of very low safety significance (SL IV) in accordance with Section 6.5 of the Enforcement Policy. The finding has a cross-cutting aspect of Change Management in the Human Performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remained the overriding priority (H.3). |
| 05000390/FIN-2016001-09 | 2016Q1 | Watts Bar | Appropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant Response | The 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-2016001-06 | 2016Q1 | Watts Bar | Failure to Track Applicable Technical Specification Action Statement for Charging Pump Inoperability | The 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-2016001-08 | 2016Q1 | Watts Bar | Charging Pump 1B-B Room Cooler Fan Bearing Failure | Inspectors 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. |
| 05000390/FIN-2016001-05 | 2016Q1 | Watts Bar | Failure to Use Approved Procedures to Place RHR Letdown In Service | The 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-2016001-04 | 2016Q1 | Watts Bar | Failure to Place the RHR System in ECCSStandby Mode Prior to Exceeding an RCS Temperature of 212 oF | The 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-2016001-03 | 2016Q1 | Watts Bar | Failure to Adequately Implement the Administration of Site Technical Procedures for TDAFW Pump Governor Calibration | The 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-2016001-01 | 2016Q1 | Watts Bar | Failure to Use a Procedure Appropriate to the Circumstances for the Auxiliary Control Air System Train A | A 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-2016001-02 | 2016Q1 | Watts Bar | Inadequate Immediate Determination of Operability for the Auxiliary Control Air System Train A | The 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-2015004-01 | 2015Q4 | Watts Bar | Failure to Perform ISI General Visual Examination of Containment Moisture Barrier Associated with Containment Liner Leak-chase Test Connection Threaded Pipe Plugs | The 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-2015004-05 | 2015Q4 | Watts Bar | Core Barrel Lift Error Resulted in Unintended high Dose Rates | A 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-2015004-07 | 2015Q4 | Watts Bar | Licensee-Identified Violation | Watts 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-2015004-02 | 2015Q4 | Watts Bar | AFWST Permanent Plant Modification | The 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-2015004-04 | 2015Q4 | Watts Bar | Shield Building Operability Requirements | The 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-2015004-03 | 2015Q4 | Watts Bar | Failure to Comply with Source Range Neutron Flux Channel Technical Specification Requiprements | A 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-2015004-06 | 2015Q4 | Watts Bar | Failure to Identify a Condition Adverse to Quality for Unacceptable Preconditioning of the 1A-A Charging Pump Discharge Check Valve | The 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-2015010-01 | 2015Q3 | Watts Bar | 420 Minute Operator Manual Action to Provide Source Range Monitoring Capability | The 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. |
| 05000269/FIN-2015002-02 | 2015Q2 | Oconee | Inadequate Acceptance Criteria for PSW Pump Surveillance Testing | The NRC identified a finding for the licensees failure to ensure that appropriate acceptance criteria was used during testing to verify PSW primary pump functionality in accordance with the Duke Energy Carolinas Topical Report, Quality Assurance Program. The licensee entered this issue into their corrective action program as PIP O-15-03190. The performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, PSW pump surveillance PT/0/A/0500/001, Protected Service Water Primary and Booster Pump Test, Rev. 0, did not incorporate acceptance limits established by design documents, and as a result, the licensee could unknowingly consider the PSW primary pump functional beyond seven percent pump degradation. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its functionality. The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of avoid complacency within the human performance area. Specifically, the licensee failed to utilize standard human error prevention tools to ensure critical reviews were performed for PSW pump testing. |
| 05000390/FIN-2015002-04 | 2015Q2 | Watts Bar | Failure to Follow Procedure during SSPS Testing | A self-revealing, Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings was identified for the licensees failure to follow procedure 1-SI-99-10-A, 62 Day Functional Test of Solid State Protection System (SSPS) Train A and Reactor Trip Breaker A, Revision 59 as amended, for troubleshooting by Procedure Control Form 070-4. Specifically, the licensee attempted to take voltage measurements which were not directed by the revised procedure. The licensee stopped testing, conducted a prompt investigation and removed the first line supervisor and foreman from their duties pending remediation. The licensee placed the issue into their corrective action program as CR 1015778 The performance was more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to follow the troubleshooting procedure resulted in drawing an arc in the SSPS cabinet and tripping an upstream supply breaker which resulted in the inoperability of the 1A-A containment spray pump. 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 of containment spray for greater than its Tech Spec allowed outage time. The performance deficiency had a cross-cutting aspect of Procedure Adherence in the area of Human Performance because crew members failed to follow the work instructions in the troubleshooting procedure (H8). |