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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000390/FIN-2018003-06 | 2018Q3 | Watts Bar | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Watts Bar Unit 1 TS 3.8.1, AC Sources - Operating, Condition A, requires, in part, that an inoperable required offsite circuit be restored to operable status within 72 hours. Contrary to the requirements of Technical Specification 3.8.1, a required offsite circuit was determined to be inoperable from May 27, 2017, to June 2, 2017. |
| 05000390/FIN-2018003-05 | 2018Q3 | Watts Bar | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Watts Bar Nuclear Plant (WBN) Unit 1 Operating License Number NPF-90, Condition 2.F, requires, in part, that TVA shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Fire Protection Report for the facility, as approved in Appendix FF Section 3.5 of Supplement 18 and Supplement 29 of the SER (NUREG-0847). The WBN Fire Protection Report was developed for WBN to ensure compliance with the requirements of this license condition. Fire Protection Report, Part II, is the Fire Protection Plan. The Fire Protection Plan, Section 14, Fire Protection Systems and Features Operating Requirements (ORs), Subsection 14.10, Fire Safe Shutdown Equipment, paragraph 14.10.4, requires a fire watch to be established in auxiliary building room 757-A10 within one hour of closing pressurizer block valve 1-FCV-68-332-B. Contrary to the above, on July 19, 2018, the licensee failed to establish a fire watch in auxiliary building room 757-A10 within one hour of closing pressurizer block valve 1-FCV-68-332-B. |
| 05000391/FIN-2018003-04 | 2018Q3 | Watts Bar | Inadequate Sensitive Equipment Control Results in Unit 2 Reactor Trip on April 12, 2018 | A self-revealed Green finding and associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, Drawings, was identified for the licensees use of a procedure that was not appropriate to the circumstances, which led to the conduct of improperly planned maintenance on sensitive equipment, ultimately resulting in a reactor trip. Specifically, an inadequacy was identified in station procedure 0-TI-12.10, Control of Sensitive Equipment, which lists the sensitive equipment defined, in part, as equipment that could cause a unit trip, on which work activities are required to be appropriately planned and conducted in a manner that will preclude a unit trip. The procedure did not list the high side reactor coolant system loop flow transmitter common drain line as sensitive equipment, which allowed the licensee to improperly perform maintenance on it without the appropriate planning and control necessary to preclude the Unit 2 reactor trip that occurred on April 12, 2018. |
| 05000390/FIN-2018003-03 | 2018Q3 | Watts Bar | Failure to Collect Compensatory Samples for an Out-of-Service Effluent Monitor | The inspectors identified a Green finding and associated NCV of TS 5.7.2.3 when the licensee failed to take compensatory samples in accordance with Table 1.1-1 of the Offsite Dose Calculation Manual when the Unit 1 steam generator blowdown effluent monitor was out of service. Specifically, radiation monitor 1-RM-90-120/121 was inoperable from April 27 to May 27, 2018, and compensatory samples were not collected and analyzed within the required frequency of at least once per 24 hours. |
| 05000391/FIN-2018003-02 | 2018Q3 | Watts Bar | Unauthorized Entry Into a High Radiation Area | A self-revealed Green finding and associated NCV of TS 5.11.1.e was identified when the licensee failed to maintain current survey information and failed to inform a worker of increased dose rates in a high radiation area. As a result, a worker received an electronic dosimeter alarm on the Unit 2 pressurizer platform due to changing radiological conditions associated with a reactor mode change. |
| 05000390/FIN-2018003-01 | 2018Q3 | Watts Bar | Configuration Control Error Results in Actual Auxiliary Building Internal Flooding Event | A self-revealed Green finding and associated NCV of Technical Specification (TS) 5.7.1, Procedures, was identified when the licensee failed to maintain adequate configuration control in the high pressure fire protection (HPFP) system in accordance with station configuration control procedure, NPG-SPP-10.2, Clearance Procedure to Safely Control Energy. Specifically, the licensee failed to restore HPFP system vent and drain valves to their appropriate configuration prior to returning the system to service which resulted in a significantly large amount of HPFP system water (on the order of 10,000 gallons) being introduced into many areas (including all levels) of the Unit 1 side of the auxiliary building and wetting numerous structures, systems, and components (SSCs) (including cables, ventilation ducts, motor-operated valves, etc.) |
| 05000390/FIN-2018003-07 | 2018Q3 | Watts Bar | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Watts Bar Unit 1 TS LCO 3.8.7, Inverters-Operating, requires that two inverters in each of the four channels shall be operable. Contrary to the above, the licensee failed to ensure that two inverters in each of the four channels were operable. Specifically, from April 9, 2017 to January 10, 2018 inverter 1-II was inoperable due to an unqualified class 1E capacitor associated with the inverter. |
| 05000390/FIN-2018002-02 | 2018Q2 | Watts Bar | Licensee-Identified Violation | LER: 05000390, 391/2017-013-00, Incorrectly Adjusted Auxiliary Building Gas Treatment System Damper Leads to a Condition Prohibited by Technical Specifications, November 6, 2017. Violation: Watts Bar Unit 1 TS 3.7.12, Auxiliary Building Gas Treatment System (ABGTS), Condition A, requires that an inoperable ABGTS train to be restored to operable status within 7 days. Condition B of TS 3.7.12 requires the plant to be in Mode 3 within 6 hours and Mode 5 within 36 hours if one train of ABGTS is inoperable longer than 7 days. Contrary to the requirements of TS 3.7.12, ABGTS, train A was determined to be inoperable from July 7, 2017, at 2030 Eastern Daylight Time (EDT) to September 5, 2017, at 1645 EDT while the plant remained in Mode 1. Significance/Severity Level: This violation was characterized using traditional enforcement because the NRC determined that this violation was not reasonably foreseeable and preventable by the licensee and, therefore, is not a performance deficiency. The violation was assessed using Sections 2.2.4 and 6.1.d.1 of the NRCs Enforcement Policy and determined to be a SL IV violation. Corrective Action Reference(s): Condition Report (CR) 1335791 |
| 05000390/FIN-2018002-01 | 2018Q2 | Watts Bar | Inadequate Procedure Results in Exceeding the Design Pressure of the RHR Piping | A self-revealed Green NCV was identified when the licensee failed to consider potentially adverse system interactions when developing procedures affecting quality. Specifically, the licensee exposed Unit 1 residual heat removal system piping to higher than its design pressure while performing two evolutions simultaneously in accordance with associated procedures. |
| 05000390/FIN-2018001-01 | 2018Q1 | Watts Bar | Misapplication of Technical Specification Limiting Condition for Operation 3.0.6 | Inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50), Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to adhere to their current licensing basis (CLB) during the implementation of procedure 0-SOI-30.05, Auxiliary Bldg HVAC Systems, which governs the operation of the engineered safety feature (ESF) coolers serving as support systems for Technical Specification (TS) equipment. Specifically, based upon the documented CLB at the time, the licensee failed to enter the appropriate TS condition and action statement for the TS supported equipment when a single train of support ESF coolers was removed from service. With a single train of ESF coolers out of service, this rendered the TS supported equipment unable to meet the single failure criterion (SFC) requirement. |
| 05000390/FIN-2017004-01 | 2017Q4 | Watts Bar | Misapplication of Technical Specification Limiting Condition for Operation 3.0.6 | (Opened) Unresolved Item 05000390/2017004-01, Misapplication of Technical Specification Limiting Condition for Operation 3.0.6 Introduction. Inspectors identified an unresolved item (URI) associated with the misapplication of LCO 3.0.6 from the licensees TS as it pertains to the functionality of engineering safety feature (ESF) coolers serving as non-TS support equipment. The item is unresolved pending the outcome of engineering analyses being performed by the licensee to determine if the ESF coolers are necessary for TS-supported systems to maintain operability. In April 2010, TVA revised the bases for the Watts Bar Unit 1 TS by adding language to expand the scope of LCO 3.0.6. The licensee evaluated the TS bases revision against the 10 CFR 50.59 criteria and determined a license amendment was not required for the change.Prior to the TS bases revision, LCO 3.0.6 provided an exception for entering a supported systems conditions and required actions due to the inoperability of a TS support systemwhich by definition is a support system that has an associated LCO in the TS. Following the TS bases revision, the scope of LCO 3.0.6 was expanded to allow another exception pertaining to non-TS support systems (i.e., support systems with no associated LCO)that are 100 percent redundant and have the capability of individually supporting both TS trains. Specifically, the revision allows both of the supported TS trains to be considered operable when one of the 100 percent redundant, non-TS support system trains is declared non-functional (i.e., the non-TS support systems do not have to meet the single failure criterion resulting in the TS systems not meeting the same criterion). This revision to the TS bases manifests itself in the operation of both Unit 1 and Unit 2 ESF coolers that serve as 100 percent redundant, non-TS support systems for both trains of a TS system such as the emergency core cooling system (ECCS), containment spray (CS) system, and CCS. There are 12 plant areas with redundant trains of ESF coolers that support both trains of a TS system. In support of maintenance and in accordance with operating procedure 0-SOI-30.05, TVA routinely removed one of the redundant coolers from service exposing the twotrains of supported TS equipment to a single failure vulnerability. When the plant was in this configuration, the licensee considered both trains of the supported TS systems to be operable, and a TS LCO condition was not entered. The licensee justified operating in this manner based upon the interpretation of LCO 3.0.6, as previously discussed.Inspectors reviewed control room logs and identified multiple occasions where the plant was operated in this manner. In response, the inspectors reviewed the licensing basis for Watts Bar and determined there was a discrepancy between: (1) the General Design Criteria (GDC) found in Chapter 3 of the facilitys UFSAR; and (2) the operation of TS systems with the single failure criterion not being met.The GDC that pertain to ECCS, CS, and CCS all contain a requirement that the system safety function can be accomplished assuming a single failure. Inspection IMC 0326, Operability Determination & Functionality Assessments for Conditions Adverse to Quality or Safety (Agencywide Documents Access & Management System (ADAMS)Accession Number ML15328A099) contains guidance for inspectors to assist their review of licensee determination of operability and resolution of degraded or nonconforming conditions. IMC 0326 specifies that failure to meet a GDC is an entry point for an operability determination. Also, based on the definition of operability, IMC 0326 states: The operability requirements for an SSC (structure, system, and component) encompass all necessary support systems (per the TS definition of operability) regardless of whether the TS explicitly specify operability requirements for the support functions. In response to this inspection discovery, the licensee took two actions. First, in the nearterm, TVA revised the applicable ESF cooler operating procedure to require entrance into the appropriate LCO condition and required action statement when one of the ESF cooler trains is nonfunctional. Secondly, while the current design bases for the affected systems indicates that the coolers are required for system operability, TVA is performing engineering evaluations to determine if the support requirement can be eliminated under certain conditions. This effort is being tracked in TVAs corrective action program by CR 1357258. Based on the ongoing engineering evaluations, the inspectors have characterized this issue as a URI pending the outcome of the results. Once the evaluations are finalized, additional inspection can be performed to determine if a PD actually exists (e.g., TS violation). This is identified asURI 05000390/2017004-01, Misapplication of Technical Specification Limiting Condition for Operation 3.0.6. |
| 05000390/FIN-2017004-02 | 2017Q4 | Watts Bar | Inadequate Procedure for Temporary Configuration Changes | A self-revealed NCV of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for inadequacies associated with TVA procedure NPG-SPP-09.5, Temporary Modifications Temporary Configuration Changes, Revision 11. Specifically, a procedural exception allowed a temporary configuration change to be installed in the spent fuel pool without a screening in accordance with 10 CFR 50.59, Changes, Tests, and Experiments. The change subsequently caused an inadvertent draining of the level of the spent fuel pool to the point the control room received the low level alarm.The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control, and it adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding affected the common spent fuel pool and the Barrier Integrity Cornerstone while Unit 1 was at power and Unit 2 was in mode 4. For Unit 1, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not result in a loss of spent fuel pool inventory below the minimum analyzed level in the site-specific licensing basis. For Unit 2, the inspectors determined that this finding was of very low safety significance (Green) because it did not involve an actual reduction in function of hydrogen control for PWR ice condenser containments. The finding had a cross-cutting aspect in the Work Management attribute of the Human Performance area as defined in NRC IMC 0310, because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. (H.5) |
| 05000391/FIN-2017004-03 | 2017Q4 | Watts Bar | Failure to Promptly Identify a Condition Adverse to Quality for a Boric Acid Leak on 2-SMV-68-548 | An NRC-identified NCV of very low safety significance associated with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure to promptly identify and correct a condition adverse to quality. Specifically, NRC inspectors identified a boric acid leak on the Unit 2 loop 1 hot leg sample valve, 2-SMV-68-548, that had been missed by licensee personnel performing boric acid corrosion control program walkdowns during Unit 2 refueling outage 1 (RFO1). The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. The inspectors performed the significance determination using NRC IMC 0609. The finding affected the Barrier Integrity Cornerstone while Unit 2 was shut down, so IMC 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, was used to determine that this finding was of very low safety significance (Green) because it did not degrade the ability to isolate a drain down or leakage path. The finding had a cross-cutting aspect in the Work Management attribute of the Human Performance area as defined in NRC IMC 0310, because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. (H.5) |
| 05000390/FIN-2017004-04 | 2017Q4 | Watts Bar | Licensee-Identified Violation | The following licensee-identified violation of NRC requirements was determined to be of very low safety significance and met the NRC Enforcement Policy criteria for being dispositioned as a Non-Cited violation. Watts Bar Unit 1 TS 3.6.3, Containment isolation Valve, Condition A, states, in part, that a penetration flow path with one containment isolation valve inoperable to be isolated by use of at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured within four hours. Contrary to this requirement, Watts Bar Nuclear Plant Unit 1 containment isolation valve 1-FCV-61-122, Glycol cooled floor return header isolation was inoperable on March 5, 2016, at 1512 EST, and the penetration associated with this containment isolation valve was not isolated until 2113 EST on March 5, 2016. TS 3.6.3, Condition A, has a required completion time of four hours; however, valve 1-FCV-61-110 was not closed until six hours into the event. The licensee has entered this event into their corrective action program under CR 1146157. This licensee-identified violation of NRC requirements was determined to be of very low safety significance, Severity Level IV, and met the NRC Enforcement Policy Section 2.3.2 criteria for being dispositioned as a non-cited violation. The performance deficiency was more than minor because it was associated with the reactor containment barrier performance attribute of the barrier cornerstone in NRC IMC 0609, Attachment 04, dated October 7, 2016. This finding was further evaluated in accordance with NRC IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. The finding screened to Green since there was no actual open pathway in the physical integrity of the reactor containment. |
| 05000250/FIN-2017003-03 | 2017Q3 | Turkey Point | Inadequate Maintenance Rule (a)(4) Risk Assessment for the High Head Safety Injection Pumps | An NRC-identified NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), was identified for the licensees failure to adequately assess and manage the Unit 3 and Unit 4 online risk associated with taking both Unit 4 high head safety injection (HHSI) pumps out of service. This issue was entered in the licensees corrective action program as AR 2193584. Corrective actions completed included providing additional training to senior reactor operators (SROs) on the maintenance rule (a)(4) implementation procedure and the definition of unavailability as used in maintenance rule (a)(4) risk assessments. The licensees failure to adequately assess and manage the Unit 3 and Unit 4 online risk associated with taking both Unit 4 HHSI pumps out of service, as required by 0-ADM-225, On Line Risk Assessment and Management, was a performance deficiency. The performance deficiency was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone. The significance of the finding was determined using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significant Determination Process. The finding was determined to be of very low safety significance (Green) because the incremental core damage probability deficit for the timeframe the HHSI pumps were unavailable was less than 1E-6 for each unit, prior to, and after, the failure of the Unit 3A 4kV switchgear bus. The finding had a cross-cutting aspect in the area of Human Performance, Training, because the control room SROs did not have an adequate understanding regarding crediting operator actions and the definition of unavailability. The SROs incorrectly considered the Unit 4 HHSI pumps as available to perform their safety functions under the maintenance rule (a)(4) risk assessments (H.9). |
| 05000250/FIN-2017003-02 | 2017Q3 | Turkey Point | Inadequate Operator Fundamentals during Diesel Driven Fire Pump Surveillance Testing | An NRC-identified finding was identified for the failure to adequately implement OP-AA-100-1000, Conduct of Operations procedure. Specifically, non-licensed operators (NLOs) failed to identify that the diesel driven fire pump (DDFP) was operating in a degraded condition. The outboard shaft gland was at elevated temperature because there was no packing leakoff established. Plant operators initiated an action request (AR) 2220785 to repair the stuffing box packing and the DDFP was declared non-functional. The electric driven fire pump (EDFP) remained functional and available to supply 100% of the fire water capacity while the DDFP was non-functional. This issue has been entered into the licensees corrective action program as ARs 2220785 and 2226305.The failure to identify that the DDFP was operating in a degraded condition was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external hazards (fire) attribute of the initiating events cornerstone and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations. Specifically, NLOs did not identify a degrading and unreliable DDFP condition. The inspectors determined that the issue had very low safety significance (Green) because the EDFP remained available to provide 100 percent of the required fire water capacity. The finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because NLOs did not recognize and consider that the DDFP was operating without adequate packing gland leakoff after a significant idle period (H.12) |
| 05000250/FIN-2017003-01 | 2017Q3 | Turkey Point | Failure to Identify and Correct CCW Pipe Corrosion | An NRC-identified NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to promptly identify and correct component cooling water (CCW) external pipe corrosion that led to a through-wall flaw and leak on the Unit 3 CCW surge tank makeup line. FPL performed an immediate operability screening and determined the condition was operable but degraded considering previous prompt operability determinations for more significant CCW system leaks that bounded the leak rate and with similarly characterized structural flaws. Plant operators later isolated the through wall leak and established an alternate makeup path. This issue has been entered into the licensees corrective action program as AR 2223132.The failure to identify and correct the significant external corrosion that occurred on the Unit 3 CCW surge tank makeup line was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, through wall corrosion affects the reliability of the CCW system. The inspectors determined the finding to be of very low safety significance because it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee failed to identify the significant external corrosion and apparent metal pipe wastage. Prior opportunities for FPL to identify the significant external corrosion and pipe wastage occurred through maintenance activities on the same pipe section and system engineer quarterly systems walkdowns (P.1). |
| 05000390/FIN-2017002-05 | 2017Q2 | Watts Bar | Licensee-Identified Violation | Watts Bar Nuclear Plant TS 5.7.1.1 states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures in Regulatory Guide (RG) 1.33, Revision. 2, Appendix A, February 1978. Procedures for surveillance tests are applicable procedures under RG 1.33 Appendix A, 8.b. Contrary to this requirement, on April 4, 2017, surveillance procedure 0-SI-82-4, 18 Month Loss of Offsite Power with Safety Injection Test DG 1B-B, Revision 63, was not implemented as written. Specifically, Step 3.1 (3) was not followed when the 1B-B safety injection pump discharge isolation valve was closed but not tagged as directed by the procedure. As a result of not being tagged, there was no programmatic control in place to return the valve to the open position upon completion of 0-SI-82-4. Therefore, the valve was left in the closed position, causing the B train of safety injection to be inoperable from April 11, 2017, until May 10, 2017, when the valve was discovered to be closed during operator rounds. Because the 1B safety injection pump was inoperable for longer than its TS allowed outage time of 72 hours, a regional senior reactor analyst conducted a detailed risk evaluation using SAPHIRE (Version 8.1.5) and the standard model for Watts Bar (SPAR Version 8.50). The resulting change in core damage frequency was less than 1E-6; therefore, the finding was determined to be of very low safety significance (Green). The licensee entered this issue into their corrective action program as CR 1294133. |
| 05000391/FIN-2017002-01 | 2017Q2 | Watts Bar | Inadequate Chemistry Procedure Results in Inoperable Containment Isolation Valves | SL IV. A self-revealed severity level (SL) IV non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when implementing an inadequate procedure resulted in rendering the steam generator chemistry sample containment isolation valves inoperable. The licensee entered this issue into their corrective action program as CR 1160910. The inspectors determined that the use of an inadequate procedure that rendered the containment isolation valves inoperable was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC-2517, Appendix C, because the use of an inadequate procedure rendered the containment isolation valves inoperable. The inspectors determined this finding to be of very low safety significance because it did not represent a breakdown of the licensees quality assurance program. This finding had a cross-cutting aspect in the work management component of the Human Performance cross-cutting area because the work process did not include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities (H.5). |
| 05000390/FIN-2017002-04 | 2017Q2 | Watts Bar | Failure to Report Multiple Examples of a Loss of Safety Function in accordance with 10 CFR 50.72 and 50.73 | Severity Level IV. The inspectors identified a Severity Level IV non-cited violation of 10 Code of Federal Regulations (CFR) 50.72 and 50.73, with multiple examples due to the licensees failure to make the required eight-hour non-emergency notification and submit a Licensee Event Report (LER) to the NRC within 60 days for conditions that, at the time of discovery, could have prevented fulfillment of a safety function. These issues have been entered into the licensees corrective action program as condition report (CR) 1310096. The inspectors determined that the licensees failure to comply with 10 CFR 50.72(b)(3)(v) and 50.72(a)(2)(v) was a performance deficiency. This performance deficiency was dispositioned under traditional enforcement because the failure to make a non-emergency notification and submit an LER within the time requirements may impact the ability of the NRC to perform its regulatory oversight function. The violation was assessed using Sections 2.2.4 and 6.9.d.9 of the NRCs Enforcement Policy and determined to be a SL IV violation. Traditional enforcement violations are not assessed for cross-cutting aspects. |
| 05000390/FIN-2017002-06 | 2017Q2 | Watts Bar | Licensee-Identified Violation | Title 10 CFR 50.72(b)(3)(v)(C) requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems or components that are needed to control the release of radioactive material. Contrary to the above, on March 9, 2017, the licensee failed to notify the NRC that reactor containment was inoperable, resulting in a condition that could have prevented fulfillment of a safety function. Specifically, an inner containment door equalizing valve was not fully shut when the outer containment door was open for entry into upper containment, thereby resulting in a direct path from containment to the auxiliary building. This failure to report 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 a report required by 10 CFR 50.72 or 50.73, and the issue was determined to be a SL IV violation. The licensee entered this issue into their corrective action program as CR 1273873. |
| 05000390/FIN-2017002-02 | 2017Q2 | Watts Bar | Failure to Implement Clearance on Containment Isolation Valve Results in TS 3.6.3 Violation | Green. A self-revealed non-cited violation of Technical Specification (TS) 3.6.3, Containment isolation Valves, was identified for a failure to properly implement a clearance for containment isolation valve surveillance testing. Clearance 1-30-1011-WW removed fuses from a different valve than the one specified in the clearance. The licensee entered this issue into their corrective action program as CR 1245529. The failure to comply with NPG-SPP-10.2, Steps 3.1.2.B.5 and 6, was a performance deficiency. The performance deficiency was more than minor because it adversely affected the configuration control attribute of the Barrier Integrity Cornerstone because the incorrectly placed clearance resulted in the inoperability of the containment isolation valve for longer than its TS allowed outage time, reducing ensurance that the containment function assumed in the safety analyses would be maintained. The inspectors determined that this violation was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters. The finding has a cross-cutting aspect in the Avoid Complacency component of the Hum an Performance area as defined in NRC IMC 0310, because multiple personnel failed to recognize and plan for the possibility of 3 mistakes and error reduction tools, such as concurrent verification, were not appropriately implemented (H.12). |
| 05000390/FIN-2017002-03 | 2017Q2 | Watts Bar | Failure to Follow Procedure Results in Reactor Coolant Pump Failure to Transfer and Unit 1 Reactor Trip | Green. A self-revealed Green finding was identified for the failure to follow procedure NPG-SPP-22.207, Procedure Use and Adherence Revision 4, which requires that applicable procedures are used for all activities controlled by a written procedure. The licensee entered this into their corrective action program as CR 1291140 The failure to follow procedure NPG-SPP-22.207, Procedure Use and Adherence, Revision 4, was a performance deficiency. The performance deficiency was more than minor because it affected the Initiating Events Cornerstone attribute of Human Performance and adversely affected the cornerstone objective in that it resulted in two reactor trips. The inspectors determined that the finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment. The finding was not assigned a cross-cutting aspect since none of the CCAs described in IMC 0310 corresponded to an apparent cause or most significant causal factor of the performance deficiency. (Section 4OA3.6) |
| 05000390/FIN-2017001-03 | 2017Q1 | Watts Bar | Licensee-Identified Violation | The following licensee-identified violation of NRC requirements was determined to be of very low safety significance or Severity Level IV and met the NRC Enforcement Policy criteria for being dispositioned as a non-cited violation.Watts Bar Nuclear Plant (WBN) Unit 1 Operating License Number NPF-90, condition 2.F, requires, in part, that TVA shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29. The WBN Fire Protection Report was developed for WBN to ensure compliance with the requirements of this licensee condition. Fire Protection Report, Part II, is the Fire Protection Plan (FPP). FPP Section 14, Fire Protection Systems and Features Operating Requirements (ORs), Subsection 14.10(b), Fire Safe Shutdown Equipment, paragraph 14.10.4, requires an hourly roving fire watch be established in auxiliary building rooms 757-A1 and 757-A10 within an hour of closure of pressurizer block valve 1-FCV-68-332-B. Contrary to the above, on January 11, 2017, the licensee failed to perform a fire watch as required for fire safe shutdown equipment. Specifically, an hourly fire watch was not established or conducted when valve 1-FCV-68-332-B was closed for WO 117615614 for 1-SI-68-93, 18 Month Channel Calibration of PORV 1-PCV-69-334 Cold Overpressure Mitigation System Actuation Channel. The licensee determined valve 1-FCV-68-332-B was closed for about 2.5 hours when it was opened to restore compliance. 25 This violation is of very low safety significance (GREEN) based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase I Screening Approach. The inspectors determined that this issue did not affect the Unit 1 reactors ability to reach and maintain safe shutdown (either hot or cold) condition. This violation was documented in the licensees corrective action program as CR 1250743 |
| 05000391/FIN-2017001-02 | 2017Q1 | Watts Bar | Inadequate Immediate Determination of Operability for Essential Raw Cooling Water Flush Valve | Green. An NRC-identified finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings was identified for the failure to follow TVA procedure OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking, Revision 23. Specifically, the licensee failed to address the seismic design bases impacts and structural integrity of the 2A-A essential raw cooling water (ERCW) strainer flush valve, 2-FCV-67-9B-A, in the basis of the immediate determination of operability (IDO). 3 The failure to document an IDO on January 16, 2016, based on information sufficient to address the capability of TS components to perform specified safety functions was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone. Using IMC 0609 Appendix A, Exhibit 2 Mitigating Systems Screening Questions, 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 Identification component of the Problem Identification and Resolution area as defined in NRC IMC 0310, because the organization failed to identify issues completely, accurately, and in a timely manner. Specifically, the Operations and Engineering department failed to fully and accurately identify the impact of the through wall flaw on 2-FCV-67-9B-A. (P.1) |
| 05000390/FIN-2017001-01 | 2017Q1 | Watts Bar | Failure to Maintain the Abnormal Operating Instruction for Tornados | Green. An NRC-identified non-cited violation (NCV) of Technical Specification (TS) 5.7.1.1.a, Procedures, was identified for a failure to maintain procedure 0-AOI-8, Tornado Watch or Warning. The entry criteria were inadequate to ensure that the required actions for a tornado watch or warning would be performed in a manner such that potential plant impact from a tornado would be mitigated or prevented. The violation was entered into the licensees corrective action program (CAP) as condition report (CR) 1280644. The licensees immediate corrective action was to install a weather radio in a continually manned security area with instructions for the security personnel to notify the control room for any tornado watch or warning declaration in Rhea County, TN. The failure to maintain procedure 0-AOI-8 was a performance deficiency. The performance deficiency was more than minor because it adversely affected the procedure quality attribute of the Initiating Events Cornerstone objective, in that failure to take required actions in accordance with 0-AOI-8 after a tornado watch is issued could result in the inability to perform those actions if the watch is upgraded to a warning resulting in potential equipment failure. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not cause a reactor trip, involve the complete or partial loss of mitigation or support equipment, or impact the frequency of a fire or internal flooding event. The finding has a cross-cutting aspect in the Identification component of the Problem Identification and Resolution area because the licensee had not identified procedure 0-AOI-8 inadequate entry criteria despite past issues with timely entry. (P.1). |
| 05000390/FIN-2016004-01 | 2016Q4 | Watts Bar | Inadequate Immediate Determination of Operability for Essential Raw Cooling Water Pumps | Green: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to base an immediate determination of operability (IDO) for essential raw cooling water (ERCW) pumps on information sufficient to conclude that a reasonable expectation of operability existed. The licensee restored compliance on November 30, 2016, when they documented an IDO that met the requirements of OPDP-8. The violation was entered into the licensees CAP as CR 1237178. The performance deficiency was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone. Specifically, reasonable assurance of operability did not exist for the ERCW pumps from November 29, 2016 until November 30, 2016. The inspectors determined the finding was of very low safety significance (Green) because it did not represent an actual loss of function for at least a single train for longer than its technical specification allowed outage time. The cause of this finding had a cross cutting aspect of Teamwork in the Human Performance area, because individuals and work groups failed to communicate and coordinate their activities within and across organizational boundaries such that nuclear safety is the overriding priority. (H.4). |
| 05000390/FIN-2016004-03 | 2016Q4 | Watts Bar | Notice of Enforcement Discretion 16-2-01 for Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications | (Opened) Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications and Notice of Enforcement Discretion 16-2-01 Introduction: The inspectors opened an unresolved item associated with a potential noncompliance with TS 3.8.1 that occurred on October 15, 2016. Notice of Enforcement Discretion 16-2-01 was granted by the NRC staff agreeing not to enforce compliance with the TS completion time for an additional 130 hours. Description: At 6:32 a.m. on October 12, 2016, Watts Bar operations staff declared the 1A-A EDG inoperable when the output breaker to the 1A shutdown board opened unexpectedly due to phase overcurrent during performance of the load test required by procedure 0-SI-82-13, 24 Hour Load Run - DG 1A-A. The 1A-A emergency diesel generator was operating normally prior to the opening of the breaker. The licensees initial assessment determined the likely cause of the breaker trip was operation of the tap changer associated with the offsite power supply transformer. A subsequent 24 hour EDG load test was started at 12:35 a.m. on October 13, 2016. At 6:45 p.m. on October 13, 2016, operations staff noted mega volt amps (reactive) swings. During subsequent troubleshooting activities, it was determined that the mega volt amps (reactive) variance could be consistently reproduced by slight movement of a potentiometer on the 1A-A EDG voltage regulator. The licensee determined that an issue in the voltage regulator circuit was the most likely cause of the output breaker trip, and made preparations to replace and calibrate the voltage regulator on which the potentiometer was located. The licensee determined that it would require more than 72 hours to complete the removal and replacement of the voltage regulator and post-maintenance testing. The licensee requested a notice of enforcement discretion and an additional 144 hours to restore EDG 1A-A. A notice of enforcement discretion for an additional 130 hours was granted by the NRC staff at 9:30 p.m. on October 14, 2016. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific TSs in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine if there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item. (URI 05000390, 391/2016004-03, Notice of Enforcement Discretion 16-2-01 for Emergency Diesel Generator 1A-A Inoperable for Longer Than Allowed by Technical Specifications) This activity constitutes completion of one event follow-up sample, as defined in IP 71153 |
| 05000327/FIN-2016004-01 | 2016Q4 | Sequoyah | Degraded Fire Barrier Penetrations | Green. The NRC identified a non-cited violation (NCV) of the facilitys operating license for the licensees failure to ensure that all fire barrier penetrations in fire zones boundaries protecting safety related areas are functional at all times. Specifically, on eight separate fire barrier penetrations, the licensee failed to recognize that the barrier had become damaged to the point of being nonfunctional. The licensee also failed to implement required compensatory measures for a nonfunctional fire barrier penetration contrary to the approved fire protection report (FPR). The licensee entered the issues into their corrective action program (CAP) as Condition Reports (CRs) 1229468, 1229470, 1243550, 1243970, 1243552, 1243554, 1243555, and 1243557. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the fire barriers being damaged to the point of declaring the fire barrier penetrations nonfunctional, there was no assurance that the fire barrier would prevent the spread of fire through the cable penetration during a design basis fire. The inspectors performed the SDP using NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and assigned a High degradation rating, giving no credit for Barrier Protection in accordance with the Fire Barrier Degradation section. The inspectors concluded, that the finding was of very low safety significance (Green) due to fully functional automatic suppression systems on either side of the fire barrier (Question 1.4.3-C). Using Manual Chapter 0310, Aspects Within the Cross-Cutting Areas, the inspectors identified a cross-cutting aspect in the Identification component of the Problem Identification and Resolution area, because the licensee failed to enter the damaged fire barrier into their CAP after it was initially damaged (P.1) |
| 05000390/FIN-2016004-02 | 2016Q4 | Watts Bar | Inadequate Immediate Determination of Operability for Containment Penetration X-65 | Green: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to address all the design criteria for check valve, 1-CHV-31-3407, in the basis of the immediate determination of operability (IDO) for containment penetration X-65 to conclude that a reasonable expectation of operability existed. On September 19, Technical Specification (TS) compliance was restored when Penetration X-65 returned to operable when it was isolated and drained. The violation was entered into the licensees corrective action program as condition report (CR) 1216892. The performance deficiency was more than minor because it adversely affected the design control attribute of the barrier integrity system cornerstone. Specifically, reasonable assurance of operability did not exist for containment penetration X-65 from September 18, 2016, until September 19, 2016. The inspectors performed an initial screening of the finding and determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components; and hydrogen igniters are not applicable. The cause of this finding had a cross-cutting aspect of Evaluation in the area of Problem Identification and Resolution, because the licensee did not consider all functions of check valve 1-CKV-31-3407 when performing the IDO after the valve failed to pass the surveillance instruction. (P.2). |
| 05000390/FIN-2016501-01 | 2016Q4 | Watts Bar | Failure to Maintain Minimum On-Shift Emergency Response Staffing Levels | Green: The NRC identified a non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50.47(b)(2) for the licensees failure to maintain the effectiveness of its emergency plan, when on more than one occasion, the number of control room operators fell below minimum staffing, as required by Appendix C of NP-REP Tennessee Valley Authority (TVA) Nuclear Power Radiological Emergency Plan (E-Plan). The licensees corrective actions included entering the issue into their corrective action program as CR 1233650. The performance deficiency was more than minor because it was associated with the emergency response organization readiness attribute of the Emergency Preparedness cornerstone and adversely impacted the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the finding in accordance with Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and using Table 5.2-1 Significance Examples for 50.47(b)(2), determined that this finding represented an example of a staffing process that would permit a shift to go below E-Plan minimum staffing requirements. The inspectors determined that the licensees process, on more than one occasion, failed to ensure that on-shift staffing met E-Plan minimum staffing requirements between March 20 and May 6, 2016. The cause of the finding was determined to be associated with the cross-cutting aspect of thorough evaluation of problems in the corrective action component of the Problem Identification and Resolution area because the organization failed to periodically analyze information from the corrective action program and other assessments in the aggregate to identify programmatic and common cause issues (P.4). |
| 05000390/FIN-2016003-02 | 2016Q3 | Watts Bar | Inappropriate Procedure used for Work Order Scope Change Results in Loss of 1B-B Shutdown Board. | A self-revealed non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to use a procedure appropriate to the circumstances when work scope changed which contributed to the loss of the 1B-B shutdown board on May 17, 2016. The violation was entered into the licensees CAP as CR 1172243. The failure to use a procedure appropriate to the circumstances, such as NPG-SPP-07.6, NPG Work Management Planning Procedure, Revision (Rev.) 14, for a work scope change associated with a design change work order on the 1B-B shutdown board on May 17, 2016, was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone objective because the loss of the 1B-B shutdown board caused the inoperability of the B train of the onsite electrical distribution system and also resulted in the inoperability of all B train structures, systems, or components (SSCs) powered from the 1B-B shutdown board. The inspectors performed an initial screening of the finding and 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 technical specification (TS) allowed outage time. The finding had 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 process of planning and executing the work activities for Design Change Notice (DCN) 64063 failed to identify and manage the risk associated with system restoration due to either equipment failure or personnel error (H.5). |
| 05000390/FIN-2016003-03 | 2016Q3 | Watts Bar | Licensee-Identified Violation | Watts Bar Nuclear Plant Units 1 and 2 Technical Specification 5.7.1.1.d requires, in part, that written procedures be established, implemented, and maintained covering the activities involved with Fire Protection Program implementation. The Watts Bar Fire Protection Report lists compensatory actions that must be implemented when there are impaired fire protection systems, including, under some circumstances, a continuous fire watch. TVA Corporate Procedure NPG-SPP-18.4.6, Control of Fire Protection Impairments, Rev. 0006, is the implementing/controlling process for all Fire Protection impairments, and establishes the process for implementing compensatory actions for fire impairments as directed by the Fire Protection Report. NPG-SPP-18.4.6, Section 3.2.6.A, states that Fire watches are utilized for the surveillance of areas where fire protection systems are impaired. The compensatory fire watch process is described in Attachment 7 of NPG-SPP-18.4.6. Contrary to the above, on April 27, 2015, the licensee failed to perform a continuous fire watch as required for fire protection systems that were impaired. Specifically, the licensee failed to establish the compensatory continuous fire watch required by Fire Protection Impairment Permit C10-0639, which authorized an impairment of fire detection systems to allow for welding work on the 713 elevation of the auxiliary building of Watts Bar Nuclear Power Plant. 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 I Screening Approach. The inspectors determined that the fire finding did not affect the Unit 1 reactors ability to reach and maintain safe shutdown (either hot or cold) condition. Therefore, the finding screened as Green. Specifically, the only equipment important to safety in the affected fire area was associated with the construction unit (Unit 2), and would not have impacted the safe shutdown of Unit 1. This violation was documented in the licensees corrective action program as CR 1019953. |
| 05000390/FIN-2016003-01 | 2016Q3 | Watts Bar | Falsified Fire Watch Records | Severity Level IV. The NRC identified a Severity Level IV violation of 10 CFR 50.9 Completeness and Accuracy of Information, for the failure to maintain continuous compensatory fire watch information that was complete and accurate in all material respects. The licensees actions of creating falsified fire watch completion records for the 713 elevation of the Auxiliary Building was a performance deficiency. The licensee entered this issue into the corrective action program as CR 1019953 and took remedial action against the involved individuals commensurate with the circumstances. The NRC evaluated this issue under the traditional enforcement process because it involved willfulness. In consideration of the fact that the individuals were contract fire watch personnel with minimal supervisory responsibilities, and that the underlying safety significance of the missed fire watch was low, the NRC concluded that this violation should be characterized at Severity Level IV in accordance with Section 2.2.1.d of the Enforcement Policy. Furthermore, because this violation involved willfulness and lack of supervisory oversight, the non-cited violation criteria of paragraph 2.3.2.a.4.(c) was not satisfied, such that this violation will be cited. This violation was evaluated under the traditional enforcement process and thus does not have a cross cutting aspect. |
| 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). |
| 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-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-2016002-01 | 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-2016002-03 | 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. |
| 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-2016002-10 | 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-2016002-09 | 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-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-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-2016001-08 | 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-2016001-04 | 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-2016001-09 | 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-2016001-07 | 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. |