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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000390/FIN-2018003-06 | 2018Q3 | Severity level IV | Licensee-identified | 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 | Green | Licensee-identified | 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 | Green | H.5 | Self-revealing | 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 | Green | H.12 | NRC identified | 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 | Green | H.4 | Self-revealing | 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 | Green | H.12 | Self-revealing | 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 | Green | Licensee-identified | 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-2018012-01 | 2018Q2 | NRC identified | Potential Failure to Implement Reviews of Adverse Employment Actions in Accordance with Confirmatory Order, EA-09-009 and EA-09-203 | Confirmatory Order EA-09-009, -203 was issued to TVA on December 22, 2009 and requires the following: By no later than ninety (90) calendar days after the issuance of this Confirmatory Order, TVA shall implement a process to review proposed licensee adverse employment actions at TVAs nuclear plant sites before actions are taken to determine whether the proposed action comports with employee protection regulations, and whether the proposed actions couldnegatively impact the SCWE. Such a process should consider actions to mitigate a potential chilling effect if the employment action, despite its legitimacy, could be perceived as retaliatory by the workforce...... The inspectors reviewed TVA procedure NPG-SPP-01.7.4, Adverse Employment Action and the Executive Review Board, Revision 0001, dated December 17, 2017. This procedure does not, in some instances, require TVA to review proposed licensee adverse actions before actions are taken to determine whether the proposed action could negatively impact the SCWE. TVA is required per this procedure to conduct an ERB and review the proposed adverse actions for those actions listed under the ERB Adverse Actions (TVA Employees Only) category. TVA is not required per this procedure to conduct an ERB or perform a SCWE impact review for those proposed adverse actions listed under the Non-ERB Adverse Actions category. In addition, three of the adverse actions listed in the Non-ERB Adverse Actions category (Demotion, Transfer to a Less Desirable Job, and Denial of Access) were previously included in TVAs adverse action process, but excluded from the required ERB review in the latest revision and are therefore no longer required to be reviewed using their current procedure. Planned Closure Action(s): Further inspection is needed in order to determine if the licensee is in compliance with Confirmatory Orders EA-09-009, -203 and EA-17-022. Specifically, inspectors need to review samples of adverse employment actions taken and the licensees review of those actions to determine if the licensee is adequately implementing their reviews in accordance with their procedures and the confirmatory order. | ||
| 05000390/FIN-2018002-01 | 2018Q2 | Green | H.1 | Self-revealing | 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-2018002-02 | 2018Q2 | Severity level IV | Licensee-identified | 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-2018050-01 | 2018Q2 | Green | Licensee-identified | 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 | |
| 05000390/FIN-2018410-01 | 2018Q2 | Green | H.12 | NRC identified | Security | |
| 05000390/FIN-2018001-01 | 2018Q1 | Green | NRC identified | 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-2018010-03 | 2018Q1 | NRC identified | Potential Failure to Maintain Design Requirements of RTDs | The TAS RTD receipt documents specified design criteria identified in EQOP-ESE-7 in WCAP 8587 and the EQTR WCAP-8687 Supp. 2-E07A. The installed condition of the RTDs did not appear to meet the qualification performance specifications. Testing performed indicated that a gap existed between the RTDs and installed thermowells creating a large thermal resistance. Westinghouse document FDR-WBT-2015-01, Field Deviation Notice for Watts Bar Unit 2, indicated that the RTD/thermowell fit for the tip was outside the maximum tolerance listed in design drawing 1847E83 Rev. 4. The performance specifications were to ensure that the RTD time response across the thermowells thermal resistance was within tolerance. The higher thermal resistance increased the response time. The deviation appeared to be accepted as is. The Westinghouse equipment qualification data package, WCAP 8587/EQOP-ESE-7 Supplement 1 Rev. 7, page D1-5 Rev. 21, specified, in part that the performance requirements for the RTDs must meet +0.2"F repeatability; first order time response 5 seconds with well for step change of at least 20F with a water flow of 7 ft/sec. ...Time response testing has been successfully performed via type testing on a sample model of this RTD. The performance requirement appears to require a 5-second RTD/thermowell response time for qualification directly measured across the thermowell medium. The inspectors are concerned that the installed RTD/thermowell configuration was not verified to meet the performance requirements for qualification. It appeared that, due to the installation issues, the RTD response time increased, as measured by LCSR testing methods, the total delay time of 9.8 seconds. However, Westinghouse used a total response time of only 9.0 seconds for their analyses at the direction of TVA specification per WBT-TVA-3027, Revision 0, PIN ELICB-055 Evaluation to Support a 9.0-second Total RTD Response Time, August 2015. The analyses determined that some accident margins were impacted at 9.0 seconds. The inspectors questioned what the impact of the additional delay would have on the accident analyses and qualification. The inspectors questioned how the LCSR test method could accurately account for increased thermal resistance between the RTDs and the thermowells and whether the original 10-percent uncertainty for LCSR testing was adequate in the currently installed configuration. The validity of the LCSR method depends on how well the temperature sensor design satisfies LCSR test assumptions. The original installations relied on specific RTD thermowell bonding to establish a predictable thermal resistance and initial response time. It is unclear how the actual response time was determined this installation. This URI is being opened to determine if a PD exists. A review of documents and specifications provided by licensee indicated that new information requests would be likely. This issue has existed since September 2015. This issue was captured in CR 1398936, RCS Narrow Range RTDs Design and Qualification Requirements. | ||
| 05000390/FIN-2018010-02 | 2018Q1 | NRC identified | Potential Failure to Perform a 50.59 Evaluation for Auxiliary Building and Containment Isolation | DCN 66459 added relays and wiring to change the actuation system that initiated the Auxiliary Building Isolation (ABI) and Containment Ventilation Isolation (CVI) protective functions. The new control elements (relay contacts and wiring) bypassed the Unit Solid State Protection System (SSPS) circuitry. The intent was to actuate the CVI function on an ABI actuation from the opposite unit while it was at full power operation and the other unit was in refueling mode. By bypassing the refueling units protection system and controlling the components in the refueling unit, the modification in effect made the protection systems (CVI) a shared system. The CVI was classified as part of the Engineered Safeguards Protection Systems (ESFAS). The ESFAS was not listed in the UFSAR Chapter 3.1.2 WBNP Conformance with GDCs (General Design Criteria), as shared systems under GDC 5. The UFSAR compliance with GDC 5 Sharing of Structures, Systems, and Components, specified all shared systems are sized for all credible initial combinations of normal and accident states for the two units, with appropriate isolation to prevent an accident condition in one unit from carrying into the other. The new control elements integrated in the ESFAS logic, apparently on both units. The licensee did not perform a failure modes and effects analysis to determine the negative effects that could degrade the ESFAS isolation functions when they are required to operate. The inspectors are concerned that the integration of the two ESFAS circuitry could have a detrimental effect. Additional failure modes appear to have been introduced into these systems. The inspectors need to determine the extent to which each units protection system and CVI were exposed to additional failures including common cause failures to determine whether there could be more than a minorissue and a potential failure to perform an adequate 50.59 evaluation in accordance with NPG-SPP-09.3 Plant Mods and Engineering Change Control, Section III, was a performance deficiency. This URI, is being opened to determine whether the PD is more than minor. This modification was complete on June 16, 2017. This issue was captured in CR 1398935, Potential violation of 10 CFR 50.59(d)(1) via DCN 66459. | ||
| 05000390/FIN-2018010-01 | 2018Q1 | NRC identified | Potential Failure to Request NRC Approval to Increase the OPT and OTT Response Times | The reactor trips that protect from fuel damage that could result from departure from nucleate boiling around the fuel are identified as over-temperature-change-in-temperature (OTT) and over-power-change-in-temperature (OPT). The trips use the temperature from the reactor coolant systems hot legs as inputs into complex equations. In 1991, the licensee requested a license amendment to upgrade the Temperature Averaging System (TAS) and protection system to digital technology (Eagle 21 protection system). The Westinghouse topical reports (TR) for the TAS and Eagle 21 was reviewed and the TAS was approved with conditions for the RTD response times, electronic delay times, and surveillance test uncertainties in NUREG 847, the Safety Evaluation Report (SER), Supplement 8 dated January 1992. The SER specified, that the overall response time (RTD response time plus electronics delay) for the new RdF RTDs is 0.5 second longer (6.5 vs. 6.0 seconds) than the former Rosemount RTDs. This leaves a margin of 0.5 second (7.0-6.5) between the analysis and overall RTD response time. The breakdown of components used to arrive at the overall response time is 5.5 seconds for the RTD/thermowell and a conservative electronics delay of 1.0 second. The applicant stated that it will use the loop current step response (LCSR) test to measure RTD response time. A 10-percent allowance for LCSR test uncertainty will be used to ensure an overall channel response time of 7.0 seconds or less. ...During initial startup testing, actions will be taken to correct any resistance temperature detector (RTD) channel with an overall response time of greater than 7.0 seconds including electronics delay and a 10-percent allowance for loop current step response test uncertainty. After any such corrective action, the channel will be retested to verify an overall response time of 7.0 seconds or less (the value assumed in pertinent safety analyses). In 1997, licensee Design Change Notice (DCN) 39293 was implemented to increase the RTD response time. It stated, the response time requirement for OPT reactor trip was increased from 7 seconds to 8 seconds. This time includes RTDs, electronic processing, and trip circuit delays. As a result, the allowance for the sensor response time can be increased from 5.5 to 6.5 seconds. The Reactor Protection System Description, N3-99-1003, and the Technical Requirements Manual (TRM) were revised to reflect the change in response time for this channel. The change appeared to account for the 1.0 second electronic delay, but did not appear to account for the 10-percent allowance for LCSR test uncertainty, which would be derived from the RTD/thermowell delay. The uncertainty margin would appear to increase from 0.5 to 0.6 seconds. This change was implemented without NRC review and approval. In 2015, during hot functional testing of Unit 2 TAS RTDs, the RTD/thermowell delay did not meet the 6.5s required by the TRM from the change in 1997. On May 23, 2015, DCN 66327 was implemented to increase the response time again. The DCN stated, this DCN increases the total Narrow Range RTD response time from 8 to 9 seconds while changing the sensor response time from 6.5 to 8 seconds. Westinghouse has evaluated this change in letter WBT-D-5476 and determined that existing analyses are not impacted by this change. In this new response time the 1.0 second electronic delay and 8 second RTD/thermowells delay appeared to be accounted for, but not the margin for LCRS test uncertainty. If the 10-percent allowance for LCSR test uncertainty were accounted for, the total response time would appear to increase to 9.8 seconds. Westinghouse used a total response time of 9.0 seconds for their analyses at the direction of TVA, per WBT-TVA-3027, Revision 0, (5.10) PIN ELICB-055 Evaluation to Support a 9.0-second Total RTD Response Time, August 2015. The 10 percent LCSR uncertainty does not appear to have been included. Westinghouse letter LTR-TA-15-92, Transient Analysis Evaluation of an Increased RTD Delay Time for Watts Bar Unit 2, Rev. 0, stated, in part, due to the limiting nature of the (Steam Line Break) SLB w/ (Rod Withdrawal at Power) RWAP event, in which no margin currently exists to the departure from nucleate boiling ratio (DNBR) safety analysis limit (SAL), the inclusion of a 9.0-second total RTD response time resulted in a 0.55% DNBR penalty. For the feed water event, defined as a reduction in feedwater temperature, the Westinghouse letter stated, in part, key event results for both of the multiple-loop cases were impacted by the delay in receiving the OPT trip. While substantial margin was maintained to the DNBR limit of 1.38, the peak core heat flux values slightly exceeded the limit value of 121%. The letter concluded that the slower responding RTDs did not significantly impact the non-LOCA transient analyses and that the acceptance criteria for the events continued to be met, with the exception of the SLB w/ RWAP. However, generic DNB margin will be allocated to offset the 0.55% DNBR penalty associated with the evaluation. As such, the non-LOCA transient analyses can support operation of Watts Bar Unit 2 with a total RTD delay time of up to 9.0 seconds. The inspectors questioned the licensee to understand why the 10-percent allowance for LCSR test uncertainty was not accounted for in the Westinghouse analyses, and to what extent it could have affected the results. In addition, the inspectors questioned whether the 10-percent uncertainty was adequate in the current installation configuration. The inspectors also questioned how the LCSR test could account for increased thermal resistance between the RTDs and the thermowells. The test may not measure the actual delay time from the hot leg across the thermowell thermal resistance to RTD. The original installations relied on specific RTD thermowell bonding to establish a predictable thermal resistance and initial response time. It is unclear how this was performed for this installation to determine the actual response time. The 10 CFR 50.59 evaluation was performed May 22, 2016. This issue has been captured in the Corrective Action Program (CAP) as CR 1398934, Potential failure to request lic. amendment to change OPdT/OTdT response time | ||
| 05000391/FIN-2017004-03 | 2017Q4 | Green | H.5 | NRC identified | 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 | Green | Licensee-identified | 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. | |
| 05000390/FIN-2017004-02 | 2017Q4 | Green | H.5 | Self-revealing | 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) |
| 05000390/FIN-2017004-01 | 2017Q4 | NRC identified | 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-2017007-04 | 2017Q2 | NRC identified | Potential Failure to Justify Qualification of O-Rings by Commercial Grade Dedication | The inspectors opened a URI to review the adequacy of critical characteristics identified for commercial grade dedication of O-rings used in an EQ application. In 2008, TVA began to dedicate commercial grade O-rings from Parker Hannifin through a vendor, Qualtech. The O-rings are used in components that must operate in harsh environments and thus are subject to the strict requirements in 10 CFR 50.49. The dedication process used Fourier Transfer Infrared Spectroscopy (FTIR) to identify chemical characteristics, durometer to measure rubber hardness, and visual inspection to verify physical dimensions and color. When asked if this FTIR or durometer test can verify activation energy or other aging mechanisms, the licensee 10 stated, that FTIR cannot by itself validate activation energy or thermal aging mechanisms. Further review is needed to determine the limitations involved with the accuracy of FTIR and durometer testing. The inspectors could not verify that this dedication process specified the required critical characteristics to ensure compliance with 10 CFR 50.49. The inspectors further questioned how uncertainties could be determined and evaluated from these techniques. Since the licensee is relying on this dedication process, the components were not aged and tested in accordance with 10 CFR 50.49 requirements. The licensee contends that the verification of the material by FTIR and durometer analysis provided reasonable assurance that the original qualification test reports by Westinghouse remained applicable to the dedicated O-rings. This URI is opened to determine if a performance deficiency exists and it is identified as URI 05000390/2017007-04 and 05000391/2017007-04, Potential Failure to Justify Qualification of O-Rings by Commercial Grade Dedication. | ||
| 05000390/FIN-2017007-05 | 2017Q2 | NRC identified | Potential Failure to Address Environmental Qualification of Barton Transmitters | The inspectors opened a URI to review the adequacy of the licensees justification for the qualified life of Barton Model 764 differential pressure transmitters, documented in EQ binder WBNEQ-XMTR-001. Description: The inspectors reviewed EQ binder WBNEQ-XMTR-001 for Barton 764 differential pressure transmitters. The qualification standards of record were IEEE 323-1974 and NUREG 0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, Revision 1. The pressure transmitters contain a cover and a housing O-ring that are specified to be replaced following calibration at least every two years to maintain qualification, and internal mineral oil seal O-rings for the sensor diaphragms that are not regularly replaced. The binder specified that the transmitter internal electronics and the O-rings were qualified for 40 years. The transmitters in question are in lower containment specified at 120F and exposed to the maximum accumulated ambient radiation absorbed dose of 2E7 Rad over 40 years. Westinghouse (WEC) aged the transmitters for a 0.5eV activation energy at 125C including the electronics inside and the O-rings. The 0.5eV activation energy was specified for conservatism to account for the known synergistic effects of simultaneous thermal exposure and radiation exposure. This synergism can be significant and was also identified as significant for the O-ring material in EQ NUREGs (e.g. NUREG/CR-3629, NUREG/CR-3588) and various research papers. During the 0.5eV EQ testing, WEC noted that the transmitter O-rings (all) failed at 1,673 hours, which equated to less than six years of aging at 120F. Westinghouse chose a new activation energy (0.92eV) for the O-rings, but continued to use 0.5eV for the electronics. The transmitter oil seal O-rings were replaced with new ones. Then, the transmitters with the original electronics and new oil seal O-rings were aged for an additional 350 hours, while the cover and housing O-rings were only replaced with new ones during the final 14 hours. The inspectors evaluated the end of the accelerated aging processes for 120F. The inspectors noted that the electronics aging was accelerated to less than 7.5 years total at 0.5ev. The oil seal O-ring aging was accelerated to 22.6 years at 0.92eV. The cover and housing O-ring aging was accelerated to 11 months at 0.92eV. Further information from the licensee is needed to evaluate the justification for the 40-year qualified life of the transmitters. Specifically, the justification for the deviation from the original 0.5eV activation energy provided by WEC and the demonstration that synergistic effects and uncertainties were accounted for when deviating from this activation energy. This URI is opened to determine if a performance deficiency exists and it is identified as URI 05000390/2017007-05 and 05000391/2017007-05, Potential Failure to Address Environmental Qualification of Barton Transmitters. | ||
| 05000390/FIN-2017007-02 | 2017Q2 | Green | NRC identified | Failure to Maintain an Adequate Record of Qualification | The NRC identified a Green NCV of 10 CFR 50.49(j), for the licensees failure to maintain a complete record of qualification for Brand-Rex cables under environmental qualification binder WBNEQ-CABL-050. Specifically, the licensee could not produce a certificate of conformance related to thermal aging test data obtained from Brand-Rex. The licensee entered this issue into their corrective action program as CR 1310230. 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 safety related cable systems. Specifically, the irretrievable loss of quality records that demonstrate the equipment is qualified for its application in conformance to Appendix B requirements, impacted the reliability and capability of safety-related cable systems. The inspectors determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating SSC and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000390/FIN-2017002-03 | 2017Q2 | Green | Self-revealing | 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-2017002-05 | 2017Q2 | Green | Licensee-identified | 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. | |
| 05000390/FIN-2017007-03 | 2017Q2 | NRC identified | Potential Failure to Address Environmental Qualification of Brand-Rex Cables | The NRC opened a URI to review the adequacy of the licensees justification for the qualified life of Brand-Rex cables, documented in EQ binder WBNEQ-CABL-050. Description: The inspectors reviewed qualification binder WBNEQ-CABL-050 for Brand-Rex cables. Tab C of the binder contained WBNAPS2-089, EQ Calculation for Brand-Rex Cable Insulation, which determined the qualified life and evaluated the operability during design basis accidents for materials used in Brand-Rex cable insulation for plant service environments. Tab D of the binder documented three cable qualification test reports, each of these specified that the qualification standard of record was IEEE 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, in accordance with the Updated Final Safety Analysis Report (UFSAR). Tab D also contained two Arrhenius plots obtained from the cable vendor. The first plot was the originally specified plot of record and used XLPO/XLPE insulation breakdown as the end of life failure mechanism. A second plot was later supplied by letter dated March 10, 1986, at the request of the licensee. This plot used XLPO/XLPE insulation elongation to break as the end of life failure mechanism. Both plots used temperatures at 113oC, 121oC, 136oC, and 150oC, and indicated that a single test sample was evaluated at each temperature periodically until the end of life was identified. From these documents, the inspectors determined that the licensee qualified the cables by a combination of type test and analysis supported by type test data in accordance with IEEE 323-1974. The inspectors reviewed the data from these plots and determined they had activation energies of 1.37eV for the insulation breakdown plot and 1.48eV for the elongation to break plot. The qualification binder did not use data from the first plot, and eliminated the 150oC test data point from the second plot. The result was an activation energy of 2.13eV, which indicated a 40-year life for the insulation. The criteria for analysis in IEEE 323-1974, Section 6.5, specified, in part, that for extrapolation, the modes of failure produced under intensified or accelerated environmental or other influences shall be the same as those predicted under the required service conditions. The inspectors believe that insulation breakdown was a known predicted failure mechanism. Section 6.5 also specified, in part, that a valid mathematical model is required and it shall be based upon established principles and verifiable test data. The inspectors determined that the licensee extrapolated an invalid mathematical model by using an invalid treatment of the test data. In the licensees development of the mathematical model for this XLPE/XLPO insulation, they removed test data that closely agreed with published 150oC test data, and they used test data that differed from published 136oC test data. When asked, the licensee could not adequately justify, verify, or validate that their treatment of the test data met 10 CFR 50.49 requirements. The inspectors determined that this treatment altered the extrapolation of the mathematical model and produced a 2.13eV activation energy that differed from published activation energies for this type of insulation. Because of this treatment and the lack of verification, validation, and justification, the inspectors believe that the licensees mathematical model did not meet the specifications for analysis in IEE 323-1974. The standard IEEE 323-1974, Section 6.5.4, specified that, the qualified life shall be based upon the known limits of extrapolation of the time dependent environmental effects if an accelerated aging test was used to determine the mathematical model, which required that the certainty of the statistical analysis bound the qualified life of the components for qualification. The limits of extrapolation was described as confidence limits in IEEE 101-1972, IEEE Guide for the Statistical Analysis of Thermal Life Test Data, which is referenced in IEEE 323-1974, Section 6.3.3, as one of the established principles at the time of these Arrhenius extrapolations. The IEEE 101-1972 standard is also referenced in NUREG 0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, (ADAMS Accession No. ML031480402) as an established principle in the resolutions to questions about aging in Section 4. The inspectors believe that the licensee did not account for uncertainties, which must be quantified in their extrapolations in accordance with established principles and NUREG 0588. In addition, established principles in consensus standards specified that the certainty of the extrapolations are strongly dependent on the data for the various test temperatures, spaced at a minimum of 10oC apart, and the total number of samples for each test temperature. For the uncertainties, either fewer test temperatures or fewer samples could produce larger uncertainties, and thus wider confidence limits that could reduce the qualified life. The inspectors noted that the Arrhenius plot used by the licensee at 113oC and 121oC were separated by only 8oC, which was less than required minimum for accurate extrapolations. The inspectors noted that the licensees extrapolations could contain wide confidence limits based on the limited test temperatures and samples. In response to these concerns, the licensee provided NUREG-0847, Supplement No. 15, Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant Units 1 and 2, Appendix DD, page 36, dated June 1995, which documented an audit of the licensees cable binder WBNEQ-CABL-050 by the NRC and their contracted audit team, Idaho National Engineering Laboratory (INEL), which concluded in March 1995. The inspectors noted that this binder had been periodically updated since the audit, including the addition of calculation, WBNAPS2-089, EQ Calculation for Brand-Rex Cable Insulation. Further review of the EQ binder reviewed by the NRC/INEL audit is needed to determine whether the conclusions were based on information similar to that which is included in the current binder. This URI is opened to determine if a performance deficiency exists and it is identified as URI 05000390/2017007-03 and 05000391/2017007-03, Potential Failure to Address Environmental Qualification of Brand-Rex Cables. | ||
| 05000390/FIN-2017007-01 | 2017Q2 | Green | H.7 | NRC identified | Failure to Replace Namco Limit Switch Gasket to Maintain EQ Qualification | The NRC identified a Green non-cited violation (NCV) of title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to implement instructions to replace Namco limit switch gaskets as required to maintain environmental qualification. The licensee entered this issue into their corrective action program as CR 1309040. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of 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, not maintaining the main steam header isolation valve limit switches in their qualified condition impacted their reliability. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. Because the finding was indicative of current licensee performance, the inspectors assigned the cross-cutting aspect of Documentation in the area of Human Performance (H.7) because the procedure did not contain accurate instructions related to the replacement of the gaskets. |
| 05000391/FIN-2017002-01 | 2017Q2 | Green | H.5 | Self-revealing | 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 | Severity level IV | NRC identified | 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 | Severity level IV | Licensee-identified | 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 | Green | H.12 | Self-revealing | 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-2017403-01 | 2017Q1 | Green | H.1 | NRC identified | Security | |
| 05000391/FIN-2017001-02 | 2017Q1 | Green | P.1 | NRC identified | 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 | Green | P.1 | NRC identified | 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-2017001-03 | 2017Q1 | Green | Licensee-identified | 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 | |
| 05000390/FIN-2016004-01 | 2016Q4 | Green | H.4 | NRC identified | 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-2016013-02 | 2016Q4 | Severity level IV | NRC identified | Failure to Provide Accurate Information | SL-IV. The NRC identified a Non-cited Violation (NCV) of 10 CFR 50.9, Completeness and Accuracy of Information for the licensees failure to provide accurate information in all material respects to the Commission. The team determined on April 22, 2016, the licensee provided inaccurate information in a letter to the NRC titled, RESPONSE TO NRC LETTER CONCERNING A CHILLED WORK ENVIRONMENT FOR RAISING AND ADDRESSING SAFETY CONCERNS AT THE WATTS BAR NUCLEAR PLANT (ML16113A228). This information was material because the NRC relied on this information to conclude that TVA was in compliance with CO-EA-09-009/203 requirements. The licensee placed this issue into their corrective action program. The NRC determined this violation constituted a more than minor traditional enforcement violation associated with failure to provide accurate information. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address violations which impede the NRCs ability to regulate using traditional enforcement. The inspector determined that the licensees failure to provide accurate information was a violation of 10CFR50.9 which had the potential to impede or impact the regulatory process, and therefore subject to traditional enforcement as described in the NRC Enforcement Policy, dated November 1, 2016. This violation is characterized as a Severity Level IV violation because it was similar to Example Section 6.9.d.1 of the NRC Enforcement Policy. | |
| 05000390/FIN-2016013-01 | 2016Q4 | Green | P.1 | NRC identified | Failure to Implement the Program Requirement to Enter Issues into the CAP | Green. The NRC identified a Finding for the licensees failure to consistently implement the program requirements of the CAP. Specifically, the licensee failed to implement NPG-SPP-22.301, section 3.2.2 which required the licensees staff to initiate a Condition Report (CR) to enter various items into their CAP. The licensee placed this issue into their corrective action program. The performance deficiency was more than minor because, if left uncorrected, issues would remain unanalyzed that could represent a more significant safety concern. The performance deficiency was screened using IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Cornerstone dated June 19, 2012. The finding screened to Green because none of the examples were related to any structure, system, component, (SSC) 3 exceeding its technical specification allowed outage time. A cross cutting aspect of Identification was assigned because the licensees threshold for identifying and entering issues into their CAP was not low enough as defined by their procedures. (P.1) |
| 05000390/FIN-2016501-01 | 2016Q4 | Green | P.4 | NRC identified | 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-2016404-01 | 2016Q4 | Green | NRC identified | Security | ||
| 05000390/FIN-2016004-03 | 2016Q4 | NRC identified | 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 | ||
| 05000391/FIN-2016013-03 | 2016Q4 | NRC identified | Failure to Implement Confirmatory Order Requirement for Adverse Employment Action | TBD. The NRC identified an Apparent Violation of Confirmatory Order Modifying License, (EA-09-009,203) Dated December 22, 2009 (ML093510993) for the licensees failure to; (1) implement a process to review proposed licensee adverse employment actions at Watts Bar Nuclear plant before actions were taken to determine whether the proposed action comports with employee protection regulations, and whether the proposed actions could negatively impact the SCWE; and (2) implement a process to review proposed significant adverse employment actions by contractors performing services at TVAs nuclear plant sites before the actions were taken to determine whether the proposed action comports with employee protection regulations, and whether the proposed action could negatively impact the SCWE. The NRC determined this violation constituted a more than minor traditional enforcement violation associated with failure to implement actions required by Confirmatory Order Modifying License, (EA-09-009,203). The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address violations which impede the NRCs ability to regulate using traditional enforcement. The inspector determined that the licensees failure to implement the requirements of the Confirmatory Order had the potential to impede or impact the regulatory process, and therefore subject to traditional enforcement as described in the NRC Enforcement Policy, dated November 1, 2016. The NRC has not made an enforcement decision on this matter. | ||
| 05000390/FIN-2016013-04 | 2016Q4 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.5.2 Emergency Core Cooling Systems (ECCS) Operating Condition A required, in part, that while in Mode 1 that if one train becomes inoperable that it be restored to an operable status in 72 hours. Condition B required action to place the unit in Mode 3 in 6 hours and Mode 4 in 12 hours if that train is not restored in 72 hours. Contrary to the above, the Unit 1 1B-B CCP was inoperable from July 24, 2016, until August 5, 2016, in excess of the allowed outage time of Condition A without the unit being placed in Mode 3 in 6 hours and Mode 4 in 12 hours as required by Condition B. This issue was documented in the licensees corrective action program as CR 1199024. The finding was screened using IMC 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The finding required a detailed risk evaluation because a single train of CCP was inoperable for greater than its allowed outage time. The regional Senior Reactor Analyst reviewed the inspector provided detailed risk evaluation that was performed using the Saphire SDP module. The finding was determined to be Green. | |
| 05000390/FIN-2016004-02 | 2016Q4 | Green | P.2 | NRC identified | 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-2016011-02 | 2016Q3 | Green | H.6 | NRC identified | 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. |
| 05000391/FIN-2016011-03 | 2016Q3 | Severity level IV | H.7 | NRC identified | 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 | Severity level IV | H.6 | NRC identified | 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-2016403-01 | 2016Q3 | Green | H.8 | NRC identified | Security | |
| 05000390/FIN-2016003-02 | 2016Q3 | Green | H.5 | Self-revealing | 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-2016003-01 | 2016Q3 | Severity level IV | NRC identified | 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. |