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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000269/FIN-2018002-02 | 2018Q2 | Oconee | Failure to Coordinate a No-later-than Arrival Time for the Shipment of a Category 2 Quantity of Radioactive Material | The inspectors identified aSeverity Level IV NCV of 10 CFR 37.75(b) when the licensee failed to coordinate a no-later-than arrival time for a Category 2 shipment of radioactive material. Specifically, the licensee failed to recognize that a package of primary resin contained a Category 2 quantity of Cobalt-60 prior to shipment, and therefore failed to arrange a no-later-than arrival time with the receiving licensee. |
| 05000287/FIN-2018002-01 | 2018Q2 | Oconee | Failure to Perform ISI General Visual Examinations of Containment Moisture Barrier Associated with Containment Liner Leak Chase Test Connection Piping | The inspectors identified a Green NCV of 10 CFR Part 50.55a, Codes and Standards, involving the licensees failure to properly apply Subsection IWE, of ASME Section XI, for conducting general visual examinations of the leak chase test connection piping at the concrete floor interface which provides a moisture barrier to the containment liner seam welds. |
| 05000400/FIN-2018002-06 | 2018Q2 | Harris | Failure to Implement Viable Compensatory Actions with Seismic Monitoring System Out of Service for Planned Preventive Maintenance | An NRC-identified Green NCV of 10 CFR 50.54(q)(2) was identified for the licensees failure to follow and maintain the effectiveness of its emergency plan that meets the requirements of the risk-significant emergency planning standard 10 CFR 50.47(b)(4). Specifically, the licensee failed to implement viable compensatory actions while conducting planned preventive maintenance that rendered both seismic monitoring systems unavailable for 53.5 hours resulting in a loss of emergency assessment capability for declaring a Notification of Unusual Event under Emergency Action Level (EAL) HU2.1 for a seismic event. |
| 05000400/FIN-2018002-05 | 2018Q2 | Harris | Failure to Follow Secondary Water Chemistry Plan for Elevated Levels of Secondary Water Impurities | An NRC-identified Green NCV of TS 6.8.4.c, Secondary Water Chemistry, was identified for the licensees failure to follow secondary water chemistry control requirements in accordance with procedure CSD-CP-HNP-0002, Harris Secondary Water Chemistry Strategic Plan. . Specifically, the licensee remained at 100% power for approximately 10 hours after entering secondary water chemistry Action Level 3 due to elevated chlorine and sulfates chemical impurity concentrations, which was contrary to the procedure requirements to downpower the unit to below 5% power as quickly as safe plant operation permits. This unit downpower delay allowed additional time for the chemical impurities to adversely affect the steam generators. |
| 05000400/FIN-2018002-04 | 2018Q2 | Harris | Failure to Implement Adequate Steam Generator Blowdown Demineralizer Control Procedures | A self-revealing Green NCV of Technical Specifications (TS) 6.8.1.a, Procedures and Programs, was identified for licensees failure to establish and implement adequate steam generator blowdown demineralizer control operating procedures resulting in exceeding secondary water chemistry Action Level 3 criteria for impurities in the steam generators. Specifically, the licensee did not implement adequate isolation valve controls between the demineralizer resin regeneration system and the feedwater system during resin regeneration activities. This open path allowed leakage of sulfates and chlorides into the feedwater system. The level of these impurities exceeded the secondary chemistry Action Level 3 threshold and resulted in an unplanned shutdown. |
| 05000400/FIN-2018002-03 | 2018Q2 | Harris | Failure to Adequately Document Changes to the Emergency Plan | The inspectors identified multiple examples of a Severity Level IV (SL-IV) NCV of 10 CFR 50.54(q)(3), for changes to the licensees radiological emergency plan (E-Plan) associated with protective action recommendation (PAR) procedures and emergency response equipment that failed to demonstrate that the changes would not reduce the effectiveness of the E-Plan. Specifically, the licensee did not provide an adequate analysis to demonstrate that the removal of the sheltering in-place PARs was not a reduction in effectiveness of the E-Plan. Additionally, the licensee did not perform an analysis demonstrating that the removal of a temporary diesel generator providing a backup source of power to the Technical Support Center (TSC) did not reduce the effectiveness of the E-Plan. |
| 05000400/FIN-2018002-02 | 2018Q2 | Harris | Inadequate Fire Brigade Performance Assessment of Announced Fire Drill | An NRC-identified Green NCV of 10 CFR 50.48(c) and National Fire Protection Association (NFPA) Standard 805, Section 3.4.3, Training and Drills, was identified for the licensees failure to adequately assess the fire brigade performance during an announced fire drill conducted March 21, 2018. Specifically, the inspectors identified several fire brigade performance deficiencies, improvement items, and lessons learned that were not identified and documented in the licensees corrective action program during the fire drill critique as required by the licensees fire drill administrative control procedure. |
| 05000400/FIN-2018002-01 | 2018Q2 | Harris | Failure to Promptly Identify and Correct a Condition Adverse to Quality For a Through-Wall Leak in the ESW Screen Wash Piping | An NRC-identified Green NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Actions, was identified for the licensees failure to promptly identify and correct a condition adverse to quality involving through-wall leakage in the B train ESW screen wash piping. Specifically, on April 30, 2018, operators failed to initiate a work request or condition report after security personnel reported through-wall leakage in the B train ESW screen wash piping. No further follow-up or corrective actions were taken until May 3, 2018, when NRC inspectors identified the same through-wall piping leakage during a plant walkdown inspection and reported the degraded condition. |
| 05000400/FIN-2018002-07 | 2018Q2 | Harris | Minor Violation | A minor, self-revealing violation of TS 6.8.1.a, Procedures and Programs,was identified for failure to follow procedure AD-OP-ALL-0200, Clearance and Tagging. On April 7, 2018, while the plant was in Mode 3 at 0 percent power, the licensee isolated breaker DP-1A-1 circuit 28 in accordance with clearance OPS-1-18-5015-DEH MODS-0093. Isolating this breaker caused an unexpected auto start signal for both motor driven auxiliary feedwater (MDAFW) pumps for a loss of last running main feed pump despite the 1B main feedwater pump still being in operation. Both MDAFWs started and operators manually secured the 1B main feedwater pump to maintain proper feedwater flow to the steam generators. TS 6.8.1.a, requires, in part, that written procedures be implemented covering activities referenced in Regulatory Guide 1.33, Revision 2, dated February 1978, including safety-related activities carried out during operation of the reactor plant. Procedure AD-OP-ALL-0200, Section 5.5, step 4, states Clearance impacts must be evaluated to ensure that effects on systems and components outside of the boundary are identified and are acceptable, or properly dispositioned. Contrary to this requirement, the licensee did not identify that the isolation of breaker DP-1A-1 circuit 28 would cause the MDAFWs to auto start in Mode 3 when developing clearance OPS-1-18-5015-DEH MODS-0093. Screening: The violation is minor because the impact to the plant was minimal; the unit was in Mode 3 throughout the event, the reactor remained subcritical, and feedwater flow to the steam generators was not lost. Enforcement: Because the performance deficiency is minor, it will not be subject to enforcement action in accordance with the NRCs Enforcement Policy. The licensee entered this issue into their CAP as NCR 02196873. The associated LER is closed. |
| 05000259/FIN-2018001-03 | 2018Q1 | Browns Ferry | Failure to Implement Controls for Locked High Radiation Area (LHRA) Access | A self-revealing, Green, NCVof TS 5.7.2, was identified for the failure to control access to a LHRA. Specifically, a worker installed and climbed a ladder in the Unit 3 drywell without Radiological Personnel (RP) present. In doing so, the worker accessed an area with dose rates >1 rem/hr that had not been posted, locked, or surveyed prior to entry |
| 05000259/FIN-2018001-02 | 2018Q1 | Browns Ferry | Unauthorized Entry into a High Radiation Area(HRA | A self-revealing, Green, NCVof Technical Specification (TS)5.7.1, was identified for a worker who entered a HRA without proper authorization. Specifically, the worker entered the Unit 3 A & C Residual Heat Removal Heat Exchanger Room using an incorrect Radiation Work Permit and without being briefed on the radiological conditions. |
| 05000259/FIN-2018001-01 | 2018Q1 | Browns Ferry | Inadequate Post-Maintenance Testing of 4kV Breaker Stationary Switches | A self-revealing,Green, NCV of 10 CFR Part 50 Appendix B, Criterion V,was identified when the licensee failed to perform an adequate post-maintenance test in accordance with NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, the post maintenance testing on the 3C diesel generator output breaker did not ensure that all contacts on replacement stationary switches successfully changed state after installation. |
| 05000296/FIN-2018001-04 | 2018Q1 | Browns Ferry | Inadequate Configuration Control of High Pressure Coolant Injection (HPCI)ValveDesign Issues | A self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion III,was identified when the licensee failed to ensure adequate control of valve design configurations in accordance with NPG-SPP-9.3, Plant Modifications and Engineering Change Control Revision 6. Specifically, the licensee changed, over time, HPCI discharge valve yoke nut and bearing components contrary to original design without documenting or evaluating the changes |
| 05000269/FIN-2017004-01 | 2017Q4 | Oconee | Failure to Identify Sensitive Equipment During Modification Results in Loss of Safety Function | A self-revealing Green non-cited violation (NCV) of Oconee Nuclear Station Technical Specification (TS), Section 5.4, Procedures, was identified for the licensees failure to identify sensitive equipment in a work area that warranted implementation of compensatory measures as required by station procedure AD-EG-ALL-1180, Engineering Change (EC) Walkdowns. During the design and planning phase of a station modification, the licensee failed to identify sensitive components located in the subject work area and subsequently failed to implement adequate protective measures as defined in station procedures to prevent plant impacts during modification installation. The licensee entered this issue into their corrective action program (CAP) as nuclear condition report (NCR)02131608 and implemented corrective actions to identify other positionable components required for emergency power source operability that would require the use of protective measures, as defined by AD-OP-ALL-0204, Plant Status Control, in order to prevent inadvertent operation. The licensee created a formal Engineering department communication which included lessons learned from the event and familiarization with the EC walkdown checklist. The signs on the governor actuator cabinets were also revised to emphasize the sensitive nature of the equipment. The licensees failure to properly identify sensitive equipment and implement compensatory measures to prevent plant impacts as required by station procedure AD-EG-ALL-1180 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the loss of the emergency AC power path function for 11 hours and 31 minutes. The finding was assessed using IMC 0609, Attachment 4 and IMC 0609, Appendix A. Inspection Manual Chapter 0609, Appendix A required a detailed risk evaluation because the finding represented a loss of system and/or function. A regional senior reactor analyst (SRA) performed the detailed risk evaluation using SAPHIRE Version 8.1.6 and a modified Version 8.50 of the SPAR Model for Oconee. The SRA developed two change sets to model the total exposure time for the finding. The first simulated a common cause failure of both Keowee units with an exposure time of 7 hours. The second simulated the failure of both Keowee units while the standby buses were energized by the Lee Station for 5 hours. The result was less than 1E-6 for each Oconee unit, which would be a finding of very low significance (Green). The inspectors utilized IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014, and determined the finding had a cross-cutting aspect of work management in the area of human performance, in that the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. (H.5) |
| 05000287/FIN-2017004-02 | 2017Q4 | Oconee | Failure to Properly Risk Screen Work Within Two Feet of a Single Point Vulnerability Component | A self-revealing Green NCV of Oconee Nuclear Station TS, Section 5.4, Procedures, was identified for the licensees failure to identify and properly risk screen work within 2 feet of a single point vulnerability (SPV) component in accordance with procedure AD-OP-ALL-0201, Protected Equipment. Specifically, the transmission and Oconee organizations failed to recognize that planned maintenance on a breaker in the 525 kilovolt (kV) switchyard was within 2 feet of an SPV component and, as a result, appropriate planning and oversight were not in place to prevent a plant trip during maintenance activities. The licensee entered this issue into their CAP as NCR 02138958. Corrective actions included revisions to station and transmission procedures to ensure inclusion of appropriate SPV program information, addition of the SY special emphasis code to all switchyard type work which require coordination of transmission resources, and the addition of the T1 trip/transient risk special emphasis code to all breaker failure relays in the 230 kV and 525 kV switchyard cabinets containing SPV components.The licensees failure to identify and properly risk screen the planned maintenance on PCB-57 as work within 2 feet of an SPV component in accordance with AD-OP-ALL-0201 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, human errors led to a Unit 3 main generator lockout, which resulted in a reactor trip. The finding was assessed using IMC 0609, Attachment 4 and IMC 0609, Appendix A. The inspectors determined the finding was of very low safety significance (Green) because the finding did not represent a transient initiator that caused both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (i.e. loss of condenser, loss of feedwater). The inspectors utilized IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014, and determined the finding had a cross-cutting aspect of work management in the human performance area, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. (H.5) |
| 05000424/FIN-2017003-05 | 2017Q3 | Vogtle | Licensee-Identified Violation | 10 CFR 20.1501 requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Contrary to the above, on June 28, 2017, the licensee failed to evaluate radiological conditions in room 1- AB -C-94, Back flushable Filter Crud Tank Pump Room, following the tank being placed in recirculation by Operations. On July 2, 2017, during a routine survey of room 1- AB- C-94, general area dose rates in the area were found to be as high as 600 mrem/hr. On the previous survey, conducted on June 19, 2017, maximum dose rates were found to be as high as 60 mrem/hr. This finding was evaluated using IMC 0609, Appendix C, Occupational Radiation Safety SDP, and was determined to be of very low safety significance (Green) because the finding is not related to ALARA dose planning, did not result in an overexposure or the substantial potential for overexposure, and the ability to assess dose was not compromised due to the use of appropriate personnel dosimetry. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This issue was entered into the licensees corrective action program as CR 10383067. |
| 05000424/FIN-2017003-04 | 2017Q3 | Vogtle | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion XI, Test Control stated, in part, that test programs shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service. UFSAR Section 8.1.4.3.C.2 stated that the onsite electrical system was designed in accordance with IEEE 308 -1974, Criteria for Class 1E Power System at Nuclear Generating Stations. IEEE 308 -1974 Section 6.3 recommended periodic tests be performed at scheduled intervals to detect deterioration of equipment to demonstrate operability of the components that are not exercised during normal operation. Contrary to the above, the licensee did not establish adequate test control measures to assure that the protective function of all 1E lockout relays were periodically verified. Specifically, there was no preventative maintenance to test the 1E lockout relays for non- MSPI loads. This condition has existed since plant initial operation and was identified during a licensee Nuclear Oversight audit on July 13, 2017. The inspectors determined this finding was of very low safety significance (Green) because the inspectors found no documented history of in- service failures of 1E lockout relays rendering safety -related equipment inoperative. This issue was documented in the licensees corrective action program as CR 10381797. |
| 05000424/FIN-2017003-01 | 2017Q3 | Vogtle | Failure to Implement and Establish Appropriate Work Instructions Affecting Safety-Related Chiller | A Self -Revealing, Green, non- cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow measurement and test equipment (M&TE) in support of essential safety features (ESF) chilled water pumps in- service testing (IST). As a result, the Unit 1 A train safety -related chiller was inadvertently rendered inoperable when technicians isolated a flow transmitter associated with the chillers auto -start control logic when installing and removing M&TE in support of the IST. The licensee entered this issue into their corrective action program (CAP) under condition report (CR) 10390340 and corrective action report 270610 and planned to revise the procedure. Failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow M&TE in support of ESF chilled water pumps IST, which can affect ESF chiller performance, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD). 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because while the unit 1 A train ESF chiller was rendered inoperable, it did not represent a loss of function of the train for greater than its TS Allowed Outage Time. The finding was assigned a cross cutting aspect of Challenge the Unknown because questions and risks regarding the use of flow M&TE for the test were not properly evaluated and managed before proceeding. (H.11) |
| 05000425/FIN-2017003-03 | 2017Q3 | Vogtle | Failure to Maintain Cleanliness of Motor Operated Valve Limit Switch Compartment | A Self -Revealing , Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to perform an adequate cleanliness inspection of the Unit 2 nuclear service cooling water (NSCW) system pump no. 6 discharge motor -operated -valve (MOV) limit switch compartment, as required by the maintenance procedure. As result , the valve failed to operate when demanded and rendered the NSCW pump inoperable. The failure to perform an adequate cleanliness inspection of NSCW pump no. 6 discharge MOV limit switch compartment following preventive maintenance, as required by maintenance procedure NMP -ES- 017- 008, was a performance deficiency (PD). The licensee cleaned affected MOV sub -components, verified proper operation, and restored operability of the pump. This issue was entered into the licensees CAP as CR10399054 . The performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The finding was determined to be of very low safety significance (Green) because although the performance deficiency affected the qualification and operability of the NSCW pump, it did not represent a loss of function of an NSCW train for greater than its TS Allowed Outage Time . The finding was assigned a cross cutting aspect of Avoid Complacency, because maintenance technicians did not recognize the possibility of making mistakes when performing routine tasks of inspecting and manipulating grease containing components inside the limit switch compartment. (H.12) |
| 05000425/FIN-2017003-02 | 2017Q3 | Vogtle | Failure to Maintain ECCS Flow Balance and Check Valve Inservice Test Procedure | An NRC- Identified, Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to maintain a Unit 2 surveillance procedure that demonstrated satisfactory performance of the forward flow safety function of emergency core cooling system ( ECCS ) check valves. The licensee revised and performed the test to verify satisfactory valve performance. This issue was entered into the licensees CAP as CR10410794. The failure to maintain procedure 14721D -2 to ensure test conditions that adequately demonstrated satisfactory performance of ECCS check valves 2- 1205- U6 -001/00 2, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD ). The performance deficiency was more than minor because if left uncorrected, it could result in degradation of ECCS check valves to go undetected. The finding was associated with the mitigating system cornerstone. The finding was determined to be of very low safety significance (Green) because the performance deficiency did not result in a loss of operability or functionality of ECCS check valves. The finding was assigned a cross cutting aspect of Resources, because the licensee did not ensure that an ECCS surveillance procedure was adequate to support nuclear safety . (H.1) |
| 05000395/FIN-2017002-02 | 2017Q2 | Summer | Failure to Provide NRC Staff Complete and Accurate Information | The inspectors identified a severity level (SL) IV NCV of 10 CFR 50.9(a), Completeness and accuracy of information, involving licensee document,RC-13-0142, dated October 14, 2013. This document was a response to a request for additional information involving a license amendment request (LAR) to adopt NFPA 805 and contained an approval request, L12, associated with oil misting from the reactor coolant pumps. The licensee entered this violation into their corrective action program as CR-17-03961. The inspectors determined that the licensees failure to provide complete and accurate information associated with approval request, L12, was a violation of 10 CFR 50.9(a). Because this violation of 10 CFR 50.9(a) impacted the NRCs ability to perform its regulatory function, the inspectors evaluated this violation using traditional enforcement (TE). Since the TE violation is associated with a previous Green reactor oversight process violation, and the misinformation was identified after the NRC relied on it for issuing a previous operating license amendment, the TE violation was determined to be a SL IV, NCV, consistent with the language of the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information. This violation involved TE; therefore a cross-cutting aspect was not assigned. |
| 05000369/FIN-2017002-01 | 2017Q2 | Mcguire McGuire | Inadequate Survey Results in Unposted HRA | Green . A self -revealing Green non- cited violation (NCV) of 10 CFR 20.1501(a)(2) was identified for the licensees failure to conduct an adequate area radiation survey in Room 619 of the auxiliary building (waste gas decay tank (WGDT) room). Specifically, on April 19, 2016 , a high radiation area (HRA) was identifi ed near WGDT A in the WGDT room when a worker entering the area received a dose rate alarm on his electronic dosimeter (ED) and follow -up surveys revealed dose rates as high as 110 mrem/hr at 30cm. Also, as a result of the licensees failure to perform a survey, the area was not barricaded and posted in accordance with plant Technical Specification (TS) 5.7.1, High Radiation Area. The licensee immediately barricaded and posted the area as an HRA, performed an apparent cause evaluation to determine additional long term actions and entered the issue into their corrective action program as Nuclear Condition Report (NCR) 02021742. The licensees failure to conduct an area radiation survey to evaluate the magnitude and extent of radiation levels near WGDT A was a performance deficiency. This finding was determined to be more than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, failure to identify, post and control HRAs could allow workers to enter HRAs without knowledge of the radiological conditions in the area and receive unintended occupational exposure. The finding was evaluated using Inspection Manual Chapter (IMC) 0609 Appendix C, Occupational Radiation Safety Significance Determination Process. The finding was not related to the a s low as reasonably achievable (ALARA) planning, did not involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding involved the cross -cutting aspect of avoid complacency in the area of human performance because the possibility of significant dose rate changes in the WGDT room during startup was a latent issue for which the licensee failed to recognize and plan. (H.12) |
| 05000395/FIN-2017002-01 | 2017Q2 | Summer | Failure to Implement Corrective Actions to Restore Compliance for Previous NRC-identified Green NCV 05000395/2013003-03 | The inspectors identified a Green finding with a cited violation of Operating Licensee Condition 2.C.(18) for failure to ensure that conditions adverse to fire protection as noted in a previous NRC-identified Green NCV, 05000395/2013003-03, Failure to Adequately Design, Install and Maintain Oil Collection Devices for Reactor Coolant Pump Motors, were corrected. Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for (1) a failure to ensure an adequate design for the oil lift pump enclosure, and (2) a failure to have oil collection components for internally leaked oil dripping from the motor air discharge ductwork flange. The licensee entered the issue in their corrective action program as condition report CR-17-03962.The inspectors determined that the failure to implement corrective actions for the oil collection system to restore compliance was a performance deficiency (PD). The inspectors used IMC 0612 and determined that the PD was more than minor and therefore a finding because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. This finding has a credible impact on safety because the failure to adequately install, maintain and design the oil collection system presented a degradation of a fire confinement component which has a fire prevention function of not allowing an oil leak to reach hot surfaces. This finding had been evaluated and screened to a low safety significance (Green) and documented in the previous NRC-identified Green NCV, 05000395/2013003-03. Because the licensee failed to implement corrective actions and restore compliance in a timely manner, this violation is being treated as a cited violation, consistent with Section 2.3.3 of the NRC Enforcement Policy. The inspectors used IMC 0310 and determined this finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the organization failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance and restore compliance (P.3). |
| 05000413/FIN-2017002-01 | 2017Q2 | Catawba | 10 CFR 50.59 Evaluation of a Change to an Engineered Safety Features Actuation (ESFAS) Test Procedure | The inspectors selected the six operability determinations or functionality evaluations listed below for review based on the risk-significance of the associated components and systems. The inspectors reviewed the technical adequacy of the determinations to ensure that technical specification operability was properly justified and the components or systems remained capable of performing their design functions. To verify whether components or systems were operable, the inspectors compared the operability and design criteria in the appropriate sections of the technical specification and updated final safety analysis report to the licensees evaluations. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. Additionally, the inspectors reviewed a sample of corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the attachment. Unit 1, Following the 1A DG periodic test on April 18, 2017, a 10 drop per minute lube oil leak from the 1A engine driven lube oil pump was identified, CR 2117628 Unit 2, DG-2A Power-driven Potentiometer pre-position circuit, CR 2116554 Unit 2, Acoustic monitors for the pressurizer safety valves, CR 2125238 Unit 1, Faint abnormal odor on 1 DG control panel A, CR 2132262Unit 1, 1A EDG kilowatts spiked from 900 kw to 7000 kw when jumper was placed in circuit for testing, CR 2118549 Unit 2, Questions on auxiliary shutdown panel supply unit, CR 2124814 b. Findings(Opened) Unresolved item (URI): 10 CFR 50.59 Evaluation of a Change to an Engineered Safety Features Actuation (ESFAS) Test ProcedureIntroduction: The inspectors identified a URI associated with the implementation of a procedure change to the Unit 1 auxiliary shutdown panel (ASP) room air conditioning system. Additional information is needed to determine if a performance deficiency exists.Description: In November 2015, the licensee changed procedure PT/1/A/4200/009, Engineering Safety Features Actuation periodic test (ESFAS), to allow testing of the 1A ASP room air conditioning unit while it was unavailable. A URI was identified because the change verified the ESFAS Sequenced On light was lit, where the previous version of the procedure confirmed the air conditioning unit was running. The licensee did not perform a 50.59 evaluation, and the inspectors determined the change may affect the intent of the surveillance requirement.The licensee has initiated a 50.59 evaluation to determine the impact of the change relative to ESFAS testing requirements. The inspectors will review the completed evaluation to determine if a performance deficiency exists. The licensee documented this issue and background information in their corrective action program as CR 2124814. (URI 05000413/2017002-01, 10 CFR 50.59 Evaluation of a Change to an Engineered Safety Features Actuation (ESFAS) Test Procedure). |
| 05000261/FIN-2017001-01 | 2017Q1 | Robinson | Failure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak -Chase Test Connection | Green . An NRC- identified Green non -cited violation ( NCV ) of 10 CFR Part 50.55a, Codes and Standards, was identified for the failure to perform general visual examinations of moisture barriers in the containment leak -chase channel test connections in accordance with the American Societ y of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC), Section XI, Subsection IWE , Requirements for Class MC and Metallic Liners of Class CC Components of Light -Water Cooled Plants . Following the inspectors identification of this issue, t he licensee initiated actions to conduct the re quired visual examinations during the March 2017 refueling outage and initiated actions to revise the containment inservice inspection (ISI) plan such that the required examinations will be performed in the future . This issue was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 02109909. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME B PVC , Section XI, Subsection IWE , was a performance deficiency (PD) . The finding was of more than minor significance because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, visual examinations of mois ture barriers associated with the containment leak -chase channel test connections provide assurance that the containment metal liner and liner seam welds remain capable of performing its intended safety function. In the absence of such examinations, corro sive conditions at the moisture barrier (concrete -to-tubing interface) could go undetected. As a result, degradation of inaccessible portions of the containment liner could progress to challenge the containment operational capability. Using IMC 0609, A ttachment 4, Initial Characterization of Findings, the finding was determined to affect the Barrier Integrity Cornerstone because it involved ISI program examinations designed to identify degradation of the containment metal liner. The inspectors screen ed the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At -Power, Exhibit 3 Barrier Integrity Screening Questions, and determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the containment. The inspectors reviewed this performance deficiency for cross -cutting aspects as required by IMC 0310, Components With Cross -Cutting Aspects. The finding was determined to be reflective of present licensee performance because in 2014, the licensee did not take effective corrective actions to implement the ASME BPVC 3 requirements in the Subsection IWE P rogram , when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014- 07, which highlighted this industry -wide problem. Therefore, the finding was assigned a cross - cutting aspect in the resolution component of the problem identification and resolution cross -cutting area (P.3) |
| 05000261/FIN-2017001-03 | 2017Q1 | Robinson | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non- confor mances are promptly identified. Contrary to the above, in March 2014, while performing examinations in steam generator C during forced shutdown RFO229F3, the licensee failed to identify a loose part lodged in contact with tube R37C22. The licensee identified the loose part in March 2017 during refueling outage RO30. The licensee verified that indications of the part were detectable during RFO229F3, retrieved the part, verified that degradation caused by the part met all structural integrity requirement s, plugged the tube, and removed it from service. This issue was identified in the licensees CAP as NCR 0210725. The inspectors evaluated this violation using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings At -Power, and determined that the violation was of very low safety significance (Green) because evaluations demonstrated that the tube could sustain three times the differential pressure across it during normal full power steady state operation and that the steam generator did not violate the accident leakage performance criterion |
| 05000261/FIN-2017001-02 | 2017Q1 | Robinson | Failure to Submit Complete and Accurate Information for a Requested License Amendment | Severity Level IV. An NRC -identified severity level IV (SL IV) NCV of 10 CFR 50.9(a), Completeness and Accuracy of Information, was identified for the licensees failure to provide complete and accurate information in a license amendment request (LAR), dated November 19, 2015, requesting extension of the containment leak rate test frequencies required by various containment technical specifications (TS s). In this LAR, the licensee incorrectly stated that they had revised their ASME BPVC, Section XI, Subsection IWE program to include visual examinations of the test connections in the leak -chase channel penetration pressurization system ( PPS) , when in fact, the program had not been revised and the examinations had not been performed . This information was material to the NRC because it was used, in part, as the basis for the approval and issuance of License Amendment 247, dated October 11, 2016, extending the TS containment leak rate test frequencies. The licensees corrective actions included conducting the visual examinations of the test connections in the leak -chase channel PPS during the ongoing refueling outage in March 2017 and initiating actions to add the visual examination requirements to their Subsection IWE program. This issue was entered into the licensees CAP as NCR 02110516. The failure to provide complete and accurate information in accordance with 10 CFR 50.9(a) for the LAR associated with License Amendment 247 is a violation of NRC requirements . This violation was screened against the ROP guidance in IMC 0612, Appendix B, Issue Screening, and no associated ROP finding was identified. The inspectors evaluated this issue using the Traditional Enforcement process because it had the potential to impact the NRCs ability to perform its regulatory function. Specifically, the violation impacted the regulatory process, in that the inaccurate information was material to the NRCs review and acceptance of licensee actions to address the industry -wide operating experience discussed in NRC IN 2014- 07. Based on licensee inaccurate information that they had addressed IN 2014 -07 by revising their containment ISI program to perform visual inspections of accessible tubing in the containment leak -chase channel PPS system, the NRC staff concluded that the licensee was properly implementing the ASME BPVC, Section XI, Subsection IWE program. In accordance with the guidance in Sections 2.2 and 6.9 of the NRC Enforcement Policy, the inspectors determined this is an SL IV violation, because had the information been complete and accurate at the time provided, it likely would have resulted in the need for further clarification of the licensees actions to address NRC IN 2014- 07 , but would not have caused the NRC to change its decision to issue the license amendment or resulted in substantial further inquiry . Also, on March 23, 2017, the licensee completed the visual examinations of the subject tubing in the leak -chase channel system and did not identify any significant degradation. In accordance with IMC 0612, Appendix B, traditional enforcement issues are not assigned a cross -cutting aspect. |
| 05000259/FIN-2016004-01 | 2016Q4 | Browns Ferry | Inadequate Reassembly Procedure for HPCI Steam Line Inboard Isolation Valve Actuator | Green. A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" was identified for the licensee's failure to provide sufficient detail, in this case, appropriate to the work activity in procedure, MCI-0-000-ACT004, Maintenance of SMB-0 through SMB-4T Limitorque Actuators, which impacted the design features of HPCI valve 1-FCV-73-2. As an immediate corrective action, the valve was repaired and corrective actions initiated to address the quality and details of motor operated valve procedures. The licensee entered the violation into their corrective action program as Condition Reports (CRs) 1228056 and 1229289. 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the reliability of the valve was reduced due to the impending worm gear teeth failure. While the valve was full open, the High Pressure Coolant Injection (HPCI) pump was able to fulfill its safety function of injecting water into the reactor. Since the valve was able to close upon entering outage U1R11, the HPCI system was able to isolate the HPCI steam supply line in the event of a HPCI steam line break. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Mitigating Systems Screening Questions. The inspectors determined the finding screened to Green as HPCI was not unavailable longer than its TS allowed outage time and the finding did not involve the loss or degradation of equipment designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect of Procedure Adherence in the Human Performance area (H.8), because individual staff members did not review procedures and instructions prior to work to validate they were appropriate for the scope of work. |
| 05000296/FIN-2016004-02 | 2016Q4 | Browns Ferry | Inadequate Prompt Determination of Operability for the HPCI System | Green. An NRC identified NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings" was identified for the licensee's failure to accomplish the Prompt Determination of Operability (PDO) for CR 1039036 in accordance with the requirements of NEDP-22, "Operability Determinations and Functional Evaluations," Sections 3.2.2.E, 3.2.2.G, and Attachment 2. As an immediate corrective action, the licensee revised the PDO to include an evaluation that supported a reasonable expectation of operability. The licensee entered the violation into the corrective action program as CR 1219620. The performance deficiency was more-than-minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, after considering the inadequacies of the PDO, additional and significant evaluation was required to maintain reasonable assurance of the HPCI system operability. The doubt stemmed from uncertainty about the actual water level in the turbine, the expected transient severity, and the unanalyzed effects of the piping configuration. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Exhibit 2 "Mitigating Systems Screening Questions," dated June 19, 2012. The inspectors determined the finding was Green because it was a deficiency affecting the qualification of HPCI, but it maintained its operability. The inspectors determined that the finding had a cross-cutting aspect of Evaluation in the Problem Identification and Resolution area (P.2), because the organization did not thoroughly investigate this issue commensurate with its potential safety significance. |
| 05000413/FIN-2016003-01 | 2016Q3 | Catawba | Licensee-Identified Violation | The licensee identified a non-compliance with Operating License Condition 2.C.(5), for Units 1 and 2, for the failure to protect one of the redundant trains of equipment needed to achieve post-fire SSD from fire damage. Specifically, the licensee failed to use one of the means described in Branch Technical Position (BTP) Chemical Engineering Branch (CMEB) 9.5-1, Item C.5.b.2 to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. Description: On June 2, 2014, the licensee submitted LER 413/2014-002-00 with Revision 01 submitted on December 1, 2014, which documented discovery of cable routing issues and postulated fire-induced circuit failures that could prevent operation or cause maloperation of equipment required to achieve SSD in the event of a fire. This condition was identified during the licensees transition to National Fire Protection Association Standard 805 (NFPA 805). During the transition to NFPA 805, the licensee identified multiple instances of cables for equipment required to achieve SSD not meeting the separation requirements of the current licensing basis. The licensee determined that this condition existed for 22 fire areas (FAs) across both units. The licensee characterized these issues as variance(s) from deterministic requirements (VFDRs). The conditions identified in the LER are related to VFDRs that met the following criteria: 1) VFDRs that required a plant modification to meet the fire risk criteria of NFPA 805, or 2) VFDRs where a potential concern existed with respect to NRC Information Notice (IN) 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, dated February 28, 1992. The licensee determined that the deficiencies existed because of latent design deficiencies in the cable routing and circuit design. This LER was applicable to Units 1 and 2. Upon discovery, the licensee entered this issue into their corrective action program as PIP C-1401427, and implemented compensatory actions in the form of fire watches and/or control of transient combustible material for the affected FAs. Analysis. Failure to protect one redundant train of cables and equipment necessary to achieve post-fire SSD from fire damage was a performance deficiency. This finding was more than minor because it was associated with the reactor safety mitigating system cornerstone attribute of protection against external events (i.e., fire). Specifically, failure to protect safe shutdown cables and equipment from fire damage negatively affected the reactor safety mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this issue relates to fire protection and this noncompliance was identified as a part of the sites transition to NFPA 805, this issue is being dispositioned in accordance with Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) of the NRC Enforcement Policy. In order to verify that this non-compliance was not associated with a finding of high safety significance (Red), inspectors reviewed qualitative and quantitative risk analyses performed by the licensee. These risk evaluations took ignition source and target information from the licensees fire probabilistic risk assessment to demonstrate that the significance of the non-compliances were less-than-Red (i.e. CDF less than 1E-4/year). Inspectors determined that cables associated with some of the VFDRs were not located in the zone of influence (ZOI) of any credible ignition source. For cables that were located in the ZOI of a credible ignition source, inspectors were able to perform a calculation to determine the change in conditional core damage probability (CCDP), based on the postulated fire-affected equipment not being available. Based on these screenings, inspectors determined that the significance of this non-compliance was lessthan-Red. A bounding risk assessment performed by a regional Senior Risk Analyst (SRA) reviewed the licensee and inspector risk evaluations and confirmed the CDF risk increase due to this condition was less than 1E-4, and therefore less than RED. The inspectors determined that no cross cutting aspect was applicable to this performance deficiency because this finding was not indicative of current licensee performance. Enforcement. Operating License Condition 2.C.(5), for Units 1 and 2, requires that the licensee implement and maintain in effect all provisions of the approved FPP as described in the UFSAR, as amended, for the facility and as approved in the SER through Supplement 5. BTP CMEB 9.5-1, which incorporated the guidance of Appendix A to BTP ASB 9.5-1 and the technical requirements of Appendix R to 10 CFR 50, established the regulatory and licensing requirements for the FPP at Catawba Nuclear Station (CNS). The CNS FPP was reviewed against and approved for conformance with BTP CMEB 9.5-1 in the SER through Supplement 5. BTP CMEB 9.5-1, Item C.5.b.1, requires that fire protection features be provided that are capable of limiting fire damage so that one train of systems necessary to achieve and maintain hot standby conditions from either the control room or emergency control station(s) is free from fire damage. BTP CMEB 9.5- 1, Item C.5.b.2 requires one redundant train to be protected from fire damage by one of the following specified methods: (a) separation of cables and equipment by a fire barrier having a 3-hour rating, (b) separation of cables and equipment by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards and with fire detectors and an automatic fire suppression system in the fire area, or (c) enclosure of cables and equipment in a fire barrier having a 1-hour rating and with fire detectors and an automatic fire suppression system in the fire area. Contrary to the above, the licensee failed to use one of the means described in BTP CMEB 9.5-1, Item C.5.b.2 to ensure that one of the redundant trains of equipment necessary to achieve and maintain hot shutdown conditions was protected from fire damage. Specifically, on April 2, 2014, the licensee identified the failure to protect equipment in accordance with the current licensing basis. The licensee determined that fire damage could prevent operation of, or cause maloperation of, components that were required to achieve and maintain SSD. This condition has existed since initial plant startup for Units 1 and 2. The licensee entered this issue into the corrective action program (PIP C-14-1427) and implemented compensatory measures in the form of fire watches and/or control of transient combustible material for the affected FAs. Because the licensee committed to adopt NFPA 805 and change their fire protection licensing bases to comply with 10 CFR 50.48(c), the NRC is exercising enforcement and reactor oversight process (ROP) discretion for this issue in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. Specifically, this issue was identified and will be addressed during the licensees transition to NFPA 805, it was entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, it was not likely to have been previously identified by routine licensee efforts, it was not willful, and it was not associated with a finding of high safety significance (Red). |
| 05000287/FIN-2016002-03 | 2016Q2 | Oconee | Degraded power cables result in inoperable startup transformer and loss of Unit 3 safety function | A self-revealing Green violation of Oconee Technical Specification 5.4, Procedures, was identified for the licensees failure to establish adequate procedures to detect degradation of the startup transformer power cables. Station procedure IP/0/A/2400/002, Substation Insulators, Lighting Arrestors, CCVT, Transformer Drop Down Line, Bus Inspection and Maintenance, lacked sufficient detail for maintenance personnel to properly inspect power cables for cracks and fraying. This allowed undetected degradation of the Oconee startup transformer power cables to develop causing the Unit 3 startup transformer to become inoperable. The licensee performed repair activities on the degraded power cables to remove areas where strands of the power cables were severed and re-established proper connections. Also, the licensee created work orders in their work management process to replace the drop down lines on the Unit 1 and Unit 3 startup transformers. The licensee entered this issue into their corrective program as NCR 01733811. The licensees failure to establish an adequate procedure to detect degradation of startup transformer power cables during periodic maintenance was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the power cable failure caused inoperability of the Unit 3 startup transformer. The finding was screened in accordance with IMC 0609, Significance Determination Process, Attachment 4 and Appendix A and determined to require a detailed risk evaluation. A senior reactor analyst performed a detailed risk evaluation of this condition and determined delta CDF was 3E-7 (Green). The finding was determined to have a cross-cutting aspect of evaluation in the problem identification and resolution cross-cutting area because the licensees corrective actions resulting from a degraded power cable in 2002 failed to incorporate sufficient detail into their procedures necessary to detect frayed cables. |
| 05000250/FIN-2016002-01 | 2016Q2 | Turkey Point | Failure to Update the Final Safety Analysis Report with Applicable Safety System Criteria | NRC inspectors identified a SL IV, NCV of 10 CFR 50.71(e), Maintenance of Records, Making of Reports. The licensee failed to include Eagle 21 licensing basis information into the Updated Final Safety Analysis Report (UFSAR). The Eagle 21 licensing basis information was specified in License Amendment (LA) numbers 135 (Unit 3) and 140 (Unit 4). The licensee entered the issue into their corrective action program (CAP) as action request (AR) 2048916 to update the UFSAR with the design and licensing basis for the Eagle 21. The failure to update the UFSAR was a performance deficiency that was determined to be minor because it did not meet the more than minor screening criteria. Because the issue impacted the NRCs ability to perform its regulatory process, the inspectors evaluated the violation using the traditional enforcement process. The inspectors determined the issue was a SL IV violation because it met violation example 6.1.d.3. The violation represented a failure to update the Final Safety Analysis Report (FSAR) as required by 10 CFR 50.71(e), but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. Cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000287/FIN-2016002-04 | 2016Q2 | Oconee | Failure to Make a Non-Emergency Eight Hour Notification of a Loss of Safety Function | An NRC-identified Severity Level IV NCV of 10 CFR 50.72(b)(3)(v) was identified for the licensees failure to make a required non-emergency eight hour notification for a loss of the emergency AC power path function. On December 7, 2015 Oconee Nuclear Station Unit 3 experienced a loss of the emergency AC power path function for approximately 21 minutes. The licensee entered this issue into their corrective action program as NCR 01981762 and will evaluate their internal reportability procedures regarding the time of discovery. The failure to make an eight hour non-emergency report for a loss of the emergency AC power path function per 10 CFR 50.72(b)(3)(v) was a performance deficiency. This performance deficiency impacted the ability of the NRC to perform its regulatory oversight function and was dispositioned using traditional enforcement. This violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy, revised February 4, 2015. Using the example listed in Section 6.9.d.9, A licensee fails to make a report required by 10 CFR 50.72, the issue was determined to be a Severity Level IV violation. In accordance with IMC 0612, because this violation involved traditional enforcement and does not have an underlying technical violation that would be considered more than minor, a cross-cutting aspect was not assigned to this violation. |
| 05000287/FIN-2016002-01 | 2016Q2 | Oconee | Failure to Perform ISI General Visual Examinations of Containment Moisture Barrier | An NRC-identified Green NCV of 10 CFR Part 50.55a, Codes and Standards, was identified for the licensees failure to conduct 100 percent general visual examinations of the moisture barriers to the containment liner in accordance with Subsection IWE of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI. Specifically, the licensee failed to conduct visual examinations of the sealant applied to interior expansion joint locations in containment. In response, the licensee repaired the identified moisture barriers and confirmed the operability of the containment liner with the satisfactory results of the containment integrated leak rate test. The licensee entered this issue into their corrective action program as NCR 02027086. The failure to conduct a general visual examination of 100 percent of the moisture barriers intended to prevent intrusion of moisture against inaccessible areas of the containment liner was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the inspectors determined that this finding was of more than minor significance because the failure to conduct required visual examinations and identify the degraded moisture barriers, which could allow the intrusion of water, if left uncorrected, had the potential to lead to a more significant concern. The inspectors used IMC-0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, Exhibit 3 Barrier Integrity Screening Questions, and determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined no cross-cutting aspect was associated with this finding because the finding was not reflective of present licensee performance. |
| 05000269/FIN-2016002-02 | 2016Q2 | Oconee | Failure to Properly Control Transient Combustible Materials in the Oconee Main Control Rooms | An NRC-identified Green non-cited violation (NCV) of Oconee Nuclear Station Units 1, 2, and 3 Renewed Facility Operating License Condition 3.D, Fire Protection, was identified for the licensees failure to adequately implement the requirements of the transient combustible material program. Specifically, the licensee failed to control the storage of transient combustible material in the Oconee main control rooms with the proper evaluation in accordance with procedure AD-EG-ALL-1520, Transient Combustible Control, Attachment 3, Allowed Combustible Materials in Level B and Level C Areas. The licensee removed the stored items from each of the main control rooms and entered this issue into their corrective program as nuclear condition reports (NCRs) 02012091, 02012290, and 02013990. The licensees failure to control the storage of transient combustible material in the Oconee main control rooms with the proper evaluation in accordance with procedure AD-EG-ALL-1520 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, uncontrolled transient combustibles challenge the habitability requirements of the main control room in the event of a fire and the ability of licensed operators to respond to events using the systems designed to prevent undesirable consequences. The finding was screened in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings and IMC 0609 Appendix F, Fire Protection Significance Determination Process Task 1.3.1, and determined to be of very low safety significance (Green) because the finding did not prevent the reactor from reaching and maintaining a safe shutdown condition. The finding was determined to have a cross-cutting aspect of procedure adherence in the human performance cross-cutting area because the licensee failed to implement the requirements of station procedure AD-EG-ALL-1520, Transient Combustible Control. |
| 05000250/FIN-2016002-02 | 2016Q2 | Turkey Point | Failure to Correct Conditions Adverse to Quality Associated with the Eagle 21 System | NRC inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for a failure to correct a condition adverse to quality. The licensee identified that the ability to test the Eagle 21 was degraded but failed to take adequate corrective actions to correct the condition. The licensee entered the issue into their CAP as ARs 2023314 and 02145155. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not using core operating limits report (COLR) specified time-constants in surveillance requirement (SR) tests to demonstrate operability of the Eagle 21 system adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the Thermal Over-Power (OPT) and Thermal Over-Temperature (OTT) reactor trip algorithms. The finding was determined to be of very low safety significance (Green) because defense in depth of the reactor protection system (RPS) existed to trip the unit via alternate and diverse means. The inspectors determined the finding was indicative of present licensee performance and was associated with the cross-cutting aspect of human performance, in the area of conservative bias, because individuals failed to evaluate a proposed action to determine if it was safe in order to proceed, rather than unsafe in order to stop (H.14). |
| 05000250/FIN-2016002-03 | 2016Q2 | Turkey Point | Failure to Post a High Radiation Area | A self-revealing, Green, NCV of Technical Specification (TS) 6.12.1, was identified by health physics inspector(s) for the failure to post a high radiation area (HRA). Specifically, on April 6, 2016, the licensee failed to post the area by the exterior wall of the Unit 4 spent fuel pool (SFP) on the Auxiliary Building roof as a HRA. This performance deficiency was determined to be more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to post and control HRAs can allow workers to enter HRAs without knowledge of the radiological conditions in the area and result in the receipt of unintended occupational exposure. The finding was evaluated using the Occupational Radiation Safety SDP. The finding was not related to the As Low As Reasonably Achievable (ALARA) planning, did not involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding involved the cross-cutting aspect of Human Performance, Work Management (H.7) because the organization failed to implement its process for planning and controlling access to HRAs on the Auxiliary Building roof when fuel bundle movement were still ongoing. The violation was entered into the licensees CAP as AR 02123851. |
| 05000296/FIN-2016001-08 | 2016Q1 | Browns Ferry | Licensee-Identified Violation | Licensee Event Report (LER) 05000296/2014-003-00 Primary Containment Isolation Valve Inoperable for Longer Than Allowed by Technical Specifications 10 CFR 50, Appendix B, Criterion 5 required, in part, that activities affecting quality be implemented in accordance with documented procedures and drawings. Contrary to the above, between March 7, 2014 and June 6, 2014, relay 3-RLY-074-10A-K98A was wired incorrectly as discussed in LER 05000296/2014-003-00. The licensee corrected the wiring and entered the issue into the licensee's corrective action program as CR 892500. Inspectors screened the violation using IMC 0609, Appendix G, Attachment 1, Exhibit 3 Mitigating Systems Screening Questions, dated May 9, 2014. Because the finding degraded a functional auto-isolation of RHR on low reactor water level, a Phase 2 screening was required. Using attachment 3, Phase 2 Significance Determination Process Template for BWR During Shutdown, dated February 28, 2005, inspectors completed Worksheet 1 for Loss of Inventory in Plant Operating State 1 (Head On) and determined the risk was approximately 1e-7/yr, which was less than the 1e-6/yr threshold for a greater than Green finding. The dominant core damage sequence was the failure to isolate a reactor coolant leak and subsequent failure by operators to open vent paths (e.g. a safety relief valve) to control RCS pressure to enable continued low pressure injection. In the evaluation, no operator recovery credit was given for leak isolation, but credit was given for the redundant isolation valve that was operable which could have satisfied the automatic isolation function. The Regional Senior Reactor Analyst performed a detailed risk review of the finding. The risk review considered both the outage related risk, and the risk associated with a trip from power that would have the plant in shutdown cooling during the recovery. A screening analysis using bounding assumptions and the risk models ISLRHR event tree was performed. The dominant cutsets involved failure of the redundant valve to operate, and operator actions to recover. Because of the short exposure time during the shutdown periods, the redundant valve with the automatic action available, and the availability of operator recovery, the Finding was determined to be Green. This violation is being treated as an NCV consistent with Section 2.3.2 of the Enforcement Policy. |
| 05000296/FIN-2016001-06 | 2016Q1 | Browns Ferry | Failure to Identify Applicable Technical Specification Action Statement for a PCIV | An NRC identified non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures, for the licensees failure to implement OPDP-8, Operability Determinations and LCO Tracking. Specifically, the licensee failed to track the applicability of condition A of TS LCO 3.6.1.3 upon discovery of the equipment failure related to the Residual Heat Removal (RHR) Shutdown Cooling (SDC) inboard suction valve as described in LER 05000296/2014-003-00. As an immediate corrective action, the licensee entered the violation into the corrective action program as CR 1115172. The performance deficiency was more-than-minor because, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, this failure was indicative of a programmatic weakness with the licensees evaluation of certain logic circuit failures which can result in misapplication of the allowances of TS LCO 3.0.6 and inappropriate TS LCO entries. The inspectors determined that this type of error was likely to recur which could lead to worse errors if uncorrected. The inspectors determined the finding was Green because the error did not result in an actual open pathway in the physical integrity of reactor containment, containment isolation system or heat removal components. The inspectors determined that the finding had a crosscutting aspect of Training in the area of Human Performance because the finding was indicative of a knowledge gap among the operations department (H.9). |
| 05000259/FIN-2016001-05 | 2016Q1 | Browns Ferry | Failure to Include the Correct Proper Shipping Name on Radioactive Material Shipping Papers | The inspectors identified a NCV of 10 CFR 71.5 for the failure to include the correct Proper Shipping Name (PSN) on radioactive material shipping papers in accordance with the requirements of Department of Transportation (DOT) regulation 49 CFR 172.202. This resulted in multiple Low Specific Activity (LSA) shipments containing quantities exceeding an A2 value being shipped as UN2915, Radioactive Material, Type A Package. The licensee documented this issue in CR 1145617 and took immediate corrective actions including updating the software used to perform shipping activities and additional training of personnel. The performance deficiency was greater than minor because it was associated with the Public Radiation Safety Cornerstone, Program & Process attribute (transportation program), and adversely affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors determined the finding to be of very low safety significance (Green) because the issue involved transportation, but there were no radiation limits exceeded, and there was no package breach. In addition, it did not involve a Certificate of Compliance or low-level burial problem, nor was there a failure to make notifications or provide emergency response information. The finding has a cross-cutting aspect in the area of Human Performance, Training (H.9), because the DOT requirements pertaining to LSA shipments were not well understood. |
| 05000259/FIN-2016001-04 | 2016Q1 | Browns Ferry | Unposted High Radiation Areas | A self-revealing, NCV of 10 CFR 20.1902(b), with two examples, was identified for the failure to post multiple HRAs. Specifically, areas within the Unit 2 (U2) Control Rod Drive Rebuild Room and U2 Reactor Water Cleanup Holding Pump Room contained dose rates exceeding 100 mrem/hr at 30 cm and remained unposted for several months during 2015. These issues were entered into the licensees corrective action program as CR 1017294, CR 1023385, and CR 1119944, and the licensee took immediate corrective actions to correctly post the areas, performed surveys to evaluate the extent of condition, and performed an Apparent Cause Evaluation. The performance deficiency was greater than minor because it was associated with the Occupational Radiation Safety cornerstone attribute of Program and Process (Monitoring and RP Controls) and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors determined the finding to be of very low safety significance (Green) because it was not related to As Low As Reasonably Achievable (ALARA) planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. This finding involved the cross-cutting aspect of Human Performance, Documentation (H.7) because the unposted high radiation areas were a direct result of the failure to identify documented radiological conditions that required additional posting and control. |
| 05000259/FIN-2016001-03 | 2016Q1 | Browns Ferry | Unauthorized Entry into a High Radiation Area | A self-revealing, Non-cited Violation (NCV) of Technical Specification (TS) 5.7.1, was identified for a worker who entered a High Radiation Area (HRA) without proper authorization. Specifically, the worker entered a posted HRA located outside the Radwaste Ventilation Equipment Room without receiving a HRA briefing, and subsequently received a dose rate alarm. This issue was entered into the licensees corrective action program as Condition Report (CR) 1072342, and the licensee took immediate corrective actions including surveys of the area, and restricting the workers access to the Radiologically Controlled Area. The performance deficiency was greater than minor because it was associated with the Occupational Radiation Safety cornerstone attribute of Program and Process (Monitoring and Radiation Protection (RP) Controls) and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors determined the finding to be of very low safety significance (Green) because it was not related to As Low As Reasonably Achievable (ALARA) planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. This finding involved the cross-cutting aspect of Human Performance, Procedural Adherence (H.8) because the event was a direct result of the workers failure to adhere to requirements for HRA access. |
| 05000260/FIN-2016001-01 | 2016Q1 | Browns Ferry | Unacceptable Preconditioning of RCIC Valve Prior to ASME In-Service Testing | An NRC identified finding (FIN) for failure to meet TVA procedure NETP-116.3, Inservice Testing Program Preconditioning Guidelines, because unacceptable preconditioning of the Unit 2 Reactor Core Isolation Cooling (RCIC) steam supply valve occurred prior to quarterly In-Service Test (IST). Specifically, the preconditioning was unacceptable because the testing sequence was avoidable, it masked the actual asfound condition of the valve, and it could possibly result in an inability to verify the operability of the valve. As an immediate corrective action, the licensee performed an evaluation that determined the valve remained operable. The finding was entered into the licensee's corrective action program as CR 1159463 . 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 and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Additionally, if left uncorrected, the performance deficiency could lead to a more significant safety concern. Specifically, the licensees justification of this particular preconditioning event could be applied to justify additional, avoidable, preconditioning events and possibly result in an inability to verify the operability of components. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined the finding was Green because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its TS allowable outage time, did not result in a loss of function of non-TS equipment, and did not involve the loss of equipment or function specifically designed to mitigate an external event. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance area of Consistent Process (H.13), because individuals did not complete the required preconditioning evaluation forms described in licensee procedure NETP-116.3, which would have challenged the validity of the licensees original determination of acceptability. |
| 05000269/FIN-2016007-01 | 2016Q1 | Oconee | Pressure Boundary of Motor Operated Valves Could be Breached Due to Fire- Induced Hot Short | An unresolved item was identified regarding the licensees evaluation of certain motor operated valves (MOVs) in the NSCA. Specifically, based on the conclusions in the licensees NSCA, as well as subsequent evaluations, MOVs that are subject to a hot short that bypasses the torque or limit switch could result in damage to the valve that causes an unmitigated loss of reactor coolant system (RCS) inventory due to leakage through the damaged valves pressure boundary or the valves associated sealing components. Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, stated that fire damage could cause an electrical hot short that bypasses thermal overload protection for MOVs, and that this hot short could result in damage to the valve. As a part of the licensees transition to NFPA 805, the licensee identified a number of MOVs that could be susceptible to IN 92-18 type damage. These valves were classified as non-compliant components or variances from deterministic requirements (VFDRs). The subsequent evaluation of these valves by the licensees Fire PRA group determined that these VFDRs met the acceptance criteria of the Fire Risk Evaluation, as documented in OSC-9314, as being acceptable "as-is" and that no further action was required. These VFDRs and their FPRA dispositions were communicated to the NRC in the April 2010 Oconee NFPA 805 license amendment request (LAR). Subsequent to NRCs issuance of the SER, Oconee Valve Engineering determined that, due to the size of the installed motor/gearbox, 10 MOVs could potentially suffer IN 92-18 damage to the extent that the integrity of the valve body or bonnet could be compromised. Loss of valve integrity of the valve pressure boundary was not an assumption used in the FPRA evaluation. The licensee documented this condition in AR 01906086. After further evaluation, the licensee documented in AR 01999309 that 9 of the original 10 valves identified could potentially suffer IN 92-18 damage to the extent that the integrity of the valve body or bonnet could be compromised. For the 9 affected valves, the licensee has undertaken additional evaluations to determine whether some portion of the valve would fail before the valves pressure boundary is compromised, or that any possible leakage that may result can be bounded by the credited RCS make-up sourcein this case, the reactor coolant make-up pump. Inspectors determined that no immediate safety concern existed with this item because the licensee had implemented compensatory measures in accordance with the sites approved FPP. This item is unresolved pending inspector receipt and review of the licensees evaluation. |
| 05000259/FIN-2016001-07 | 2016Q1 | Browns Ferry | Licensee-Identified Violation | Licensee Event Report (LER) 05000259/2015-005-00 Inboard Main Steam Isolation Valve Actuators Inoperable for Longer Than Allowed by Technical Specifications. TS 3.6.1.3 condition A required, in part, that when one or more penetration flow paths with one Primary Containment Isolation Valve (PCIV) inoperable except due to MSIV leakage not within limits that within 4 hours the affected penetration flow path be isolated by use of at least one closed and de-activated automatic valve with flow through the valve secured. TS 3.6.1.3 condition E required, in part, that when the Required Action and associated Completion Time of Condition A was not met in MODE 1, that the Unit must be placed in Mode 3 within 12 hours and Mode 4 within 36 hours. Contrary to the above, on multiple occasions between December 1, 2012 and October 29, 2015, the inboard MSIVs PCIV function was inoperable on all main steam lines on all three Units longer than the allowed outage time and the follow on action completion time. This violation is documented in the licensees CAP as CR 1098857. This finding was screened to Green using IMC 0609 Appendix H dated May 6, 2004. Table 6.2 Phase 2 Risk Significance was used to screen the finding to Green because at no point during the time period between December 1, 2012 and October 29, 2015 did any outboard MSIV leakage on any Unit exceed 10,000 scfh. |
| 05000259/FIN-2016001-02 | 2016Q1 | Browns Ferry | Failure to adequately maintain emergency plan implementing procedures | The inspectors identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(2), for the licensees failure to maintain the effectiveness of its emergency plan by ensuring procedures for use by the emergency response organization are maintained and up-to-date as required by 10 CFR 50.47(b)(16). Corrective actions already taken were implementation of a revision (49) to EPIP-5, effective January 7, 2016, essentially replacing Section 3.6 and references to appropriate Appendices, and a broader scope EOC to review all site EPIPs to ensure no other inadvertent omissions were made. The inspectors determined that the performance deficiency was more than minor because it was associated with the procedure quality attribute of the Emergency Preparedness (EP) cornerstone, adversely affected the associated cornerstone objective, and may have been used had an emergency been declared. The finding was evaluated using the EP significance determination process and was identified as having very low safety significance (Green) because it was a failure to comply with NRC requirements and was not a loss of the planning standard function. The finding was associated with a cross-cutting aspect in the Evaluation component of the Problem Identification and Resolution area because the licensee failed to thoroughly evaluate a similar issue at one of its other sites to ensure extent of conditions commensurate with their safety significance are thoroughly resolved. (P.2) |
| 05000296/FIN-2016013-02 | 2016Q1 | Browns Ferry | Records of Fire Watch Patrol Not Complete or Accurate | 10 CFR 50.9, Completeness and Accuracy of Information, states, in part, information required by the Commissions regulations, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. Contrary to the above, on multiple occasions in May 2015, TVA maintained records of hourly fire watch patrols that were not complete and accurate in all material respects. Specifically, fire watch patrol documentation required by NPG-SPP-18.4.6 contained incomplete documentation of fire watches for a fire impairment in the Unit 3 Diesel Building and Unit 3 4kV Shutdown Board Room. Procedure NPG-SPP-18.4.6, Rev. 0006, Control of Protection Impairments, requires the licensee to complete Attachment 6, NPG-SPP-18.4.6-6, Hourly Compensatory Fire Watch Route Sheet, by entering the time, printing name, and signing as each area is patrolled, and returning it to the foreman at the end of the shift. The procedure allows the use of an equivalent form which shall include at least: impairment number, date of coverage, area covered, and start time. However, the equivalent form intended to document completion of the fire watch patrols of the Unit 3 Diesel Building and Unit 3 4kV Shutdown Board Room in May 2015, NPG-SPP-18.4.6-2(12-10-2010), Roving Fire Watch Route/Coverage Sheet, was not filled out completely. Specifically, the start/stop times and employees initials were not entered. Complete hourly fire watch patrol data is material to the NRC in that it provides evidence of compliance with the regulatory requirements of 10 CFR 50.48. Therefore, there was an apparent violation of 10 CFR 50.9 because the record the licensee is required to maintain to document completion of hourly fire watch patrols in the Unit 3 Diesel Building and Unit 3 4kV Shutdown Board Room was incomplete. |
| 05000259/FIN-2016001-10 | 2016Q1 | Browns Ferry | Licensee-Identified Violation | Licensee Event Report (LER) 05000259/260/296/2015-004-00 Containment Atmosphere Dilution B Train Supply System Inoperable Longer Than Allowed by Technical Specifications: Technical Specification LCO 3.6.3.1, Containment Atmosphere Dilution System, Condition B required that when Two CAD subsystems are inoperable that the licensee verify by administrative means that the hydrogen control function is maintained and to restore one CAD subsystem to OPERABLE status within 7 days. Condition C required action to place the affected unit in Mode 3 within 12 hours if the Condition B completion time was not met. Contrary to Technical Specification LCO 3.6.3.1 condition C, completion times were not met to place the units in Mode 3 within 12 hours when both trains of CAD were considered unavailable. This licensee identified violation is documented in the licensees CAP as CR 1087766. This finding was screened to Green using IMC 0609 Appendix H,Table 6.1 because the finding did not affect any of the listed Systems, Structures, or Components important to LERF. |
| 05000296/FIN-2016001-09 | 2016Q1 | Browns Ferry | Licensee-Identified Violation | Licensee Event Report (LER) 05000296/2015-002-00 Switch Failure Rendered Automatic Startup of Some Emergency Core Cooling System Pumps Inoperable Longer than Allowed by Technical Specifications: TS 3.3.5.1 condition A required, in part, that when one or more channels of Emergency Core Cooling System (ECCS) Instrumentation were inoperable that the condition listed in table 3.3.5.1-1 be immediately entered for that channel. MJ(STA 52) switch on breaker BFN-3-BKR-211-03ED/008 failed rendering automatic start sequence timing for the 3B and 3D Core Spray pumps, the 3D RHR Pump, and the D1 RHRSW Pump sequence time to become inoperable for conditions where normal power was maintained. This resulted in the licensee not meeting the TS completion times from September 17, 2014 until January 24, 2015, for TS 3.3.5.1 condition C (Core Spray Pumps 3B and 3D), TS 3.5.1 condition B (3D RHR pump), and TS 3.7.1 condition G (D1 RHRSW pump). This licensee identified violation is documented in the licensees CAP as CR 980277. This finding was able to be screened to Green using IMC 0609 Appendix A dated June 9, 2012 because although these pumps were inoperable, their respective systems did not lose their function as emergency starts were not affected. |
| 05000250/FIN-2015004-02 | 2015Q4 | Turkey Point | Licensee-Identified Violation | 10 CFR 50.48 states that each operating nuclear power plant must have a fire protection plan that satisfies Criterion 3 of Appendix A of this part. Turkey Point Renewed Operating License condition D, for Units 3 and 4, states that the licensee shall implement and maintain in effect all provisions of the approved FPP as described in the UFSAR Appendix 9.6A. The approved FPP is implemented, in part, by 0-ADM-016, Fire Protection Program, as referenced in Section 7.2 of UFSAR Appendix 9.6A. Section 5.6 of 0-ADM-016 requires that, for non-functional post-fire safe shutdown components, engineering evaluations should identify appropriate compensatory actions, including hourly fire roves. Contrary to the above, between May 1st, 2014, and April 23rd, 2015, hourly fire watch patrols were not conducted on numerous occasions in fire zones that required regular hourly tours due to fire protection equipment impairment. The failure to perform the fire watch tours did not cause the inoperability of any equipment but resulted in the loss of a defense-in-depth feature for fire detection in fire zones affected by an impaired or non-functional fire safety component or feature. This violation was associated with the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of the systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding to be of very low safety significance (Green) after performing a detailed risk evaluation in accordance with Manual Chapter 0609, Appendix A, because the missed fire watch tours reflected a low degradation of the Fire Prevention and Administrative Controls FPP element in that other area fire protection defense-in-depth features such as automatic fire detection (smoke detectors), automatic fire suppression capability (sprinklers), manual suppression capability (fire brigade), and safe shutdown capability from the main control room were still available. The licensee entered this violation into their CAP as AR 02056905. |