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05000390/FIN-2018003-072018Q3Watts BarLicensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Watts Bar Unit 1 TS LCO 3.8.7, Inverters-Operating, requires that two inverters in each of the four channels shall be operable. Contrary to the above, the licensee failed to ensure that two inverters in each of the four channels were operable. Specifically, from April 9, 2017 to January 10, 2018 inverter 1-II was inoperable due to an unqualified class 1E capacitor associated with the inverter.
05000390/FIN-2018003-062018Q3Watts BarLicensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Watts Bar Unit 1 TS 3.8.1, AC Sources - Operating, Condition A, requires, in part, that an inoperable required offsite circuit be restored to operable status within 72 hours. Contrary to the requirements of Technical Specification 3.8.1, a required offsite circuit was determined to be inoperable from May 27, 2017, to June 2, 2017.
05000390/FIN-2018003-052018Q3Watts BarLicensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Watts Bar Nuclear Plant (WBN) Unit 1 Operating License Number NPF-90, Condition 2.F, requires, in part, that TVA shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Fire Protection Report for the facility, as approved in Appendix FF Section 3.5 of Supplement 18 and Supplement 29 of the SER (NUREG-0847). The WBN Fire Protection Report was developed for WBN to ensure compliance with the requirements of this license condition. Fire Protection Report, Part II, is the Fire Protection Plan. The Fire Protection Plan, Section 14, Fire Protection Systems and Features Operating Requirements (ORs), Subsection 14.10, Fire Safe Shutdown Equipment, paragraph 14.10.4, requires a fire watch to be established in auxiliary building room 757-A10 within one hour of closing pressurizer block valve 1-FCV-68-332-B. Contrary to the above, on July 19, 2018, the licensee failed to establish a fire watch in auxiliary building room 757-A10 within one hour of closing pressurizer block valve 1-FCV-68-332-B.
05000391/FIN-2018003-042018Q3Watts BarInadequate Sensitive Equipment Control Results in Unit 2 Reactor Trip on April 12, 2018A self-revealed Green finding and associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, Drawings, was identified for the licensees use of a procedure that was not appropriate to the circumstances, which led to the conduct of improperly planned maintenance on sensitive equipment, ultimately resulting in a reactor trip. Specifically, an inadequacy was identified in station procedure 0-TI-12.10, Control of Sensitive Equipment, which lists the sensitive equipment defined, in part, as equipment that could cause a unit trip, on which work activities are required to be appropriately planned and conducted in a manner that will preclude a unit trip. The procedure did not list the high side reactor coolant system loop flow transmitter common drain line as sensitive equipment, which allowed the licensee to improperly perform maintenance on it without the appropriate planning and control necessary to preclude the Unit 2 reactor trip that occurred on April 12, 2018.
05000390/FIN-2018003-032018Q3Watts BarFailure to Collect Compensatory Samples for an Out-of-Service Effluent MonitorThe inspectors identified a Green finding and associated NCV of TS 5.7.2.3 when the licensee failed to take compensatory samples in accordance with Table 1.1-1 of the Offsite Dose Calculation Manual when the Unit 1 steam generator blowdown effluent monitor was out of service. Specifically, radiation monitor 1-RM-90-120/121 was inoperable from April 27 to May 27, 2018, and compensatory samples were not collected and analyzed within the required frequency of at least once per 24 hours.
05000391/FIN-2018003-022018Q3Watts BarUnauthorized Entry Into a High Radiation AreaA self-revealed Green finding and associated NCV of TS 5.11.1.e was identified when the licensee failed to maintain current survey information and failed to inform a worker of increased dose rates in a high radiation area. As a result, a worker received an electronic dosimeter alarm on the Unit 2 pressurizer platform due to changing radiological conditions associated with a reactor mode change.
05000390/FIN-2018003-012018Q3Watts BarConfiguration Control Error Results in Actual Auxiliary Building Internal Flooding EventA self-revealed Green finding and associated NCV of Technical Specification (TS) 5.7.1, Procedures, was identified when the licensee failed to maintain adequate configuration control in the high pressure fire protection (HPFP) system in accordance with station configuration control procedure, NPG-SPP-10.2, Clearance Procedure to Safely Control Energy. Specifically, the licensee failed to restore HPFP system vent and drain valves to their appropriate configuration prior to returning the system to service which resulted in a significantly large amount of HPFP system water (on the order of 10,000 gallons) being introduced into many areas (including all levels) of the Unit 1 side of the auxiliary building and wetting numerous structures, systems, and components (SSCs) (including cables, ventilation ducts, motor-operated valves, etc.)
05000251/FIN-2017004-012017Q4Turkey PointFailure to Perform an Adequate ASME BPVC Section XI Repair/Replacement Plan for a Code Class 1 and 2 ReplacementA NRC-identified NCV of 10 CFR 50.55a, Codes and Standards, was identified for the failure to adequately perform a Boiler and Pressure Vessel Code (BPVC) class 1 and 2 replacement activity in accordance with the Turkey Point Plant American Society of Mechanical Engineers (ASME) Section XI Repair/Replacement Program. Specifically, the licensee did not ensure a system leakage test conducted on October 19, 2017, was appropriately evaluated to meet the requirements of ASME Section XI for pre-service leakage testing of a Unit 4 high head safety injection (HHSI) cold leg injection check valve that was replaced on October 15, 2017. This issue was entered into the licensees Corrective Action Program (CAP) as ARs 2235484 and 2239149. Corrective actions included documenting a formal bases for current operability via a prompt operability determination and updating work order (WO) documentation to fully comply with ASME BPVC Section XI requirements. This performance deficiency was determined to be more than minor because an inadequate inservice inspection repair/replacement plan adversely affected the Reactor Coolant System (RCS) Equipment and Barrier Performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the issue had very low safety significance because there was no actual degradation of the RCS boundary. This finding was assigned a cross-cutting aspect in the Procedure Adherence component of the Human Performance cross-cutting area, in that the licensee did not effectively evaluate and appropriately implement the ASME BPVC requirements in the 4-873A Repair/Replacement Plan which were reiterated in licensee administrative procedure 0-ADM-532, ASME Section XI Repair/Replacement Program (H.8).
05000251/FIN-2017004-032017Q4Turkey PointInadequate Installation of Outdoor Use Electrical Enclosures Results in Manual Reactor TripA self-revealing finding (FIN) was identified for failure to ensure the 4B and 4C main feedwater regulating valve (MFRV) control circuits remained free from the effects of water intrusion or condensation in electrical enclosures. Specifically, a hand selector switch (HSS) enclosure for the 4C MFRV redundant positioners was flooded during wind-driven rain and resulted in the 4C MFRV failing closed, lowering 4C steam generator water level, and a subsequent Unit 4 manual reactor trip initiated by control room operators.Engineering Change (EC) 246879 appropriately selected NEMA-4X rated enclosures for the HSSs but associated SPEC-C-065 did not provide critical configuration details for the enclosure installations. Water collected in the 4B and 4C MFRV positioner HSS enclosures because the penetrations were on top of the enclosures and not properly sealed and the bottom of the enclosure did not have a weep hole.This performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, because the failure resulted in lowering steam generator water levels and caused control room operators to complete a fast load reduction and manually trip the reactor. In accordance with NRC IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, the inspectors determined that the issue had very low safety significance because it only caused a reactor trip and did not cause the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Since EC 246879 and associated work orders were completed in 2013, the inspectors determined the finding was not indicative of current licensee performance and was not assigned a cross-cutting aspect.
05000251/FIN-2017004-022017Q4Turkey PointFailure to Identify and Correct a Deficient CCW Penetration Seal Configuration that Exacerbates External Piping Corrosion ConditionsA NRC-identified NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to promptly identify and correct an adverse condition to quality that led to continued corrosion and significant scaling and pitting of the Unit 4 component cooling water (CCW) 18-inch headers at the penetration seals from the CCW heat exchanger room to the 10-foot pipeway. This issue was entered into the licensees CAP as ARs 2217942, 2227877, 2211843, 2236687, and 2239632. Corrective actions included removing protective boots that were inappropriately installed and not in accordance with design drawings and work order instructions, and were collecting hypersaline water that wetted carbon steel piping.The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences because corrosion and pipe wastage was ongoing and unmonitored for the Unit 4 CCW headers. In accordance with IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined the finding to be of very low safety significance because it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee failed to identify the adverse condition that allowed corrosion to continue unmonitored (P.1).
05000400/FIN-2017003-012017Q3HarrisIncomplete and Inaccurate Emergency Action Level SubmittalsThe NRC identified a Severity Level (SL ) IV non- cited violation (NCV) of 10 CFR 50.9, Completeness and accuracy of information, for failure to provide complete and accurate information for prior approval of a new emergency action level (EAL) scheme. The documents submitted to the NRC were, Shearon Harris Nuclear Power Plant, Unit 1 Changes to the Emergency Action Level Scheme, dated April 25, 2010, and License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99- 01, Revision 6, dated April 30, 2015. The submit ted documents specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, which contained declaration EAL threshold values for the containment high range radiation monitor that were lower than the correct values due to use of a n improper calculation methodology. The calculation methodology that was used was not in accordance with the license. It was used to calculate the loss of fuel clad barrier and potential loss of containment threshold values. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017 -017 to inform operators and emergency response organization decision- makers of the proper application of the EAL scheme and appropriate threshold values to be implemented. Additionally, the licensee plans to submit a license amendment request to update the EAL scheme. The licensee entered this violation into their corrective action program (CAP) as nuclear condition report (NCR) 02155272. The inspectors evaluated the underlying technical issue and determined that the licensees failure to maintain the effectiveness of its emergency plan was a performance deficiency. The issue was documented as a Green licensee- identified violation (LIV) in Section 4OA7 of this report. The reactor oversight process (ROP) , significance determination process , does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it was necessary to address this violation which impeded the NRCs ability to regulate, using traditional enforcement to adequately deter non- compliance. Using the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information, and Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, this issue was determined to be a SL IV violation. Though the NRC would have questioned the issue with a request for additional information, it would not have resulted in substantial further inquiry. 3 Additionally, the associated technical violation was determined to be of very low safety significance. Traditional enforcement violations are not assessed for cross -cutting aspects
05000400/FIN-2017003-022017Q3HarrisReview of Removal of the Technical Support Center (TSC) Temporary Diesel GeneratorThe inspectors conducted a detailed review of NCR 02123373, Emergency Action Level Document Calculation Assumptions. The inspectors chose the sample because the EAL issue initially appeared to be potentially more significant than finally determined. The inspectors evaluated the following attributes of the licensees actions: complete and accurate identification of the problem in a timely manner evaluation and disposition of operability and reportability issues consideration of extent of condition, generic implications, common cause, and previous occurrences classification and prioritization of the problem identification of root and contributing causes of the problem 19 identification of any additional condition reports completion of corrective actions in a timely manner 2. The inspectors conducted a detailed review of NCR 00520918, Loss of Offsite Power Impact on Technical Support Center (TSC). The inspectors chose the sample because it was discovered that on July 17, 2017, the licensee had removed a temporary diesel generator that was intended to provide a back -up reliable power source to the TSC until a permanent solution was implemented. The inspectors evaluated the following attributes of the licensees actions: complete and accurate identification of the problem in a timely manner evaluation and disposition of operability and reportability issues consideration of extent of condition, generic implications, common cause, and previous occurrences classification and prioritization of the problem identification of root and contributing causes of the problem identification of any additional condition reports completion of corrective actions in a timely manner b. Findings 1. Incomplete and Inaccurate Emergency Action Level Submittals Introduction: The NRC identified a Severity Level IV NCV of 10 CFR 50.9 , Completeness and accuracy of information, for failure to provide complete and accurate information for prior approval of a new EAL scheme. The documents submitted to the NRC were, Shearon Harris Nuclear Power Plant, Unit 1 Changes to the Emergency Action Level Scheme, dated April 25, 2010, and License Amendment Request to Adopt Emergency Action Level Scheme Pursuant to NEI 99- 01, Revision 6, dated April 30, 2015. The first submittal to the NRC in 2010 was not complete and accurate in all material respects , and the submittal in 2015 was a missed opportunity to identify the errors made in the first submittal in 2010. Description : On May 10, 2017, Shearon Harris identified the hot operating mode EAL thresholds were calculated incorrectly using a NUREG -0654 methodology vice the required NEI 99- 01 Rev. 6 method, as specified in the current facility licensing basis. When employing the NUREG -0654 methodology to calculate the EAL threshold values, the reactor coolant system (RCS) inventory was assumed to be released at a 50 gallons per minute (gpm) RCS leak rate and activity of 300 micro -Curies per gram (ci/gm) dose equivalent iodine (DEI), over a six -hour period of time. In comparison, when employing the NEI 99- 01 Rev. 6 methodology, the assumption as part of calculating the EAL threshold values was that the entire RCS inventory was released instantaneously at an activity of 300ci/gm DEI. Both of the licensees submittals to the NRC, specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, contained declaration EAL threshold values for the containment high range radiation monitor for loss of fuel clad barrier and potential loss of containment , that were significantly lower than the correct values , due to use of the improper calculation methodology. The submittal dated April 30, 2015, was submitted to provide a complete change to the EAL scheme. This submittal was a missed opportunity by the licensee to identify that the wrong methodology to calculate the EAL threshold values had been used. 20 These submittals were not correct in material content and impacted the NRC s regulatory processes. The NRC evaluated the licensees failure to provide complete and accurate information to determine if there were any unresolved issues. The inspectors concluded that the incomplete and inaccurate information in the license submittal was material to the NRC because, had the NRC staff known the actual methodology used was inaccurate, the staff would have required the licensee to modify the EAL threshold values . The licensee appropriately revised the EAL threshold values utilizing the correct calculation methodology. The licensee issued NC R 02123373, dated May 10, 2017, for EAL thresholds that were calculated without using the correct methodology described in the facility licensing basis. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017- 017 to inform operators and emergency response organization decision - makers of the proper application of the EAL scheme and revised threshold values to be implemented until a permanent change is made to the license. Additionally, the licensee issued NCR 02155272, dated October 3, 2017, for the incomplete and inaccurate EAL submittal, specifically addressing and resolving the completeness and accuracy issues identified by the inspectors. The final significance determination of the underlying technical issue for the licensees failure to maintain the effectiveness of its emergency plan was documented in NRC Inspection Report 05000400/2017003, Section 4OA7, as a Green LIV. Analysis : The inspectors evaluated the underlying technical issue and determined that the licensees failure to maintain the effectiveness of its emergency plan was a performance deficiency. The issue was documented as a Green LIV in Section 4OA7 of this report. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it was necessary to address this violation which impeded the NRCs ability to regulate, using traditional enforcement to adequately deter non- compliance. Using the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information, and Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report , this issue was determined to be a SL IV violation. Though the NRC would have questioned the issue with a request for additional information, it would not have resulted in substantial further inquiry. Additionally, the associated technical violation was determined to be of very low safety significance. Traditional enforcement violations are not assessed for cross -cutting aspects . Enforcement : Section 50.9 of 10 CFR states, in part, that, information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on April 25, 2010, and on April 30, 2015 , information was submitted by the licensee to the NRC that was not complete and accurate in all material respects. Specifically, the submitted documents specified the licensee s EAL scheme for Category F Fission Product Barrier EAL, contained EAL declaration threshold values for the containment high range radiation monitor , that were lower than the actual correct values , due to use of an improper calculation methodology. This was not in accordance with the license. It was used to calculate the loss of fuel clad barrier and potential loss of containment thresholds values. The licensee implemented compensatory corrective actions by issuing Standing Instruction 2017 -017 to inform operators and emergency response organization decision -makers of the proper application of the EAL scheme and appropriate threshold values to be implemented. Additionally, the licensee plans to submit a license amendment request to update the 21 EAL scheme. Because this violation was not repetitive or willful, and was entered into the licensees CAP as NC R 02155272, it is being treated as a SL IV NCV, consistent with Section 2.3.2 .a of the NRC Enforcement Policy. ( NCV 05000400/2017003- 01, Incomplete and Inaccurate Emergency Action Level Submittal s) 2. Adequacy of Process for Removal of the TSC Temporary Diesel Generator Introduction: The inspectors opened an Unresolved Item (URI) to complete a review of the licensees removal of a temporary diesel generator on July 17, 2017, that was previously installed to provide reliable backup power to the TSC in the event of a Loss of Offsite Power (LOOP) coincident with a Loss of Coolant Accident (LOCA) event. This temporary diesel generator was originally intended to be installed until a reliable backup power source could be implemented under a permanent modification. Description : The licensee initiated NCR 00520918 on March 1, 2012, to address the consequences of a LOOP/LOCA event on the T SC functionality. Since the TSC is designed with two sources of electrical power, both from offsite power sources, it was recognized that a complete loss of offsite power to the TSC could result in long term TSC operational concerns. Specifically, with t he loss of both offsite power sources, the TSC emergency ventilation system, which provides required radiation protection for event response personnel, would be non- functional, as well as other critical TSC equipment following the loss of short -term (~1 -2 hour s) back -up battery power supplies. The inspectors noted that the operability/functionality section of NCR 00520918 stated that the TSC was functional based on the (current) availability of both of the offsite power sources; however, should a LOOP event occur, then the TSC would be considered non -functional since offsite power would be rendered non -functional. This statement demonstrated the licensees understanding of the vulnerability of continued TSC functionality during a LOOP event. In recognition of this vulnerability, the NCR implemented a short -term solution for procuring and installing a temporary diesel generator in late 2012 under modification EC 85350. The inspectors noted that an emergency preparedness change review evaluation was conducted in accordance with 10 CFR 50.54(q) under action request 00568695. This change request stated that it was necessary to provide the infrastructure for an additional reliable power source for the TSC habitability systems. NCR 00520918 stated that the long- term solution was to provide a permanent backup power supply to the TSC , at which time the temporary diesel generator would be removed. While an action item was initiated to install this TSC permanent backup power source under modification EC 85145, the modification was later revised, removing the intended implementation of a permanent backup power source to the TSC. The inspectors were concerned that the TSC could have equipment and habitability issues during design basis LOOP/LOCA events when the normal TSC offsite power would be non- functional. In addition, the inspectors determined that the TSC temporary diesel generator was removed from the site on July 17, 2017, without implementing the originally intended reliable permanent backup power to the TSC and without conducting a 10 CFR 50.54(q) evaluation specific to its removal to demonstrate that this action did not reduce the effectiveness of implementing the emergency plan. The inspectors requested additional information from the licensee related to the documentation, basis, and process used for the removal of the TSC temporary diesel generator, and evidence that the TSC facility would still be capable of performing all of its intended functions during a LOOP/LOCA event. This issue of concern requires more information to 22 determine if a performance deficiency exists, and if the performance deficiency potentially constitutes a violation of regulatory requirements . Pending review of additional information from the licensee, this issue is identified a s URI 05000400/2017003 -02, Review of Removal of the Technical Support Center ( TSC ) Temporary Diesel Generator.
05000400/FIN-2017003-032017Q3HarrisLicensee-Identified ViolationSection 50.54(q)(2) of 10 CFR requires, in part, that a licensee shall follow and maintain the effectiveness of an emergency plan which meets the planning standards of 10 CFR 50.47(b) and the requirements of 10 CFR Part 50, Appendix E . Section 50.47(b)(4) of 10 CFR requires that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, from April 2010 to May 2017, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, the licensee's emergency classification scheme action levels for Category F Fission Product Barrier EAL , contained declaration threshold values for the containment high range radiation monitor , which were lower than the correct values due to an improper methodology used in calculating the loss of fuel clad barrier and potential loss of containment barrier threshold values and rendered the EALs ineffective. The licensee implemented compensatory actions by issuing Standing Instruction 2017- 017 to inform operators and emergency response organization decision- makers of the proper application of the EAL scheme and appropriate threshold values to be implemented until a permanent change can be made to the license. The issue was entered into the licensees CAP as NCR 02123373. The inspectors evaluated this issue as an ineffective EAL per IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process , Figure 5.4 -1. The inspectors concluded that the violation was of very low safety significance (Green). Although the incorrect EAL would alone render an early EAL classification of a General Emergency (GE) based upon the specific radiation monitor, other EALs would provide a GE classification in an accurate and timely manner aligned with the incorrect threshold values of the containment high range radiation monitor .
05000327/FIN-2016004-012016Q4SequoyahDegraded Fire Barrier PenetrationsGreen. The NRC identified a non-cited violation (NCV) of the facilitys operating license for the licensees failure to ensure that all fire barrier penetrations in fire zones boundaries protecting safety related areas are functional at all times. Specifically, on eight separate fire barrier penetrations, the licensee failed to recognize that the barrier had become damaged to the point of being nonfunctional. The licensee also failed to implement required compensatory measures for a nonfunctional fire barrier penetration contrary to the approved fire protection report (FPR). The licensee entered the issues into their corrective action program (CAP) as Condition Reports (CRs) 1229468, 1229470, 1243550, 1243970, 1243552, 1243554, 1243555, and 1243557. The performance deficiency was determined to be more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the fire barriers being damaged to the point of declaring the fire barrier penetrations nonfunctional, there was no assurance that the fire barrier would prevent the spread of fire through the cable penetration during a design basis fire. The inspectors performed the SDP using NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and assigned a High degradation rating, giving no credit for Barrier Protection in accordance with the Fire Barrier Degradation section. The inspectors concluded, that the finding was of very low safety significance (Green) due to fully functional automatic suppression systems on either side of the fire barrier (Question 1.4.3-C). Using Manual Chapter 0310, Aspects Within the Cross-Cutting Areas, the inspectors identified a cross-cutting aspect in the Identification component of the Problem Identification and Resolution area, because the licensee failed to enter the damaged fire barrier into their CAP after it was initially damaged (P.1)
05000321/FIN-2016003-022016Q3HatchFailure to Ensure Work Hours are Within Work Hour LimitsAn NRC-identified non-cited violation (NCV) of 10 CFR Part 26, Fitness for Duty Programs, was identified when the licensee failed to ensure that personnel subject to work hour controls did not exceed 72 hours in a work week. The licensee entered this condition into their corrective action program as Condition Report 10214872 and restored compliance when the affected individuals received an adequate rest period. The failure to ensure that work hours for personnel subject to work hour controls were tracked in accordance with licensee procedures was a performance deficiency. The finding was more than minor because, if left uncorrected, the failure to appropriately implement work hour limitations for covered workers could adversely impact the conduct and oversight of work on safety significant components. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not result in an adverse impact to plant safety due to worker fatigue. The inspectors determined this performance deficiency had a cross-cutting aspect of Consistent Process in the Human Performance area because the licensee failed to assess which workers were subject to work hour limits. (H.13)
05000321/FIN-2016003-012016Q3HatchUnit Downpower Caused by RFP Vent Line FailureA self-revealing finding was identified when the licensee failed to install a reactor feed pump (RFP) vent line weld in accordance with plant procedures resulting in a failure that required an unplanned Unit 1 power reduction greater than 20%. Failure to install the correct weld thickness on the unit 1 B RFP vent line, as required by procedures, was a performance deficiency. This performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective in that an unplanned reactor power reduction was required from 100 percent to 60 percent RTP. The inspectors determined this finding was of very low safety significance (Green) because there was not a reactor trip or loss of mitigation equipment. The inspectors determined that this finding had a cross-cutting aspect in the Resolution aspect of the problem identification and resolution area, because the organization did not take effective corrective actions to address the previous weld configuration issue. (P.3)
05000424/FIN-2013004-022013Q3VogtleLicensee-Identified ViolationThe licensee identified a violation of Title 10 CFR 50.74, Notification of change in operator or senior operator duties, for the failure of the facility licensee to notify the Commission of changes in the medical status of numerous licensed operators within 30 days of learning of the medical change, as required. Contrary to the above, from November 2012 to July 2013, the licensee failed to provide the Commission the necessary medical information to insure licensed operators met the requirements of 10 CFR 55.33, Disposition of Initial Application. The licensee entered this violation into their corrective action program as CR 207318. The finding was determined to be of very low safety significance because the individual operators were found to have been physically capable of meeting the requirements as required, as well as, being in the presence of other medically qualified operators. This violation was characterized as a Severity Level IV non-cited violation, consistent with Example 6.4.d.1(c) of the Enforcement Policy.
05000424/FIN-2013004-012013Q3VogtleLicensee-Identified ViolationOn October 11, 2012, the NRCs Office of Investigations (OI) initiated an investigation to review whether a contract employee willfully failed to implement a maintenance procedure involving the independent verification of landing electrical leads during maintenance of a safety-related Motor Operated Valve. Based in part on the investigation, completed on July 2, 2013, the NRC concluded that the actions of the contract employee were willful, and his actions caused VNP to be in violation of regulatory requirements. This issue was identified by the licensee. Technical Specification 5.4, Procedures, requires that written procedures, specified in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, shall be established, implemented, and maintained. Section 9.a of Regulatory Guide 1.33, Appendix A, requires that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Licensee procedure NMP-ES-017-004, MOV Diagnostic Procedure for Gate and Globe Valves, provides instructions for maintenance of safety-related motor operated valve 1HV8105. Step 6.10.3 of the procedure provides instructions to re-connect control wiring per the Data Sheet. Per the Data Sheet the lead (LS-16) from point 39 needed to be re-landed and independently verified. Contrary to the above, on September 26, 2012, a contract laborer from Crane Nuclear willfully failed to independently verify that lead LS-16 was correctly landed per procedure NMP-ES-017-004. Specifically, the contract laborer failed to conduct an independent verification but signed-off the data sheet stating that he had verified the landing of leads for MOV 1HV8105. The inspectors determined that the underlying technical significance of the failure to perform an independent verification on landing a wire from a control switch of a safety-related MOV was a minor violation. However, based on an assessment of the factors described in Section 2.2.1.d of the Enforcement Policy, this violation is disposition as a non-cited violation. The licensee entered this issue into its corrective action program as CR 524641.
05000250/FIN-2012004-022012Q3Turkey PointLicensee-Identified ViolationThe licensee identified that Unit 3 train 2 auxiliary feedwater flow control valve FCV-3- 2832 was rendered inoperable when a maintenance technician installed a cap over the solenoid vent port. The cap was installed after removal of test equipment. Turkey Point Technical Specification 6.8.1 requires that procedures required by the FPL Quality Assurance Topical Report (QATR) be maintained and implemented. The topical report includes procedures for control of maintenance and specifies that maintenance procedures contain instructions in sufficient detail to permit maintenance work to be performed correctly. The licensee met this requirement, in part, with work order 40181373-01, written for the investigation and testing of train 2 auxiliary feedwater flow control valve (FCV-3-2832) following observed erratic operation. After the testing was completed, the work order required the maintenance technician to un-install the test equipment. Contrary to the above, on September 18, 2012, work order 40181373-01 did not contain instructions in sufficient detail to un-install the test equipment correctly, and a technician mistakenly placed a cap over a solenoid vent line for FCV-3-2832, making the valve unable to close after being opened by an actuation signal. The error was discovered by the licensee during a planned auxiliary feedwater test conducted the next day. When discovered, the licensee entered the appropriate technical specification action, removed the cap to restore operability to the valve, and demonstrated operability by completing a surveillance test. The inspectors evaluated the event using NRC Inspection Manual 0612, Power Reactor Inspection Reports; Inspection Manual Chapter 0609.04, Initial Characterization of Findings; and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. The finding was screened as being of very low safety significance (Green) when all screening questions in IMC 0609 Appendix A were answered no . Because this violation was of very low safety significance and was entered in the licensees corrective action program as AR 1804442, this violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000250/FIN-2012004-012012Q3Turkey PointOperation at power with Unit 3 feedwater flow transmitter connected incorrectlyA self-revealing, non-cited violation (NCV) of Turkey Point Technical Specification (TS) 3.3.1 Reactor Trip System Instrumentation was identified when process tubing to a Unit 3 feedwater flow transmitter was found incorrectly installed. As a result, one channel of reactor protection was not operable when required. When control room indications of erratic feedwater flow were noted, the applicable technical specification action was entered, bistables were tripped, and the process tubing misalignment was corrected. The problem was documented in the corrective action program as action request (AR) 1800833. Failure to adequately perform maintenance and to verify proper alignment of flow transmitter FT-3-476 process tubing after replacement was a performance deficiency. The performance deficiency was determined to be more than minor because it affected the configuration control attribute of the Mitigating Systems Cornerstone which ensures the reliability of systems that respond to initiating events, such as the reactor protection system. The finding was screened using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2. Because the finding affected only a single reactor protection system (RPS) trip initiator and other redundant trips or diverse methods of reactor shutdown were not affected, the finding was determined to be of very low safety significance (Green). The finding was assigned a cross-cutting aspect in the Work Practices component of the Human Performance area (H.4.a) because the licensee did not establish human error prevention techniques, such as self and peer checking and proper documentation of activities to prevent incorrect installation of the flow transmitter.
05000250/FIN-2012002-032012Q1Turkey PointLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for disposition as an NCV. 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization. FPL implements this requirement, in part, with procedure ENG-QI 1.7, Quality Instruction Design Input/Verification, which states engineering methods employed shall ensure that design inputs are correctly translated into new designs and design changes, and that design verification activities are correctly performed. Contrary to the above, engineering methods employed did not ensure that design inputs were correctly translated into the A auxiliary feedwater pump design change nor were design verification activities correctly performed on engineering design change package PCM 2005-029. As a result, on February 3, 2012, during a design review while developing a modification package for the A auxiliary feedwater pump, FPL identified a design calculation error in the 2005 modification package for the A auxiliary feed water pump. The pump modification raised the pump power requirements. The revised design horsepower output specified for the turbine accounted for the increased pump power demand, but failed to account for recirculation flow, turbine lube oil coolers flow, and instrument uncertainties. When identified by FPL, a prompt operability determination was completed. FPL determined that although there was a reduction in margin, the required auxiliary feed water turbine horsepower remained bounded by vendors design limits. This issue was entered into the corrective action program as AR 1731117. The finding was screened as having very low safety significance (Green) using NRC Inspection Manual Chapter 0609 SDP Phase 1 screening because the finding did not result in an inoperable auxiliary feedwater pump, did not affect functionality of the system, and the design basis continued to be met.
05000250/FIN-2012002-022012Q1Turkey PointControl power cables repeatedly submerged in ground water, contrary to designA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified when FPL did not maintain safety-related power cables in the environment for which they were designed and tested. Specifically, 125 volt DC control power cables feeding various safety related components and cables supporting other risk significant equipment had been repeatedly submerged in ground water for extended periods of time and this submergence had the potential to affect the ability of the cables to perform safety related functions. The issue was entered into the licensees CAP as AR 1717619. Although predominantly Unit 3 cables were submerged, because equipment is shared, both units were affected. Allowing water accumulation in the manhole(s) after disabling of the sump pump without compensatory measures to keep the safety related and risk significant cables dry resulted in subjecting the cables to an environment for which they were not designed, and was a performance deficiency. The finding was more than minor because it challenged the reliability of systems that respond to initiating events to prevent undesirable consequences, which is an attribute of the Mitigating Systems cornerstone. The inspectors evaluated the finding in accordance with IMC 0609.04, Phase 1, Initial Screening and Characterization of Findings. The finding was of very low safety significance because it did not represent an actual loss of safety function or contribute to external event core damage sequences. The finding had a cross-cutting aspect in Problem Identification and Resolution, Corrective Action Program, (P.1(c)), because FPL did not thoroughly evaluate submerged cables such that the resolutions addressed causes and extent of conditions, including evaluating for operability.
05000250/FIN-2012002-012012Q1Turkey PointEmergency lighting to auxiliary feedwater area disabledThe inspectors identified a non-cited violation of the Units 3 and 4 operating licenses condition 3.D, Fire Protection, when the licensee failed to provide emergency lighting in the common auxiliary feedwater (AFW) cage and other areas. The electrical panel that supported normal lighting in the area was taken out of service for maintenance thus placing the emergency lights on battery power until the batteries depleted and the areas became dark, impacting the ability of operators to complete manual actions in the area, if needed. The licensee documented the issue in the corrective action program (CAP) as AR 1738082. The inspectors determined that the failure to provide emergency lighting in areas requiring local manual actions to safely mitigate certain fire events, and the associated access/egress routes, was a performance deficiency. The issue was more than minor because the objective of the Mitigating System Cornerstone to ensure the availability of fire protection equipment was affected when emergency lighting was not provided. The inspectors assessed the finding using NRC Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and assigned a low degradation rating because of the reasonable likelihood that plant operators would obtain alternate lighting and complete the prescribed manual actions. The finding screened as having very low safety significance. The cross cutting aspect of Work Control Planning, (H.3(a)), was assigned because the licensee did not use risk insights, did not assess environmental conditions (lighting) that may have impacted human performance, and did not plan for contingencies nor compensatory actions when the normal lighting was removed from service leading to loss of emergency lighting
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