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05000250/FIN-2011002-032011Q1Turkey PointNoneTechnical Specification 5.2.1.f requires the containment steel liner to have a nominal thickness of 0.25 inches. Contrary to the above, on October 22, 2010, during a planned inspection, FPL found corrosion in the lower liner plate at the -15 foot level with areas of the liner to be below 0.25 inches. The liner was backed by concrete and no direct path to the environment was identified. FPL repaired the liner and planned to recoat the lower cavity area in a future outage. FPL contracted an evaluation of the liner holes to radiological consequence of an accident and found the contribution to dose from the degradation to be negligible. The finding was screened as Green using NRC Inspection Manual Chapter 0609, Attachment 0609.04, SDP Phase 1 screening because the finding did not result in any loss of containment barrier function.
05000250/FIN-2011003-012011Q2Turkey PointFailure to properly perform a procedure results in damage to an RHR pumpA self-revealing, non-cited violation (NCV) of Technical Specifications 6.8.1.a, Procedures, was identified when operators did not properly align the RHR system from shutdown cooling mode to injection mode. As a result, the 4A RHR pump was left running with no suction source causing a failure of the pump mechanical seal and minor flooding in the Unit 4, A RHR pump room. The pump was not available for either injection or shutdown cooling operations until the seal was replaced. The issue was documented in the corrective action program as AR 1644427 and a root cause investigation was initiated. Failure to properly align the RHR system to the injection lineup was contrary to plant procedures and was a performance deficiency. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating System Cornerstone and resulted in damage to an RHR pump. The finding was screened using IMC 0609, Appendix A, Phase 1, and because there was no loss of safety function with the alternate RHR pump remaining operable, the finding was determined to be of very low safety significance. The finding affected the cross-cutting area of Human Performance, Work Practices because personnel did not adequately implement error prevention techniques, such as pre-job briefings, self and peer checks, and proper documentation of activities
05000250/FIN-2011004-012011Q3Turkey PointFailure to control defective component results in safety system surveillance failureA Self-revealing Non-cited violation of Technical Specification requirements was identified for failure to implement procedures to control a defective component and prevent its use in a safety-related system. Specifically, the licensee installed a solenoid valve, known to be defective in the valve actuator for the Unit 3 B emergency containment cooler and the valve subsequently failed a surveillance requirement. The issue was documented in the licensees corrective action program as CR1682798 and corrected by replacing the defective solenoid valve prior to returning the system to service. The failure to identify and control the solenoid valve after having received information that the valve was defective was a performance deficiency. The performance deficiency was determined to be more than minor because it adversely affected the safety related emergency containment cooler system which is used to protect the public from radionuclide releases caused by accidents. The finding was screened using IMC 0609, Significance Determination Process (SDP), Attachment 0609.04 for the Containment Barrier and was screened as of very low safety significance (Green). The inspectors determined that the cross-cutting aspect of Problem Identification and Resolution was affected when the licensee did not identify the defective component in the corrective action program in a timely manner after having received notification from the vendor of a component defect.
05000250/FIN-2011004-022011Q3Turkey PointLicensee-Identified ViolationTS 3.3.3.3(a) requires, in part, the accident monitoring instrumentation channels shown in Table 3.3-5, including the main steam line (MSL) high range-noble gas effluent monitor, to be operable. Contrary to this, on October 1, 2010, licensee evaluation of design bases for the proposed replacement of the MSL high range-noble gas effluent monitor common to Units 3 and 4, i.e., Radiation Monitor (RAD)-6426 with Data Acquisition Monitor (DAM)-1 and High Range Noble Gas Detector Assembly SA-9, determined that the current configuration of the sample line and monitor system failed to meet the TS operability requirements. Specifically, the licensee determined that noble gas samples collected from each of the U3 and U4 steam lines would not be representative of noble gases released from the MSL safety valves and/or atmospheric dump valves during postulated emergency plan scenarios. Further, licensee evaluations indicated that the subject monitoring system had not met the TS requirement since it was installed in 1981. The inspectors determined that this finding is more than minor. Initial NRC concerns and licensee commitment for proposed sampling and estimating noble gas quantities released via steam pathways in accordance with NUREG 0578 were documented in letters dated March 10, 1980, March 28, 1980, and August 20, 1980, from Robert E. Uhrig, Vice President, Florida Power and Light to the Office of Nuclear Reactor Regulation Projects and Licensing Offices. The inspectors noted that the subject correspondence documented that subsequent to installation, operating tests were to be conducted for the purpose of correlating noble gas activities in steam samples with flow out of the system through the MSL safety relief or atmospheric dump valves to demonstrate proper operation. However, licensee representatives stated that their reviews of monitor operability determined that neither test records nor other correlation data were found which demonstrated completion of the proposed initial operating tests for the installed monitoring system. The inspectors noted that proper oversight and review of those initial post startup tests, if conducted, potentially could have identified the design inadequacies subsequent to sample line and monitoring systems installation. Further, the inoperable monitor had a credible impact on equipment maintained to support emergency response dose calculation capabilities in accordance with Emergency Plan Implementing Procedure 20126, Offsite Dose Calculations. The finding was considered to have very low safety significance (Green) because the licensee had alternate methods for estimating effluent releases from the MSL atmospheric dump and/or release valves. This issue and corrective actions were documented in the licensees corrective action program as Condition Report (CR) Numbers 572823, 585330, and 596361.
05000250/FIN-2011005-012011Q4Turkey PointFailure to Correct Valve Deficiency Results in Both Headers of Intake Cooling Water InoperableA self-revealing non-cited violation of 10 CFR 50 Criterion XVI was identified when the licensee failed to repair a degraded butterfly valve in the Unit 3 intake cooling water system. On August 11, 2011, failure of this valve led to a loss of intake cooling water (ICW) flow to the component cooling water heat exchangers. The licensee documented the failure in their corrective action program as AR 01680272 and initiated a cause investigation. An NRC special inspection of this occurrence was documented in NRC Inspection Report 05000250/2011013. The licensees failure to take prompt corrective actions for a degraded valve, though it had been identified in 2007 as vibrating excessively, was a performance deficiency. This performance deficiency was considered more than minor because it could be reasonably viewed as a precursor to a significant event, the loss of all intake cooling water. A Senior Reactor Analyst in a Phase 3 risk assessment, determined the increase in risk to either unit was of very low risk significance i.e., Green. Unit 3 risk was assessed because the event occurred on that unit; however Unit 4 risk was also assessed because the same vulnerability existed on the ICW valves on that unit (e.g., similar design, maintenance history, etc.). The main contributors to the low risk results were: 1) the recovery probability of the ICW system, given the extended time available to operators before a RCP seal LOCA could occur; and 2) the multiple redundant sources available to cool the core should the CCW system fail. The dominant core damage scenarios were valid demands for a reactor trip followed by the failure to recover ICW proceeding to a RCP seal LOCA and core damage. The inspectors determined that the cause of this finding was related to the Problem Identification and Resolution cross cutting area when the licensee failed to take appropriate corrective action to address safety issues (valve fluttering) in a timely manner, commensurate with the safety significance.
05000250/FIN-2011005-022011Q4Turkey PointFailure to maintain TSC habitabilityThe licensee identified an Apparent Violation (AV) of 10 CFR Part 50.54(q), for failure to follow and maintain in effect emergency plans which require that adequate emergency facilities and equipment to support the emergency response are provided and maintained. Specifically, during the periods from December 4, 2010 to July 13, 2011, and from October 10 to October 28, 2011, the licensee failed to maintain a fully functional Technical Support Center when portions of its ventilation system were removed from service without compensatory measures. As a result, had the facility been required, personnel assigned to respond in the TSC would not have been protected from radiological hazards that would occur in some accidents. The licensee documented this issue in their corrective action program as AR 1701357. The finding was more than minor because it affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The Emergency Preparedness cornerstone was affected in that during the time the Technical Support Center was not functional, it did not meet 10 CFR 50.47(b)(8) Planning Standards program elements in that personnel assigned to the TSC during an emergency may not have been protected from radiological hazards. This finding was evaluated in accordance with Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Section 4.8 and Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply, and determined to be a finding of low to moderate safety significance (White) because there was a loss of the planning standard. The two events, December 2010 to July 2011, and October 2011, were assessed as a single finding with a common performance deficiency. The cause of the finding is related to the Problem Identification and Resolution cross-cutting area, in that the licensee did not thoroughly evaluate problems with the TSC ventilation system as necessary, including properly classifying, prioritizing, and evaluating for operability and reportability, conditions adverse to quality.
05000250/FIN-2011005-032011Q4Turkey PointFailure to make a required 8 hour NRC report for major loss of emergency assessment capabilityThe inspectors identified an Apparent Violation of 10 CFR 50.72(b)(3)(xiii) when a major loss of emergency assessment capability was not reported to the NRC within 8 hours. The TSC ventilation system was identified as being in a degraded condition from December 4, 2010 until July 13, 2011, affecting the habitability of the TSC for emergency responders, and the occurrence was not reported. The issue was identified to the licensee by the inspectors after review of NRC Event Notification 47387. The finding was more than minor because it impacted the NRCs regulatory process, which relies on certain events being properly reported to the NRC. Because this finding impacted the regulatory process, it was evaluated using traditional enforcement and is being considered for escalated enforcement action in accordance with NRCs Enforcement Policy. No cross-cutting aspect associated with this issue was identified.
05000250/FIN-2011008-022011Q3Turkey PointMolded Case Circuit Breaker TestingThe age range of approximately 511 safety-related MCCBs at Turkey Point is twenty to forty years - some are original plant equipment, some were installed in the 1980s, and the remainder in the early 1990s. With the exception of bench testing prior to installation, no testing/maintenance has been performed on the breakers. MCCBs are susceptible to age related failures such as, overheating due to loose connections and long term grease hardening. Overheating can exceed material temperature ratings, distort motor control center case and operating mechanism tolerances, and result in hardening/baking of grease. Long term grease hardening can result in the breaker failing to open or a delay in opening during a downstream electrical fault. NRC Information Notice (IN) 93-64, Periodic Testing and Preventive Maintenance of Molded Case Circuit Breakers, states, in part: MCCB preventive maintenance practices can mitigate the effects of aging and help ensure continued MCCB reliability, and that certain standard MCCB tests (such as individual pole resistance, 300 percent thermal overload and instantaneous magnetic trip tests) performed periodically were found effective along with the additional techniques of infrared temperature measurement and vibration testing. Also, EPRI NP-7410, Section 7.3.1, states, that ensuring that all MCBs are periodically exercised is considered a vital part of a maintenance program, applicable to all breakers regardless of their safety classification. In addition, EPRI/NMAC NP-7410-V3, Section 7.3.1, states that safety related MCCB cycling/trip testing should be performed on a 4 to 6 year frequency. In 2005 and 2006, during Turkey Points preventive maintenance optimization (PMO) project, the licensee identified the lack of a testing program for safety-related 120vac and 120vdc MCCBs which resulted in the creation of a preventative maintenance (PM) program for the breakers. The PMs for the 120Vac breakers were to include a periodic inspection and electrical test to verify functionality. The PMs for the 125Vdc breakers were to replace each individual breaker. However, the licensee suspended the PMs, in part, because of scheduling challenges associated with Technical Specification (TS) restrictions the TS has a two hour action statement associated with the deenergization of the ac or dc load centers. In 2008, in response to the cancelled PMs, CAR 08-069 was created and assigned as a Turkey Point Excellence (TPE) project. TPE considered several options, and decided on a one-time replacement of the vital 120Vac and 125Vdc breakers. In 2010, the licensee initiated AR 1649834 because the funding for the TPE project was terminated. This AR created a new long term asset management initiative to re-target the project in future years. In 2011, ECR 1657020 was created for a one-time replacement of the MCCBs and entered into the licensees long term asset management program (PTN-11-0177 (U3) and PTN -11-0179(U4). The team identified that since 2005/2006 when the lack of periodic testing of the MCCBs was identified; no interim measures were taken to correct the nonconforming condition. Specifically, on multiple occasions since 2005, the licensee failed to take adequate actions to ensure the reliability and capability of the MCCBs to perform their design function while pursuing long term strategies. Additionally, the team identified that the licensee failed to scope the protective tripping function of the MCCBs in the Maintenance Rule program. These issues were entered into the licensees corrective action program as ARs 1675539 and 1676808 which include developing an interim strategy that is to consider visual inspections augmented by thermography, planned cycling, and testing of the MCCBs.
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
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-2012002-032012Q1Turkey 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-2012003-012012Q2Turkey PointFailure to Perform an Analysis for the Permanent Removal of Main Steam Pipe Whip RestraintsThe inspectors identified a non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to perform an analysis for the removal of the Unit 3 main steam pipe whip restraints. These restraints are credited for mitigating high energy line breaks with a potential consequence of an unrestrained pipe break outside of containment. The licensee entered the issue into the corrective program as action request AR1757120 and revised the modification package to reinstall the pipe whip restraints prior to Unit 3 start-up. The team determined that the licensees failure to perform an analysis, as required by procedure ENG-QI 1.0, Design Control, for the permanent removal of main steam pipe whip restraints is a performance deficiency. The performance deficiency was more than minor because it affected the Mitigating Systems cornerstone attribute to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the lack of pipe whip restraints would adversely affect the capability of equipment required to mitigate high energy line break events. The team screened the finding in accordance with IMC 0609, Significance Determination Process, Attachment 4, Phase 1-Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of safety function, since the deficiency was identified and corrected before the modification was implemented. The team identified a crosscutting aspect in the decision making component of the human performance area.
05000250/FIN-2012003-022012Q2Turkey PointLicensee-Identified ViolationTS 6.12.2 requires, that each High Radiation Area (HRA) with dose rates greater than 1000 mrem/hour at 30 centimeters shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry. Contrary to the above, on April 28, 2012, while performing assigned locked high radiation area door checks in the Unit 3 containment, a Senior RP Technician identified the Unit 3 reactor sump area door could be partially opened due to the placement of the padlock and chain near the hinge. Immediate corrective actions were taken upon discovery and documented in AR01760652. The violation was evaluated using the Occupational Radiation Safety Significance Determination Process and was determined to be of very low safety significance (Green) because this finding did not involve ALARA planning or work controls, was not an over-exposure, did not have a substantial potential for overexposure, and the ability to access dose was not compromised.
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-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-2012005-012012Q4Turkey PointFailure to Verify 1B Feedwater Heater Drain Valve ClosedA self-revealing finding was identified when the licensee failed to follow procedure 0-ADM-222, Drain and Vent Rig Controls, while installing a temporary drain hose on Turkey Point Unit 4 in-service equipment. Operations and maintenance workers failed to verify a drain line flow path was isolated on the 1B feed water heater prior to removing a pipe valve cap that resulted in an unexpected lowering of condenser vacuum. Operators took action to close the open drain line isolation valve and terminate the plant transient. The licensee captured this condition in their corrective action program as AR 1819010. The licensees failure to verify the closed position of 1B feed water heater drain valve 4- 30-128, as required by procedure 0-ADM-222, prior to removing the pipe cap was a performance deficiency. The inspectors determined the performance deficiency was more than minor using IMC 0612, Appendix B, Issue Screening, because the performance deficiency was associated with the configuration control 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 power operations. Specifically, the failure to verify the position of 4-30-128 resulted in lowering condenser vacuum that could have led to a reactor trip and the unavailability of the main condenser. The inspectors evaluated the finding using the significance determination process for findings at power of IMC 0609, Appendix A, Exhibit 1, Transient Initiators. The inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a reactor trip and a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. The finding was associated with a cross-cutting aspect in the work practices component of the human performance area because the licensee did not define and effectively communicate expectations, or follow the procedural requirement to physically verify valve position during the drain hose installation work.
05000250/FIN-2012011-012012Q3Turkey PointFailure to Translate Design Basis Requirements Into Plant Procedures and Calculations for CCW Heat Balance EquationAn NRC identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to translate the worse-case total post-accident ICW flow rate for CCW heat exchangers, as documented in calculation PTN-4FSM-04-003 Revision 2, into surveillance, 3/4-OSP-030.4, CCW Heat Exchanger (HX) Performance Test. In addition, the licensee failed to incorporate seasonal salinity variances into calculation PTN-BFJM-96-004, HX3 and HX4 Computer Code Verification. The effects of these two discrepancies was a reduction in maximum allowed canal temperature margin by approximately1.5% or 1.5 degrees Fahrenheit. The licensee entered this issue into their corrective action program (CAP) as Condition Report (CR) 1789995.The failure to maintain the CCW heat balance calculation to ensure the plant could meet their design basis to perform heat removal for normal cool down of the facility, and to mitigate the effects of accident conditions within acceptable limits is a performance deficiency. The inspectors determined that the performance deficiency was more than minor because the calculation errors impacted the Mitigating Systems cornerstone objective to ensure the capability of the CCW system to respond to initiating events to prevent undesirable consequences and affected the cornerstone attribute of Design Control. The inspectors determined that this finding did not have a cross-cutting aspect, because the finding was determined not to be indicative of current licensee performance.
05000250/FIN-2012011-022012Q3Turkey PointInadequate Corrective Actions Following Identification of a NON-CONSERVATIVE Technical SpecificationAn NRC identified non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified when the licensees failure to take timely corrective action to address a nonconforming condition of Technical Specification (TS) 3/4.5.2 S R4.5.2a. The non-conservative TS was identified and placed in the corrective action program in 2006 as CR 2006-22868. TS 3.5.2 SR 4.5.2a was determined to be non-conservative and the corrective action to submit a TS amendment to address the non-conservative TS was not implemented. The licensee is scheduled to submit the license amendment in the fourth quarter of 2012, as referenced in AR 1790829. The inspectors determined that the licensees failure to timely correct a condition adverse to quality associated with the non-conservative TS was a performance deficiency. The performance deficiency was more than minor because if left uncorrected the failure to implement timely corrective actions has the potential to lead to a more significant safety event in that the unit could be placed in an unanalyzed condition for up to 24 hours. The inspectors determined that the finding was of very low safety significance because there has been no loss of safety system function. The inspectors determined that this finding directly involved the crosscutting area of Problem Identification and Resolution, component of the CAP and an aspect in taking appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity.
05000250/FIN-2013002-012013Q1Turkey PointFailure to Implement Timely Corrective Actions to Test Molded Case Circuit BreakersThe NRC identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish a test program to demonstrate that safety-related 120 VAC and 125 VDC molded case circuit breakers (MCCBs) would be able to reliably perform their intended safety functions, specifically protective tripping. The team identified that since 2005 and 2006, when the lack of periodic testing of the molded case circuit breakers was identified, no interim measures were taken to correct the nonconforming condition. Additionally, the team identified that the licensee failed to scope the protective tripping function of the MCCBs in the maintenance rule program. Upon identification by the team, the licensee entered these issues into their correction action program as ARs 1675539, 1676808, 1788355, and 1852219. As immediate corrective actions, the licensee tested 35 breakers which performed satisfactorily. The results of this testing and an action to develop a long-term test program for the entire 120 VAC and 125 VAC MCCBs were documented in AR 1852219. A license amendment will also be pursued to allow for more TS outage time in order to remove and replace the more difficult MCCBs. The licensees failure to implement prompt and effective corrective actions to ensure that safetyrelated molded case circuit breakers were adequately tested was a performance deficiency. The performance deficiency was more than minor because it adversely affected the mitigating systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609.04, Initial Screening and Characterization of Findings, the inspectors conducted a Phase 1 Significance Determination Process screening using Exhibit 2 of Appendix A to Manual Chapter 0609 and determined the finding to be of very low safety significance (Green) because it was a qualification deficiency confirmed not to result in the loss of operability or functionality. Because the licensee did not ensure that the necessary resources were available and adequate to maintain long term plant safety through the minimization of preventative maintenance deferrals, this finding is assigned a cross-cutting aspect in the resources component of the human performance area.
05000250/FIN-2013002-022013Q1Turkey PointNoncompliance with Radiological BarrierA self-revealing non-cited violation (NCV) of Technical Specification (TS) 6.12.1 was identified when a worker did not comply with a radiological barrier and entered a high radiation area (HRA) without proper authorization. Specifically, the worker entered the HRA without receiving a HRA briefing, and subsequently received a dose rate alarm. Upon identification, the licensee immediately restricted the workers access to the Radiological Controlled Area (RCA). This condition has been placed into the licensees Corrective Action Program (CAP), under Action Request (AR) 01852456. The finding was determined to be more than minor because it was related to the Occupational Radiation Safety cornerstone attribute of Program and Process, and adversely affected the cornerstone attribute to ensure the adequate protection of worker health and safety, because the worker was not made knowledgeable of the radiological conditions. Additionally, the finding was similar to IMC 0612, Appendix E, Example 6.h, which describes an improper entry into an HRA. The finding was evaluated in accordance with IMC 0609, Appendix C, where it was determined to be Green because it did not involve ALARA planning or work controls, was not an overexposure, did not contain a substantial potential for an overexposure, and the ability to assess dose was not compromised. The inspectors determined that this issue had a crosscutting aspect in the Work Practices component of the Human Performance area because the licensee did not communicate radiological conditions to the worker through a pre-job brief.
05000250/FIN-2013002-032013Q1Turkey PointWillful Violation of Radiological BarrierA self-revealing Severity Level (SL) IV non-cited violation (NCV) of Technical Specification (TS) 6.8, Procedures, was identified on June 6, 2012, when a worker willfully bypassed a radiological barrier and entered a posted high radiation area (HRA) without proper authorization. Specifically, the worker entered the HRA without receiving a HRA briefing and being issued a key as required by licensee procedure RP-SR-103-1002, High Radiation Area Controls and subsequently received a dose rate alarm. Upon identification, the licensee immediately restricted the workers access to the radiological controlled area (RCA) and placed this issue into the corrective action program (CAP) as action request (AR) 01773513. Due to the willful nature of the workers actions, the inspectors determined the performance deficiency was more than minor in accordance with the guidance contained in Chapter 2 of the Enforcement Manual, Revision 8. This willful finding involved an isolated act of a low-level nonsupervisory individual. It was addressed promptly by appropriate corrective actions, there was no actual safety significance and the underlying technical significance was low. Therefore, the inspectors concluded this finding was Severity Level IV, consistent with Section 2.2.2 of the Enforcement Policy, dated January 28, 2013. There was no cross-cutting aspect because this performance deficiency was dispositioned using traditional enforcement.
05000250/FIN-2013002-042013Q1Turkey PointFailure to Correct FLOW-INDUCED Vibration Leads to CCW Piping Weld FailuresA self-revealing non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified when the licensee failed to implement corrective actions that addressed low stress high cycle fatigue of component cooling water (CCW) relief valve RV-4- 747B piping caused by flow induced vibration. As a result, CCW system flow induced vibration resulted in weld cracks and system pressure boundary leakage in November 2012. The licensee repaired the weld failures and installed a pipe support on the line to minimize flow induced vibration on the associated pipe in February 2013 during a scheduled refueling outage. The licensee documented this condition in their corrective action program as action request (AR) 1824939. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to implement corrective actions to address CCW system flow induced vibration resulted in weld cracks and CCW system pressure boundary leakage in November 2012. The inspectors evaluated the finding under the mitigating systems cornerstone and used Inspection Manual Chapter (IMC) 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1, Checklist 4, PWR Refueling Operation, dated May 25, 2004. The inspectors determined the finding was of very low safety significance (Green) because the finding did not require a quantitative assessment of risk significance since each item on the Checklist 4 was met during the time the condition existed and while the 4B residual heat removal (RHR) train was removed from service to repair the weld leak. The finding was associated with a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not complete engineering evaluations necessary to support modifications that would prevent CCW system RV-4-747B piping weld failures caused by flow induced vibration.
05000250/FIN-2013003-012013Q2Turkey PointFailure to Promptly Identify and Correct a Pressure Boundary Through Wall Leak on the 3A CCW Pump Casing Vent PipeThe NRC identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a through wall pressure boundary leak on the 3A component cooling water (CCW) pump casing vent piping that affected system operability. The inspectors determined that the licensees failure to identify and correct a through wall leak on an ASME Code Class pressure boundary was a performance deficiency. The condition was entered in the licensee corrective action program (CAP) as action request 01883690 and the pipe was replaced. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the mitigating systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors determined that the licensees failure to identify a system pressure boundary leak precluded evaluations and repairs necessary to assure the reliability of the component cooling water system. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, dated June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 for the Mitigating Systems Cornerstone, dated June 19, 2012. The inspectors answered yes to the Exhibit 2 question A.1 because the system maintained its functionality. As a result, the inspectors determined the finding to be of very low safety significance (Green). This finding was associated with a cross-cutting aspect in the corrective action program component of the problem identification and resolution area. Specifically, the licensee failed to consider the potential for system pressure boundary leakage when evaluating the operability of the component cooling water system
05000250/FIN-2013003-022013Q2Turkey PointLicensee-Identified ViolationTurkey Point Unit 3 and 4 Technical Specification 6.8.1.a requires, in part, that written procedures shall be implemented as referenced in Florida Power and Light (FPL) Quality Assurance Topical Report (QATR). FPL QATR states that Regulatory Guide 1.33, Quality Assurance Program Requirements, is applicable in establishing procedural controls. Regulatory Guide 1.33 states in part, that safety related activities will be covered by written procedures. Turkey Point instrumentation and controls maintenance procedures 3-SMI-041.11A, Pressurizer Level Protection Operational Test Channel I LT- 459 and 3-SMI-041.104, Pressurizer Level Protection Channel I Loop Calibration LT- 459 both specified the use of a Fluke Model 8842A multimeter designed for the application. Contrary to the above, the maintenance technicians used a Fluke Model 8846A multi-meter which had different impedance characteristics than the Model 8842A and had not been evaluated for use in this application. The operational test with the wrong meter resulted in unsatisfactory results requiring an instrument calibration. The calibration and return to service of the instrument with the incorrect Fluke meter resulted in resetting the instrument set point to 92.262 percent level, which exceeded the technical specification limit of 92.2 percent. This finding is of very low safety significance because it did not affect the function of other systems used to shutdown the reactor, did not add positive reactivity, or result in mismanagement of reactivity by the operators as screened in IMC 0609 Appendix A, Exhibit 2, Section C, Reactivity Control Systems. This event is documented in the licensee corrective action program as action request number 01836648.
05000259/FIN-2010006-012010Q3Browns FerryFailure To Correct The EECW Valves Throttled Below Analyzed ConditionThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to adequately evaluate and take prompt corrective actions to address a condition adverse to quality related to two Emergency Equipment Cooling Water (EECW) system flow control valves determined to have been throttled below the analyzed 0.125 inch gap for a period of approximately three months. This condition restricted the flow to the cooler due to flow blockage which could have resulted in inoperability of the downstream safety-related Core Spray (CS) pump room heat exchangers. This finding was entered into the licensees corrective action program as PER 257029. The inspectors determined that the licensees failure to promptly address an identified deficiency associated with safety related equipment 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 affects the cornerstone objective to ensure the availability, reliability, and capability of the Core Spray system to respond to initiating events to prevent undesirable consequences; (i.e., core damage) , since it resulted in 2 valves in the core spray system remaining throttled below their analyzed seat to disc clearance for several months after the licensee became aware of this condition, thus subjecting these valves to an increased likelihood of clogging with debris and affecting the reliability of the system. The inspectors determined that the finding was of very low safety significance because the finding was not a design deficiency, did not result in an actual loss of system or single train function, and was not potentially risk significant due to external events. The inspectors determined that this finding directly involved the cross-cutting area of Problem Identification and Resolution, component of the Corrective Action Program and aspect of Through Evaluation of Identified Problems because the licensee did not perform a thorough evaluation of identified problems such that the resolutions address causes and extent of conditions.
05000259/FIN-2010006-022010Q3Browns FerryFailure to Implement the Provisions of Preventative Maintenance (PM) Program Which Contributed to a Manual Reactor ScramThe inspectors identified a finding for the licensees failure to implement the applicable provisions of the Tennessee Valley Authority (TVA) Preventative Maintenance (PM) Program to replace the coil in the solenoid valve controlling the opening of the Unit 3 Condensate Demineralizer bypass valve on the specified PM frequency. Failure of this coil was identified as a contributing cause in Root Cause Analysis for PER 200203, Unit 3 Manual Scram Due to Lowering Reactor Water Level. This finding was entered into the licensees corrective action program as PER 245390. The inspectors determined that the licensees failure to implement the TVA PM program was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during at power operations, since failure to implement the provisions of the PM program increased the likelihood of a component failure which contributed to a plant transient. Specifically the failure of the solenoid coil contributed to a reactor trip. The inspectors determined that the finding was of very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions were not available. The inspectors determined that this finding directly involved the cross-cutting area of Human Performance, component of Work Practices and aspect of Procedural Compliance because licensee personnel failed to follow the guidance contained in the Preventive Maintenance program resulting in a plant transient.
05000259/FIN-2010006-032010Q3Browns FerryFailure to Correct a Condition Adverse to Quality Associated with the 2D Residual Heat Removal (RHR) Room CoolerThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality by failing to implement adequate corrective actions to address degradation in the performance of the 2D RHR room cooler. On July 17, 2009, the 2D RHR room cooler thermal overload failed due to high mechanical vibrations, which the licensee failed to identify and correct prior to a subsequent failure on August 19, 2009. This finding was entered into the licensees corrective action program as PER 261728. The inspectors determined that the licensees failure to implement adequate corrective actions after the 2D RHR motor trip on July 17, 2009 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 in that it adversely affected the reliability of the 2D RHR room cooler to respond to initiating events. The inspectors determined that the finding was of very low safety significance because it did not result in inoperability of a safety function for greater than the allowed technical specification outage time. The inspectors determined that this finding directly involved the cross-cutting area of Problem Identification and Resolution, component of the Corrective Action Program and aspect of Appropriate and Timely Corrective Actions because the licensee did not implement appropriate and timely corrective actions to resolve a condition adverse to quality. Specifically, the problem with the 2D RHR room cooler was not adequately addressed after the motor trip on July 17, 2009.
05000259/FIN-2010006-042010Q3Browns FerryFailure to Correct a Condition Adverse to Quality Associated Cooling Water Flow Degradation in the 1B Core Spray Room CoolerThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality and implement adequate corrective actions for the degraded 1B Core Spray (CS) room cooler. The licensee failed to implement adequate correct actions to address the inability of the room cooler perform its design function with degraded cooling water flow prior to its loss of function on June 25, 2010. The licensee has since replaced the cooler in order to provide additional flow margin. The failure to take adequate corrective actions to address the potential high river temperature along with degraded heat exchanger flow was a performance deficiency. The performance deficiency was more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the availability of the 1B CS room cooler to respond to initiating events. The inspectors determined that a Phase 2 screening was required because the 1B division of core spray was inoperable for greater than the 7 day technical specification allowed out of service time. Using the pre-solved Phase Two significance determination worksheet, the inspectors determined that the finding was of very low safety significance. The inspectors determined that this finding directly involved the cross-cutting area of Problem Identification and Resolution, component of the Corrective Action Program and aspect of Appropriate and Timely Corrective Actions because the licensee did not implement appropriate and timely corrective actions to resolve a condition adverse to quality. Specifically, the licensee failed to address the debris fouling of the 1B CS room cooler prior to its failure on June 25, 2010.
05000259/FIN-2010006-052010Q3Browns FerryFailure to maintain an Adequate Surveillance Procedure to Prevent an Unplanned HPCI IsolationThe inspectors identified a self-revealing non-cited violation of Technical Specifications 5.4.1.a, Procedures, for an inadequate surveillance procedure used to test High Pressure Coolant Injection (HPCI) pressure switches that led to an unplanned HPCI system isolation and HPCI system being declared inoperable. This finding was entered into the licensees corrective action program as PER 239313. The inspectors determined the failure to establish an adequate procedure used for connecting and disconnecting VOMs during testing of pressure switches on the HPCI system was a performance deficiency. The performance deficiency was more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that the licensee did not ensure reliability and availability of the HPCI system to respond to initiating events to prevent undesirable consequences. The inspectors determined the finding was of very low safety significance because HPCI was out of service for a total of about 12 hours and did not exceed its TS allowed outage time per TS 3.5.1.c. The inspectors determined that this finding directly involved the cross-cutting area of Human Performance, component of Resources and aspect of Complete Documentation because the licensee failed to provide an adequate procedure to perform the HPCI surveillance test.
05000259/FIN-2010006-062010Q3Browns FerryInadequate Maintenance Procedure for Siemens Horizontal Vacuum Circuit Breakers Circuit BreakersThe inspectors identified a non-cited violation of Technical Specification (TS) 5.4.1 for the licensees failure to have adequate preventative maintenance procedures for Siemens Horizontal Vacuum Circuit Breakers. Plant procedure EPI-0-000-BKR015, 4KV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and Compartment Maintenance, Revision 28, did not provide specific guidance for checking the tightness of the closing spring charging motor mounting bolts. As a result, on June 15, 2010 while the 3C RHR pump was in service for suppression pool cooling, the charging motor in the pump breaker cubicle became detached from its mount. The charging spring failed to recharge and the pump would not have restarted if needed following a trip of the circuit breaker. The licensee reattached the charging motor and restored the 3C RHR pump to service. The licensee also revised procedure EPI-0-000-BKR015 to include instructions for ensuring the charging motor was securely fastened to the circuit breaker. This finding was entered into the licensees corrective action program as PER 234443. The inspectors determined that the failure to have an adequate maintenance procedure for circuit breaker maintenance was a performance deficiency. This performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and adversely affected the cornerstone objective in that the PM procedure for the breaker did not assure the 3C RHR pump could perform its intended safety functions. The inspectors determined that the finding was of very low safety significance because it did not result in inoperability of a safety function for greater than the allowed technical specification outage time and was not potentially risk-significant due to external events. The inspectors determined that this finding directly involved the crosscutting area of Human Performance, component of Resources and aspect of Complete Documentation because the licensee did not maintain adequate plant procedures for equipment maintenance. Specifically, procedure EPI-0-000-BKR015, Revision 28 did not contain guidance for checking the charging motor bolt tightness resulting in the 3C RHR pump charging motor becoming detached and adversely affecting train operability.
05000302/FIN-2009005-012009Q4Crystal RiverFailure to Follow a Plant Procedure Resulted in an Inoperable HPI SystemA self-revealing Non-Cited Violation (NCV) of Improved Technical Specification (ITS) 5.6.1.1.a was identified for the failure to follow a plant procedure which resulted in a loss of a 480 volt engineered safeguards motor control center (ES MCC)-3B1. Concurrent with pre-existing conditions, the high pressure injection (HPI) system was declared inoperable and ITS 3.0.3 was entered for a period of one hour and 24 minutes. The licensee entered this issue into the corrective action program as nuclear condition report (NCR) 333515. The finding was more than minor since it affected the equipment availability attribute of the mitigating system cornerstone and resulted in ITS 3.0.3 entry for the HPI system being inoperable. The finding was evaluated against NRC Phase 1 Significance Determination Process (SDP) and Phase 2 SDP was required due to a loss safety function of the HPI system. A Regional Senior Reactor Analyst performed a Phase 3 SDP evaluation and concluded this finding was of very low safety significance (Green). The major assumptions of the evaluation were that the HPI function was out of service for exposure period (1 .5 hours) and there would be no recovery of the de-energized motor control center. The dominant accident sequence involved a support system failure of the Emergency Feedwater (EF) Indication and Control System rendering Main Feedwater and automatic control of EF unavailable, operators were unable to manually control EF flow causing its failure and with the HPI function lost due to the performance deficiency, core damage ensued. The inspectors determined the cause of the finding is related to the cross-cutting area of Human performance with a work practices aspect H.4 (c)). Specifically, work scope changes involving safety-related equipment did not receive the appropriate level management oversight resulted in a plant procedural violation
05000302/FIN-2009005-022009Q4Crystal RiverManual Reactor Trip Due to Group 7 Control Rods Insertion Caused by Inadequately Protected Test JumperA self-revealing NCV of Improved Technical Specification (ITS) 5.6.1.1.a was identified for the failure to follow the provisions of preventative maintenance procedure PM-126, Electrical Checks of CRD (Control Rod Drive) Power Train. Failure to follow PM-126 caused the failure of the Group 7 control rod programmer during maintenance and resulted in the unexpected insertion of the Group 7 control rods fully into the core. This unexpected insertion of these control rods into the core caused control room operations personnel to manually trip the reactor from 100 percent power. The licensee entered this issue into the corrective action program as NCR 351705. This finding was determined to be more than minor because it was associated with the initiating events cornerstone attribute of Human Performance, and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during at-power operations. The finding was evaluated using Phase 1 of the At-Power SDP, and was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions were not available. The cause of this finding was directly related to the cross-cutting area of Human Performance with a work practices aspect (H.4 (b)). Specifically, the workers failed to follow the preventative maintenance procedure
05000302/FIN-2009005-032009Q4Crystal RiverLicensee-Identified ViolationThe following issue of very low safety significance (Green) was identified by the licensee and was a violation of NRC requirements. This issue met the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation. 10 CFR 26.205(d) requires, in part, that individuals subject to work hour controls do not exceed 26 work hours in any 48-hour period and 72 work hours in any 7-day period; requires a 34-hour break in any 9-day period; and a 10-hour break between successive work periods. During the period of October 12 to October 19, 2009, one worker exceeded 26 hours in a 48-hour period; nine workers exceeded 72 hours in a 7-day period; five workers did not have a 34-hour break in a 9-day period; and two workers did not have the required 10-hour break between successive work periods. The violation was limited to one work group, Florida Transmission Personnel, who were on-site to support outage work. The licensee determined that the Transmission personnel did not have a firm understanding of the revised 10 CFR Part 26 requirements. The finding was more than minor because, if left uncorrected, it would become a more significant safety concern. Specifically, the excessive work hours would increase the likelihood of human performance errors during plant maintenance activities that could affect equipment performance. The finding is of very low safety significance because no significant events or human performance issues were directly linked to personnel fatigue as a result of the hours worked. This issue was documented in the licensees corrective action program as NCR 361777
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