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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000250/FIN-2007003-03 | 2007Q2 | Turkey Point | Incorrect Connnector Components Cause Inoperability of Multiple Rod Position Circuits | On June 6, 2007, the licensee placed Unit 3 in Mode 3 pursuant to Technical Specification 3.0.3 requirements due to multiple inoperable rod position indication circuits. The licensee initiated an event review to determine the cause of the faulty rod position indications. At the end of the inspection period, information from the license regarding the rod position indication problems, such as the Unit 4 extent of condition and evaluation of human performance aspects of using neoprene inserts in quality related connectors, were needed to resolve this issue. Therefore, pending additional inspection this will remain open as Unresolved Item URI 50-250&251/2007-03-03, Incorrect Connnector Components Cause Inoperability of Multiple Rod Position Circuits. |
| 05000250/FIN-2007003-04 | 2007Q2 | Turkey Point | Availability and Functionality of Unit 3 and Unit 4 Alternate Shutdown Systems | On June 10, 2007, in response to inspector questioning, the licensee found that the alternate shutdown dedicated communications system had not been tested satisfactorily since 2001. In response, the licensee started a review of system operability. At the end of the inspection period, information from the licensee regarding the nature of the individual failures and ability to mitigate communication problems were needed to resolve this issue. Therefore, pending additional inspection this will remain open as Unresolved Item URI 50-250&251/2007-03-04, Availability and Functionality of Unit 3 and Unit 4 Alternate Shutdown Systems |
| 05000250/FIN-2007004-02 | 2007Q3 | Turkey Point | Inappropriate Blanket Overtime Authorization | The inspectors identified a non-cited, SL IV violation of TS 6.8.5 when inappropriate blanket overtime was authorized for thirty-eight electrical maintenance personnel for the entire Unit 3 fall 2007 refueling outage. This issue was promptly discussed with licensee management, the authorization was rescinded, and action was taken by the licensee to manage overtime in accordance with the technical specification requirements. The licensee entered this issue into their corrective action program for resolution. This finding was evaluated using traditional enforcement since it impacted the regulatory process in that the non-compliance with technical specifications was authorized at an executive level, which could become a more significant safety concern. This finding is of very low safety significance because there were no 3 Enclosure actual adverse plant or equipment conditions attributed to worker fatigue |
| 05000250/FIN-2008002-01 | 2008Q1 | Turkey Point | Failure to implement procedures regarding overtime hours for plant operations personnel. | The inspectors identified a Non-cited violation of Technical Specification (TS) 6.8.5 for the failure of the licensee to properly implement procedure requirements regarding the control of overtime for operations personnel that may perform safetyrelated duties. To address this issue the licensee has revised the operations shift scheduling to reduce or eliminate the need for overtime beyond the limits specified in licensee procedures. The finding is greater than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening. If left uncorrected, the excessive work hours could adversely affect the stations defense-indepth and increase the likelihood of human errors during response to plant events and would become a more significant safety concern. The failure to implement requirements for controlling the use of overtime is contrary to TS and is a performance deficiency which could adversely the impact operability to monitor safe operation of the plant and other onsite activities. This issue has a cross-cutting aspect in the area of Human Performance, Resources (Item H.2.(c) of IMC 0305, because sufficient qualified personnel were not available to maintain working hours within working hour guidelines. (4OA2 |
| 05000250/FIN-2008002-02 | 2008Q1 | Turkey Point | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meet the criteria of Section VI of the NRC Enforcement policy, NUREG-1600, for being dispositioned as an NCV. 10 CFR 55.46(c) states, in part, a plant-referenced simulator used for the administration of the operating test or to meet experience requirements in 55.31(a)(5) must demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. 10 CFR 55.46(d) states, in part, that facility licensees that maintain a simulation facility shall conduct performance testing throughout the life of the simulation facility in a manner sufficient to ensure that paragraphs (c)(2)(ii), as applicable, and (d)(3) of 10 CFR 55.46 are met. The intent of the performance testing is to ensure that no noticeable differences exist between the simulator control room and the Unit 3 control room. The results of performance tests must be retained for four years after the completion of each performance test or until superseded by updated test results. Contrary to the above, the licensee identified three simulator performance testing deficiencies as required by 10 CFR 55.46(c) and (d). This finding was considered more than minor because of the potential for negative training. Negative training could have occurred because of the simulator fidelity testing deficiencies. This finding was of very low safety significance because the discrepancy was on the simulator verses the actual plant. Furthermore, no negative training occurred as a result of these performance testing deficiencies. The licensee has entered these deficiencies into their corrective action program (Condition Reports 2008-462, 2008-460, and 2008-1812) |
| 05000250/FIN-2008003-01 | 2008Q2 | Turkey Point | Maintenance causes smoke and fumes to enter the control room causing fire alarms. | A Self-Revealing finding of very low safety significance was identified after smoke and welding fumes from maintenance entered the control room through the ventilation system causing smoke alarms. When identified, the licensee stopped the maintenance and entered the issue into the corrective action program as CR 2008- 17166. The Initiating Events cornerstone was affected when smoke alarms occurred requiring the operators to initiate actions to protect themselves and the plant. The event screened as Green when mitigating systems remained unaffected and would have functioned, if needed. The cause of the finding is related to the cross-cutting area of Human Performance, Work Practices, (H.4.b) when personnel did not follow procedures in developing the work package for metalizing operations outside of the control room. (1R05 |
| 05000250/FIN-2008003-02 | 2008Q2 | Turkey Point | Failure to implement procedures that assure component lineups prior to power escalation. | The inspectors identified a non-cited violation of Technical Specification (TS) 6.8.1, Procedures for failure to implement Unit 4 plant startup requirements regarding alignment of components that support operability of the recirculation sump. When identified, the licensee corrected the alignments and entered the issues into the corrective actions program as CR 2008-15444 and 2008-15505. The Mitigating Systems cornerstone was affected when standby equipment was not in the specified ready lineup. The finding screened to be of very low safety significance when no loss of safety function occurred. The cross-cutting area of Human performance Work Practices (H.4.c) was affected when the licensee did not assure supervisory oversight of work activities (valve lineup and debris gate position) to assure that nuclear safety was supported. (1R20 |
| 05000250/FIN-2008003-03 | 2008Q2 | Turkey Point | Failure to take timely corrective actions leads to emergency diesel generator failure. | A Self-revealing Non-cited violation of 10 CFR 50, Appendix B, Criterion XVI was identified when external corrosion of a Unit 3 emergency diesel radiator was not promptly repaired resulting in a diesel failure. The licensee repaired the radiator and entered the event into their corrective action program as CR 2008-11134. The finding affected the equipment performance attribute of the Mitigating System cornerstone due to the impact on availability and reliability of the EDG system. The finding screened to be of very low safety significance, Green, when the loss of safety function for the single train did not exceed the allowed outage time. The finding involved the cross-cutting area of Problem Identification, and Resolution, (P.1.c), when the licensee did not thoroughly evaluate the radiator corrosion such that the issue could be resolved prior to failure. (4OA2 |
| 05000250/FIN-2008004-01 | 2008Q3 | Turkey Point | Failure to Implement Technical Specification Requirements Regarding Structural Integrity of Reactor Coolant system Components | The inspectors identified a Non-cited violation of Technical Specification (TS) 3.4.10, when the licensee failed to either isolate a flawed ASME Code Class 1 crack or place Unit 4 in a condition where the TS did not apply. As a result, plant operation was continued with a crack leaking boric acid on components, challenging the integrity of the reactor coolant system. When identified to the licensee, Unit 4 was shut down and cooled to a condition where the requirement did not apply and the crack was repaired. The licensee documented the failure to enter the TS as condition report (CR) 2008- 27020. The finding was more than minor because the un-isolated crack challenged the integrity of the reactor coolant system and affected the objective of the Reactor Safety/Initiating Events Cornerstone 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 inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for LOCA Initiators. Because of check valve protection downstream, and with the unimpeded ability to isolate charging upstream, the finding screened as having very low safety significance (Green) |
| 05000250/FIN-2008004-02 | 2008Q3 | Turkey Point | Licensee-Identified Violation | Turkey Point Technical Specification (TS) 3.5.2.a requires four operable safety injection pumps with discharge aligned to the reactor coolant system (RCS) cold legs when RCS temperature is greater than 380 degrees F. TS 3.0.3 requires that when a limiting condition for operation is not met, within one hour, action shall be initiated to place the unit in Mode 5. Contrary to the above, on May 5, 2008, Unit 4 temperature exceeded 380 degrees F with no safety injection pumps aligned to the RCS, and appropriate action was not initiated. When discovered after about 5 hours of operation above 380 degrees, safety injection isolation valve 4-867 was opened and the technical specification actions were no longer applicable. The licensee documented the event in their corrective action program as CR 2008-15431 and initiated a comprehensive root cause evaluation. In their evaluation, the licensee identified as one cause the fact that valve alignments had not been completed as required by licensee procedures. Having a mitigating system valve in the incorrect position was a performance deficiency and the misalignment was evaluated using the NRC significance determination process |
| 05000250/FIN-2008004-03 | 2008Q3 | Turkey Point | Licensee-Identified Violation | Turkey Point TS 3.6.1.3 requires that both containment air lock doors be closed except for transit entry and exit. The licensee implements this requirement, in part, with procedure 0-ADM-009, Containment Entries When Containment Integrity is Established, which states, in step 5.1.10, that to comply with technical specification 3.6.1.3, both personnel hatch airlock doors shall be maintained closed when actual transit is not in progress. Contrary to the above, on September 18, the inner personnel hatch airlock door was not maintained closed when two personnel entered the Unit 4 containment for plant operations (operate lighting, search for and operate a valve). The condition lasted for approximately 25 minutes until the individuals exited containment and closed the inner door. The finding was more than minor because it was associated with the integrity of the reactor containment, and was of very low safety significance because one containment door remained closed during the entry. When identified to the licensee on September 24, 2008, after an individual questioned procedure compliance from the entry, the licensee documented the occurrence in condition report 2008-29525, and initiated an investigation. |
| 05000250/FIN-2008004-04 | 2008Q3 | Turkey Point | Licensee-Identified Violation | Turkey Point TS.9.11 requires that the spent fuel storage pool level be maintained greater than or equal to 56 feet, 10 inches (56.83 feet). The licensee implements this requirement by implementing procedure 4-OSP-201.3, Nuclear Plant Operator Daily Logs, which requires level be maintained above 56.91 feet and that when an abnormal condition that could affect a technical specification is logged, the shift manager will be notified and corrective actions will be noted. Contrary to the above, on September 20 and 21, 2008, abnormal spent fuel pool level (56.90 feet) was identified in the operator daily logs and neither the shift manager was notified nor were corrective actions noted. As a result, on September 22, 2008, a level below the technical specification minimum was logged (56.81 feet). The finding was more than minor because it was associated with the spent fuel cooling system radiological barrier and was of very low safety significance because the level was corrected within the allowed technical specification time. When identified to supervision, the level was raised by adding water to the pool and the issue was documented in the corrective action program as CR 2008-26431. A cause evaluation was initiated as part of the corrective action report |
| 05000250/FIN-2008005-01 | 2008Q4 | Turkey Point | Failure to Accomplish An Activity Affecting Quality in Accordance with Procedures | NRC issued a Severity Level (SL) IV violation to FPL on December 23, 2008, for failure to accomplish an activity affecting quality in accordance with procedures. Specifically, a supervisor failed to follow licensee procedure 0-OSP - 040.8, Reactivity Deviation from Design Calculation, when he reviewed and approved an incorrect (i.e., not current) boron sample that was collected several hours before the reactivity calculation was performed. This violation is being tracked in this inspection report as SL IV VIO 05000250, 251/2008-005-01: Failure To Accomplish An Activity Affecting Quality In Accordance With Procedures |
| 05000250/FIN-2008005-02 | 2008Q4 | Turkey Point | Licensee-Identified Violation | Technical Specification 3.6.1.7, Action b. requires that with a containment purge supply valve having a measured leakage exceeding limits, restore the inoperable valve to operable status or be in hot standby within 6 hours and cold shutdown within the next 30 hours. Contrary to the above, on March 1, 2008, during a local leak rate test, the licensee found that purge valve POV-3-2600 exceeded allowed leakage limits and the reactor had not been placed in cold shutdown as required. The violation existed for about 61 hours while the reactor was in hot standby (Mode 3). When identified during the surveillance test, the valve was exercised, lubricated and satisfactorily returned to operable status within about one hour. The violation was of very low safety significance because the redundant isolation valve, POV-3- 2601, remained closed during the testing and was available to satisfy the safety function |
| 05000250/FIN-2009002-01 | 2009Q1 | Turkey Point | Licensee-Identified Violation | Turkey Point Technical Specification 3.6.3 requires three operable emergency containment filtering units when Unit 3 is operated in Modes 1 thru 4. Further, with one filtering unit inoperable, restore the inoperable unit to operable status within 7 days or be in Hot Standby within 6 hours and Cold Shutdown within the following 30 hours. Contrary to the above, as of August 27, 2008, Unit 3 containment filtering unit 3B was inoperable in excess of 7 days because of an inadequate electrical design that had occurred years before and no action was taken to place the unit in the required configuration. The problem was discovered by the licensee during an engineering review in preparation for circuit modification and was corrected on August 28, 2008 by modifying the circuit to eliminate the design flaw. Because redundant filtering units were not affected, and the affected unit would be manually started if required, the issue was of very low safety significance (Green). The issue was documented in the licensee corrective action program as CR 2008-27014 and reported to the NRC in Licensee Event Report 05000250/2008-004-00 |
| 05000250/FIN-2009002-02 | 2009Q1 | Turkey Point | Licensee-Identified Violation | Turkey Point Technical Specification 6.8.5 requires that administrative procedures be implemented to limit the working hours of personnel who perform safety related functions, and that any deviation from the guidelines be authorized by department managers or higher. The licensee implements these requirements with procedure QI 1- PTN-1 which states in paragraph 5.8.1 that to the extent practicable, personnel are not assigned to shift duties while in a fatigued condition that could significantly reduce their mental alertness or their decision making ability. Additionally, in paragraph 5.8.6, the procedure states that the circumstances of the extraordinary action shall be documented on an overtime deviation request form and that each deviation requires a separate deviation form. Contrary to the above, the licensee had identified multiple examples (listed below) where deviations from the working hour guidelines had occurred without documenting the circumstances of the extraordinary action on a separate deviation form. The issue was entered into the corrective actions program as CR 2008-31143. The licensee has planned and implemented actions noted in the following CRs to prevent exceeding the working hour limits on any routine basis: 1. (CR 2008-17179) Deviation request completed for meeting of up to 33 operations personnel held on May 14, 2008 for work authorization to exceed 24 hours in a 48 hour period or 72 hours in a 7 day period, but no authorization was written for the same individuals exceeding these limits in the subsequent shifts. 2. (CR 2008-17180) Deviation request completed for meeting of up to 35 operations personnel held on May 20, 2008 for work authorization to exceed 24 hours in a 48 hour period or 72 hours in a 7 day period, but no authorization was written for the same individuals exceeding these limits in the subsequent shifts. 3. (CR 2008-21249) Deviation request completed for two operations personnel on June 27, 2008 for work authorization to exceed 72 hours in a 7 day period, but no authorization was written for these individuals subsequently exceeding these limits on June 29 and June 30, 2008. 4. (CR 2008-15659, and 2008-14968) Deviation request completed after the fact for two operators assigned to exceed 24 hours in a 48 hour period or 72 hours in a 7 day period, on April 30, 2008 and no deviation request was completed for exceeding these limits on May 1, 2008 |
| 05000250/FIN-2009003-01 | 2009Q2 | Turkey Point | Failure to Implement Procedures for Conducting a Valve Alignment Causes Spill of Reactor Coolant And Contamination Of a Plant Employee | A Self-revealing Non-cited Violation of Technical Specification (TS) 6.8.1 was identified for failure to follow procedures that assure that valves are maintained in the proper positions. As a result of mis-positioning of letdown system valves, a spill of reactor coolant from the Unit 3 letdown system occurred onto the auxiliary building roof and a security officer was contaminated. The licensee documented this in CR 2009-14469.The finding was more than minor because it affected the Human Performance attribute of Initiating Events cornerstone and if failure to implement valve position controls were left uncorrected it would have the potential to lead to a more significant safety concern. The inspectors evaluated the finding using NRC Inspection Manual 0609, Attachment0609.04, SDP Phase 1. Because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available, the finding was screened as Green. The cross-cutting element of Human Performance, Work Practices, Human Performance & Error Prevention (H.4(a)), was affected when the licensee did not properly document activities regarding the failure to position valves in accordance with a specified valve lineup |
| 05000250/FIN-2009003-02 | 2009Q2 | Turkey Point | Failure to Implement TS Requirements Resulting From Loss of Configuration Control of the 3C Main Steam Isolation Valve | A Self-Revealing Non-cited violation of TS 3.7.1.5 requirements was identified when the Unit 3 C main steam isolation valve (MSIV) failed to close on demand on May4, 2009. Licensee evaluation has found the root cause of the failure to be an inadequate post maintenance test after maintenance that resulted in the air throttle valve for the MSIV being left in the closed position. When identified, the licensee placed the throttle valve in the correct position and tested the valve stroke time satisfactorily. The licensee documented this in CR 2009-13568.The finding was more than minor because it affected the Configuration Control attribute of the Mitigating Systems cornerstone and the failure of the MSIV to close when demanded challenged the integrity of the main steam system for isolating steam system or generator tube ruptures. The inspectors evaluated the finding using NRC Inspection Manual 0609, Attachment 0609.04, SDP Phase 1 and SDP Phase 2. An initial SDP Phase 2 screening of the finding revealed a greater than green result for Large Early Release Probability (LERF) and Phase 3 was required. A Regional Senior Reactor Analyst performed a Phase 3 evaluation of the performance deficiency and classified the finding of very low safety significance (Green). The major assumption was predicated on the information in NUREG 1806, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10CFR50.61), which indicated that the possibility of core damage was remote following an extreme cool down due to a Main Steam Line Break without isolation. The cross-cutting aspect of Human Performance, Work Practices, Human Performance & Error Prevention (H.4(a)) was affected when personnel did not practice error prevention techniques such as self and peer checking, and properly document activities |
| 05000250/FIN-2009003-03 | 2009Q2 | Turkey Point | Failrue to Maintain Lighting Impedes Compensatory Measure For Failed Fire Detection | The inspectors identified a Green finding for failure to correct failed lighting in a Unit 4 electrical penetration room that prevented the hourly rover from adequately compensating for fire detection that was out of service. The inspectors determined that maintaining lighting in areas of degraded fire protection features is not a specific NRC requirement. The licensee documented this in CR 2009-17533.The finding was more than minor because it affected the External Event attribute of the Mitigating Systems cornerstone and failure to correct a problem that impacted the ability of fire watch personnel to adequately compensate for out of service fire detection equipment could reasonably be viewed as a precursor to a significant fire event. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination. The finding was screened as Green because the assigned fire degradation rating was low. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, Appropriate & Timely Corrective Actions (P.1(d)) because the licensee did not document and correct a problem that was previously identified |
| 05000250/FIN-2009003-04 | 2009Q2 | Turkey Point | Failure to Assure That Design Controls Were Maintained During Maintenance On The 3B Main Steam Isolation Valve (MSIV) | The inspectors identified a Non-cited violation of 10 CFR50, Appendix B, Criterion III, Design Control when maintenance personnel failed to follow procedure during reassembly of 3B main steam isolation valve and did not maintain proper configuration of a safety-related component. The licensee documented this in CR 2009-11481.The finding was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone, and it affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences, such as the 3B MSIV. Using Manual Chapter 0609, Attachment 0609.04, Phase 1 screening, this issue was determined to be of very low safety significance because the design deficiency did not result in loss of operability. The cross-cutting element of Human Performance, Work Practices (H.4.(b)) was affected when the licensee did not effectively communicate expectations regarding procedural compliance and contractor personnel did not follow procedures |
| 05000250/FIN-2009003-05 | 2009Q2 | Turkey Point | Failure to Implement Design Controls When Modifying Safety Equipment During Painting Activities | A Self-revealing Non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V was identified for failing to implement procedures that assure design control during an alteration to the 4C intake cooling water pump motor, a safety-related component. As a result, the running Unit 4 C intake cooling water pump experienced a high temperature condition and was stopped by operators. The pump may not have been able to complete its design function with the alteration that restricted the cooling air flow for the motor during painting activities. The licensee documented this in CRs 2009-15970 and2009-16336.The finding was more than minor because it affected the Human Performance attribute of the Mitigating Systems cornerstone and the licensee did not complete an engineering evaluation of the modification causing a high temperature condition on the motor to assure that the motor could perform its design functions. Also, NRC Inspection Manual Chapter 0612, Appendix E, Example 4.a was applicable (failure to perform an engineering evaluation with missed opportunities for licensee identification) and the finding was more than minor. The finding screened as Green using NRC Inspection Manual Chapter 0609, Attachment 0609.04, SDP Phase 1 screening because the finding did not result in a loss of function of a single train of TS equipment for greater than the allowed outage time of 14 days. The finding affected the cross-cutting area of Human Performance, Work Practices, Supervisory & Management Oversight (H.4(c)) because the licensee did not ensure supervisory oversight of work activities, including contractors, such that nuclear safety is supported |
| 05000250/FIN-2009003-06 | 2009Q2 | Turkey Point | Inadequate Evaluation of Damaged Rod Control Extension Results in High Risk Condition and Condition Yellow | A Self-revealing Finding was identified when the licensee did not manage maintenance activities adequately to identify and repair a damaged rod control drive component on Unit 3 prior to setting the reactor vessel closure head on the reactor vessel flange. As a result, the subsequently filled reactor coolant system had to be drained again to 2 feet below the reactor vessel flange (a high risk activity) placing the unit in the licensees risk condition Yellow for repairs. The licensee documented this in condition report (CR) 2009-10284.The finding was more than minor because it affected the Human Performance attribute of Initiating Events cornerstone and the licensee=s risk assessment failed to anticipate that the maintenance activity could result in another plant draining evolution with its inherent risk of an initiating event of loss of inventory or shutdown cooling. With appropriate mitigating equipment available, the finding screened to be of very low safety significance (Green). The finding affected the cross cutting area of Human Performance, Work Practices, Supervisory & Management Oversight (H.4(c)) because the licensee did not appropriately provide oversight of work activities, including contractors, such that nuclear safety is supported |
| 05000250/FIN-2009003-07 | 2009Q2 | Turkey Point | Failure to Implement TS Requirements Regarding Structural Integrity of Code Class 2 Main Steam Isolation Components | The inspectors identified a Non-cited violation of TS 3.4.10 requirements on Unit3 regarding required components, when plant operation continued although a structural flaw in Class 2 main steam isolation valve steam trap piping had been identified. As a result of using an incorrect drawing in assessing the leak, plant operation continued although a plant shutdown should have been initiated. The licensee documented this in CR 2009-15284.The finding was more than minor because it affected the RCS equipment and barrier performance attribute of the Barrier Integrity cornerstone and the un-isolable through wall leak challenged the integrity of the main steam system for isolating steam generator tube ruptures. Using Manual Chapter 0609, Attachment 0609.04, Phase 1 screening, this finding was determined to be of very low safety significance because all containment barrier characterization answers marked as No. The cross-cutting element of Human Performance, Decision Making, Conservative Assumptions & Safe Actions (H.1 (b)) was affected when the licensee did not use conservative assumptions in evaluating a Class 2component flaw and its TS implications, and did not demonstrate that continued operation with the crack was safe in order to proceed |
| 05000250/FIN-2009004-01 | 2009Q3 | Turkey Point | Licensee-Identified Violation | Technical Specification Table 3.3-1, functional Unit 20, requires the reactor trip system trip logic to be operable. Technical specification 3.0.3 requires that action be taken within one hour to place the unit in Hot Standby within the next six hours. Contrary to the above, for the period from July 14 2008, until October 11, 2008, reactor trip system logic for undervoltage protection was not operable, and action was not taken to shutdown the unit as required. When discovered on October 11, 2008, an investigation was initiated, the reactor protection trip logic circuitry was altered and relays were replaced to restore the system to an operable configuration. The issue was documented in the licensees corrective action program as Condition Report 2009-28058. This finding was of very low safety significance because redundant reactor protection features remained available to assure safety should an undervoltage condition occur |
| 05000250/FIN-2009004-02 | 2009Q3 | Turkey Point | Licensee-Identified Violation | Technical Specification 3.6.4 requires each containment isolation valve be Operable or, Either restore the valve to operable status or isolate the affected penetration within 4 hours by use of at least one closed manual valve. Contrary to the above, on June 1, 2008, secondary system containment isolation valve CV-3-6275C failed a stroke test, was declared inoperable, and the affected penetration was not closed within 4 hours as required. When identified by the licensee during investigation of the failed test, on June 9, Technical specification 3.6.4 was invoked and the penetration was isolated by closing a manual isolation valve. The issue was documented in the corrective action program as CR 2008-18474. The finding was of very low safety significance because manual isolation of the penetration remained available, if needed in an event, and redundant mitigating trains of auxiliary feedwater remained available |
| 05000250/FIN-2009005-01 | 2009Q4 | Turkey Point | Failure to Implement Required TS Controls for a High Radiation Area with Dose Rates in Excess of 1000 mrem/hr | A Self-revealing Non-cited Violation of Technical Specification (TS) 6.12.2, was identified for failure to meet high radiation area (HRA) control requirements for an accessible location, i.e., Unit 4 (U4) reactor auxiliary building (RAB) roof, with radiation levels greater than 1000 millirem per hour (mrem/hr) during refueling activities. Specifically, on November 3, 2009, general area dose rates exceeding 1000 mrem/hrwere identified outside of an established HRA posted barricade on the RAB roof adjacent to the outside wall of the Spent Fuel Pool (SFP) building. The HRA posted barricade, i.e., locked-HRA (LHRA) barrier, was established to delineate an area outside of which dose rates would not exceed 1000 mrem/hr. The licensee documented this issue in condition report (CR) 2009-31494.The finding was more than minor because it affected the Program and Process(exposure control) attribute of the Occupational Radiation Safety cornerstone and the failure of the licensee to implement proper HRA controls which could have led to unanticipated worker exposures. The inspectors evaluated the finding using the Occupational Radiation Safety Significance Determination Process and determined the issue to be of very low safety significance (Green) based on High Radiation Area controls in place for the subject area. The cross-cutting element of Human Performance, Decision-Making (H.1(b)) was affected when the licensee failed to conduct adequate radiological surveys needed to demonstrate compliance with TS HRA requirements for locations potentially having dose rates exceeding 1000 mrem/hr during current Unit 4 refueling activities (2OS1) |
| 05000250/FIN-2009005-02 | 2009Q4 | Turkey Point | Evaluate Inappropriate Characterization of Reactor Coolant System (RCS) Filters for Transportation and Disposal | The inspectors identified an Unresolved Item (URI) regarding the significance of the inappropriate characterization of Reactor Coolant System (RCS) filters for transportation and disposal. During review of records related to a shipment of RCS filters (shipment W-09-36), the inspectors noted that the filter radionuclide concentrations were based on samples of Chemical and Volume Control System (CVCS) resin rather than representative samples of the filter media. This is contrary to the guidance in NRCs BTP on Waste Classification, Information Notice 86-20, and various industry reports. These documents describe spent resin and primary filters as separate waste streams that require independent, representative, sampling of each. This is due to the different properties of ion exchange resins and mechanical filters which tend to concentrate radioactive contaminants in differing concentrations. Discussions with shipping/radwaste staff indicated that this has been the practice for approximately five years. In order to disposition the significance of this finding, the NRC requires a comparison of10 CFR Part 61 analyses for the CVCS resin and RCS filter waste streams. The BTP states that, The staff considers a reasonable target for determining significant differences between measured or inferred radionuclide concentrations in separate samples is that the concentrations are accurate to within a factor of 10. If significant differences are identified, an analysis of the impact (significance) on any previous filter shipments to determine if any waste was misclassified also may be necessary. URI05000250/251, 2009005-02, Evaluate Inappropriate Characterization of Reactor Coolant System (RCS) Filters for Transportation and Disposal. |
| 05000250/FIN-2009005-03 | 2009Q4 | Turkey Point | Violation of Technical Specification 5.5.1.1 regarding Unit 3 spent fuel storage with degrading Boraflex poison | The inspectors identified an apparent violation of Technical Specification 5.5.1.1requirements regarding storage of fuel assemblies in the Unit 3 spent fuel pool when Keff limits for fuel configurations were not maintained using methods described in the Final Safety Analysis Report, potentially leading to a loss of shutdown margin should a dilution event occur in the pool. When identified to the licensee, the spent fuel pool boron concentration was administratively increased and other actions were planned to restore compliance. This finding was considered more than minor because the design control attribute that assured fuel assemblies remain subcritical in the spent fuel pool was affected. The finding was determined to potentially have greater significance because of the lack of both criticality monitoring capability in the spent fuel pool and procedures for responding to an inadvertent criticality. The inspectors evaluated this finding against NRC IMC 0609Phase 1 Screening Worksheet for Initiating Events, Mitigation Systems, and Barriers Cornerstones. The inspectors determined that IMC 0609, Appendix M is required to determine the level of safety significance of this finding because the existing SDP guidance is not adequate to provide reasonable estimates of the finding significance within the established SDP timeliness goal of 90 days. NRC staff is currently reviewing this finding to determine the level of safety significance or enforcement aspect of the issue. |
| 05000250/FIN-2010002-01 | 2010Q1 | Turkey Point | Failure to implement design controls in a temporary modification. | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for failing to maintain control of temporary equipment installed on unit 4 A residual heat removal pump piping when the permanent component cooling water flow indication to the pump seal failed high. Operators were using a controlotron as a compensatory measure to verify adequate cooling flow to the unit 4A residual heat removal pump seal and to assure operability of the unit 4A residual heat removal pump. If the controlotron had failed, the operators would not have received a component cooling water low flow alarm in the control room, lack of cooling flow to the pump would have gone undetected, and operability of the residual heat removal pump could have been affected. The inspectors identified the licensee failed to follow the temporary system alteration procedure to ensure design adequacy and to determine if the alteration required a 10 Code of Federal Regulations (CFR) 50.59 evaluation and NRC approval. The licensee documented this in the corrective action program as condition report 2010- 479. The finding is more than minor because it affected the configuration control attribute of the Mitigating Systems Cornerstone in that it reduced the reliability of the 4A residual heat removal pump with the permanent flow indicator out of service while using an unevaluated controlotron to determine continued operability of the 4A residual heat removal pump. The inspectors screened the finding using NRC Inspection Manual Chapter 0609, Significance Determination of Reactor Inspection Findings for At Power Operations, Phase 1 screening. The finding was of very low safety significance because the design or qualification deficiency did not result in actual loss of operability or functionality of the pump. The cross cutting aspect of Human Performance, Work Practices (H.4(b)) was affected. (1R18) |
| 05000250/FIN-2010002-02 | 2010Q1 | Turkey Point | Licensee-Identified Violation | Technical Specification 3.9.13 requires that with one containment radiation monitor out of service during core alterations, core alterations may continue as long as the containment ventilation isolation valves be maintained shut and within one hour, operate the control room ventilation system in the recirculation mode. Contrary to the above, on April 1, 2009, and on prior occasions, core alterations continued with one channel of control room isolation actuation out of service and without the control room ventilation in the recirculation mode. The non-compliance was identified during review of plant conditions while performing engineered safeguards integrated testing with the plant in Mode 6 refueling. The redundant radiation monitoring and actuation channel remained available and had an event occurred, operators would have been able to use standby Self-contained breathing apparatus (SCBA) assuring the safety function. The issue was screened to be of very low safety significance (Green). When identified, the licensee placed the control room in recirculation and isolated the ventilation valves. The issue was documented in condition report 2009-9899 and additional corrective actions were specified. Because the licensee identified the issue and documented it into their corrective action program, and because the finding is of very low safety significance, this violation is being treated as a licensee identified NCV consistent with Section VI.A of the NRC Enforcement Policy |
| 05000250/FIN-2010003-01 | 2010Q2 | Turkey Point | Failure to perform adequate surveys to ensure proper estimation of radionuclide concentrations in mechanical filter waste shipments | The inspectors identified a Green non-cited violation (NCV) of 10 CFR Part 20.1501(a) for the failure to perform adequate surveys to meet the requirements of 10 CFR Part 20 Appendix G. 10 CFR Part 20 Appendix G states that shippers of radioactive waste must identify and quantify radionuclides contained in each waste container. Specifically, the inspectors determined that the use of resin samples to characterize three shipments of mechanical filters in calendar years 2008 and 2009 was inadequate to ensure proper identification and quantification of the radionuclides present in each container. The licensee entered the issue into their corrective action program as condition report (CR) number 2009-32955 The finding is more than minor because it is associated with the Public Radiation Safety cornerstone attribute of Programs and Processes and adversely affects the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The finding was assessed using the Public Radiation Safety Significance Determination Process (SDP). Based on the fact that subsequent follow up analyses demonstrated that none of the filter waste was under-classified, the finding was determined to be of very low safety significance (Green). This finding has a crosscutting aspect of Human Performance, Decision Making (H.1(b)), because the decision to use resin samples to characterize filter shipments was based on incorrect assumptions, i.e., that spent resin samples would be representative of the filter waste stream, and those assumptions were not demonstrated to be conservative prior to implementation. |
| 05000250/FIN-2010003-02 | 2010Q2 | Turkey Point | Failure to implement TS requirements regardin rod position indication | A Self-Revealing Non-cited Violation of Technical Specification 3.1.3.1.b requirements was identified on Unit 3 when position indication for two rod control cluster assemblies (RCCs) drifted out of tolerance with the associated rod group position indication. Contrary to technical specification requirements, rod positions were neither re-aligned with the group counter nor was reactor power reduced to less than 90 percent within the allowed one hour action time with a potential consequence of challenging accident analysis assumptions. The issue was documented in the corrective action program as CR 2010-14724 The finding was more than minor because if inaccurate rod position indication was left uncorrected, there was a possibility of an adverse affect of an actual rod misalignment beyond that assumed in accident analyses. The Initiating Events cornerstone was affected because rod position alignment assures that accident analysis assumptions are maintained. The inspectors evaluated the finding using NRC Inspection Manual 0609, Attachment 0609.04, Initial Screening and Characterization of Findings and classified the finding of very low safety significance (Green) using the Transient Initiator tool. The cross-cutting aspect of Human Performance, Decision Making (H.1.a) was affected when supervisory personnel did not implement their roles and authorities to ensure safety by implementing Technical Specification requirements. |
| 05000250/FIN-2010004-01 | 2010Q3 | Turkey Point | Failure to provide adequate instructions when working on the reactor protection system results in reactor trip | A self-revealing non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee started corrective maintenance on the Unit 4 reactor protection system with an inadequate procedure. As a result, a reactor trip occurred when a reactor trip circuit was not placed on bypass as an initial condition needed to safely complete the work. During the event investigation, the licensee determined that neither the work order, nor the pre-job review identified the need to place the affected train of the reactor protection system on the bypass breaker. The finding was determined to be more than minor because it affects the Initiating Events cornerstone attribute of procedure quality and adversely affected the cornerstone objective to limit the likelihood of an event that upsets plant stability by resulting in a reactor trip. The finding was evaluated in accordance with IMC 0609, Attachment 4, and determined to be of very low safety significance (Green) per SDP Phase 1 determination because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. This finding has a cross-cutting aspect in the area of Human Performance, Work Control H.3(b) because the licensee did not appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of the operational impact of work and plant conditions that may affect work activities, resulting in a reactor trip. |
| 05000250/FIN-2010004-02 | 2010Q3 | Turkey Point | Licensee-Identified Violation | Technical Specification 3.8.1.1.b requires restoration of an inoperable diesel generator to operable status within 14 days or be in at least hot standby within the next 6 hours. Contrary to the above, during the period July 28, 2009 thru August 12, 2009, a period in excess of 14 days, the Unit 4 B emergency diesel generator was inoperable because of a faulty fuel strainer and Unit 4 was not placed in hot standby as required. When identified by the licensee during a surveillance test, the licensee entered the issue in the corrective action program as CR 2009-13740 and corrected the condition by replacing the entire fuel strainer assembly. A regional Senior Reactor Analyst performed a Phase III evaluation under the Significance Determination Process. The dominant internal events accident sequence was a single unit loss of offsite power followed by the common cause failure of all the emergency diesel generators with a failure to recover offsite power or an emergency diesel generator before core damage two hours later. Assumptions of the evaluation included that common cause would be considered and, recovery of the failed diesel generator would not be considered. The exposure time for the evaluation was 14.3 days. Based upon this evaluation the performance deficiency was characterized as a finding of very low safety significance (Green). Further corrective actions to modify the strainer were planned. |
| 05000250/FIN-2010004-03 | 2010Q3 | Turkey Point | Licensee-Identified Violation | Technical Specification 3.8.1.1.b requires restoration of an inoperable diesel generator to operable status within 14 days or be in at least hot standby within the next 6 hours. Contrary to the above, during the approximate 1.5 month period prior to June 7, 2010, a period in excess of 14 days, the Unit 3 B emergency diesel generator was inoperable because of a failed fuel oil transfer pump and Unit 3 was not placed in hot standby as required. When identified by the licensee during a surveillance test, the licensee entered the issue in the corrective action program as AR 406564 and corrected the condition by replacing the failed transfer pump. This violation was of very low safety significance (Green) because alternate methods of filling the fuel oil day tank using available transfer pumps were proceduralized and available. Further corrective actions to develop a preventive maintenance activity to prevent similar failures were planned. |
| 05000250/FIN-2010004-04 | 2010Q3 | Turkey Point | Licensee-Identified Violation | Technical Specification 3.8.1.1.b requires restoration of an inoperable diesel generator to operable status within 14 days or be in at least hot standby within the next 6 hours. Contrary to the above, from April 27 through May 10, 2010, the unit 4A emergency diesel generator was inoperable approximately 16.9 days because of speed sensing magnetic pickup damage due to it being set too close to the engine flywheel during maintenance. When identified by the licensee during a surveillance test, the licensee entered the issue in the corrective action program as AR 406620 and corrected the condition. Additional corrective actions were implemented under the AR. A regional Senior Reactor Analyst performed a Phase III evaluation under the Significance Determination Process. The dominant internal events accident sequence was a single unit loss of offsite power followed by the common cause failure of all the emergency diesel generators with a failure to recover offsite power or an emergency diesel generator before core damage two hours later. Assumptions of the evaluation included that common cause would be considered and, recovery of the failed diesel generator would not be considered. The exposure time for the evaluation was 16.9 days. Based upon this evaluation the performance deficiency was characterized as a finding of very low safety significance. |
| 05000250/FIN-2010004-05 | 2010Q3 | Turkey Point | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion V, requires, in part, that activities affecting quality shall be prescribed by documented instructions or procedures of a type appropriate to the circumstances. The licensee implements this requirement, in part, with administrative procedure 0-ADM-503, Temporary System Alteration, which requires that a temporary system alteration design review be performed to ensure that the design of the alteration is consistent and compatible with the system and complies with the component design basis. Contrary to the above, from September 2 to 6, 2010, a temporary alteration was installed on the 3B and 4A vital batteries without a design review to assure the installation was consistent with the design basis. When the unauthorized modification was identified to the licensee by a plant operator during rounds, the structures were removed and an evaluation initiated. The licensee determined that the batteries affected remained capable of their design function and this issue was documented into the corrective actions program as AR 577944. The finding was more than minor using NRC Manual Chapter 0612, Appendix E, Example 4.a because there were previous examples where the licensee failed to perform engineering evaluations for plant activities (NRC Inspection Reports. |
| 05000250/FIN-2010005-01 | 2010Q4 | Turkey Point | Inappropriate procedure guidance results in degradation of boration flow path and loss of charging flow | The inspectors identified a Non-cited violation (NCV) of Technical Specification 6.8.1, Procedures, when plant alarm response and off-normal procedures were not adequate to prevent lifting of a charging relief valve. As a result, during operations to assure adequate seal injection flow, a charging throttle valve was shut causing lifting of a charging system relief, diversion of charging flow, and degradation of the boration flow path. When identified to the licensee by the inspectors during review of charging system anomalies, the licensee documented the occurrence in the corrective action program as CR 595200 and upgraded procedures. Although the event occurred on Unit 3, similar procedures existed on Unit 4. While attempting to regulate RCP seal injection flow, operators shut charging throttle valve HCV-3-121. This was a performance deficiency, in that it caused lifting of the charging relief valve(s), diversion of charging flow, and subsequent failure of a charging relief valve. The relief valve failure reduced the reliability of charging flow to the loops and affected the ability of the charging system to perform its design functions including providing for reactivity control, maintaining the proper water inventory in the reactor coolant system, and providing RCP seal injection flow. The issue was more than minor. 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 function, with some level of charging or seal flow remaining throughout the event. All screening questions were answered No. The Mitigating Systems cornerstone was affected when charging capability and the boration flow path was degraded by diversion of flow through the relief valve back to the charging pump suction. The finding affected the cross-cutting aspect of Human Performance, Resources, when operating procedures did not adequately provide accurate guidance to prevent mis-operation (shutting) of the charging throttle valve. |
| 05000250/FIN-2010005-02 | 2010Q4 | Turkey Point | Scaffold blocked access to fire areas used in a control room evacuation event | The inspectors identified a Non-cited violation (NCV) of Turkey Point License Condition 3.D, Fire Protection, when scaffolding was placed as a barricade against personnel access to doors to fire zones 108B and 104. The barricade impeded access to the 3B and 3A DC Equipment rooms through doors that are used in the event of a control room evacuation event and may have delayed or prevented operator actions to mitigate a potential fire. When identified to the licensee, the scaffolding was promptly removed and the problem was documented in AR 594112. The issue was more than minor because the objectives of the Mitigating Systems Cornerstone were affected. Using NRC Manual Chapter 0609, Appendix F, the inspectors assigned a moderate degradation rating to the deficiency because of the likely inability of the plant operators being able to implement the procedural actions within the licensee stipulated time. A regional Senior Reactor Analyst evaluated the performance deficiency under the Phase 3 protocol of the Significance Determination Process. Based upon the results of that evaluation, the performance deficiency was characterized as of very low safety significance (Green) for both units. The evaluation was performed via hand calculation using elements of NRC Manual Chapter 0609, Appendix F, NUREG-6850 as amended by Frequently Asked Questions under the National Fire Protection Association 0805 pilot program. A simplified Reactor Coolant Pump (RCP) seal Loss of Coolant Accident (LOCA) failure probability based upon Westinghouse high temperature seals was used. Key human failure probabilities were estimated using standard techniques. Conditional core damage probabilities, due to a spurious Safety Injection, were derived from the licensee\'s most current model results. Major assumptions and dominant accident sequence for Units 3 and 4 were discussed and included in analysis section of 1R05 in the inspection report. The cause of the finding was related to the cross-cutting aspect of Human Performance, Work Control (H.3(a)) when the scaffold-barricade was constructed without a planned contingency or compensatory measure to assure that the fire mitigation activity could be accomplished within design time constraints. |
| 05000250/FIN-2010005-03 | 2010Q4 | Turkey Point | Welders failed to measure preheat and interpass temperatures | The inspectors identified a Non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings associated with licensee contract personnels failure to adhere to welding procedures during the 2010 Unit 3 refueling outage. Specifically, welders failed to measure preheat and interpass temperatures in ASME safety class containment spray pump lines using contact pyrometers, thermocouples, or temperature indicating crayons as required by procedure. As part of the immediate corrective actions, the licensee conducted a stand-down for welders to reinforce procedural compliance expectations. The licensee performed an extent of condition evaluation and entered the issue into their corrective action program as AR 585550. The inspectors determined that the finding was more than minor because if left uncorrected, it would have become a more significant safety concern. Specifically, the failure to adhere to the welding procedures for temperature measurement affected the assurance that appropriate welding temperatures were maintained. Inadequate temperatures during welding can result in stainless steel sensitization and susceptibility of the weld to failure from intergranular stress corrosion cracking (IGSCC) affecting the containment spray system. The inspectors also determined that this finding impacted the Barrier Integrity Cornerstone human attribute and affected the cornerstone objective of ensuring the physical barriers protect the public from radionuclide releases caused by accidents. The finding was determined to be of very low safety significance because the finding did not result in an actual loss of operability or functionality of containment spray system per Table 4a, NRC Inspection Manual Chapter 0609, Attachment 4. The cause of the finding is related to the cross-cutting aspect of Human Performance, Work Practices (H.4(c)), because licensee personnel failed to ensure supervisory and management oversight activities of their contractors such that nuclear safety was ensured. |
| 05000250/FIN-2010005-04 | 2010Q4 | Turkey Point | Inadequate implementation of corrective actions fail to correct a condition adverse to quality | The inspectors identified an NCV of 10 CFR, Part 50, Appendix B, Criterion XVI, for the licensees failure to implement timely corrective actions to address conditions adverse to quality on the Unit 3 fuel handling manipulator crane. As a result, a lack of calibration on the manipulator crane load cell affected fuel handling interlock setpoints that protect the fuel during fuel handling activities. In addition, an inadequate testing procedure led to a procedure change implemented in the field without proper review and approval. The licensee entered this violation in their corrective action program as AR 592683. Although the event occurred on Unit 3, similar procedures existed on Unit 4. The inspectors determined that the licensees failure to implement timely corrective action for lack of calibration on the manipulator crane load cell affecting fuel handling interlock setpoints and other deficiencies to be a performance deficiency. The finding was greater than minor because the Barrier Integrity Cornerstone was affected which provides reasonable assurance that physical design barriers protect the public from radionuclide releases. The finding affects the attributes of configuration control and procedure quality. The inspectors evaluated the finding using Manual Chapter 0609 SDP Phase 1 and determined that it was of very low safety significance because there were no actual challenges to the fuel barrier. The finding had cross-cutting aspect in the area of problem identification and resolution (P.1(d)) because the licensee failed to implement prescribed corrective actions to address adverse trends in a timely manner when the load cell interlock setpoints drifted. |
| 05000250/FIN-2010005-05 | 2010Q4 | Turkey Point | Licensee-Identified Violation | Turkey Point Technical Specification 6.8.1.a, states that written procedures required by the Quality Assurance Topical Report (QATR) shall be implemented. The QATR commits to use the procedures in Appendix A of Regulatory Guide 1.33, which includes in Section 1.c, Equipment control (tagging). FPL implements this requirement, in part, with procedure 0-ADM-212.1, Operations In-plant Equipment Clearance Orders, which requires in Step 5.1.9, that Prior to approving an equipment clearance order, it shall be determined the impact on equipment availability to meet technical specifications. Contrary to the above, during preparation and execution of equipment clearance order 3- 10-01-001, for the Unit 3 high head safety injection system, the impact on equipment available to meet Unit 4 Technical Specifications requirements was not determined prior to approval. As a result, while implementing the clearance order, the Unit 4 high head safety injection system was rendered inoperable for a period of 36 minutes, until the manual isolation valve 3-867 was shut, as required by the clearance. The technical specification impact, diversion of Unit 4 high head safety injection to Unit 3 and entry of Unit 4 into TS 3.0.3 for 36 minutes, was determined after the clearance was implemented. When identified by the licensee during operator surveillance of control room indications, the manual valve was promptly shut in accordance with the clearance. The event was documented in the corrective action program as AR 584026 and an investigation was initiated. A regional Senior Reactor Analyst evaluated the performance deficiency under the Phase 3 protocol of the Significance Determination Process. Based upon the results of this evaluation, the performance deficiency was characterized as of very low safety significance (Green). The NRCs most current Probabilistic Risk Assessment model for Turkey Point was used to perform the evaluation. The basic event for the common cause failure of the High Head Safety Injection valves, 843A and B, was set to always occur in the model as the surrogate for the performance deficiency. The major evaluation assumptions included a one hour exposure time and no potential to re-position either of the two valves during the exposure time. The dominant accident sequence was a Small Break Loss of Coolant Accident followed by operators failing to use the High Head Safety Injection hot leg injection path, given a failure of the cold leg injection path due to the performance deficiency. |
| 05000250/FIN-2010006-01 | 2010Q2 | Turkey Point | Inadequate procedure implementation resulting in snubber failure | The NRC identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, for the licensees failure to implement procedures during a visual inspection of safety related seismically qualified snubber SN-4-1039. Specifically, the licensee failed to identify missing, detached, loosened support items, or full thread engagement of all mechanical connections that led to a snubber failure as prescribed in procedure 0-OSP-105.1, Visual Inspection, Removal and Reinstallation of Mechanical Shock Arrestors, section 7.2.1.3.d. The snubber would not have been able to perform its design function to arrest shocks of the main steam piping to the C Steam Generator during seismic events or transients, such as sudden isolation of the main steam isolation valve. The licensee implemented immediate corrective actions which included replacing the snubber in containment, adding specific instructions in procedure 0-OSP-105.1 to specifically inspect the locking ring and correct installation, and to include emphasis on FPL expectations from vendor provided snubber inspection services. The licensee documented this in condition report CR 2008-31372. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone in that the licensee did not ensure reliability of the snubber to respond to initiating events to prevent undesirable consequences in that the snubber would not have been able to perform its design function to arrest shocks of the main steam piping to the C Steam Generator during seismic events or transients. The finding was screened using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" and was determined to have a very low safety significance (Green) because the system remained operable and capable of meeting its design function with no loss of safety function of the C main steam system. This finding was reviewed for cross-cutting aspects and none were identified. (4OA2). |
| 05000250/FIN-2010006-02 | 2010Q2 | Turkey Point | Licensee-Identified Violation | Technical Specification 6.8.1 requires that procedures required by the Florida Power and Light Quality Assurance Topical Report (QATR) be implemented. The QATR includes procedures listed in Appendix A of NRC Regulatory Guide 1.33 Revision 2. Contrary to the above, on November 26, 2009, during PMT of the New Analog Rod Position Indication System (NARPI), the control room received indication that the H6 and H10 control rods dropped from the fully withdrawn position and did not enter the required off-normal procedure ONOP-28.3, Dropped RCC, when the two control rods (H6 and H10) were confirmed to be dropped during the Unit 4 Outage. The licensee eventually entered the procedure when directed by management and tripped the reactor as required by the procedure. The non-compliance was identified by the licensee following issuance of LER 2010-001-0 on January 25, 2010, and entered into the corrective action process. During the post modification test Unit 4 was in Mode 3 (Hot Standby) and was borated such that all control rods could be withdrawn and the reactor would not go critical, eliminating any safety concern with two dropped control rods. The issue was screened to be of very low safety significance (Green). The issue was documented in condition report 2010-3782 and additional corrective actions were identified. Because the licensee identified the issue and documented it in their corrective action program and because the finding is of very low safety significance, this violation is being treated as a licensee-identified NCV consistent with Section VI.A of the NRC Enforcement Policy. |
| 05000250/FIN-2010008-01 | 2010Q1 | Turkey Point | Failure to perform adequate written 50.59 evaluation. | The inspectors identified an AV of 10 CFR 50.59(d)(1) for failure to maintain records that include a written evaluation which provides the bases for the determination that a change, test, or experiment does not require a license amendment. Specifically, the licensee received NRC approval to make changes to the facility via license amendment No. 234 dated July 17, 2007, involving the design of the spent fuel pool storage racks, including the use of Metamic inserts and other hardware, administrative controls and testing methods, to assure that the spent fuel remains within design limits. Subsequent to the NRCs approval, the licensee determined that Metamic inserts could not be installed by the date approved by the NRC. However, the licensee maintained no written evaluation which provided the bases for the determination that the change to the design of the spent fuel pool storage racks, without the use of Metamic inserts, did not require a license amendment pursuant to paragraph (c)(2) of 10 CFR 50.59.The finding was more than minor because it impacted the regulatory process which depends on plant activities being properly evaluated and, when required, reviewed and approved by 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. The inspectors determined that the cross-cutting aspect of Human Performance, (H.4(c)) is applicable to this issue because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported, when errors in administering Technical Specification requirements and programmatic controls which assure safety were not effectively implemented. |
| 05000250/FIN-2010008-02 | 2010Q1 | Turkey Point | Failure to correct Boraflex degradation in the SFP in a timely manner | The inspectors identified an AV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to effectively correct a condition adverse to quality involving degradation of Boraflex neutron absorber material in the Unit 3 SFP, such that in November 2009 two spent fuel pool storage cells L38, F19 with Boraflex degradation greater than that assumed in the criticality analyses had been allowed to remain inservice even after the licensee had revised SFP management controls. When brought to the attention of the licensee by the NRC, condition report 2009-34470 was written to document the non-compliance. The finding was more than minor because, if left uncorrected, it would become a more significant safety concern since it could not be determined if other, untested storage rack locations could be more degraded. In addition, the finding impacted the initiating event cornerstone objective of limiting events that challenge safety functions; for example, preventing criticality in an area not designed for criticality. Because probabilistic risk assessment tools were not suited for this finding, the inspectors evaluated the finding using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. Because the Boraflex degradation resulted in a significant loss of margin to criticality, NRC management concluded the finding was preliminarily greater than Green. The inspectors determined that the cross-cutting aspect of Problem Identification and Resolution (P.1(d)) is applicable to this issue because the licensee did not implement effective corrective action for degradation of Boraflex neutron absorber material |
| 05000250/FIN-2010008-04 | 2010Q1 | Turkey Point | Failure to update the FSAR to reflect SFP management practices | The inspectors identified an AV of 10 CFR 50.71(e) for failure to update the Final Safety Analysis Report (FSAR) so that the report accurately reflects significant changes made to the facility. As of December 2009, changes made to manage the Unit 3 spent fuel pool since 2001, including the use of alternate means of assuring that the spent fuel remains shutdown such as use of rod control cluster assembly inserts and water holes, use of neutron attenuation testing methods and results, and use of computer programs such as RACKLIFE, were not described in the Updated FSAR. When identified to the licensee by the inspectors, the licensee documented the condition in condition report 2009-34470, and informed the NRC (in letter L-2009-295, dated December 31, 2009) of plans to make appropriate updates to the FSAR descriptions by March 15, 2010.The finding was more than minor because it impacted the regulatory process, which relies on licensees properly maintaining their FSAR up to date. 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-2010008-05 | 2010Q1 | Turkey Point | Failure to maintain Keff per 10 CFR 56.68 and TS 5.5.1.1.a | The inspectors identified an AV of Technical Specification 5.5.1.1.a and 10 CFR50.68(b)(4) for failure to assure that the effective neutron multiplication factor (Keff) would be maintained equivalent to less than 1.0, for all cases in the Unit 3 spent fuel pool(SFP) when flooded with unborated water. The finding was more than minor because, if left uncorrected, the racks would continue to degrade further reducing the neutron absorption capability and become a more significant safety concern. In addition, the finding impacted the initiating event cornerstone objective of limiting events that challenge safety functions; for example, preventing criticality in an area not designed for criticality. Because probabilistic risk assessment tools were not suited for this finding, the inspectors evaluated the finding using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. Because the Boraflex degradation resulted in a significant loss of margin to criticality, NRC management concluded the finding was preliminarily greater than Green. The inspectors determined that the cross-cutting aspect of Problem Identification and Resolution, (P.1(c)) is applicable to this issue because the licensee did not properly evaluate the problems associated with Boraflex degradation to assure operability and reportability was adequately addressed |
| 05000250/FIN-2011002-01 | 2011Q1 | Turkey Point | Failure to Monitor a Reactivity Change Results in Power Operation Above 100 Percent | The inspectors identified a non-cited violation (NCV) of Technical Specifications 6.8.1.a, Procedures, when operators did not adequately monitor reactor power nor the position of valve TC-3-144A, a valve which affects reactivity, during a letdown valve inservice test. As a result, the Unit 3 hourly average reactor power increased above 100 percent for about 40 minutes. When identified to the licensee by the inspectors, the issue was documented in the corrective action program as AR 1643603. Failure to maintain positive control of reactor power was contrary to plant procedures and was a performance deficiency. The issue was more than minor because it resulted in reactor operation at 100.05 percent power for about 40 minutes. The finding involved configuration control affecting reactivity and was assigned to the Barrier Integrity Cornerstone. In accordance with screening criteria in IMC 0609, Appendix A, Phase 1, for degraded fuel barrier, the issue screened as Green. The finding was determined to be of very low safety significance because throughout the incident, thermal power remained bounded by the reactor safety analyses limit of 102% and no safety limits were exceeded. The finding affected the cross-cutting area of Human Performance, Work Practices, (H.4(a)) when operating personnel were not aware of reactor status, and human error prevention techniques, such as holding pre-job briefings, self and peer checking, and proper documentation of activities were not adequate to assure plant activities were properly performed. |
| 05000250/FIN-2011002-02 | 2011Q1 | Turkey Point | None | 10 CFR Part 50, Appendix B, Criterion V, requires, in part, that activities affecting quality shall be prescribed by documented instructions or procedures of a type appropriate to the circumstances. Contrary to the above, on September 21, 2010, an unplanned reactor trip occurred while the quarterly surveillance for the Channel II High Pressurizer Pressure Protection Loop (P-4-456) was in progress. The licensee determined the root cause to be inadequate inspection and installation criteria used for ELCO connectors because acceptance criteria and method of verification where not addressed by procedures. The issue was screened to be of very low safety significance (Green). When identified, the licensee took corrective actions to add ELCO connector inspection requirements to a plant procedure and conduct formal training for maintenance personnel to properly inspect and mate connectors. The issue was documented in AR 00581322. Because the licensee identified the issue and documented it into their corrective action program, and because the finding is of very low safety significance, this violation is being treated as a licensee identified NCV consistent with the NRC Enforcement Policy. |