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	Quarter	Site	Title	Description
05000275/FIN-2010005-01	2010Q4	Diablo Canyon	Failure to Maintain a Fire Barrier	The inspectors identified a noncited violation of Diablo Canyon Facility Operating License Condition 2.C (5), Fire Protection, after Pacific Gas and Electric failed to maintain the integrity of Door 155 in the rated condition. On December 9, 2010, the inspectors identified that the fire door was inoperable. Equipment Control Guideline 18.7, Fire Rated Assemblies, required the licensee to maintain Door 155 in a configuration that would provide at least a 112-hour rated fire barrier. The inspectors previously identified that Door 155 was degraded as a fire barrier in 2009. The licensee entered the violation into the corrective action program as Notification 50367381 and took immediate corrective actions to restore the fire barrier to the rated condition and to implement weekly plant fire door walkdowns The inspectors concluded that the finding was more than minor because the degraded fire barrier affected the Mitigating Systems Cornerstone external factors attribute and objective to prevent undesirable consequences due to fire. The inspectors determined that the finding was within the fire confinement category and that the fire barrier was moderately degraded. The inspectors concluded that the finding was of very low safety significance (Green) because there was a non-degraded automatic full area water-based suppression system in the exposed fire area. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not take effective corrective actions to following the previous occurrence of the violation (P.1(d)).
05000275/FIN-2010005-02	2010Q4	Diablo Canyon	Inadequate Transient Combustibles Procedure	The inspectors identified a noncited violation of Diablo Canyon Unit 2 Facility Operating License Condition 2.C.(5), Fire Protection, after Pacific Gas and Electric failed to ensure procedures for controlling flammable and combustible materials adequately incorporated requirements of the fire hazard analysis. On October 18, 2010, the inspectors identified that transient combustible materials staged in the Unit 1 12 kilovolt switchgear room did not have an approved transient combustibles permit. The licensee stated that the combustibles permit procedure did not require a permit for the room while Unit 1 was shutdown. However, the plant fire hazards design basis described safe shutdown equipment in the room that would be needed to support a safe shutdown of the operating unit, specifically the Unit 2 startup bus located in the room. The inspectors determined that the licensees transient combustibles permit procedure was inadequate because the procedure did not require a permit for the Unit1 12 kilovolt switchgear room when Unit 2 was operating. The licensee entered the issue into the corrective action program as Notification 50366302 and performed an evaluation of the transient combustibles stored in the area. The inspectors concluded that this finding was more than minor because it affected the Mitigating Systems Cornerstone external factors attribute objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The inspectors determined that the finding was within the fire prevention and administrative controls category and represented a low degradation level due to the minimal impact on the effectiveness and reliability of the affected systems. The inspectors concluded that the finding was of very low safety significance (Green) based on a qualitative screening, the low degradation rating, and only equipment needed to reach and maintain cold shutdown conditions was affected. This finding had a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure that the design documentation adequately identified the Unit 2 startup bus as equipment required for safe shutdown for Unit 2 (H.2(c)).
05000275/FIN-2010005-03	2010Q4	Diablo Canyon	Corrosion of Containment Fan Cooler Unit Cooling Coil Casings	The inspectors identified an unresolved item concerning the degradation of the containment fan cooler unit cooling coil casings due to corrosion. Specifically, the issue concerns the licensees actions to verify the heat removal capability of the containment fan coolers under degraded conditions and the failure to take corrective actions for the repair or replacement of the corroded cooling coil assemblies. The containment fan cooler units function during normal plant operation to maintain the containment atmosphere at design conditions. During accident conditions, the cooler units automatically initiate to maintain containment operability. Diablo Canyon Units 1 and 2 each have five cooler units installed inside the containment building. Each cooler unit has two cooling coil banks with six coils stacked in each bank. Each of the coils is mounted on sheet metal casings and the casings are mounted within the cooler unit frame. The casings act to prevent air bypass between the coils in the banks and as structural support for the coil tubes and fins. The inspectors reviewed Diablo Canyon Power Plant Health Issue Reports 2002-S023-002 and 2002-S023-003 which identified corrosion of the containment fan cooler unit cooling coil casings. The power plant health issue reports acknowledged that continued casing corrosion would decrease the available design margin of the heat removal capacity of the cooler unit cooling coils. Containment Fan Cooler Unit Coil Study, Phase 1, Revision 0, recommended that the corroded coil casing be repaired or replaced whenever possible to avoid impacting the cooler unit heat removal capacity. The licensees initial recommended corrective action, as described in the power plant health issue reports, was to replace the cooling coil assemblies in Units 1 and 2 beginning in Refueling Outage 1R13 (Fall 2005). The replacement plans were not implemented and the licensee currently plans to begin replacement of the cooling coil assemblies during upcoming Refueling Outages 1R18 and 2R18. All cooling coil assemblies in Units 1 and 2 are scheduled to be replaced by Refueling Outage 1R20 and 2R20, respectively. The inspectors determined that additional information was needed to resolve this issue. The inspectors were unable to clearly determine the design basis function of the cooling coil casings based upon documentation provided by the licensee during the inspection. Additionally, the licensee has not quantified the effect of the corrosion to verify that the cooling coil casing functions would be maintained under the current degraded conditions and has not provided a technical justification for the acceptability of the proposed coil assembly replacement schedule. The licensee has agreed to provide this information for NRCs review. Because more information is necessary to resolve this issue, it is considered an unresolved item pending further NRCs review. The NRC will review the licensees evaluation to determine If the licensees failure to verify the heat removal capability of the containment fan cooler units is a performance deficiency If the licensees decision to delay taking corrective actions for repair or replacement of the corroded cooling coils constitutes a violation of NRC requirements The licensee entered this issue into its corrective action program as Notification 50364991. This unresolved item is identified as URI 05000275; 323/2010005-03, Corrosion of Containment Fan Cooler Unit Cooling Coil Casings.
05000275/FIN-2010005-04	2010Q4	Diablo Canyon	Inadequate Operability Determinations	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, after Pacific Gas and Electric failed to adequately evaluate two nonconforming conditions for operability as required by Procedure OM7.ID12, Operability Determination. On October 15, 2010, the inspectors identified a less than adequate technical evaluation supporting Prompt Operability Assessment 50350918, Unit 2 - Insulation in Bio-Wall Penetration. Engineering personnel failed to adequately evaluate the extent of condition after technicians identified about 632 pounds of Temp-Mat and 60 pounds of Min-K fibrous insulation in the Unit 1 reactor coolant loop biological shield wall penetrations. This fibrous material could have potentially been transported and plugged the emergency core cooling containment sump screen. The licensee performed the prompt operability assessment for Unit 2, which was operating at the time. The inspectors concluded that the engineering personnel inappropriately applied the leak-before-break methodology to exclude about 87 percent of this material from the extent of condition review in the prompt operability assessment. The second example involved Prompt Operability Assessment Notification 50355265, RHR Sump Margin, which was completed by the licensee on October 23, 2010. In this example, engineering personnel failed to identify and demonstrate that the specified safety function of the refueling water storage tank could be maintained as required by the plant operability procedure. The inspectors identified that the post accident flow path from the reactor cavity to the containment sump was blocked by a large shield plug. This blockage reduced the amount of post accident inventory available at the containment sump at the time of transition from injection to recirculation mode of emergency core cooling operation. Engineering personnel failed to demonstrate that the safety function to ensure full sump submergence was maintained with the blocked flow path. Full submergence of the sump was used by the NRC as the basis for approval of Technical Specification 3.5.4, Refueling Water Storage Tank, inventory requirements. The licensee entered the violation into the corrective action program as Notification 50369117 and revised the prompt operability assessments using assumptions consistent with the current licensing bases. The inspectors concluded that the performance deficiency was more than minor because the finding affected the Mitigating Systems Cornerstone initial design control attribute and objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded that the finding was of very low safety significance (Green) because the finding was confirmed not to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because Pacific Gas and Electric did not use conservative assumptions in decisions to demonstrate component operability in either example (H.1(b)).
05000275/FIN-2010005-05	2010Q4	Diablo Canyon	Less than Adequate Containment Recirculation Sump Design Control	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after Pacific Gas and Electric failed to ensure Calculation STA-255, Minimum Required Refueling Water Storage Tank Level for GE Sumps, Revision 2, demonstrated adequate available refueling water storage tank inventory. On October 19, 2010, the inspectors identified that emergency core cooling post accident flow path from the reactor cavity to the containment sumps was blocked by a large steel plug on Unit 1. The accident analysis assumed this 35 square foot path was open to allow coolant from a pipe break inside the biological shield to communicate with containment sumps during the recirculation mode of emergency core cooling. The licensee credited the inventory from the reactor cavity when determining the minimum required refueling water storage tank volume in Calculation STA-255. Pacific Gas and Electric used Calculation STA-255 as the basis for determining the minimum required refueling water storage tank volume specified by Technical Specification 3.5.4, Refueling Water Storage Tank. The inspectors identified that the recirculation flow path was also blocked on Unit 2. The inspectors concluded that the most significant contributor to the violation was inaccurate plant drawings used by plant engineers during the performance of Calculation STA-255. The licensees corrective actions included completion of a prompt operability assessment justifying continued operation of Unit 2 and replacement of the shield plug with a movable platform on Unit 1 prior to plant restart The inspectors concluded that the performance deficiency was more than minor because the finding affected the Mitigating Systems Cornerstone plant modification design control attribute and objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded that the finding was of very low safety significance (Green) because the performance deficiency involved a design deficiency confirmed not to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of human performance associated with the resources component because Pacific Gas and Electric failed to use complete, accurate and up-to-date drawing for Calculation STA 255 (H.2(c)).
05000275/FIN-2010005-06	2010Q4	Diablo Canyon	Inadequate Emergency Diesel Generator Surveillance Testing	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, after Pacific Gas and Electric failed to develop and implement an adequate testing program for the emergency diesel generators that met design requirements and recommendations. Specifically, in December 2008, the inspectors identified that the diesel generator loading calculations were inadequate to demonstrate that the design bases were met. Pacific Gas and Electric updated the load calculations, but failed to make the necessary revisions to Surveillance Test Procedure STP M-9D1, Diesel Generator Full Load Rejection Test. As a result, Pacific Gas and Electric failed to test several of the emergency diesel generators at the complete load as required by Regulatory Guide 1.108, Revision 1, which is part of the current licensing bases. The licensee entered this into the corrective action program as Notification 50368801, determined there was no loss of safety function for the affected components, and applied the provisions of Surveillance Requirement 3.0.3 for a missed surveillance test. The inspectors concluded the most significant contributor to the finding was less than adequate diesel generator loading evaluations to support corrective action from previous violations associated with the emergency diesel generator testing. The inspectors concluded that the performance deficiency was more than minor because the finding affected the equipment control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green) because it did not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time. This finding had a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to perform an adequate evaluation of the nonconservative surveillance test such that the resolution addressed the fundamental basis for the surveillance (P.1(c)).
05000275/FIN-2010005-07	2010Q4	Diablo Canyon	Inadequate Quality Verification Audits	The inspectors identified a noncited violation of 10CFR Appendix B, Criterion XVIII, Audits, which required that a comprehensive system of planned and periodic audits be carried out to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program as well as follow up action, including re-audit of deficient areas, where indicated. Contrary to this requirement, Pacific Gas and Electric failed to ensure that a comprehensive system of planned and periodic audits were carried out to verify compliance with all aspects of the quality assurance program, determine the effectiveness of the program, and perform necessary follow up actions. Specifically, the 2008 Quality Verification audit of the corrective action program failed to adequately address an adverse trend in the problem evaluation process documented in NRC Inspection report 2008005, which identified eleven examples of an adverse trend in problem evaluation. The licensee entered this into their corrective action program as Notification 50365083 and determined there was no loss of safety function for the affected components. The inspectors concluded the most significant contributor to the finding was a less than adequate evaluation of the corrective action trending program. This finding was more than minor because it was associated with the equipment control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the performance deficiency was of very low safety significance (Green) it was a deficiency confirmed not to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program component, because the licensee failed to coordinate and communicate the results from assessments to affected personnel, and track the corrective actions to address issues commensurate with their significance (P.3(c)).
05000275/FIN-2010005-08	2010Q4	Diablo Canyon	Licensee-Identified Violation	Technical Specification 5.4.1.a required that the licensee establish, implement, and maintain written procedures covering the applicable activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33 recommended procedures required for shutdown of pressurized water reactors. Contrary to this, Pacific Gas and Electric failed to properly implement procedures for shutdown, as required by Technical Specification 5.4.1.a. Specifically, on October 5, 2010, operators failed to follow Operating Procedure OP L-6, Cold Shutdown/Refueling, Step 5.9.6 and establish a reactor coolant system vent path before reducing reactor coolant system temperature below 90F. As a result, operators declared the Low Temperature Overpressure Protection System inoperable in accordance with Technical Specification 3.4.12 and restored temperature above 90F. Pacific Gas and Electric entered the issue into the corrective action program as Notification 50347713. Using Appendix G of Inspection Manual Chapter 0609, Shutdown Operations Significance Determination Process, the inspectors concluded this finding was of very low safety significance because the licensee maintained an adequate mitigation capability during shutdown and the issue did not require a quantitative assessment.
05000275/FIN-2010005-09	2010Q4	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR 50.59 c(2)(vi) required that a licensee obtain a license amendment pursuant to Sec. 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would create a possibility for a malfunction of a structures, systems, and components important to safety with a different result from any previously evaluated in the FSARU. Contrary to this, Pacific Gas and Electric failed to obtain a license amendment prior to implementing revisions to plant operating procedures. Specifically, on June 29, 2010, Pacific Gas and Electric identified that revisions to Operating Procedure OP L-4, Normal Operation at Power, and OP L-5, Plant Cooldown from Minimum Load to Cold Shutdown allowed operators to control pressurizer level above the program control band during brief periods. The licensee concluded that operation above the program control band could result in a pressurizer overfill condition that could challenge the performance of the pressurizer power-operated relief valves if a spurious safety injection occurred. The licensee entered the issue into the corrective action program as Notification 50320032. The inspectors concluded this finding was of very low safety significance because it is a design deficiency confirmed not to result in the loss of operability or functionality.
05000275/FIN-2010005-10	2010Q4	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Section III, Design Control, required, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to this, Pacific Gas and Electric failed to establish measures to ensure that applicable regulatory requirements and the design bases were correctly translated into specifications, drawings, procedures, and instructions. Specifically, on May 14, 2010, Pacific Gas and Electric discovered a design deficiency during the final preparation of the Diablo Canyon Power Plant Unit 1 design change package for the safety injection test line optimization modification. The package lacked the required flow restrictors for the reactor coolant pressure boundary isolation valves, resulting in the potential to create loss of reactor coolant system inventory in Mode 4 which would exceed normal charging makeup capability. On June 14, 2010, the licensee determined that this design deficiency impacted the Unit 2 safety injection modification that had been implemented during the most recent refueling outage. Pacific Gas and Electric entered this issue into the corrective action program as Notification 50316384 and initiated corrective actions to install travel limiting stops on all of the new safety injection test line manual valves to limit the opening size prior to valve testing. The inspectors concluded this finding was of very low safety significance because it is a design deficiency confirmed not to result in the loss of operability or functionality.
05000275/FIN-2010005-11	2010Q4	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Section XI, Test Control, required, in part, that the test program demonstrate that systems and components will perform satisfactorily in service and is performed in accordance with written test procedures that incorporate the requirement and acceptance limits contained in applicable design documents. Contrary to this, Pacific Gas and Electric failed to ensure that testing demonstrated that systems and components would perform satisfactorily in service and perform in accordance with written test procedures which incorporated the requirements and acceptance limits into applicable design documents. Specifically, on November 14, 2010, the licensee performed comprehensive pump testing of Unit 1 turbine-driven auxiliary feedwater pump following replacement of the turbine governor during the most recent refueling outage. The licensee adjusted governor speed settings to meet the pump speed criteria of Procedure STP P-AFW-A11, Comprehensive Testing of Turbine-Driven Auxiliary Feedwater Pump 1-2, Revision 6. Testing personnel identified that the turbine speed settings differed from the previous governor. Following an engineering evaluation, operators performed a surveillance test of the pump and discovered that turbine speed exceeded the test acceptance criteria. The licensee concluded that the new governor had different performance characteristics from the previous governor. Pacific Gas and Electric failed to revise the test procedures to reflect the new governor performance characteristics. As a result, the licensee inappropriately adjusted the governor settings. Pacific Gas and Electric entered this issue into the corrective action program as Notification 50316384 and initiated corrective actions to re-perform comprehensive testing to the pump and adjust the governor to meet pump acceptance criteria. The inspectors concluded this finding was of very low safety significance because it did not result in an actual loss of safety function of the pump.
05000275/FIN-2010005-12	2010Q4	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Section XVI, Corrective Action, required measures be established to ensure that conditions adverse to quality are promptly identified and corrected. Contrary to this, Pacific Gas and Electric failed to ensure corrective actions to identify and correct fibrous material inside containment, as described in licensee Letter DCL-08-059, Supplemental Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors (July 10, 2008). Subsequently, on October 14, 2010, the licensee identified approximately 632 pounds of Temp-Mat and 60 pounds of Min-K fibrous material installed in the Unit 1 reactor coolant loop biological shield wall penetrations. Pacific Gas and Electric entered this issue into the corrective action program as Notification 50355265 and initiated corrective actions to remove the fibrous insulation material. The inspectors concluded that this finding was of very low safety significance because it was a design deficiency confirmed not to result in the loss of operability or functionality.
05000275/FIN-2011003-01	2011Q2	Diablo Canyon	Inadequate Fire Hazard Evaluations	The inspectors identified a noncited violation of Diablo Canyon Facility Operating License Condition 2.C (5), Fire Protection, after Pacific Gas and Electric failed to implement the required compensatory actions described in Equipment Control Guideline 18.7, Fire Rated Assemblies. On December 28, 2010, the licensee blocked open Fire Doors 175 and 182-2, entrances to the Unit 1 and 2 safety injection pump room to address auxiliary building ventilation flow balance problems. The supporting engineering evaluation failed to identify that the doors were rated fire barriers as described in the fire hazard analysis. If a fire had occurred, these blocked open doors would have allowed smoke and hot gases to pass from fire area AB-1 to impact equipment in adjacent fire areas 3-B-2 (Unit 1) and 3-D-2 (Unit 2). Equipment Control Guideline 18.7 required the licensee to either establish a continuous fire watch on at least one side of the inoperable fire doors or verify that the fire detection or automatic suppression system on at least one side of the fire doors was operable and establish an hourly fire watch. The licensee took corrective actions to establish the required fire watches and enter the finding into the corrective action program as Notification 50409975. The inspectors concluded that the failure of Pacific Gas and Electric to maintain the fire doors in the rated configuration as described in the Final Safety Analysis Report Update Fire Hazard Analysis, was a performance deficiency. This finding was more than minor because the degraded fire barriers affected the Mitigating Systems Cornerstone external factors attribute objective to prevent undesirable consequences due to fire. The inspectors concluded that the finding was of very low safety significance (Green) because the finding only affected the ability to reach and maintain cold shutdown conditions. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems associated with modification of the safety injection pump room fire doors such that the resolutions addressed causes and extent of conditions, as necessary.
05000275/FIN-2011003-02	2011Q2	Diablo Canyon	Unplanned Loss of Preferred Offsite Power Due to Less than Adequate Work Planning	The inspectors identified a self-revealing finding following the unplanned loss of 230 kV preferred offsite power to Unit 1 due to inadequate work planning. On May 17, 2011, Unit 1 lost preferred offsite power after a technician began cutting a hole in a startup bus control panel using a reciprocating saw. The reciprocating saw induced vibration on the control panel and caused the phase differential protection relay to actuate which separated the startup bus from preferred offsite power. All three Unit 1 emergency diesel generators automatically started after offsite power was lost to the plant vital loads. Procedure AD7.DC8, Work Control, stated that when performing nonroutine work, including modifications on electrical or instrument equipment, the equipment shall be isolated to prevent any unintended equipment actuations. The licensee had authorized the cutting work while the Unit 1 startup bus was in service. The licensee took corrective action to restore offsite power and entered the finding into the corrective action program as Notification 50402706. The inspectors determined that the failure to adequately evaluate the effect of the cutting activity on the energized plant equipment was a performance deficiency. This performance deficiency was more than minor because the finding was associated with the Mitigating Systems Cornerstone human performance attribute and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The senior reactor analyst utilized Table 3.7 from the plant specific risk-informed notebook and determined that the risk based on Phase 2 estimation was Green. Additionally, the analyst performed a bounding analysis that corroborated the Phase 2 result based on three complete losses of preferred power during the refueling outage with a total exposure time of 2.9 hours. Using the standardized plant analysis risk model for Diablo Canyon Units 1 and 2, the analyst quantified the conditional core damage probability for any initiator resulting in a consequential loss of offsite power as 1.2 x 10-4. Given these conditions, the analyst noted that the change in core damage frequency could be approximated as the product of these two values (3.9 x 10-8). This indicated that the subject finding was of very low risk significance (Green). This finding has a crosscutting aspect in the area of human performance associated with the work control component, in that Pacific Gas and Electric failed to appropriately plan work activities by incorporating risk insights, job site conditions, and plant structures, systems, and components.
05000275/FIN-2011003-03	2011Q2	Diablo Canyon	Unplanned Loss of Preferred Offsite Power Due to the Failure to Follow Work Instructions	The inspectors identified a self-revealing finding following two unplanned losses of 230 kV preferred offsite power to Unit 1 due to personnel errors. On May 26, 2011, Unit 1 lost preferred offsite power after a technician incorrectly installed test equipment on the Unit 2 startup bus control circuit during a post-modification test. The Unit 1 phase differential protection relay actuated and separated the startup bus from preferred offsite power after the technician energized the test circuit. On May 27, 2011, Unit 1 again lost preferred offsite power after a technician incorrectly installed test equipment on a Unit 1 wiring termination when the post-modification test specified that the test equipment was to be installed on Unit 2. The Unit 1 phase differential protection relay actuated and separated the startup bus from preferred offsite power. In each event, all three emergency diesel generators automatically started after offsite power was lost to the plant vital loads. The licensee took corrective action to reestablish offsite power and entered the finding into the corrective action program as Notifications 50405004 and 50405010. The inspectors concluded that the failure to follow post-modification testing work instructions was a performance deficiency. This performance deficiency was more than minor because the finding was associated with the Mitigating Systems Cornerstone human performance attribute and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The senior reactor analyst utilized Table 3.7 from the plant specific risk-informed notebook and determined that the risk based on Phase 2 estimation was Green. Additionally, the analyst performed a bounding analysis that corroborated the Phase 2 result based on three complete losses of preferred power during the refueling outage with a total exposure time of 2.9 hours. Using the standardized plant analysis risk model for Diablo Canyon Units 1 and 2, the analyst quantified the conditional core damage probability for any initiator resulting in a consequential loss of offsite power as 1.2 x 10-4. Given these conditions, the analyst noted that the change in core damage frequency could be approximated as the product of these two values (3.9 x 10-8). This indicated that the subject finding was of very low risk significance (Green). This finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively communicate human error prevention techniques; and consequently, these techniques were not used commensurate with the risk of the assigned task.
05000275/FIN-2011003-04	2011Q2	Diablo Canyon	Failure to Follow Procedures for Testing HEPA Ventilation Units	The inspectors identified a noncited violation of Technical Specification 5.4.1(a) for the failure to follow procedures for testing and using the high-efficiency particulate air ventilation units used to prevent personal contamination. Licensee immediate actions included removing all high-efficiency particulate air ventilation units installed for the Unit 2 outage and testing all high-efficiency particulate air ventilation units as required by procedure. This matter was placed in the licensees corrective action program as Notifications 50399479, 50399560, and 50399682. This failure to follow procedures was a performance deficiency. The finding was more than minor because it was associated with the program and process attribute of the occupational radiation safety cornerstone. The finding affected the objective to ensure adequate protection of the workers health and safety from exposure to unintended radiation from radioactive material during routine civilian nuclear reactor operation. Using the Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding was of very low safety significance because (1) it was not associated with as low as is reasonably achievable (ALARA) planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This finding was determined to have a crosscutting aspect in the area of human performance, associated with work practices, because the licensee did not effectively communicate expectations regarding procedural compliance and the personnel following the procedures.
05000275/FIN-2011003-05	2011Q2	Diablo Canyon	Inadequate Review of Severe Accident Management Guidelines	The inspectors identified a finding after Pacific Gas and Electric failed to periodically review and update the severe accident management guidelines. Procedure OM10.ID5, Severe Accident Management, required the licensee to review and update the severe accident management guidelines biennially to ensure that any changes in plant design or procedures, experience in severe accident management requalification training, and any changes in industry understanding of severe accidents were incorporated into the severe accident management guidelines. As a result of the licensees failure to implement the periodic review, the severe accident management guidelines did not incorporate the latest owners group guidance or recent plant design and hardware changes. The licensee took corrective actions to implement the biennial reviews and entered this finding into the corrective action program as Notification 50399554. Pacific Gas and Electrics failure to follow procedural requirements for periodic review of the severe accident management guidelines was a performance deficiency. The finding was more than minor because if left uncorrected, the failure to review and update the severe accident management guidelines has the potential to lead to a more significant safety concern. This finding affected the barrier integrity cornerstone because the severe accident management guidelines are procedures that would be used to maintain the functionality of the containment should a severe accident occur. The inspectors concluded that the finding was of very low safety significance because it did not represent a degradation of the radiological, smoke, or toxic atmosphere barrier function; or represent an actual open pathway in the physical integrity of the reactor containment; or involve the function of the containment hydrogen igniters. The finding did not have any crosscutting aspects because the performance deficiency occurred more than three years ago and is not indicative of current licensee performance in that the licensee has improved the design review process since the performance deficiency occurred.
05000275/FIN-2011003-06	2011Q2	Diablo Canyon	Less than Adequate Evaluation of New Security Modifications	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, after Pacific Gas and Electric failed to adequately evaluate the impact of protected area boundary modifications. These modifications affected the ability of plant operators to transfer water from the raw water storage reservoirs to the auxiliary feedwater system using temporary hoses. Plant engineers authorized a series of security modifications which included the installation of physical intrusion barriers, including delay fences and razor wire between the raw water reservoirs and the auxiliary feedwater system. The licensing basis evaluation did not address raw water makeup to the auxiliary feedwater system using temporary hoses as described in Final Safety Analysis Report Update Section 6.5, Auxiliary Feedwater System, and Section 3.7.6, Seismic Evaluation to Demonstrate Compliance with the Hosgri Earthquake Requirements Utilizing a Dedicated Shutdown Flowpath. The licensee took immediate corrective actions to establish a route for the temporary hoses, including preplanned security compensatory measures, and entered this finding into the corrective action program as Notification 50410997. The failure to adequately evaluate the impact of the security modifications on the plant licensing and design bases was a performance deficiency. This performance deficiency was more than minor because the finding affected the Mitigating Systems Cornerstone design control attribute and objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded that the finding was of very low safety significance (Green) because the finding was confirmed not to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of Problem Identification and Resolution, associated with the Corrective Action Program component, because the licensee failed to thoroughly evaluate the security modifications such that the resolutions addressed causes and extent of conditions, as necessary.
05000275/FIN-2013002-01	2013Q1	Diablo Canyon	Failure to Provide Adequate Guidance to Address General Welding Standard Requirements	On February 14, 2013, the inspectors observed field welders add a partial circumferential weld on one side of the pipe in efforts to repair the pipe misalignment prior to the completion of the final visual inspection. This action represents a violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, because the licensees procedure established special controls for critical distortions but failed to adequately define what situations fit that category. The licensee reviewed the stress calculation for the piping in question and concluded that the addition of the weld filler material did not affect the fatigue resistance of the weld, but acknowledged that a definition and additional guidance for the term critical was missing in the procedure and could have adverse effects on future final welds. The licensee entered the finding into their corrective action program as Notification 50542347. The inspectors determined that the failure of the sites welding standard to provide adequate guidance to identify what constitutes a weld distortion during welding activities was a performance deficiency. The finding was more than minor because if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, Procedure GSW-ASME did not provide the necessary guidance for welders and quality assurance personnel to identify and understand what constitutes critical distortion of a weld. The welding process can introduce effects of weld shrinkage (stresses) and distortion that could adversely affect the final condition of the weld, potentially leading to a service induced failure. Using Manual Chapter 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. The inspectors determined the finding had a cross-cutting aspect in the human performance area associated with work practices and procedural compliance, because the licensee did not adequately define or train welders to know what constituted a critical distortion, and did not effectively communicate the expectation of questioning the procedure if the welding activity required skill of the craft
05000275/FIN-2013002-02	2013Q1	Diablo Canyon	Failure to Identify Existing Indications During Prior Ultrasonic Examinations of Pressurizer Structural Weld Overlays	The inspectors identified a Green non-cited violation of 10 CFR 50.55a(a)(3)(i), which requires that proposed alternatives to industry codes and standards provide an acceptable level of quality and safety. The NRC staff approved relief request REP-1 U2 dated March 28, 2007, for installing six structural weld overlays on the pressurizer safety, relief, spray and surge nozzles. The request established acceptance criteria of laminar flaws during weld acceptance examinations limited to only the third 10-year inservice inspection interval. Contrary to the above, the licensee failed to identify unacceptable flaws as defined by the approved request following completion of these welds in 2008. The licensee entered the finding into their corrective action program as Notification 50540188. The inspectors determined that the licensees failure to identify indications that exceeded the acceptable linear dimension of laminar flaws prior to placing the system in service is a performance deficiency. The performance deficiency was more than minor because it is associated with the Initiating Events Cornerstone attribute of equipment performance, and adversely affects the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, during the months of February and March 2013, the licensee identified that three out of the six pressurizer structural weld overlays exhibited laminar flaws that exceeded the linear dimensions approved by the safety evaluation. Using Manual Chapter 0609, Attachment A, The Significance Determination Process (SDP) for Findings At Power, the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. This issue did not have a cross-cutting aspect associated with it because it is not indicative of current performance
05000275/FIN-2013002-03	2013Q1	Diablo Canyon	Failure to Effectively Evaluate Design Change for High Voltage Bushing	The inspectors reviewed a Green self-revealing finding for failure to effectively and accurately evaluate all available resources to procure appropriate equipment for plant modifications. Specifically, design engineering staff was not effective in using applicable station design documents, in conjunction with industry standards to determine minimum creepage distance for high voltage insulators when replacing ceramic bushings with polymer bushings on the main bank transformer. As a result, the licensee approved installation of an insulator stack that did not provide adequate ground protection, which caused a plant trip on October 11, 2012. The licensee entered the condition in their corrective action program as Notification 50518473. Failure to effectively and accurately evaluate all available resources to procure appropriate equipment for plant modifications was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenged critical safety functions during power operations, and is therefore a finding. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, Exhibit 1, Initiating Events Screening Questions, this finding was determined to be of very low safety significance (Green) because, although it resulted in a reactor trip, it did not result in the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding had a cross-cutting aspect in the area of human performance, associated with the decision making component, because the licensee did not use conservative assumptions in decision making when considering the suitability of the design for the environment.
05000275/FIN-2013002-04	2013Q1	Diablo Canyon	Licensee-Identified Violation	A violation of 10 CFR 50.55a(g)(4) was identified involving the failure to perform a system pressure test of the reactor vessel flange leak-off line of Units 1 and 2 in accordance with the applicable edition of Section XI of the ASME Code. The identified violation was entered into the corrective action program as Notifications 50524370 and 50524575. The violation was more than minor because it is associated with the Barrier Integrity Cornerstone attribute of systems, structures, components and barrier performance, and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, the violation was determined to be of very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function
05000275/FIN-2013002-05	2013Q1	Diablo Canyon	Licensee-Identified Violation	The licensee identified a violation of Technical Specification 3.8.1 because EDG 2-3 was inoperable for greater than 7 days. On August 18, 2012, an operator discovered that the fuel oil booster pump belt on EDG 2-3 was broken. The licensee subsequently determined that during the engine shutdown on August 3, 2012, the fuel oil booster pump had seized, which then caused the belt to snap. On August 20, 2012, the licensee completed replacement of the pump and drive belt. This violation has no associated performance deficiency because the licensee had set the drive belt tension in accordance with the manufacturers recommendation, and there was no internal or industry operating experience that indicated the drive belt tension level was inappropriate. In accordance with IMC 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this violation required a detailed risk evaluation because it represented an actual loss of diesel generator function for greater than the Technical Specification allowed outage time. Using the Diablo Canyon Units 1 and 2 Standardized Plant Analysis Risk model, Version 8.20, modified to account for offsite power recovery and the licensees procedures for intertrain crosstie, the senior reactor analyst determined that the incremental conditional core damage probability from internal initiators was 1.9 x 10-7. As best available information, the analyst utilized the results from the licensees fire and seismic models as the external initiators contributor. The final change in core damage frequency was calculated to be 6.5 x 10-7. Therefore, this violation was of very low safety significance (Green). The licensee entered the issue into the corrective action program as Notification 50507816. Corrective actions include lowering the drive belt tension specification to minimize side deflection force and modifying the procedure for placing a diesel generator to standby to specifically include a visual check of fuel oil booster pump drive belt integrity.
05000275/FIN-2013004-01	2013Q3	Diablo Canyon	The failure to use procedures to perform corrective maintenance on an emergency diesel generator	The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, after the licensee performed corrective maintenance on a diesel fuel oil system leak without appropriate documentation or procedures. This resulted in the fuel oil header not being properly primed or vented, which rendered an emergency diesel generator inoperable. The licensee entered the condition into the corrective action program as Notification 50561918. The failure to use procedures to perform corrective maintenance on an emergency diesel generator was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency, was not a loss of the system or function, and did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time. The finding had a cross-cutting aspect in the area of human performance, associated with the work practices component, because licensee staff did not communicate human error prevention techniques, such as proper documentation of activities, and did not use this technique commensurate with the risk of the assigned task, such that work activities are performed safely. Specifically, the system engineer recognized the possibility of introducing air into the system, but assumed that operators would have filled and vented the system using the appropriate procedure, while operators did not use a procedure to tighten the leaking fitting and refill the priming tank.
05000275/FIN-2013004-02	2013Q3	Diablo Canyon	Valid EDG 2-1 start Signal Caused by a Loss of 4 kV Class 1E Bus G	The inspectors reviewed a self-revealing non-cited violation 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with troubleshooting of the Unit 2, 4kV bus G that resulted in an unplanned de-energization. This caused an unplanned entry into a 72-hour shutdown technical specification action statement due to diesel fuel oil transfer pump 0-2 becoming unavailable. The licensee entered the condition into the corrective action program as Notification 50544198. The failure to plan and coordinate emergent maintenance such that it would not impact other mitigating systems was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. This finding was evaluated for each unit separately. For Unit 1, which was at power, using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this finding was determined to be of very low safety significance (Green) because, it was not a design or qualification deficiency, was not a loss of the system or function, and did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time. For Unit 2 this finding did not require evaluation using Inspection Manual Chapter 0609, and Appendix G because the unit was defueled. The finding had a cross-cutting aspect in the area of human performance, work practices component, because workers failed to use multiple human error prevention techniques.
05000275/FIN-2014005-02	2014Q4	Diablo Canyon	Failure to Control Access to a High Radiation Area with Dose Rates Greater Than 1 Rem/Hour	The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.7.2 because the licensee failed to control access to a high radiation area with dose rates greater than 1 rem/hour. A radiation protection technician assumed responsibility for guarding the area and reestablished compliance with technical specification requirements. Licensee representatives documented the occurrence in the corrective action program as Notification 50590243 and performed an apparent cause evaluation. The failure to control access to a high radiation area with dose rates greater than 1 rem/hour is a performance deficiency. The requirement not met was Technical Specification 5.7.2. The significance of the performance deficiency was more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern if workers had entered an uncontrolled, high radiation area and received unintended radiation dose. The Occupational Radiation Safety Cornerstone was affected; therefore, the inspectors used Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, August 19, 2008, to determine the significance of the violation. The violation had very low safety significance because: (1) It was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with avoiding complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk and did not implement appropriate error reduction tools (H.12).
05000275/FIN-2014005-03	2014Q4	Diablo Canyon	Failure to Effectively Implement Risk Management Actions Associated with Safety-Related Emergency Diesel Generators	The inspectors identified a self-revealing non-cited violation of 10 CFR 50.65(a)(4) for failing to manage risk when a protected train emergency diesel generator was unexpectedly rendered inoperable while another train was being returned to service. Specifically, the installed and administrative operational barriers failed to prevent a loss of safety function to an operable emergency diesel generator resulting in two inoperable emergency diesel generators for a period of two hours. The licensee took immediate actions to adequately implement the physical and administrative operational barriers, repair the damage to the protected emergency diesel generator, and entered the condition into the corrective action program as Notification 50600810. The inspectors determined that the licensees failure to adequately implement risk management actions associated with maintenance on emergency diesel generator EDG 1-2 was a performance deficiency. The performance deficiency was more than minor and therefore a finding because it was associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the performance deficiency involved the licensees assessment and management of risk associated with performing maintenance in accordance with 10 CFR 50.65(a)(4). The inspectors reviewed the results of Calculation RA 13-11, Evaluation for Unit 1, EDG 1-3 Inoperable while EDG 1-2 is in Maintenance, Revision 0, for impact to incremental core damage probability. Th inspectors used Inspection Manual Chapter 0609, Attachment 04, Initial Characterization o Findings, and Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined because the incremental core damage probability deficit (ICDPD) was not greater than 1E -06/year, the finding was determined to be of very low safety significance (Green). The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the work practices component, in that personnel work practices are used commensurate with the risk of the assigned task, such that work activities are performed safely. Specifically, the operator did not consider potential undesired consequences, such as damage to the fuel line, and perform adequate self or peer checks prior to performance of an inspection of protected equipment to ensure risk managemen action would provide appropriate protection (H.11).
05000275/FIN-2014005-04	2014Q4	Diablo Canyon	Licensee-Identified Violation	Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. The reactor coolant system is a safety-related system, and therefore requires measures to assure the design basis is correctly translated into specifications, drawings, procedures, and instructions. Contrary to this requirement, on March 28, 2008, the reactor coolant system design basis was not correctly translated into specifications, drawings, procedures, and instructions because it did not combine the loads from a loss-of-coolant-accident (LOCA) and an earthquake. This finding was identified by the licensee and entered into the corrective action as Notifications 50403188 and 50404966. The finding was determined to be of very low safety significance because the reactor coolant system maintained its operability.
05000275/FIN-2017002-01	2017Q2	Diablo Canyon	Inadequate Expansion Scope of Risk - Informed Welds	Green . The inspectors identified a non -cited violation of the licensees risk -informed inservice inspection program (which is their alternative to portions of the ASME Code, Section XI inservice inspection program approved in accordance with 10 CFR 50.55a(z)) for the failure to properly expand the scope of additional welds to inspect. Specifically, a rejectable flaw on a pipe weld in the pressurizer spray line was identified during refueling outage 1R19 while performing an ultrasonic examination. The licensee expanded the inspection scope by four additional welds, but failed to select those assigned with the same degradation. For immediate corrective actions, the licensee identified and intended to inspect four additional welds assigned to the same degradation mechanism as required by the risk -informed inservice inspection program. This issue was entered into the licensees corrective action program as Notification 50920222. The licensees failure to properly expand the weld examination scope as required by the risk -informed inservice inspection program was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to select additional welds that were susceptible to the same degradation mechanism as weld WIB -378 placed the plant at an increased risk due to the potential of having an active degradation mechanism that could affect additional components. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP ) for Findings At-Power, dated June 19, 2012, the inspector s determined the finding screened as having very low significance (Green) because: (1) it was not a design deficiency; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and (4) did not result in the loss of a high safety -significant non -technical specification train. This finding had a cross -cutting aspect in the area of human performance associated with 3 change management because leaders failed to use a systematic process for evaluating and implementing the change to a risk -informed inservice inspection program. The implementing procedure failed to include the reference to degradation mechanism allowing for a misinterpretation of weld expansion requirements once a flaw was identified in a weld WIB -378 (H.3).
05000275/FIN-2017002-02	2017Q2	Diablo Canyon	Failure to Conduct Required Biennial Medical Examinations Within Two Years	SL -IV. The inspectors identified a Severity Level IV, non -cited violation of 10 CFR 55.21, Medical Examination, for the licensees failure to ensure that a medical examination by a physician to determine satisfaction of 10 CFR 55.33(a)(1) requirements was conducted every 2 years for two licensed senior operators. Specifically, one licensed senior operator exceeded the two- year medical examination requirement by approximately 16 months between November 27, 2015, and April 6, 2017. A second licensed senior operator exceeded the 2 -year medical examination requirement by 4 months between November 19, 2016, and April 6, 2017. As a corrective action, the licensee has conducted the required medical examination for one senior operator and initiated a license termination request for the other senior operator. This issue was entered into the licensees corrective action program as Notification 50912407. The failure of the facility licensee to conduct required biennial medical examinations for two licensed senior operators was a performance deficiency. This issue was evaluated using the traditional enforcement process because it negatively impacted the NRCs ability to perform its regulatory oversight function. Specifically, the failure to comply with medical testing requirements for two operators compromised the facility licensees ability to assure conformance to medical standards, detect non -conforming medical conditions, and report non-conformances to the NRC. This performance deficiency was determined to be Severity Level IV because it fits the Severity Level IV example of Enforcement Policy Section 6.4.d.1, Violation Examples: Licensed Reactor Operators. This section states, Severity Level IV violations involve, for example ... (b) an individual operator who did not meet the American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, Section 5, Health Requirements and Disqualifying Condit ions, as certified on NRC Form 396, Certification of Medical Examination by Facility Licensee, required by 10 CFR 55.23, Certification, but who did not perform the functions of a licensed operator or senior operator while having a disqualifying medical condition. No cross -cutting aspect was assigned because the violation was processed using traditional enforcement.
05000275/FIN-2017002-03	2017Q2	Diablo Canyon	Failure to Report a Permanent Medical Condition Within 30 Days	SL -IV. The inspectors identified a Severity Level IV, non -cited violation of 10 CFR 55.25, Incapacitation Because of Disability or Illness, for the licensees failure to notify the NRC within 30 days of a change to one licensed senior operators medical condition. Specifically, the licensed senior operator developed a permanent medical condition which caused him to permanently leave the site on December 1, 2014, and transition into a long- term disability program on April 23, 2015. The licensee did not notify the NRC of this change in medical condition. As a corrective action, the licensee initiated a license termination request for the affected operator, effective April 6, 2017. This issue was entered into the licensees corrective action program as Notification 50912407. The failure of the facility licensee to notify the NRC within 30 days of a change in a licensed senior operators medical condition was a performance deficiency. This issue was evaluated using the traditional enforcement process because it negatively impacted the NRCs ability to perform its regulatory oversight function. Specifically, the failure to report 4 changes in a licensed senior operators medical condition prevented the NRC from taking action to issue either a license amendment or termination, as appropriate. This performance deficiency was determined to be Severity Level IV because it fits the Severity Level IV example of Enforcement Policy Section 6.4.d.1, Violation Examples: Licensed Reactor Operators. This section states, Severity Level IV violations involve, for example (b) an individual operator who did not meet the American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, Section 5, Health Requirements and Disqualifying Conditions, as certified on NRC Form 396, Certification of Medical Examination by Facility Licensee, required by 10 CFR 55.23, Certification, but who did not perform the functions of a licensed operator or senior operator while having a disqualifying medical condition. No cross -cutting aspect was assigned because the violation was processed using traditional enforcement
05000275/FIN-2017002-04	2017Q2	Diablo Canyon	Failure to Follow Procedures Results in Partial Loss of Cooling Flow to Shutdown Cooling	Green . The inspectors reviewed a self -revealing, non- cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because PG&E personnel failed to follow the requirements of AD7.ID14, Assessment of Integrated Risk, Revision 11. Specifically, PG&E personnel failed to obtain shift manager permission, conduct a protected equipment briefing, and document shift manager approval prior to performing work on protected equipment. This resulted in a loss of flow of cooling water to one of two in- service shutdown cooling residual heat removal heat exchangers and subsequent perturbation in reactor coolant system temperature during refueling outage 1R20. The inspectors determined that PG&E s failure to follow AD7.ID14, Assessment of Integrated Risk, Section 5.14 Performing Work on Posted Protected Equipment, was a performance deficiency within PG&Es ability to foresee and correct. This performance deficiency was considered to be more than minor because it impacted the configuration control attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the loss of cooling flow to the RHR heat exchanger while in shutdown cooling mode resulted in a perturbation in RCS temperature of approximately 8 degrees Fahrenheit. The finding was evaluated in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined to be of very low safety significance (Green) since it did not represent a loss of system safety function of at least a single train for greater than four hours. The finding had a cross- cutting aspect in the area of human performance associated with conservative bias because PG&E personnel did not use decision- making practices that emphasize prudent choices over those that are simply allowable. Specifically, despite being authorized to close component cooling water cross connect valves by the work control process, PG&E personnel did not question the impact of their actions on shutdown cooling (H.14 ).
05000285/FIN-2009007-02	2009Q2	Fort Calhoun	Failure to Perform Vendor and Industry Recommended Testing on Safety-Related and Risk Significant 4160 and 480 V Circuit Breakers	The team identified an unresolved item associated with inadequate maintenance procedures for 4160 and 480 V safety-related breakers. The team determined that maintenance procedures used to ensure that 4160 and 480 V safetyrelated breakers were being maintained and overhauled in a timely manner were inadequate. The licensee had no engineering analysis or technical basis to justify the deviation from vendor/Electric Power Research Institute guidance. At the end of the inspection, the licensee identified approximately 20 breakers that had failed over the last15 years and the team was waiting for additional information to determine if the failures were related to the inadequate maintenance. The team identified that the licensee was not performing the maintenance on the breakers as recommended by the vendor or Electric Power Research Institute guidelines. The licensee had completed a review of its breaker maintenance programs in November 2007 and modified it based on Electric Power Research Institute Documents TR-106857-V2 and TR-106857-V3, which are preventive maintenance program bases for low and medium voltage switchgear. The licensee only implemented portions of the recommended maintenance program, and had no engineering analysis or technical basis to justify the changes. Additionally, the guidance states in part that, this program assumes breakers are in nominally good condition to begin with. Breakers that have not been serviced for a very long time may need an overhaul or have a detailed inspection performed before this program is applied. The licensee had not been performing the entire vendor or Electric Power Research Institute recommended tests, inspections, and refurbishments on the breakers since installation. The team reviewed the licensee\'s circuit breaker maintenance procedures and records. The team determined that the licensee had not refurbished Asea Brown Boveri 4160 or General Electric 480 V safety-related and risk significant non-safety-related circuit breakers within the vendor specified 10-year maximum overhaul periodicity or the Electric Power Research Institute guidance of 12 years and had no engineering basis or evaluation to justify the deviation. The team compared the Electric Power Research Institute guidance and vendor-recommended maintenance requirements against the licensee\'s maintenance procedures and found that the licensee was not performing some of the recommended activities or had extended the periodicity of some inspections beyond even the Electric Power Research Institute recommended guidelines. The Fort Calhoun Station program for medium and low voltage switchgear and circuit breakers did not include most of the recommended testing and trending. Specifically, no testing of the operation of the 125-V DC control circuitry was performed at the voltages postulated to exist at the device terminals during design basis events. Contemporary industry standards and Electric Power Research Institute guidance recommend reduced control voltage testing as part of breaker maintenance. Vendor overhaul procedures include reduced control voltage testing on the as-found and as-left control circuit. While there is not an explicit requirement to perform reduced voltage testing on breaker control circuitry, the Electric Power Research Institute guidance recommends reduced voltage testing on breaker control circuitry in order to have reasonable assurance of reliable operation of control circuitry at the postulated minimum control voltage. Additional recommended testing per the preventative maintenance program basis DocumentsTR-106857-V2 and TR-106857-V3 that were not being performed included: Thermography inspections of the breakers and switchgear at recommended periodicity and trending, and: Measurement of the electrical resistance of coils and relays, trended over time to detect progressive failure of winding insulation and give an indication of the condition of these electrical devices. As a result, the team requested the basis for not performing all of the recommended maintenance activities. The licensee was unable to produce an engineering evaluation that allowed the use of the Electric Power Research Institute guidance versus the vendor guidance. Additionally, the team found that the licensee failed to update their in-use guidance when operating experience or new vendor information were issued. Because the licensee was unable to produce documentation demonstrating recommended maintenance had been performed at the appropriate intervals or which qualified the practice of extending the maintenance and refurbishment intervals, the team was concerned about the reliability of the safety-related and safety significant breakers that had not been overhauled within 10 years.n The licensee stated that the 10-year vendor requirement was based on breakers manufactured and lubricated with petroleum-based grease and that their Asea Brown Boveri circuit breakers were lubricated with synthetic-based grease, Anderol 757, which does not dry out as fast and extends the useful life of the lubrication. The licensee cited a May 11, 1995, letter from Asea Brown Boveri/Combustion Engineering that implied grease hardening was not an issue with Anderol 757 lubricant. The team identified operating experience which showed that other licensees had experienced grease hardening in Asea Brown Boveri breakers that contained the Anderol 757.Following the10 CFR Part 21 report issued by D. C. Cook on March 3, 1989, Asea Brown Boveri established the 10 year overhaul frequency. This report was issued after two Asea Brown Boveri 4160 V breakers failed to close because of hardened grease in their operating mechanism. Additional operating experience from Perry supported that grease hardening can occur in less than ten years, pertaining to the 4160 V C residual heat removal (RHR) pump breaker. It stated in part, Various anomalies were identified during the process of disassembling the breaker, and the lubricant within the operating mechanism appears to be hardened. Based on the breaker serial number it was determined that this breaker would have used the synthetic lubricate. This provided further evidence that synthetic grease can degrade in less than 10 years. Asea Brown Boveri breaker historical industry data showed that the lubrication in the operating mechanism tended to harden within 10 years and that this condition can cause sluggish breaker operation. The issue was entered into the licensees corrective action program- 14 V Enclosure and was being evaluated under Condition Report 2009-2306. This issue is unresolved pending review of the causes of the breaker failures as related to the improperly performed maintenance (Unresolved Item 05000285/2009007-02)
05000285/FIN-2010003-06	2010Q2	Fort Calhoun	Failure to Perform a Proper 50.59 Evaluation	On April 9, 2010, the licensee repaired a section of power cable for motor control center MCC-3A1 with cable splices. Approximately 17 feet of 500 MCM cable was removed from each of the three phases for the supply to MCC-3A1 and Burndy compression type butt splices were used to splice new cables to the remaining existing cables. The inspectors reviewed Section 8.5, Initial Cable Installation Design Criteria of the USAR. USAR 8.5 states, in part: The Cable and Conduit Schedule Notes, Figure 8.5-1, provides the standard design criteria for cables and conduits. Deviation from the standard criteria is acceptable provided an analysis has been completed which justified the deviation. USAR Figure 8.5-1, Cable and Conduit Schedule Notes, Note 19 states: Splicing in cable trays is not allowed unless specifically called for on drawings. Exceptions to this requirement shall require the written approval of the engineer. USAR Figure 8.5-1, Note 26 states: Deviations from the standards stated above is (are) acceptable provided an analysis has been performed to justify the deviation. USAR Section 8.5.4.c states: Cable splicing in cable trays is used only for connection of incoming and outgoing cables with containment electrical penetration conductors. The licensee performed a 50.59 Screen in accordance with the guidance provided in FCSG-23, 10 CFR 50.59 Resource Manual. The guidance adopts NEI 96-07, Revision 1 Guidelines for 10 CFR 50.59 Implementation which includes five screening questions to determine if a complete evaluation of 10 CFR 50.59 is required. The licensee determined that a cable splice was an equivalent replacement for cable, and thus it screened out in accordance with NEI-96-07 and no evaluation of 10 CFR 50.59 was required. The inspectors determined that a cable splice is not an equivalent replacement, thus a violation of 10 CFR 50.59 occurred for failure to perform an evaluation of the cable splice against the criteria set forth in 10 CFR 50.59. The violation would be greater than minor only if prior NRC approval was required. The inspectors are reviewing the technical aspects of this issue to determine if prior NRC approval would have been required. In accordance with the guidance in Inspection Manual Chapter 0612, an unresolved item is warranted if more information is required to determine if the performance deficiency is more than minor, URI 05000285/2010003-06, Failure to Perform a Proper 50.59 Evaluation
05000285/FIN-2010004-01	2010Q3	Fort Calhoun	Inadequate Documentation of the Adequacy of Design for the Pumps that Transfer Fuel Oil from Storage Tank FO-10 to FO-1	The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III due to the failure of the licensee to perform suitable testing to determine the adequacy of the design of equipment related to transferring diesel fuel from one storage tank to another. Specifically, the inspectors questioned whether fuel oil transfer pump FO-37 or a portable hand pump to be used in the event that FO-37 was unavailable to transfer fuel from storage tank FO-10 to FO-1 would be able to perform the design function. No calculations or previous testing documentation could be provided and when tested to demonstrate that the portable hand pump could perform the intended design function, the portable hand pump failed. Subsequently, the licensee evaluated that fuel oil transfer pump FO-37 is adequately designed to transfer fuel oil from FO-10 to FO-1. The licensee entered this issue into the corrective action program as Condition Reports 2010-3123, 2010-3921, and 2010-4315. The inspectors determined that the licensees failure to provide calculations or testing documentation that fuel oil transfer pump FO-37 or the designated portable hand pump could perform the intended design function was a performance deficiency. This finding is greater than minor because it affected the Mitigating System Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the equipment performance attribute to maintain availability and reliability of the diesel generators. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has a very low safety significance (Green) because it was not a design or qualification deficiency, does not represent an actual loss of safety function nor did it screen as potentially risk significant for external events. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding (Section 1R15).
05000285/FIN-2010004-02	2010Q3	Fort Calhoun	Failure to Submit a Required Licensee Event Report	The inspectors identified a Severity Level IV noncited violation for the failure to submit a licensee event report within 60 days as required by 10 CFR 50.73. Specifically, the diesel fuel oil storage system was inoperable for approximately 24 hours from January 6, 2010, until January 7, 2010. On January 6, 2010, fuel oil transfer pump FO-37 was inoperable due to a fire main rupture submerging the pump for approximately 24 hours. With no other means to transfer fuel from storage tank FO-10 to FO-1, the fuel oil storage system was inoperable, and the fuel volume in FO-10 was unavailable. This was reportable condition required by 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by technical specifications. The licensee performed a reportability evaluation, and the violation was entered into the corrective action program as Condition Report 2010-3865. The inspectors determined that the licensees failure to submit a licensee event report was a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was potentially affected. Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for traditional enforcement only in accordance with the NRC Enforcement Policy. This is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy (Section 1R15).
05000285/FIN-2010004-03	2010Q3	Fort Calhoun	Failure to Update the Updated Safety Analysis Report Solid Waste	The inspectors identified a Severity Level IV, noncited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, paragraph (e) which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Contrary to the above, the licensee failed to update periodically the Updated Safety Analysis Report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Specifically, since December 2006, the licensee stored a significant source of radioactivity in the original steam generator storage facility but failed to describe the source, volume, and storage of radioactive equipment in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2010-3636. The inspectors determined that the failure to update the Updated Safety Analysis Report as required by 10 CFR 50.71(e), Maintenance of Records, Making of Reports was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding was more than minor because it had a material impact on licensed activities in that a radioactive solid waste storage facility was relocated from the plant radiological controlled area to the owner controlled area without being described in the Updated Safety Analysis Report. The finding was characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy (Section 1R17).
05000285/FIN-2010004-04	2010Q3	Fort Calhoun	Failure to Translate Calculation into Calibration Procedure	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, the licensee failed to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, since January 2009, the licensee failed to correctly translate results of Calculation FC 05561, CCW Relief Valve Setpoints, into calibration procedures used to calibrate pressure control switches PCS-412 and PCS-413. The licensee has entered this violation into their corrective action program as Condition Report 2010-3658. The inspectors determined that the failure to correctly translate the results of the setpoint calculation into calibration procedures and instructions as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control is a performance deficiency. The finding was more than minor because it adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was more than minor because the finding resulted in a condition where there was a reasonable doubt on the operability of the component cooling water system containment isolation valves. Using Phase 1 of Inspection Manual Chapter 0609, Significance Determination Process, the finding was determined to have very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building. This finding has a crosscutting aspect in the area of human performance work practice because the licensee failed to define and effectively communicate expectations regarding procedural compliance and personnel following procedures. Specifically, in January 2009, the licensee failed to effectively communicate expectations regarding personnel following procedures to implement calculation changes (H.4(b))(Section 1R17).
05000285/FIN-2010004-05	2010Q3	Fort Calhoun	Failure to Perform a 10 CFR 50.59 Evaluation	The inspectors identified a Severity Level IV violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the facility as described in the Updated Safety Analysis Report. On April 9, 2010, the licensee changed the facility as described in the Updated Safety Analysis Report to install a cable splice in a safety related cable without determining if prior NRC approval was required. The licensee took actions to make the modification temporary until a permanent repair could be made and entered the issue into the corrective action program as Condition Report 2010-4466. Fort Calhoun Station utilizes NEI 96-07 as their process to meet 10 CFR 50.59 requirements. Their failure to perform a 10 CFR 50.59 evaluation, in accordance with NEI 96 07, prior to changing the facility as described in the Updated Safety Analysis Report is a performance deficiency. The inspectors reviewed this issue in accordance with NRC Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC\'s regulatory ability was potentially affected. Specifically, the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done the regulatory function is impacted, and is therefore more than minor. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated for Traditional Enforcement only in accordance with the NRC Enforcement Policy. The inspectors concluded that the 10 CFR 50.59 evaluation would have likely identified that prior NRC approval would have been required, unless the change to the facility was for a short duration of time. This was due to the introduction of additional potential failure mechanisms of the splices that are age-dependent. Since the licensee subsequently classified the cable splice as a temporary modification, and scheduled to be removed during the next refueling outage, the aging mechanisms would no longer be applicable. Therefore, this is a Severity Level IV violation as defined in Section 2.2.1.c of the NRC Enforcement Policy (Section 1R20).
05000285/FIN-2010004-06	2010Q3	Fort Calhoun	Failure to Follow Radiation Work Permit Requirements	The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.8.1, for failure to follow radiation work permit requirements. On November 13, 2009, two individuals became contaminated while cleaning the gasket seating surface on the endbell of the letdown heat exchanger because they did not use face shields as required by the radiation work permit. The licensee immediately restricted the two individuals from entry into the radiologically controlled area, conducted a coaching session with the individuals involved and placed this issue into the corrective action program as Condition Report 2009-5688. The failure to follow the instructions listed on a radiation work permit was a performance deficiency. The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to follow radiation work permit instructions increased personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work practices, human error prevention techniques, because the individuals failed to use self and peer checking to ensure they were signed onto the appropriate task for the work to be performed (H.4(a))(Section 2RS01).
05000285/FIN-2010004-07	2010Q3	Fort Calhoun	Failure to Properly Plan a Maintenance Activity	The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.8.1, for failure to appropriately control radiation exposures due to improperly planned maintenance activities associated with Work Package 09-AP-20. The maintenance work involved valve modifications and boric acid system cleanups. These activities resulted in exceeding the original dose estimate by more than 50 percent. The licensee entered this issue into the corrective action program as Condition Reports 2009-6171, 2009-6264 and 2010-1696. The failure to properly plan maintenance activities to minimize personnel radiation dose is a performance deficiency. This finding is greater than minor because it affected the Occupational Radiation Safety Cornerstone attribute of program and process in that ALARA planning or radiological controls did not prevent unplanned, unintended dose for a work activity. This caused increased collective radiation dose for the job activity to exceed the planned dose of approximately 14 rem by more than 50 percent. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this finding to be of very low safety significance because the finding involved ALARA planning and controls and the licensees latest rolling 3-year average does not exceed 135 person-rem. This finding had an associated human performance crosscutting aspect in the work practices component because the licensee did not ensure supervisory and management oversight of work activities, including the contractor, to maintain doses ALARA (H.4(c))(Section 2RS02).
05000285/FIN-2010004-08	2010Q3	Fort Calhoun	Inadequate Maintenance Procedure Results in Water in East Switchgear Room and Room 19	The inspectors reviewed a self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1, for the licensees failure to provide an adequate maintenance procedure for fire protection system flushing. Specifically, while performing OP-PM-FP-1000 on August 19, 2010, water backed up the VA-87 drain line and spilled onto the east switchgear room floor, into Room 19 below, as well as pooling on top of and inside of cable trays. The licensee has entered this issue into their corrective action program as Condition Report 2010-4423. The inadequate maintenance procedure is a performance deficiency. This finding is more than minor because if left uncorrected the performance deficiency could have the potential to lead to a more significant safety concern. Specifically the use of OP-PM-FP-1000 allows the potential wetting of safety related equipment in the east switchgear room and Room 19. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609, Appendix A, to determine its significance. Using Attachment 4 of that appendix, the inspectors determined that the finding has very low safety significance 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. Through conversations with the fire protection system engineer and other licensee members and the fact that similar issues have occurred in the past, the inspectors determined that the primary cause of this finding was the failure to adequately assess the significance of previous condition reports which would have required them to perform a more thorough cause evaluation. Therefore, this finding has a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems, such that, the resolutions address causes and extent of conditions, as necessary (P.1(c))(Section 4OA2).
05000285/FIN-2010004-09	2010Q3	Fort Calhoun	Failure To Perform Vendor And Industry Recommended Testing On Safety-Related And Risk Significant 4160 V And 480 V Circuit Breakers	The inspectors identified a Green noncited violation of Technical Specification 5.8.1(a) for inadequate procedures associated with 4160 V and 480 V safety-related breaker maintenance procedures. The inspectors determined that maintenance procedures used to ensure that 4160 V and 480 V safety-related breakers were being maintained and overhauled in a timely manner were inadequate. The licensee did not have an engineering analysis or technical basis to justify the deviation from vendor and/or Electric Power Research Institute guidance. The inspectors determined that this issue affected the procedure quality attribute for maintenance procedures of the Mitigating System Cornerstone of reactor safety. Specifically, the issue was more than minor because the failure to incorporate the vendor required maintenance and frequency or fully incorporate Electric Power Research Institute maintenance recommendations for extending the service interval into maintenance procedures for safety related breakers. If left uncorrected, this failure affected the availability, reliability, and capability of mitigating systems that respond to initiating events to prevent undesirable consequences because the reliability of safety-related breakers refurbished using the deficient procedures cannot be predicted. This issue was entered into the licensees corrective action program as Condition Report 2009-2306. Using the Significance Determination Process, Phase 1 Screening Worksheet, for the Initiating Events, Mitigating Systems, and Barriers Cornerstones the finding was potentially risk significant for multiple systems. Because the probability of multiple system effects is not effectively addressed by a Phase 2 analysis, a Phase 3 analysis was performed. The analyst determined that while the licensee failed to perform adequate maintenance on the breakers, the actual failure rate of the breakers was no greater than the theoretical design failure rate. The finding was determined to be of very low safety significance because the deficiency did not result in any loss of function. The finding was not risk significant due to a seismic, flooding, or severe weather-initiating event and because other plant-specific analyses that identify core damage scenarios of concern were not impacted. This finding has a crosscutting aspect in the area of problem identification and resolution because the licensee did not effectively incorporate pertinent industry operating experience into the preventive maintenance programs for the 4160 V and 480 V safety-related and risk significant non-safety-related circuit breakers (P.2(b))(Section 4OA2)
05000285/FIN-2010004-10	2010Q3	Fort Calhoun	Inadequate Maintenance Procedure Results in Plant Shutdown	A self-revealing Green noncited violation of Fort Calhoun Station Technical Specification 5.8.1 occurred for an inadequate procedure for verifying the connection between cable lugs and cables. This inadequacy resulted in the loss of Motor Control Center MCC-3A1 and a subsequent plant shutdown. The licensee repaired the affected equipment and entered this issue into the corrective action program as Condition Report 2010-4423. The inspectors determined that the licensees inadequate maintenance procedure was a performance deficiency. This finding was greater than minor because it was similar to a non-minor example 4.b in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, in that a procedural error caused a reactor trip or other transient. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. Using Attachment 4 of that chapter, the inspectors determined that this finding has very low safety significance because all of the items in Table 4a, of the Mitigating Systems Cornerstone checklist, were answered in the negative. Since the finding is not indicative of current licensee performance, there is no crosscutting aspect assigned to this finding (Section 4OA3).
05000285/FIN-2010004-11	2010Q3	Fort Calhoun	Licensee-Identified Violation	Technical Specification 5.8.1 requires written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, February 1978. Section 7.e of Appendix A to Regulatory Guide 1.33 requires radiation protection procedures. Procedure RP-307, Use and Control of Temporary Shielding, Revision 18, Step 5.3 states that, No temporary shielding shall be installed, removed or modified unless authorized. Step 7.4.4.c of this procedure states that, Radiation Protection personnel are NOTIFIED prior to removing shielding. Contrary to these requirements, on December 8, 2009, the containment coordinator removed ten lead shielding blankets hanging on a hand rail without notifying radiation protection. The removal of the blankets increased the dose rate on that side of the railing resulting in increased dose rates. The containment coordinator was counseled by the radiation protection supervisor and the ALARA coordinator. The inspectors determined this finding to be of very low safety significance because: (1) it did not involve ALARA - 55 Enclosure planning and controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This issue was entered into the licensee\'s corrective action program as Condition Report 2009 6454.
05000285/FIN-2010004-12	2010Q3	Fort Calhoun	Licensee-Identified Violation	Licensee Event Report 05000285/2010-004 identified that accelerometer flow elements for both pressurizer safety valves were inoperable from April 28 to June 2, 2010. This condition is prohibited by technical specifications after 7 days, therefore meeting the criteria for a condition prohibited by technical specifications on May 5, 2010. The licensee event report was submitted 17 days later, on August 30, 2010. This is a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(i)(B) for failure to submit a required licensee event report within 60 days of a condition prohibited by technical specifications.
05000285/FIN-2011003-01	2011Q2	Fort Calhoun	Failure to Adequately Design a Reactant Coolant Pump Lube Oil Collection System	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix R, Section III.O for the failure to ensure an adequate seismic design of the reactor coolant pumps oil collection system. The licensee used 2-inch copper pipe with brazed joints in the lube oil collection system. The seismic analysis of the system assumed the use of ASME Section IX during the installation of the system, but no codes or standards were used by the licensee for the brazed joints. The inspectors determined that the failure to design and install an adequate oil collection system which included provisions for the drain lines to the oil collection tank was a performance deficiency. This finding had a credible impact on safety because the inadequate installation and design of the oil collection systems presented a degradation of a fire confinement component, which had a fire prevention function of not allowing an oil leak. The inspectors determined the finding was more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors, such as a fire. The inspectors reviewed Inspection Manual Chapter 0609, Appendix F, and determined the finding was of very low safety significance, because of the low degradation rating of the fire confinement category related to the as found condition of the oil collection piping, the extremely low frequency of reactor coolant pump oil leaks, minor actual reactor coolant pump oil leaks during the past operating cycle, and other area fire protection defense-in-depth features such as automatic fire detection, manual suppression capability, and safe shutdown capability from the main control room. This finding involved a legacy issue associated with a modification for original installation; therefore, there were no assigned cross-cutting aspects.
05000285/FIN-2011003-02	2011Q2	Fort Calhoun	Failure to Follow Scaffolding Procedure	The inspectors identified a noncited violation of Technical Specification 5.8.1.a for failure to follow scaffold specification and construction Procedures SO-M-35 and PED-CSS-12. This led to the licensee declaring a number of emergency core cooling components inoperable and entering technical specification 2.0.1. The inspectors determined that not following a procedure required by Technical Specification 5.8.1.a was a performance deficiency. The finding was more than minor because if left uncorrected it would have the potential to lead to a more significant safety concern. The licensee routinely failed to perform seismic evaluations of scaffolds erected near safety-related equipment not constructed in accordance with Procedures PED-CSS-12 or SO-M-35 for preconfigured seismic scaffolding. The finding was associated with the Mitigation Systems Cornerstone while the reactor was operating; therefore, Inspection Manual Chapter 0609, Attachment 4 screening checklist was used. The finding was determined to have very low safety significance because it did not involve the total loss of any safety function, and did not contribute to external event initiated core damage accident sequences. The inspectors determined the primary cause of the finding was lack of the licensees oversight of the scaffolding program. The finding had a crosscutting aspect in the area of human performance, specifically, work practices, in that, the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported.
05000285/FIN-2011003-03	2011Q2	Fort Calhoun	Failure to Follow Radiation Work Permit Procedure	Inspectors identified a noncited violation of Technical Specification 5.8.1a for the failure to follow procedural requirements to plan and carry out decontamination work in the spent fuel pool transfer canal. On January 24, 2011, decontamination work was performed in the spent fuel pool transfer canal, using Radiation Work Permit 11-3317. While planning and controlling the work, the licensee failed to follow multiple procedure steps. Specifically, the licensee did not prepare an ALARA planning worksheet as the initial step of generating the radiation work permit, did not document justification for changing the electronic dosimeter set points which were eventually determined to be inappropriate, and did not perform an ALARA briefing before the entries were made into the spent fuel pool transfer canal, which was posted as a restricted locked high radiation area. The inspectors also determined that there were aspects of the procedure that contained vague expectations, which contributed to decisions being made without using the procedure. The failure to follow a procedure was a performance deficiency. The finding was more than minor because it negatively impacted the Occupational Radiation Safety Cornerstones attribute of program and process, in that, by not following the procedure; radiological safety attributes built into the radiation work permit program were circumvented. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the violation was of very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This deficiency had a crosscutting aspect in the area of human performance related to work practices. Specifically, the licensee did not communicate human error prevention techniques, such as, holding pre-job briefs, self- and peer- checking, and proper documentation of activities.
05000285/FIN-2011003-04	2011Q2	Fort Calhoun	Failure to Provide Procedural Guidance to Replace or Evaluate Age Degraded Components	A self-revealing noncited violation of Fort Calhoun Technical Specification 5.8.1, Procedures, occurred due to the failure of the licensee to ensure that adequate procedures were available for maintenance which was conducted on the reactor protective systems power supplies. Specifically, there was no procedural guidance to require replacement of power supplies, or an engineering justification for continued operation, once power supplies exceeded their vendor recommended life, and/or showed signs of failure and degradation. The inspectors determined that the licensees failure to provide procedural guidance to evaluate and/or replace age-degraded components was a performance deficiency. This was a result of the licensees failure to properly implement a required procedure, and was within the licensees ability to foresee and correct and could have been prevented. This performance deficiency was more than minor because it could be reasonably viewed as a precursor to a significant event, it could lead to a loss of the reactor protective system. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Attachment 4, and determined that this finding was associated with the Mitigating Systems Cornerstone, specifically the primary degraded reactivity control contributor. Because this finding occurred while the unit was operating at full power, the inspectors used Inspection Manual Chapter 0609 to determine its significance. The inspectors determined that the finding represented a qualification deficiency confirmed not to result in a loss of functionality because none of the failures to date prevented a reactor protective systems channel from tripping. Therefore, in accordance with the Phase 1 screening, the finding was of very low risk significance. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the component of operating experience because the licensee failed to adequately evaluate and communicate relevant internal and external operator experience.

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