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	Quarter	Site	Title	Description
05000275/FIN-2009003-01	2009Q2	Diablo Canyon	Corrective Action Following Degraded Offsite Power System	The inspectors identified an unresolved item related to the acceptability of the 230 kV preferred offsite power system to meet design basis requirements. Additional NRC review is needed to determine if the preferred offsite system has sufficient capacity and capability to supply the engineered safety features buses for all required accidents and transients. On April 10, 2009, the inspectors identified that the plant electrical design analysis may not be adequate to demonstrate that the 230 kV preferred offsite power system had sufficient capacity and capability to meet station loads following an accident on one unit and concurrent safe shutdown on the remaining unit or for a concurrent safe shutdowns on both units. The Diablo Canyon offsite power sources include a normal supply from the 230 kV distribution system and a delayed supply from the 500 kV distribution system. The normal supply is required to immediately power the engineered safety feature systems following a station accident or a reactor trip. The delayed supply backs up the normal supply and can be aligned to power the engineered safety feature systems in about 30 minutes. NRC Safety Evaluation Report, Safety Evaluation by the Directorate of Licensing U.S. Atomic Energy Commission in the Matter of Pacific Gas And Electric Company Diablo Canyon Nuclear Power Station, Units 1 and 2 San Luis Obispo County, California Docket Nos. 50-275 And 50-323, and Final Safety Analysis Report Update, Section 8.1, Offsite Power Systems, established IEEE Standard 308-1971, Class IE Electrical Systems, as part of the preferred offsite power system design basis. IEEE Standard 308-1971, Section 8.1.1, Multi-Unit Station Considerations, stated: Capacity. A multi-unit station may share preferred power supply capacity between units. In such a case, as a minimum the total preferred capability must be sufficient to operate the engineering safety features for a design basis accident on one unit and those systems required for concurrent safe shutdown on the remaining units. The type of accident and shutdown and the unit assumed to have the accident, shall be those which give the largest total preferred capability requirements. Pacific Gas and Electric used Design Calculation 357AA-DC, Units 1 and 2 Load Flow, Short Circuit and Motor Starting Analysis, September 24, 2007, to ensure that the preferred offsite power system was capable of meeting design basis electrical load requirements. The inspectors identified that Calculation 357AA-DC did not include load flow cases representing the largest total capability for an accident on one unit and concurrent safe shutdown of the other unit or concurrent safe shutdown of both units. Calculation 357AA-DC modeled the limiting load flow cases as an accident (or unit trip) on one unit while assuming a previous shutdown had occurred on the other unit. The load flow modeling was based on the assumption that plant operators would perform an orderly shutdown entailing the manual transfer of electrical loads to the 230 kV system at a time of low electrical demand from the accident or tripped unit. On June 26, 2009, the licensee completed a preliminary re-evaluation of preferred offsite power supply load flow assuming an accident on one unit and a concurrent safe shutdown on the remaining unit, and for an assumed concurrent safe shutdown on both units. The licensee concluded the voltage at the 4160 Class 1E vital buses would be less than adequate to support operation of the engineering safety features under design conditions. The licensee also analyzed the plant response based on actual available 230 kV switchyard voltages between November 2008, and June 26, 2009. For these cases, the licensee concluded that 4160 Class 1E vital bus voltages would have intermittently dropped below the minimum voltage required for operability of the engineering safety features. The inspectors concluded that actual 230 kV system voltage recovered prior to exceeding the 72-hour action time for Technical Specification 3.8.1, AC Sources - Operating, for any single occurrence. Pacific Gas and Electric had previously identified that the 230 kV offsite power source had insufficient voltage (reported as Licensee Event Report 1-95-007, 230 kV System May Not Be Able to Meet its Design Requirements for all Conditions Due to Personal Error). The corrective actions included increasing the capability of the startup transformers and installation of large capacitor banks at the plant switchyard and Mesa Substation. When sizing the replacement transformers, the licensee assumed that the preferred offsite power system only needed to have the capacity and capability for an accident or trip on one of the two units. In a licensing position paper for the 230 kV system loading requirements (Letter File 227961, from the Director Licensing to Director, Electrical I&C Engineering September 27, 1995), the licensee added the word orderly to the safety shutdown requirements specified in IEEE Standard 308-1971. The licensee did not perform a 10 CFR 50.59 evaluation of this change nor seek prior NRC approval. As a result, the licensees previous corrective actions were insufficient to restore the preferred offsite power system to compliance with the design basis. This issue is unresolved pending NRC review of the 230 kV preferred offsite power system design basis requirements. Unresolved Item: 05000275;323/2009003-01
05000275/FIN-2009005-01	2009Q4	Diablo Canyon	Less Than Adequate Work Planning Resulted in the Release of Two Gas Decay Tanks	The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1, Procedures, after Pacific Gas and Electric inadvertently released the contents of two gas decay tanks into the auxiliary building. Gas Decay Tank 2-2 was in purge mode. On October 11, 2009, plant operators were implementing an equipment control clearance to drain the emergency core cooling systems. A second group of operators were implementing a core offload master clearance. The parallel implementation of both equipment clearances resulted in Gas Decay Tank 2-2 to be vented into the auxiliary building. The auxiliary building operator received a low gas header pressure alarm after the pressure dropped to 15 psig. Per procedure, the operator aligned Gas Decay Tank 2-3 to purge mode. As a result, the second gas decay tank was released into the auxiliary building through the open vent path. The inspectors concluded that the radiological consequence of the event did not result in a potential for overexposure because the reactor had been shutdown since October 3, 2009. The inspectors concluded that the failure to properly implement the core offload master equipment control clearance was a performance deficiency. The finding is more than minor because the performance deficiency could be reasonably viewed as a precursor to a significant event. The inspectors determined the finding to have very low safety significance because the performance deficiency only represented a degradation of the auxiliary building radiological barrier function. This finding has a crosscutting aspect in the area of human performance associated with the work control component because the licensee did not adequately plan and coordinate the two clearance activities or fully consider the impact the work had on different job activities and the need for the two work groups to maintain interfaces (H.3(b)
05000275/FIN-2009005-02	2009Q4	Diablo Canyon	Failure to Properly Plan a Maintenance Activity	The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1(a) for failure to properly plan numerous outage maintenance activities including the disassembly of the Unit 2 reactor head. Specifically, Work Orders 68004363 (disassembly of the old head) and 68003988 (scaffolding activities) were not properly planned, thereby requiring those maintenance activities to be changed and/or repeated, which resulted in increased radiation exposure. Radiation Work Permits 09-2233 and 09-2237 for the disassembly of the Unit 2 old reactor vessel closure head and supporting activities during Refueling Outage 15 had an initial combined dose estimate of 5.869 rem and 1102 man-hours. However, the job ended with an actual combined dose of 17.378 rem and 1882 man-hours, which exceeded the initial dose estimate by 296 percent. The overarching reason for exceeding the original dose estimate was improper planning and control for the maintenance, which increased the man-hours to complete the task by 170 percent. The licensee entered this deficiency in the corrective action program as Notification 50275107 and plan to perform an apparent cause evaluation. The failure to properly plan maintenance activities 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 the inadequate ALARA planning caused increased collective radiation dose for the job activity to exceed 5 person-rem and the planned dose 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 although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 135 person-rem per unit. Furthermore, the finding had an associated human performance cross-cutting aspect in the work control component because the licensee did not fully incorporate job site conditions, plant structures, systems, and components, as well as human-system interface and the need for planned contingencies to maintain doses ALARA (H.3(a))
05000275/FIN-2009005-03	2009Q4	Diablo Canyon	Inadequate 50.59 Evaluation for Steam Generator Tube Rupture Analysis	The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after Pacific Gas and Electric failed to perform an adequate evaluation of a change to the facility as described in the Final Safety Analysis Report Update. In 1992, the licensee identified that auxiliary feedwater and steam generator power-operated relief valve flow rates assumed in the steam generator tube rupture accident analysis were non-conservative. To address the non-conforming condition, Pacific Gas and Electric changed the accident analysis to include a new time critical operator action to terminate turbine-driven auxiliary feedwater flow 5.54 minutes after the reactor trip and credit motor driven auxiliary feedwater automatic level control to the ruptured steam generator. The licensee did not perform a 10 CFR 50.59 safety evaluation of these changes. The NRC basis of approval of the accident analysis include four time critical operator actions, each assumed to occur after the first 10 minutes following the accident. The inspectors concluded that NRC approval was required before the licensee added the new time critical manual action under the 10 CFR 50.59 Rule in effect at the time because the change reduced the margin to safety to the basis of Technical Specification 3.7.4, 10% Atmospheric Dump Valves. The inspectors also concluded that prior NRC approval was required under the current 50.59 Rule because the change result in a departure from a method of evaluation described in the Final Safety Analysis Report Update. The performance deficiency, a less than adequate 50.59 evaluation, was the result of a latent issue. However, the inspectors concluded that the licensee had reasonable recent opportunities to identify the problem. The inspectors also concluded that plant programs, processes or organizations have not changed such that the problem would not reasonably occur today and that the most significant contributor to the performance deficiency was reflective of current plant performance. The licensee entered this issue into their corrective action program as Notification 50270786. The failure of Pacific Gas and Electric to perform a 10 CFR 50.59 evaluation of the changes to the steam generator tube rupture accident analysis was a performance deficiency. The inspectors evaluated this issue using traditional enforcement because the performance deficiency had the potential for impacting the NRCs ability to perform its regulatory function. The issue was more than minor because of reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The inspectors also evaluated the significance of this issue under the Significance Determination Process using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding affected the Mitigating Systems Cornerstone because the change described the operator actions required to mitigate steam generator tube rupture accident. The inspectors concluded the finding screened Green because the finding was a design deficiency that did not result in the loss of operability or functionality. The inspectors concluded that the violation was a Severity Level IV because the issue screened Green under the Significance Determination Process. The inspectors concluded that 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 the steam generator tube rupture analysis such that the resolutions addressed causes and extent of condition (P.1(c
05000275/FIN-2009005-04	2009Q4	Diablo Canyon	Less than Adequate Replacement Reactor Head Modification Design Control	The inspectors identified a noncited violation of Title 10 CFR, Part 50, Appendix B, Criterion III, Design Control, after the design contractor failed to perform adequately calculations demonstrating that the replacement reactor head met ASME Code acceptance criteria. The contractor failed to use the critical seismic damping values specified in the plant design basis for the design of the integrated head assembly and the control rod drive mechanism housing assembly and when calculating component stress during a postulated design basis earthquake. The licensee entered this condition into the corrective action program as Notifications 50276107 and 50276288. The inspectors concluded that the failure to properly implement the plant design basis in the replacement head design was a performance deficiency. The finding is more than minor because the performance deficiency is associated with the Initiating Events Cornerstone design control attribute and adversely affected the cornerstone objective to limit the likelihood of loss of a coolant accident during a seismic event. The inspectors determined the finding is of very low safety significance because assuming worst case degradation, the finding would not result in exceeding the Technical Specification limit for reactor coolant system leakage nor have likely affected other mitigation systems resulting in a total loss of their safety function. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not identify the use of improper damping values with a low threshold for identifying issues during oversight of contractor activities and design reviews, (P.1(a)
05000275/FIN-2009005-05	2009Q4	Diablo Canyon	Less than Adequate Change Evaluation to the Facility as Described in the Final Safety Analysis Report Update	The inspectors identified a Severity Level IV noncited 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 Final Safety Analysis Report Update. In October 2009, the inspectors identified that the replacement reactor head contractor used incorrect damping values in the replacement head design. The contractor was unable to demonstrate that the design met ASME Code using the damping values specified in the plant design basis. On November 5, 2009, the licensee incorporated the new damping values and revised the method for determining the seismic response spectra. Using NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, Revision 1, the inspectors concluded that these changes resulted in a departure from a method of evaluation described in the Final Safety Analysis Report Update establishing the facility design bases. The licensees 50.59 evaluation, Licensing Basis Impact Evaluation LBIE 2009-021, Integrated Head Assembly, was less than adequate to conclude that prior NRC approval was not required for the changes. The licensee entered this issue into their corrective action program as 50276288. The failure of Pacific Gas and Electric to perform an adequate 10 CFR 50.59 evaluation prior to changing the facility as described in the Final Safety Analysis Report Update is a performance deficiency. The inspectors evaluated this issue using the traditional enforcement process because the performance deficiency had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors concluded that the issue was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The inspectors also evaluated this issue using the Significance Determination Process. The inspectors concluded that the violation affected the Initiating Events Cornerstone because the change potentially decreased the structural integrity of the control rod drive mechanism reactor coolant pressure barrier and screened Green because assuming worst case degradation, the finding would not result in exceeding the technical specification limit for reactor coolant system leakage nor have a likely effect on other mitigation systems resulting in a total loss of their safety function. The inspectors concluded that the violation was a Severity Level IV because the issue screened Green under the Significance Determination Process. The finding has 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 the original problem associated with the replacement reactor head design such that the resolutions address causes and extent of conditions, as necessary (P.1(c)
05000275/FIN-2009005-06	2009Q4	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, states, in part, that activities affecting quality shall be prescribed by documented procedures and shall be accomplished in accordance with those procedures. On August 12, 2009, plant operations management failed to implement Section 5.4.5 of safety-related Procedure OP1.ID3, Reactivity Management Program. The procedure states information provided by reactor engineering such as rod movement recommendations, reactivity management plans, ramp plans, and reactivity briefing sheets shall be provided to the control room via an operations shift order. Contrary to this, an unapproved Reactor Engineering Ramp Plan was utilized by control room staff for approximately two hours during a downpower and subsequent shutdown of Unit 2 reactor for emergent transformer repairs. The unapproved Reactor Engineering Ramp Plan was a draft copy utilized for the operations crew just-in-time simulator training for the shutdown, and was subsequently used in the control room by mistake. The approved Reactor Engineering Ramp Plan was available in the control room at the start of the ramp; however, it was not initially utilized. Upon discovery of the unapproved plan by the control room staff, the approved (correct) plan was then used for the remainder of the shutdown. Pacific Gas and Electric entered the issue into their corrective action program as Notification 50262580. The finding is of very low safety significance because no reactivity manipulations outside of the approved plan had been made prior to discovery by the control room staff
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-2012003-01	2012Q2	Diablo Canyon	Inadequate Preferred Offsite Power System Design Control	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after plant engineers failed to adequately translate regulatory requirements and the design bases into the offsite power interface calculation on May 6, 2011. As a result, the licensee failed to demonstrate that the 230 kilo-Volt preferred offsite power source had adequate capacity and capability to supply the minimum required terminal voltage to plant engineering safety features following a limiting transmission system contingency. The licensee took corrective actions to limit the plant load that would automatically transfer to the preferred power source following a unit trip and entered the condition into the corrective action program as Notification 50492766. The failure to ensure that the 230 kV power system had adequate capability and capability as defined in the current licensing basis requirements was a performance deficiency. This performance deficiency was more than minor because it was associated with the modification design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded this finding was of very low safety significance because the duration of potential losses of a single offsite power source safety function was less than the technical specification allowed outage time, did not represent an actual loss of safety function of risk significant non-technical specification equipment, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance, associated with the decision making component, because the licensee did not demonstrate that the proposed action was safe in order to proceed while assessing the CLB requirement during decision making.
05000275/FIN-2012003-02	2012Q2	Diablo Canyon	Failure to Perform a 50.59 Evaluation	The inspectors identified a non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, because the licensee failed to document an evaluation providing a basis that changes made to the facility and associated changes to Procedure OP J-2:VIII, Guidelines for Reliable Transmission Service for DCPP, did not require prior NRC approval. When a 50.59 review was performed, the licensee incorrectly concluded that only a screening was needed. Plant operators use Procedure OP J-2:VIII to determine the operability of the preferred offsite power system for various transmission system configurations. This change accepted a reduction in the preferred offsite power capacity and capability, below the minimum specified by the current licensing basis, due to local service area load growth. This condition would have likely required prior NRC approval had a 50.59 evaluation been performed. The licensee entered this finding into the corrective action program as Notification 50492767. The failure to perform a 50.59 evaluation was also a performance deficiency. The inspectors concluded that this issue involved traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. This performance deficiency is more than minor because it was associated with modification design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded this finding was of very low safety significance because the duration of potential losses of a single offsite power source safety function was less than the technical specification allowed outage time, did not represent an actual loss of safety function of risk significant non-technical specification equipment, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance, associated with the decision making component, because the licensee did not use conservative assumptions to adopt the licensing basis requirement during decision making.
05000275/FIN-2012003-03	2012Q2	Diablo Canyon	Failure to Follow Procedure for the Control of Tools for Use on Stainless Steel	Inspectors identified a non-cited violation of Technical Specification 5.4.1.e, for the failure to follow procedures that ensured hand files and wire brushes designated for stainless steel weld preparation were stored and maintained separately from hand files and wire brushes used on carbon steel. Specifically, the inspectors determined that the licensee was not segregating tools as required by Procedure MA1.ID12, Control of Tools for Use on Stainless Steel, Revision 1, because inspectors observed rust deposits on stainless steel components in the plant. This indicated that carbon steel contaminated tools may have been used on these systems. The licensee took corrective actions to segregate the stainless steel tools that were mixed with tools used on carbon steel. The licensee established segregated locations in tool rooms for the separation of abrasive tools, trained tool room attendants to properly store and mark abrasive tools designated for use on stainless steel and evaluated the systems with indications of rust deposits. This issue was entered into the licensees corrective action program as Notifications 50475217 and 50475779. Failure to assure that hand files and wire brushes designated for exclusive use on stainless steel were stored separately from tools used on other materials was a performance deficiency. This finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and, if left uncorrected, could become a more significant safety concern. Using Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance because the issue would not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding has a cross-cutting aspect in the area of human performance, work practices, in that the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported.
05000275/FIN-2012003-04	2012Q2	Diablo Canyon	Feedwater System Weld Flaw	The inspectors identified a finding for failure to follow applicable ASME Code requirements prior to returning the feedwater system to service after code repairs for flow accelerated corrosion. The licensee failed to recognize a rejectable indication in feedwater piping weld 2K16-550-30 FW 33 observable in the original acceptance radiography film. The licensee entered the issue into their corrective action program as Notifications 50473769 and 50475897 and re-examined the radiographic films for welds performed during Refueling Outage 2R16. A random re-examination of other radiographic films will be completed at a later date. This finding was more than minor because it is associated with the human performance attribute of the Initiating Events Cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of the engineering evaluation that was performed when the flaw was recognized, the inspectors determined that the structural integrity of the feedwater piping was not affected. Based on the results of a significance determination process Phase 1 evaluation, the finding was determined to be of very low safety significance (Green) because it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or an internal/external flood. This finding has a cross-cutting aspect in the area of human performance, work practices, in that the licensee failed to ensure human error prevention techniques, such as self- and peer-checking were used so that work activities are performed safely.
05000275/FIN-2012003-05	2012Q2	Diablo Canyon	Control Room Habitability Operability Issues	Technical Specification 3.7.10, Control Room Ventilation System (CRVS), required the licensee to maintain two independent and redundant control room ventilation trains and the control room envelope operable. The specified safety function of each control room ventilation train, in conjunction with the control room envelope, was to maintain operator dose below 5 rem equivalent. Technical Specification 5.5.19, Control Room Envelope Habitability Program, and Surveillance Requirement 3.7.10.5 required the licensee to perform control room envelope in-leakage testing in accordance with Positions C.1 and C.2 of Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors. These positions required the in-leakage test to be performed in the most limiting configuration for operator dose, consistent with the plant design and licensing basis. In November 2011, the licensee completed testing in accordance with Surveillance Requirement 3.7.10.5. During the test, the licensee concluded that control room envelope in-leakage was greater than the in-leakage assumed in the licensing basis accident analysis. The licensee declared the control room envelope inoperable and applied Technical Specification 3.7.10, Actions B.1, B.2, and B.3. Action B.3 allowed continued reactor operation up to 90 days provided that the licensee implemented mitigating actions per Actions B.1 and B.2 to ensure control room envelope occupant exposures would not exceed limits. PG&E conducted additional in-leakage testing using alternate system alignments. The licensee observed that unfiltered in-leakage was reduced to an acceptable value in an alternate alignment with one control room ventilation system train plus one control room ventilation system booster fan from the opposite train in operation. The licensee subsequently established mitigating actions/compensatory measures to maintain a control room ventilation system booster fan from the opposite train available and declared the control room envelope operable. The inspectors were concerned that the licensees operability determination was inconsistent with NRC inspection guidance contained in Regulatory Issue Summary 2005-20, Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety. While the results from the alternate alignment in-leakage test demonstrated that control room occupant exposures would not exceed limits, satisfying the requirements for Technical Specification Action B.3, the system alignment for this testing did not appear to satisfy either Technical Specification 5.5.19 or Surveillance Requirement 3.7.10.5. These requirements were not satisfied because the testing was did not use the most limiting configuration for operator dose. Technical Specification Surveillance Requirement 3.0.1 stated that the failure to meet a surveillance requirement, whether the failure was experienced during the performance of the surveillance or between performances of the surveillance, was a failure to meet the Technical Specification Limiting Condition for Operation. On January 3, 2011, PG&E submitted Licensee Event Report (LER) 1-2011-008-00 Diablo Canyon Power Plant - Control Room Ventilation System Design Vulnerability. The licensee reported the failure of the control room envelope as an unanalyzed condition. However, the licensee did not report the failed surveillance test as a condition prohibited by technical specifications. Title 10 CFR 50.73 required the licensee to make a 60 day report to the NRC following discovery of a condition prohibited by technical specifications. These issues are considered unresolved pending NRC review of current and past control room habitability system operability, Unresolved Item: 05000275; 323/2012003-05 Control Room Habitability Operability Issues.
05000275/FIN-2012003-06	2012Q2	Diablo Canyon	Failure to Follow Procedure Resulted in the Loss of Low Temperature Overpressure Protection System Safety Function	The inspectors identified a non-cited self-revealing violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, following the unplanned loss of the Unit 1 low temperature overpressure protection system during Mode 5 operations on June 7, 2012. One train of the low temperature overpressure protection system safety function was lost after a maintenance technician mistakenly opened the breaker providing power to the functioning train performing troubleshooting activities on the other train. The licensees corrective actions included promptly restoring power to temperature overpressure protection system and entering the condition into the corrective action program as Notification 50488636. The failure of the plant technician to follow troubleshooting work instructions was a performance deficiency. This performance deficiency was more than minor because the performance deficiency is associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors concluded that the finding is of very low safety significance (Green) because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G, Shutdown Operations Significance Determination Process. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to use human error prevention techniques, such as self- and peer-checking, commensurate with the risk of the assigned task such that work activities were performed safely.
05000275/FIN-2012003-07	2012Q2	Diablo Canyon	Entering a High Radiation Area with Dose Rates Greater than 1.0 Rem/Hour Without Knowing the Dose Rates in the Area	The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.7.2, which was the result of a worker entering a high radiation area with dose rates greater than 1 rem/hour without knowing of the dose rates in the area. In response, licensee representatives suspended fuel movement, posted the area as a locked high radiation area, documented the occurrence in the corrective action program as Notification 50478716 and evaluated the occurrence. Entering a high radiation area with dose rates greater than 1 rem/hour without knowing the dose rates in the area was a performance deficiency. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because the failure exposed workers to high dose rates. Using the occupational radiation safety significance determination process, the inspectors determined the finding to be of very low safety significance because: (1) it was not an as low as is reasonably achievable 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 finding has a cross-cutting aspect in the human performance area, resources component, because the licensee did not have adequate facilities and equipment in the form of physical or visual barriers to preclude moving fuel into the vicinity of the spent fuel pool door with the transfer canal drained.
05000275/FIN-2012003-08	2012Q2	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR 50.55a(g)4 requires in part, that ASME Code Class 1, 2, and 3 components be inspected throughout the service life of the reactor. Contrary to the above, until November 2011, the licensee failed to enter the reactor vessel supports, a Class 1 component, into the inservice inspection program and failed to perform required code inspections of accessible portions of reactor vessel supports. The licensee entered this issue into their corrective action program and performed the nondestructive examinations required by ASME Code. This finding is more than minor because if left uncorrected it would become a more significant safety concern. The failure to enter required components into the inservice inspection program and perform required inspections of safety-related components could have allowed undetected flaws to remain in service. These undetected flaws could grow in size until failure of the component, degraded system reliability, or if sufficient general corrosion occurred, a gross failure of the component could occur. The finding was of very low safety significance because the finding did not represent a loss of safety function and the nondestructive examination for the Unit 1 reactor vessel supports did not identify any relevant indications. The licensee has scheduled the examination for the Unit 2 reactor vessel supports for the next refueling outage. This issue was entered into the licensees corrective action program as Notification 50433947.
05000275/FIN-2012003-09	2012Q2	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, required that activities affecting quality be accomplished in accordance with written procedures. Diablo Canyon Power Plant Procedure OP1 L-7, Plant Stabilization Following Reactor Trip, Section 6.9.2 required that the bypass valves for the main steam isolation valves be opened to provide a drain path to the installed steam traps. Contrary to this, from April 25-27, 2012, the licensee failed to open the bypass valves while aligning the steam system following a reactor shutdown. On April 27, 2012, plant operators identified that the steam plant had not in been in the correct valve lineup. This condition could have resulted in the loss of turbine-driven auxiliary feedwater pump safety function due to accumulation of condensation in the steam supply line. This finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding or severe weather initiating event. Pacific Gas and Electric entered the issue into the corrective action program as Notification 50477779.
05000275/FIN-2012003-10	2012Q2	Diablo Canyon	Licensee-Identified Violation	Technical Specification 5.7.2 requires each entryway to an area with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or from any surface penetrated by the radiation be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and doors and gates shall remain locked except during periods of personnel or equipment entry or exit. Contrary to this requirement, on March 23, 2012, during routine walkdowns, the licensee identified the locked high radiation area door into the reactor coolant pump room 2-2 area, on the 115-foot elevation, was not secured. Although the mechanism locked, the door was ajar and opened when pulled. The licensee confirmed the locking mechanism operated properly. The condition had existed since the first week of June 2011. The licensee acknowledged dose rates in the area during operation were as high as 2 rem/hour because of the presence of nitrogen-16. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding had very low safety significance because: (1) it was not an as low as is reasonably achievable 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. The licensee documented the violation in the corrective action program as Notification 50468048.
05000275/FIN-2014004-01	2014Q3	Diablo Canyon	Failure to Document Degraded Conditions in the Corrective Action Process	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and drawings, in that the licensee did not enter degraded conditions into the corrective action process. The inspectors identified two examples. Specifically, on May 12-13, 2014, the licensee experienced high temperatures in the 480 volt vital bus rooms and did not initiate a notification to document the unexpected condition. Second, on May 20, 2014, the licensee failed to document that a 480 volt vital bus room ventilation system register louvers was found closed. The failure to enter problems into the corrective action process on the 480 volt busses 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 inspectors determined this finding has a human performance cross-cutting aspect associated with challenging the unknown attribute, specifically in that licensee personnel did not maintain a questioning attitude to resolve unexpected conditions.
05000275/FIN-2014004-02	2014Q3	Diablo Canyon	Inadequate Maintenance Procedure Resulted in Improper Configuration of Safety Related Equipment	The inspectors reviewed a Green self-revealing, non-cited violation of Technical Specification 5.4.1.a, Procedures, for failure to implement properly preplanned maintenance procedures affecting the performance of safety-related equipment. Specifically, inspectors reviewed the licensee performance associated with surveillance and maintenance activities and identified two examples of improper configuration of safety-related equipment returned to service, because of inadequate preplanned maintenance procedures. The failure to implement properly preplanned maintenance procedures affecting the performance of safety-related equipment is a performance deficiency. The inspectors determined that the finding was more than minor because it is associated with the procedure quality 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 undesired consequences. Specifically, the restriction of airflow caused by inadvertent closure of ventilation registers following the damper inspection resulted in the undesired consequences of higher ambient 480 volt switchgear room temperatures. In addition, the misconfiguration of the source range N-32 nuclear instrumentation impacted the functioning of the P-6 permissive and prevented it from performing properly during Unit 2 reactor startup such that operator action was necessary to prevent damage to the detector. 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 the finding was a deficiency affecting the design or qualification of mitigating structures, systems, and components that did not affect operability or functionality. The inspectors concluded that this finding affected the cross-cutting aspect of human performance associated with documentation, because the licensee did not ensure plant activities are governed with comprehensive maintenance procedures which are complete, accurate, and up to date to ensure work processes did not affect the performance of safety-related equipment.
05000275/FIN-2014004-03	2014Q3	Diablo Canyon	Failure to Provide Adequate Procedural Guidance Resulting in a Loss of Unit 1 230 kV Off-site Power	The inspectors reviewed a Green self-revealing finding for the licensees failure to provide appropriate acceptance criteria to ensure work activities were satisfactorily accomplished. Specifically, the licensee failed to provide acceptance criteria for torqueing or verification of acceptable torqueing during the re-assembly of the load tap changer in Work Order 64006965, Reinhausen Tap Changer Overhaul, for the re-termination of the Unit 1 startup transformer load tap changer diverter switch flex lead terminations. The licensee documented this issue in Notification 50578636. The licensee replaced the load tap changer and revised the procedure as part of their corrective actions. The licensees failure to provide appropriate acceptance criteria in Work Order 64006965 for the re-termination of the Unit 1 Startup Transformer load tap changer diverter switch flex lead terminations was a performance deficiency. Specifically, the work order did not provide acceptance criteria for torqueing or verification of acceptable torqueing during the re-assembly of the load tap changer diverter switch flex lead terminations. This performance deficiency was more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone objective and adversely affected the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609, Attachment 04, 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, it did not result in a reactor trip or a loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a human performance cross-cutting aspect associated with work management, specifically in that the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority.
05000275/FIN-2014004-04	2014Q3	Diablo Canyon	Inadequate Procedure Results in Unnecessary Main Steam Safety Valve Lift	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensee failure to prescribe a procedure appropriate to the circumstances with respect to safetyrelated atmospheric dump valves and main steam safety valves. Specifically, control of atmospheric steam dump valves was not appropriate for a rapid plant shutdown resulting in unnecessary lifting of a spring-loaded main steam safety valve. The inspectors determined that the licensees failure to ensure appropriate procedures to properly control steam generator pressure and prevent unnecessary lifting of main steam safety valves was a performance deficiency. This performance deficiency was determined to be more than minor because it affected the Mitigating Systems cornerstone attribute of procedural quality and the objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component that did not affect operability or functionality. The inspectors concluded that this finding affected the cross-cutting aspect of human performance associated with avoiding complacency, because the licensee failed to recognize during rapid load reductions the inherent risk of lifting a main steam safety valve and did not recognize or plan with adequate procedures, for a condition with a potential latent problem.
05000275/FIN-2014004-05	2014Q3	Diablo Canyon	Notice of Enforcement Discretion 14-4-001 for a Loss of Both Required Offsite Power Circuits	On August 10, 2014, at 6:56 a.m., emergency diesel generator 2-2 was removed from service for a planned maintenance outage. During the maintenance, a diesel fuel oil inlet to fuel header capscrew was discovered broken. An extent of condition review was performed and a similar capscrew was discovered to have an ultrasonic test indication on diesel generator 2-3. Diesel generator 2-3 was declared inoperable August 14, 2014, at 4:31 p.m., and DCPP Unit 2 entered Technical Specification 3.8.1, Condition E, Required Action E.1, to ensure at least two diesel generators were operable. The capscrew on diesel generator 2-3 was replaced, but during preparations to return the diesel generator to service, a separate, non-related failure of the engine driven fuel oil booster pump shaft seal occurred. As required by Technical Specification 3.8.1, Condition H, Action H.1, operators shut the unit down and placed the unit in Mode 3, Hot Standby. Technical Specification 3.8.1, Condition H, Required Action H.2 also required the unit to be in Mode 5 in 36 hours. Enforcement discretion was requested by the licensee to permit additional time to make repairs and restore diesel generator 2-3 to operable status before entry into Mode 5 within 36 hours, as required. An additional 3 hours was requested to restore diesel generator 2-3 such that the completion time of Required Action H.2 would expire at 9:31 a.m. on August 16, 2014. A notice of enforcement discretion (NOED) was granted by the NRC staff at 2:45 p.m. on August 14, 2014. The condition causing the need for this NOED was corrected by the licensee with the restoration of diesel generator 2-3 to operable status, allowing the licensee Unit 2 to exit Technical Specification 3.8.1, Required Action H.2, and the NOED on August 14, 2014, at 6:00 p.m. On August 15, 2014, emergency diesel generator 2-2 was restored to operable status at 2:21 p.m. on August 17, 2014. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific technical specifications in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine if there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item (URI) 05000275/2014004-05; Notice of Enforcement Discretion 14-4-001 for a Loss of Both Required Offsite Power Circuits.
05000275/FIN-2015004-01	2015Q4	Diablo Canyon	Failure to Properly Evaluate for Aggregate Impact of Fire Impairments	The inspectors identified a non-cited violation of Technical Specification 5.4.1.d, Procedures, for the failure to follow approved fire protection program procedures to review the fire impairments list to assess the aggregate impact on the fire protection design and safe shutdown analysis. Specifically, from August 31 to September 2, 2015, the licensee failed to evaluate the aggregate impact of having three fire doors simultaneously blocked open in adjacent Unit 1 vital battery charger rooms. The licensee implemented immediate corrective actions by assigning a continuous fire watch to the area and documented the issue in the corrective action program as Notification 50826793. The failure to follow approved fire protection program procedures to review the fire impairments list to assess the aggregate impact on the fire protection design and safe shutdown analysis was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the Initiating Events cornerstone attribute of Protection against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during plant operations. Specifically, the failure to evaluate the aggregate impact of multiple fire system impairments affected the licensee ability to limit the impact of a potential fire. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1Initial Screening and Characterization of Findings. Because the finding involved fire protection, the inspectors transitioned to IMC 0609, Appendix F Fire Protection Significance Determination Process. The inspectors characterized the finding using IMC 0609, Appendix F, Attachment 1, "Fire Protection SDP Phase 1 Worksheet," dated September 20, 2013. The finding screened as very low safety significance (Green), per Attachment 1, Question 1.4.3-A since the fire finding category was determined to be fire confinement, due to the fire doors being propped open, and the combustion loading on both sides of the door was determined to be a duration of 30 minutes as documented in licensee calculation M-824, Controlled Combustion Loading Tracking. In addition, the inspectors determined this finding had a cross-cutting aspect in human performance associated with the teamwork component because the licensees work groups did not properly communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, the work planners did not properly communicate to the fire protection department that all three fire doors would be open at the same time during battery charger load testing. (H.4)
05000275/FIN-2015004-02	2015Q4	Diablo Canyon	Failure to Identify a Cause and Implement Actions to Prevent Recurrence of a Significant Condition Adverse to Quality	The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI Corrective Action, for the failure to identify the cause and take corrective action to prevent recurrence of a significant condition adverse to quality impacting both trains of the Unit 1 safety-related residual heat removal (RHR) system. Specifically, the licensee failed to identify a definitive cause and implement corrective actions to prevent recurrent failures of the socket weld for relief valve RHR-1-RV-8708 for both trains of the RHR system. As immediate corrective actions, the licensee installed additional piping supports to mitigate the vibrations at the socket weld and documented this issue in the corrective action program as Notification 50680750. The failure to identify the cause of the RHR vibration-induced problems and to take adequate corrective actions to prevent recurrence of the weld failures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it could lead to a more significant safety concern. Specifically, no additional supports were installed and no actions were taken to reduce or eliminate the vibrations to prevent recurring weld failures, which could affect the availability of the RHR system. The lack of corrective actions to prevent recurrence could leave RHR components and other components physically connected to the system susceptible to future failures. Using Inspection Manual Chapter 0609, Appendix A, the inspectors determined the issue to have very low safety significance (Green) because the performance deficiency, which affected the mitigating systems cornerstone, did not result in a loss of safety function and did not result in an actual loss of function for greater than the technical specification allowed outage time. The licensee entered this into their corrective action program as Notification 50680750. In addition, this finding has a cross-cutting aspect in the human performance area associated with conservative bias decision making component because individuals failed to use decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee chose to only install a fatigue resistance weld rather than install additional pipe supports as were in the Unit 2 system (H.14).
05000275/FIN-2015004-03	2015Q4	Diablo Canyon	Failure to Design the Emergency Diesel Generators to operate under Worst Case Environmental Conditions	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for the failure to implement design control measures to verify the adequacy of the Unit 1 emergency diesel generators (EDGs) cooling system design to ensure operation of the EDGs under worst-case environmental conditions. Specifically, since initial licensed operations began in 1984, the licensee failed to ensure the Unit 1 EDGs were designed and built to operate under worst-case high wind and temperature conditions. As a result, sustained high winds from specific directions could have impacted EDG radiator performance resulting in the unavailability of the Unit 1 EDGs. Immediate corrective actions included issuing shift orders to the reactor operators to monitor for specific weather conditions (high air temperature, high wind speed and direction) and provide additional room cooling using established procedures, as necessary. The licensee documented the issue in the corrective action program as Notification 50599190. The failure to implement design control measures to ensure the emergency diesel generators could perform their design basis function was a performance deficiency. The performance deficiency was more than minor, and is therefore a finding, because it was associated with the design control attribute of the mitigating system cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in a condition where sustained high winds from specific directions could have impacted EDG radiator performance resulting in the unavailability of the Unit 1 EDGs. The inspectors evaluated the finding using Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, dated June 19, 2012. The inspectors determined that a detailed risk evaluation by an NRC senior reactor analyst was required since the finding was associated with a loss of EDG function. The regional senior reactor analyst performed a Phase 3 SDP analysis for the finding. The results of analysis established the incremental conditional core damage probability (ICCDP) was 2.74E-07, less than 1 x 10-6, and therefore the analyst determined that the subject finding was of very low safety significance (Green). A cross-cutting aspect was not assigned to the finding since the finding did not represent current licensee performance. The condition existed since original construction of the plant.
05000275/FIN-2018003-01	2018Q3	Diablo Canyon	Multiple Examples of Scaffolding in Place Greater Than 90 Days Without Required Evaluation	The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V, Procedures, because PG&E personnel failed to follow the requirements of AD7.ID5, Scaffold Material Structure. Specifically, 20 instances of scaffold structures installed in the plant were identified that had been in place for greater than 90 days without required 10 CFR 50.59 reviews being completed.
05000275/FIN-2018003-02	2018Q3	Diablo Canyon	4 kV Vital Switchgear Room Ventilation Degraded or Non-Conforming Condition and Associated Compensatory Measure Not Corrected in a Timely Manner	The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because PG&E personnel failed to promptly correct a degraded or non-conforming condition associated with an open operability condition. Specifically, PG&E personnel did not promptly correct a degraded condition associated with an open operability determination and corresponding compensatory measure related to Unit 1 and Unit 2, 4 kV vital switchgear ventilation for a period of over 4 years. This time period included two refueling outages for Unit 1 and three refueling outages for Unit 2.
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-2011014-03	2012Q1	Fort Calhoun	Cause of Breaker 1B3A Trip Not Identified	The team identified an unresolved item related to an apparent lack of 480 Vac electrical bus protection and coordination associated with the unexpected tripping of feeder breaker 1B3A as a result of a fire in the 1B4A switchgear. During the fire event in the 1B4A switchgear on June 7, 2011, the feeder breaker to the 1B3A switchgear tripped unexpectedly, de-energizing a redundant train of safe shutdown equipment. The licensee performed a root cause analysis of the events associated with the fire in switchgear 1B4A and originally concluded that breaker 1B3A tripped on overcurrent based on inspection of the breaker following the event; however, additional investigations could not confirm this conclusion. Six safety-related feeder breakers and six safety-related bus-tie breakers had been replaced in November, 2009 under permanent plant modification EC 33464. The modification replaced General Electric AK-50 low voltage power circuit breakers with Nuclear Logistics Incorporated/Square-D Masterpact circuit breaker/cradle assemblies and digital trip devices. The 480 Vac electrical distribution system is illustrated in Figure 1 of Attachment 4 of this report, and is comprised of nine load centers; three load centers are fed from the 4160 Vac bus 1A3 and three load centers are fed from 4160 Vac bus 1A4. There are three island buses which can be energized from either 480 Vac bus via bus-tie breakers. The 480 Vac electrical distribution system design was such that an electrical fault in the 1B4A load center should trip the normally-closed bus-tie breaker BT-1B3A, isolating the fault from the 1B3A bus. The bus-tie breakers had electronic trip settings with time-overcurrent trip values coordinated with those of the bus feeder breakers. The team reviewed Calculation EC-91-084, Breaker and Fuse Coordination Study, Revision 8, which was developed to show that adequate overcurrent protection and coordination existed on the safety-related buses. The team reviewed the time-current characteristic curves, breaker vendor materials, licensee breaker calibration data and time-voltage plots of the 4160 Vac bus voltages but was unable to confirm the licensees original conclusions that breaker 1B3A tripped on overcurrent. The licensee elevated Condition Report CR 2011-6621 to condition level A, requiring a root cause analysis to investigate the breaker 1B3A spurious trip. Condition Report CR 2011-5613 was written to document the unexpected tripping of breaker 1B3A. The licensee removed breaker 1B3A from service and on October 12, 2011 and sent it to the vendor for additional testing and analysis. The licensees analysis of breaker 1B3A had not been completed during the inspection period. Further inspection is required to determine whether performance deficiencies exist and if they are more than minor.
05000285/FIN-2012004-01	2012Q3	Fort Calhoun	Failure to report an event to the NRC within 60 days for an operation prohibited by Technical Specifications	The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a) for the failure to submit a Licensee Event Report within 60 days after the discovery of performing an operation prohibited by technical specifications. The licensee failed to report to the NRC that they moved fuel while the Spent Fuel Pool Area Charcoal Filtration System, VA-66, was not in operation, contrary to Technical Specification 2.8.3(4). The licensee discovered in September 2011 that the fuel movement in December 2009 was inappropriate based on technical specifications, but failed to submit Licensee Event Report, 2012-008-0 until July 27, 2012. This issue was entered into the licensees corrective action program and evaluated with an Apparent Cause Analysis under Condition Report, 2012-08521 and 2012-08386 The failure to make an official report to the NRC regarding an operation prohibited by the Technical Specifications is a performance deficiency. The issue was dispositioned using traditional enforcement because failing to submit the Licensee Event Report had the potential to adversely impact the NRCs ability to perform its regulatory function. The issue is characterized as a Severity Level IV violation in accordance with the NRC Enforcement Policy, Section 6.9.d.9. Since this issue was dispositioned using traditional enforcement, there is no cross-cutting aspect.
05000285/FIN-2012004-02	2012Q3	Fort Calhoun	Fuel Move with SFP Ventilation Inoperable a Condition Prohibited by Technical Specification 2.8.3(4)	The inspectors identified a non-cited violation of very low safety significance of Technical Specification 2.8.3(4), the limiting condition for refueling operations in the spent fuel pool. In December 2009, the licensee performed refueling operations with the Spent Fuel Pool Area Charcoal Filtration System, VA-66, declared inoperable. The failure to establish an operable Spent Fuel Pool Area Charcoal Filtration System, VA-66, before moving spent fuel was a performance deficiency and a violation of Technical Specification 2.8.3(4). The licensee entered this issue into the corrective action program as Condition Reports 2012-08521, 2012-0836 and Licensee Event Report 2012-008-0. The performance deficiency was determined to be more than minor because it adversely impacted the attribute of the Barrier Integrity Cornerstone objective to maintain radiological filtration functionality during operations in the spent fuel pool to protect the public from radionuclide releases caused by accidents or events. Using IMC 0609 Appendix A, Barrier Integrity Significance Determination Process, the inspectors determined this finding to be of very low safety significance (Green). Although fuel movements were contrary to the licensees technical specifications limiting condition for refueling operations, the finding represented a degradation of the radiological barrier function provided for the spent fuel pool fuel building. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not effectively incorporate internal operating experience and lessons learned from previous VA-66 ventilation system failures during spent fuel pool refueling operations and plant safety. Specifically, the licensee failed to systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience.

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