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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000389/FIN-2018003-01 | 2018Q3 | Green | H.5 | NRC identified | Failure to meet the Transient Combustible Requirements Specified by NFPA 805 | The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.48(c), National Fire Protection Standard NFPA 805, requirements. Specifically, the licensee failed to comply with transient combustible control requirements in high risk fire zones as required by NFPA 805 and implemented by licensee procedure ADM-19.03, Transient Combustible Control. |
| 05000335/FIN-2018450-01 | 2018Q1 | Green | NRC identified | Security | ||
| 05000335/FIN-2018411-01 | 2018Q1 | Severity level Enforcement Discretion | NRC identified | Security | ||
| 05000335/FIN-2018001-01 | 2018Q1 | Green | Self-revealing | Improper Evaluation of LCV-9005 position setpoints Leads to AFAS | On November 19, 2013, during reactor startup activities, feedwater bypass valves, A (LCV-9005) and B (LCV-9006), were found to be operating at different throttle positions while maintaining their respective steam generator water levels. Valves LCV-9005 and 9006 were both originally installed in April 1978. LCV-9005 was replaced in 1994, with an equivalent valve, due to obsolescence. The original valve had a full open stroke length of 1.5 inches (in.), while the new equivalent valve had a full open stroke length of 2 in. to provide the same flow as the original valve. When installed, LCV-9005 was set up to limit its stroke length to 1.5 in., matching the replaced valve, and the associated drawings were never revised to show that the new valve had a full 2 in. open stroke length. In 2009, the distributed control system (DCS) was installed utilizing these drawings and was setup under the assumption that both valves, LCV-9005 and LCV-9006, were the same model valves and stroke lengths.The DCS system was designed to provide a signal to throttle the feedwater bypass valves following a reactor trip to 20 percent open to provide approximately 5 percent feed flow in order to recover steam generator water levels utilizing main feedwater. During Unit 2 startup activities in November 2013, the licensee noted a discrepancy in the valve positions for LCV-9006 and LCV-9005 when they were providing steam generator water level control. The licensee placed the issue in the corrective action program under Action Request (AR) 1921720 and determined that it was necessary to evaluate a revision of the LCV-9005 DCS setpoint, which was accomplished by an engineering condition evaluation under AR 1925428. The engineering condition evaluation was inadequate in that it failed to recognize the differences in the two different model valves, and therefore failed to provide adequate corrective actions to address performance issues associated with these differences.The final recommendation from AR 1925428 was that the current LCV-9005 setting did not impose any risk to the plant operation, as the 2A steam generator level had been within acceptable range with no control room alarm observed. Therefore, no setpoint change was required at that point.On October 26, 2017, following a Unit 2 trip, LCV-9005 was sent a digital DCS demand signal to be 20 percent open. Since the valve was locally set to have a maximum stroke of 1.5 in. instead of 2 in. open, the actual flow through the valve was less than 5 percent. This resulted in flow lower than needed to maintain 2A steam generator level, and caused level to lower, which eventually resulted in an actuation of the A train auxiliary feedwater actuation system (AFAS). Corrective Action(s):The licensee implemented corrective actions to: 1) properly set up LCV-9005 in order for it to have a full stroke length of 2 inches so that it could provide the required feedwater flow and, 2) update associated drawings to include correct stroke lengths.Corrective Action Reference(s): This issue was entered into the licensees CAP as AR 2232869 | |
| 05000335/FIN-2017004-03 | 2017Q4 | Green | NRC identified | Failure to Identify and Correct a Condition Adverse to Quality | The NRC-identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for failure to identify and correct a condition adverse to quality. The licensee failed to identify that their procedures lacked actions to install control power jumpers that are required to defeat the reactor coolant systems (RCS) pressure interlocks for the shutdown cooling (SDC) suction line motor operated valves (MOVs) when aligning the plant for hot leg injection (HLI) and then correct the condition. Following the identification of this procedural vulnerability, the licensee fabricated control power jumpers and revised procedure 1-GME-100.03, Installation and Removal of Temporary Power Jumpers for MOV V3481, V3652, V3432 AND V3444, to provide direction for installation of power jumpers. In addition, the licensee performed a more detailed failure modes and effects analysis to ensure that the revised procedures accounted for all possible single failures. This issue has been entered into the licensees corrective action program (CAP) as CR 2217631.The PD was more than minor because it was associated with the Design Control attribute of the Mitigating System cornerstone objective of ensuring the capability of the low pressure safety injection (LPSI) system to perform its required long term cooling safety function (HLI). The condition was evaluated by a Regional Senior Reactor Analyst and determined to have very low safety significance (Green) based on the low likelihood of a loss of coolant accident (LOCA) and low likelihood of electrical failures requiring jumpers to be installed. This issue and corrective actions were documented in the licensees CAP as Action Request (AR) 2217631. This finding was not assigned a cross-cutting aspect because the underlying cause was a legacy issue and not indicative of current performance. | |
| 05000389/FIN-2017004-02 | 2017Q4 | Green | H.12 | Self-revealing | Failure to Follow Surveillance Maintenance Procedure Resulting in a Condition Prohibited by Technical Specifications | A Green, self-revealing, NCV of TS 6.8.1 was identified for the licensees failure to adequately implement a maintenance procedure during a monthly flow channel check for the 2C Auxiliary Feedwater (AFW) pump. Specifically, the licensee failed to implement as-written surveillance maintenance procedure 2-SMI-09.05C, 2C Auxiliary Feedwater Pump Flow Channel Check, when performing the channel checks for both 2C AFW pump flow transmitters. The licensees failure to follow surveillance maintenance procedure 2-SMI-09.05C, was a PD. Upon discovery, the flow transmitters were declared inoperable and subsequently, the condition was promptly restored to normal. The PD 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. The PD adversely affected the licensees ability to monitor 2C AFW flow during a design basis accident. The inspectors determined that the finding was not greater than Green because it did not represent a deficiency affecting the design or qualification of a mitigating system; it did not represent a loss of system and/or function; it did not represent an actual loss of function for at least a single train for more than its TS allowed outage time; and it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding involved the cross-cutting area of human performance, with an aspect of avoiding complacency (H.12), in that, the licensee failed to ensure that personnel effectively used human performance tools during the AFW pump flow channel check to ensure procedure steps were completed as required. |
| 05000335/FIN-2017004-01 | 2017Q4 | Green | H.1 | Self-revealing | Inadequate Reactor System Trip Process for Inoperable Channel Results in Operation in a Condition Prohibited by Technical Specifications | A Green, self-revealing NCV of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to have an adequate procedure for reducing the trip setpoint of the B channel of the reactor protection system (RPS) high startup rate (HSUR) bistable. The licensees failure to establish an adequate procedure, as required by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, to place the "B" channel wide range nuclear instrument in a tripped condition was a performance deficiency (PD). This deficiency resulted in a violation of Technical Specification (TS) Limiting Condition for Operation (LCO) 3.3.1.1. Following discovery of the condition, the licensee initiated immediate corrective actions to place the B channel RPS HSUR in trip, meeting the TS requirement. The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of procedural quality and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, there was no procedure to perform the setpoint reduction method as identified in 1-AOP-99.01. The only direction was to Contact I&C in the step. The Instrumentation and Control (I&C) processes used to implement the HSUR reduced setpoint reduction method were inadequate, in that, they did not evaluate all potential failure conditions when setting the HSUR bistable. The finding did not screen as greater than Green because while the degradation affected a single RPS trip signal, it did not affect the function of other redundant trips; and the finding did not involve control manipulations that unintentionally added positive reactivity; and finally the finding did not result in a mismanagement of reactivity by operators. Using IMC 0310, Aspects Within the Cross-Cutting Areas, the inspectors determined that the finding had a cross-cutting aspect in the area of human performance. Specifically, the cross- cutting aspect of resources (H.1) was assigned to the finding because the licensee did not ensure an adequate procedure was available to implement the HSUR setpoint reduction. |
| 05000335/FIN-2017002-01 | 2017Q2 | Severity level Enforcement Discretion | NRC identified | Reactor Coolant Pressure Boundary Leak on the 1B2 Reactor Coolant Pump Lower Seal Heat Exchanger | On January 31, 2017, Unit 1 was shutdown to investigate and repair the source of RCS leakage in the vicinity of the 1B2 RCP seal package. The unidentified leakage rate measured was 0.17 gallons per minute (gpm), which is well below the TS limit of 1 gpm of unidentified leakage. Typical RCS unidentified leak rates are in the range of 0.05 - 0.07 gpm. The licensees investigation revealed the source of the leakage as RCS pressure boundary leakage from the RCP lower seal cooler. St. Lucie Unit 1 TS 3.4.6.2, Reactor Coolant System Operational Leakage, Action a was entered and the unit was placed in cold shutdown (Mode 5, less than 200 degrees F) in accordance with the TS. The 1B2 RCP rotating assembly and pump cover with the integral lower seal heat exchanger were replaced during the fall refueling outage which occurred between September 26 and November 8, of 2016. The RCP integral lower seal heat exchanger was a tube-in-tube heat exchanger that was permanently attached to the pump cover. The inner tube contained high pressure RCS water and the outer tube contained low pressure CCW. The heat exchanger was connected to the CCW supply and return piping utilizing flanges with the flange nuts torqued to 225-230 foot-pounds (ft-lbs,) as specified by the manufacturer. The manufacturer specified a change in the torque requirements in 2015 from a previous value of 125 ft-lbs when it was identified that the 125 ft-lbs specification was not the proper torque value for the size of the flange used. The leakage emanated from a crack in the inner tube material near the toe of a weld where the inner tube exits from the outer tube. The location was in the vicinity of a CCW system connection flange. Based on a review of containment atmospheric particulate monitor data and reactor cavity leakage flow instrument data, the licensee determined that the RCS pressure boundary leak started on November 9, 2016 or shortly thereafter. This was approximately one week after the RCP was started near the conclusion of the refueling outage.The licensee determined that the most probable cause of the cracked seal cooler tubing was due to a deficiency in the lower seal heat exchanger design that allowed stresses to approach or exceed the yield strength of the tubing when the flanges were torqued to connect the CCW piping to the cooler. The resultant plastic deformation of the tubing and associated flaw formation allowed low stress; high cycle fatigue from normal RCP operation, to propagate the flaw until it was through-wall, causing the pressure boundary leakage. A finite element analysis model, developed by an outside engineering firm for the RCP seal cooler, was used to support this conclusion. The finite element analysis model determined that when the CCW flange connection was torqued to 230 ft-lbs, a tensile stress was imparted that approached or exceeded the minimum yield strength of the lower seal heat exchanger tubing and possibly caused plastic deformation and subsequently an outside diameter surface flaw in the failure region. A counter torque could not reasonably be applied during installation due to the design of the CCW flange connection.This issue was documented in the licensees corrective action program as AR 2182938. Licensee corrective actions included; 1) removing the 1B2 RCP seal cooler heat exchanger flaw and completing a weld repair of the heat exchanger outlet tubing; 2) visually inspecting all Unit 1 and Unit 2 RCP lower seal heat exchangers to identify any leakage and the presence of any outer diameter surface flaws, and; 3) determining whether a lower torque value can be used when connecting CCW to the seal cooler heat exchanger, or by implementing a different method of torqueing the CCW flanges that would reduce the stress on the tubing to an acceptable level. Enforcement: St. Lucie Unit 1 TS limiting condition for operation 3.4.6.2, Reactor Coolant System Operational Leakage, required, in part, that RCS operational leakage shall be limited to no pressure boundary leakage during plant operations in Mode 1 through 4. With any pressure boundary leakage, Unit 1 had to be placed in hot standby (Mode 3) within 6 hours, and in cold shutdown (Mode 5) within the following 30 hours. Contrary to the above, Unit 1 experienced RCS pressure boundary leakage from approximately November 9, 2016, until the unit was shut down on January 31, 2017, and later cooled down to Mode 5 on February 1, 2017. The inspectors utilized the enforcement policy examples of Section 6.1, and available ris k- informed tools to assess the safety significance of the RCS pressure boundary leakage and related violation. Based on the fact that the through-wall crack leak rate was stable, was within the capacity of the charging system, and would not impact other systems used to mitigate a loss of coolant accident, the inspectors concluded the safety significance of the violation was very low and consistent with Severity Level IV. Additionally, the risk aspects were discussed and confirmed with a regional Senior Risk Analyst. This issue was documented in the licensees corrective action program as AR 2182938.The NRC exercised enforcement discretion in Enforcement Action (EA)-2017-117, in accordance with Section 3.10 of the Enforcement Policy because the violation was not associated with a licensee performance deficiency. Specifically, the violation was not attributable to an equipment failure that was avoidable by reasonable licensee quality assurance measures or management controls and therefore inspectors concluded that there was no performance deficiency associated with the RCS boundary leakage. The RCP cover with its integrated lower seal cooler was replaced with a new component and installed in accordance with vendor instructions. This enforcement discretion will not be considered in the assessment process or the NRCs Action Matrix. This LER is closed. | |
| 05000335/FIN-2017001-01 | 2017Q1 | Green | H.12 | NRC identified | Inadequate Procedure Results in Adding an Incorrect Lubrication Oil to the 1B CS Motor Inboard Bearing | An NRC-identified Green, non-cited violation (NCV) of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for the licensees failure to establish, implement, and maintain written procedures covering activities referenced in NRC Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensees failure to maintain a plant lubrication manual with correct lubrication oil specifications for the 1B containment spray (CS) pump motor resulted in adding unacceptably low viscosity lubrication oil to the inboard bearing of the 1B CS pump motor. Immediate corrective actions included restoring the 1B CS pump inboard bearing with the correct lubrication oil and placing the issue in the licensees corrective action program.The licensees failure to correctly specify the 1B CS pump motor inboard bearing lubrication requirements in licensee general maintenance procedure GMP-22 was a performance deficiency (PD). The PD was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate procedure resulted in adding the incorrect lubrication oil to the 1B CS pump motor bearing, causing the pump to be declared inoperable for approximately 56.5 hours. The finding screened to Green because the failure did not: (1) affect the design or qualification of the systems, structures and components, (2) represent an actual loss of function, and (3) represent an actual loss of function of at least a single train for greater than its TS allowed outage time. The finding involved the cross-cutting area of human performance, in the aspect of avoid complacency, in that, the individuals involved with the procedure revision did not implement appropriate error reduction tools to ensure the procedure was appropriately changed to reflect the new lubrication oil requirement (H.12). |
| 05000335/FIN-2017403-01 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000335/FIN-2016012-01 | 2016Q4 | White | H.12 | Self-revealing | Failure to Maintain Component Configuration Control Resulted in a Complicated Reactor Trip | To Be Determined (TBD). A self-revealing finding was identified for the licensees failure to maintain configuration control of the inadvertent energization lockout relay manual synchronization circuitry as required by licensee procedures MA-AA-100 and ADM-08.12, during the October 2013 modification to the Unit 1 automatic main generator synchronization circuit. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone and it adversely affected the associated cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions because it resulted in an actual plant trip. The inspectors screened the finding under the initiating events cornerstone using Attachment 4 (October 7, 2016) and Appendix A (June 19, 2012) of Inspection Manual Chapter 0609, Significance Determination Process (April 29, 2015). The inspectors determined the finding required a detailed risk evaluation because the finding caused a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser and loss of feedwater). A preliminary significance characterization of White has been assigned. The preliminary finding involved the cross-cutting area of human performance associated with the cross-cutting aspect of avoiding complacency because the individuals involved failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk and failed to implement human error reduction tools associated with configuration control. (H.12) |
| 05000335/FIN-2016003-01 | 2016Q3 | Green | H.14 | NRC identified | Reactor Coolant System Leakage Technical Specification Violation | An NRC-identified Green non-cited violation (NCV) of Unit 1 Technical Specification 3.4.6.2 Reactor Coolant System Leakage was identified. Specifically, the licensee failed to enter TS 3.4.6.2 Action c for reactor coolant system pressure isolation valve (V3217) when the valve experienced operational seat leakage of approximately 30 gpm during flushing and cooling the shutdown cooling system. Immediate corrective actions were not required since the valve was later determined to be inoperable and repaired. The licensee entered this issue into the licensees corrective action program. The licensees failure to recognize that gross seat leakage from check valve V3217 indicated of a major problem with valve seat alignment and that higher differential pressure would not help seat the valve was a performance deficiency (PD). The performance deficiency is more than minor because it is associated with the barrier integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers such as the containment, protected the public from radionuclide releases caused by accidents or events. The PD resulted in 46 additional hours of operation with V3217 seat leakage outside of TS acceptance criteria which required the unit to be in cold shutdown. The finding involved the cross-cutting area of human performance and specifically within that area was associated with conservative bias because the operability evaluation did not demonstrate it was safe to proceed with valve V3217 experiencing gross seat leakage (H.14). |
| 05000335/FIN-2016003-02 | 2016Q3 | Green | Licensee-identified | Licensee-Identified Violation | Licensee identified violation (LIV) - T.S.6.8.1 requires written procedures be established, implemented, and maintained covering applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Rev 2, 1978. Appendix A, Section 9, Procedures for Performing Maintenance, states, in part, that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to this, Unit 1 Pressure Isolation Valve (PIV) V3217 was rebuilt in October 2013, using Licensee procedure 0-GMM-80.22, Swing Check Valve Inspections. 0-GMM-80.22 did not provide specific detail to ensure consistency and first time work quality and directly resulted in V3217 being reassembled incorrectly. Specifically the disc arm bushings were installed backwards, as well as no spacers in the bushing bores. The period of concern was from the achievement of Mode 4 on August 5, 2016 at 09:43 hours, to declaration of entry into the TS action statement and entry into Mode 5 on August 4, 2016 at 20:03 hours, resulting in 82 hours of operation with V3217 seat leakage outside of TS acceptance criteria. The inspectors characterized the safety significance of the issue utilizing Manual Chapter 0609.04, Significance Determination Process Initial Characterization of Findings, and determined the issue affected the barriers cornerstone due to leakage past an isolation valve. Manual Chapter 0609 Appendix A, The significance determination process (SDP) for Findings At-Power, Exhibit 3 was used to further evaluate this finding which screened as Green because the finding represented neither an actual open pathway in the physical integrity of the reactor containment and does not involve an actual reduction in the function of the hydrogen igniters in the reactor containment. This issue has been entered into the licensees CAP as AR 2148252. | |
| 05000335/FIN-2016007-01 | 2016Q2 | NRC identified | Intake Cooling Water Pump House Transient Combustible Fire Loading Calculation | The inspectors identified an unresolved item (URI) associated with the transient combustible heat load calculation for both Units ICW pump houses and the basis for exclusion of treated or fire retardant wood. The URI is being opened to review the licensees evaluation and determine if a performance deficiency exist. Three ICW pumps and motors are located in each house. Each pump motor is 600 horsepower. During a walkdown of both units ICW pump houses, inspectors noted that the scaffolding around the ICW pumps consisted of metal and wood planks. The inspectors determined that the wood was not included in heat load calculation for the respective pump houses. The licensee stated that the wood was treated or fire retardant and did not need to be included in the sites transient combustible heat load calculations. The inspectors questioned the licensee on the basis for not including the treated wood in the transient combustible heat load calculation. The licensee entered this issue into the CAP as 2133079 and 2134308, and initiated corrective actions to evaluate the basis for not performing a combustible heat loading calculation for fire retardant wood. The licensee also took corrective actions to replace the wood with a non-combustible material. Additional inspection time is required to review the licensees evaluation and determine if a performance deficiency exist. This issue will be tracked as URI 05000335,389 / 2016007-01, Intake Cooling Water Pump House Transient Combustible Fire Loading Calculation. | ||
| 05000335/FIN-2016010-03 | 2016Q2 | Green | NRC identified | Failure to Define, Justify, and Document Activation | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify, justify, and document an activation energy used to determine the thermal lifespan of safety related cable insulation. In response to this issue, the licensees immediate corrective actions included an immediate determination of operability, in which the licensee concluded that affected components remained operable. The licensee entered this issue in the corrective action program for resolution as AR2128756. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, using incorrect activation energies provided erroneous environmental qualification of Class 1E components, which affected the reliability of the acoustic monitor when called upon. The team used IMC 0609 Attachment 4, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the findings were a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained their operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000335/FIN-2016011-02 | 2016Q2 | Severity level IV | NRC identified | Failure to modify the Diesel Oil Storage Tank Overflow Line as Required by a Fire Protection License Requirement | Inspectors identified a Severity Level IV violation of 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, for the licensees failure to modify the Unit 2A and 2B diesel oil storage tank (DOST) overflow lines as required by a fire protection license requirement. The issue was entered into the sites corrective action program as AR 2140024. The licensees failure to notify the NRC of changes to a licensed activity that was stipulated in the fire protection license condition (Table S-1) was a performance deficiency. The inspectors determined the PD was more than minor because the licensee failed to notify the NRC that the Unit 2 DOSTs overflow lines would not be modified; and, subsequently failed to request an exemption from the requirements of NFPA 30. Traditional enforcement was applied because the PD impacted the ability of the NRC to perform its regulatory oversight function. In accordance with the NRC Enforcement Manual, Part II, Section 2.2, Actions Involving Fire Protection, the inspectors evaluated this finding to be a Severity Level IV violation. The inspectors determined that a cross-cutting aspect was not applicable because the issue was associated with a traditional enforcement violation. | |
| 05000335/FIN-2016011-01 | 2016Q2 | Severity level IV | NRC identified | Failure to Meet the Quality Requirements Specified By NFPA 805 | Inspectors identified a Severity Level IV violation of 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, for failing to maintain adequate documentation and quality of analyses. Specifically, the NRC identified multiple examples when the licensee failed to comply with site quality assurance procedures. The issue was entered into the sites corrective action program as ARs 2139768, 2139986, and 2139993. The licensees failure to maintain adequate documentation and quality of analyses to maintain configuration control, such that they could be checked for adequacy and accuracy, was a performance deficiency (PD). The inspectors determined that the issue was more than minor because the ability of the NRC to verify aspects of the licensees NFPA 805 program was impacted. The inspectors determined that the Fire Protection Significance Determination Process (IMC 0609, Appendix F) was not suitable for screening this issue. Traditional enforcement was applied because the PD impacted a regulatory oversight function. In accordance with the NRC Enforcement Manual, Part II, Section 2.2, Actions Involving Fire Protection, the inspectors evaluated this finding to be a Severity Level IV violation. A cross-cutting aspect was not applicable because the issue was associated with a traditional enforcement violation. | |
| 05000335/FIN-2016010-01 | 2016Q2 | Green | NRC identified | Failure to comply with TS requirements for CHRRMs | The inspectors identified a green non-cited violation of Technical Specification (TS) 3.3.3.1 for failing to take the required TS actions after identifying a condition adverse to quality that affected the operability of the containment high range radiation monitors (CHRRMs) (RD-26-40 and RD-26-41). The licensee declared the CHRRMs for both Unit 1 and Unit 2 inoperable and identified alternate methods for assessing emergency action levels, performing core damage assessment and dose assessment. The licensee entered these issues in the corrective action program for resolution as AR2128751 and AR2135780. The performance deficiency was determined to be more than minor because it was associated with the Emergency Response Organization Performance attribute of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was evaluated using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process. The finding is of very low safety significance (Green) because the finding affected an EAL that was rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event, but because of other EALs, an appropriate declaration could be made in a degraded manner (e.g., delayed). This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000335/FIN-2016011-03 | 2016Q2 | Green | H.5 | NRC identified | Failure to Meet the Combustible Control Requirements Specified By NFPA 805 for Work Platforms Located in the Intake Cooling Water Pump House | Inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, for the licensees failure to comply with the combustible control requirements for work platforms that were located in the Intake Cooling Water (ICW) Pump House. The issue was entered into the sites corrective action program as AR 2137088. The licensees failure to adequately implement combustible material control requirements in procedures ADM-27.11 and Procedure 0010434 was a performance deficiency (PD). The (PD) adversely impacted the Initiating Events cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during plant operations. Additionally, if left uncorrected, the deficiencies in the combustibles control program could result in wood platforms being staged in other areas of the plant. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, dated June 19, 2012, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, which determined that, an IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, review was required because it was a fire prevention finding. The finding was determined to be of very low safety significance (Green), at Step 1.4.1.B because the impact of a fire would be limited to no more than one train of equipment important to safety. The inspector identified a cross-cutting aspect in work management because the licensee failed to ensure that the sites combustible control requirements were met during the installation and use of wood platforms in the ICW pump house (H.5). |
| 05000335/FIN-2016010-02 | 2016Q2 | Green | H.6 H.6(e) | NRC identified | Failure to Implement Qualification Procedures and Methods in Accordance with IEEE 323-1974 | The inspectors identified three examples of a green non-cited violation of Title 10 Code of Federal Regulations (CFR) Part 50.49.e.(5) aging for the licensees failure to assure conformance with the qualification procedures and methods specified in IEEE 323-1974 IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations as amended by RG 1.89 Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants. In response to this issue, the licensees immediate corrective actions included an immediate determination of operability, in which the licensee concluded that that for the specific examples documented in this violation, the affected components were operable. The licensee entered these issues in the corrective action program for resolution as AR2128753, AR02128366, AR2128755, and AR2135777. The three performance deficiencies were determined to be more than minor because they were associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with time in service, significant aging degradation of SSCs increases the likelihood these SSCs could unpredictably fail when called upon to perform their designed safety function. The team used IMC 0609 Attachment 4, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the findings were a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained their operability or functionality. This finding was assigned a cross-cutting aspect of H.6 Design Margins in the Human Performance Area because the finding was indicative of current licensee performance and the licensee did not operate and maintain equipment within design margins and margins were not carefully guarded and were changed without a systematic and rigorous process (WP.2). |
| 05000389/FIN-2016001-02 | 2016Q1 | Green | H.11 | Self-revealing | Failure to Provide Detailed Work Instructions Resulted in a Unit Transient | A self-revealing finding was identified for the licensees failure to provide adequate work instructions for the circulating water system 1B1 traveling water screen drive motor replacement. Specifically, the inadequate work instructions resulted in a plant transient in order to remove the associated circulating water pump (CWP) from service. This issue was placed in the licensees corrective action program (CAP) as action request (AR) 2095560. The licensee completed the following corrective actions: (1) Counsel all maintenance supervisors in regard to having a questioning attitude and to seek guidance if unsure; (2) Rewire the 1B1 traveling screen drive motor for the proper rotation; (3) Install labels indicating the proper rotation for all eight traveling screen drive motors; (4) Submit document change requests to update the total equipment database; (5) Update all work orders (WO) for the remaining screen drive starter replacements to provide motor rotation direction and mark the post-maintenance test (PMT) step as a critical step, and; (6) Change clearance requests for traveling screen work to include directions to have electricians on station prior to returning the control switch to automatic. The failure to provide adequate work instructions for replacement of the 1B1 traveling screen motor was a performance deficiency (PD). The PD was more than minor because it was associated with the procedure quality attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, the inadequate WO instructions resulted in installing the 1B1 traveling screen drive motor incorrectly on December 4, 2015. After the maintenance, the system automatically started and the screen rotated backwards. The backward rotation allowed accumulated debris to be transported to the 1B1 debris filter system (DFS) filter and caused it to overload. The resulting high differential pressure (DP) on the DFS filter necessitated the need to lower unit power (plant transient) and required removal of the 1B1 CWP from service. The finding was determined to be of very low safety significance (Green) based on Exhibit 1, Initiating Events Screening Questions, found in IMC 0609, Significance Determination Process, Appendix A, Significance Determination Process (SDP) for Findings At-Power (June 19, 2012). This was due to the fact that the finding did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined the cause of this finding was associated with a cross-cutting aspect of ensuring risks are evaluated and managed before proceeding in the Challenge the Unknown component of the human performance area. Specifically, the licensee did not have a healthy questioning attitude and did not recognize the need to seek guidance when installing a new circulating water system traveling screen motor (H.11). |
| 05000335/FIN-2016001-04 | 2016Q1 | Green | Licensee-identified | Licensee-Identified Violation | Licensee Identified Violation (LIV) - T.S.6.8.1 requires written procedures be established, implemented, and maintained covering applicable procedures recommended in Appendix A in RG 1.33, Rev 2, 1978, section 7 c.(4) PWR Gaseous Effluent System Ventilation Air Monitoring. Specifically, procedure, 1-NOP-25.08, Unit 1 FHB Ventilation System Operation, step 4.5 provides instructions to stop (isolate) exhaust fan numbers HVE-15 & HVE-17 in order to discontinue gaseous effluent releases from the FHB when Unit 1 FHB gaseous effluent monitor (1RSC-26-4) is inoperable and 8-hour compensatory sampling has not been established as required by ODCM 3.3.3.10. Contrary to this, on October 7, 2014, with 1RSC-26-4 declared inoperable and without establishing 8-hour compensatory sampling as required by ODCM 3.3.3.10, the licensee failed to isolate FHB fans HVE-15 and HVE-17 as required by step 4.5 of -NOP-25.08, Unit 1 FHB Ventilation System Operation, and effluent releases continued via the FHB pathway for 16 hours. This violation was evaluated using the guidance in IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, and was determined to be of very low safety significance (Green) because it did not represent a substantial failure to implement the effluent release program and post-release data indicated that the release did not exceed 10 CFR 50 Appendix I dose values. | |
| 05000389/FIN-2016001-03 | 2016Q1 | Green | P.2 | Self-revealing | Inadequate Corrective Actions to Prevent Failure of the 2C ICW Pump Motor | A self-revealing, NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to implement corrective actions to prevent failure of the 2C intake cooling water (ICW) pump. The failure was a result of several air box baffle bolt-heads breaking off due to corrosion and impacting the motor stator winding, which caused an electrical ground on the winding. Corrosion of the bolts was attributed to not having functional motor heater elements. Corrective actions included repairing the motor heater elements on the 2A and 2C ICW pump motors. This issue was entered into the licensees CAP as AR 02077661. The licensees failure to implement adequate corrective actions to prevent the Unit 2C ICW pump motor winding failure that resulted from extensive corrosion of the baffle bolts was a PD and was within the licensees ability to prevent. The PD was more-than-minor because if left uncorrected, the PD has the potential to lead to a more significant safety concern. Specifically, not repairing a degraded or non-functioning motor winding heater in a timely manner prohibits protection against the humid salt water environment which the motor windings are exposed to during standby operational conditions and creates an environment for accelerated corrosion on the baffle bolts and motor winding leading to premature failure of the motor. Manual Chapter 0609 Appendix A, The Significance Determination (SDP) Process for Findings At-Power, Exhibit 2 Mitigating Systems Screening Questions. dated June 19, 2012, was used to further evaluate this finding. The finding screened as Green because the finding represented neither an actual loss of function of at least a single train for greater than its technical specification (TS) Allowed Outage Time, nor two separate safety systems out of service (OOS) for greater than its TS Allowed Outage Time. Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, was used to further evaluate the shutdown safety significance of this finding. The finding screened to Green because the inspectors answered no to all the screening questions listed under Exhibit 3 - Mitigation System Screening Questions. The finding involved the cross-cutting area of the evaluation component in problem identification and resolution (PI&R) because the organization did not thoroughly evaluate the function of the motor winding heater to ensure that resolutions address causes and extent of conditions commensurate with the long term operability of the ICW pump motors. Specifically, after identifying that the motor winding heater on the 2C ICW pump motor was not functioning, the licensee entered this issue into the CAP but did not adequately evaluate the significance of having a non-functional heater on the motor winding and instead deferred the heater repairs to be completed at the next motor overhaul which was scheduled to be performed in four years (P.2). |
| 05000335/FIN-2016008-01 | 2016Q1 | Green | P.2 | NRC identified | Failure to Consider Elevated Temperature Effects on MOV Actuator Output Capability | The NRC identified a non-cited violation of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to consider the impact of elevated ambient temperatures on motor operated valve (MOV) actuator output. The licensee entered the issue into the corrective action program and also evaluated the elevated ambient temperature effects on several affected station MOVs and determined the MOVs remained operable. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure the capability of several MOVs scoped into their MOV program because they did not consider reduced actuator output torque due to elevated temperatures. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was assigned a cross-cutting aspect of Evaluation in the Problem Identification and Resolution Area because the finding was indicative of current licensee performance, and the licensee did not thoroughly evaluate the issue identified in AR 2030822, such that the design issue of accounting for elevated temperature was resolved (P.2). |
| 05000335/FIN-2016008-02 | 2016Q1 | Severity level IV | NRC identified | Failure to Update UFSAR to Reflect Station Blackout Coping Time Basis | The NRC identified a non-cited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, for the licensees failure to update the Updated Final Safety Analysis Report (UFSAR) to reflect the offsite power design characteristic group and emergency alternating current power configuration group for station blackout coping duration. The licensee entered the issue into the corrective action program in order to update the information. The failure to update the UFSAR was dispositioned using the traditional enforcement process because it had the potential to impact the regulatory process. The team determined the violation was more than minor because not accurately classifying the offsite power design characteristic group and emergency alternating current power design characteristic group could have a material impact on licensed activities. The team determined the violation to be a Severity Level IV violation because the lack of upto- date information has not resulted in any unacceptable change to the facility or procedures. This violation was not assigned a cross-cutting aspect because crosscutting aspects are not assigned to traditional enforcement violations. | |
| 05000335/FIN-2016008-06 | 2016Q1 | Green | NRC identified | Failure to Provide a Missile-Protected Intertie | The NRC identified a finding for the licensees failure to properly provide a completely missile-protected intertie from the Unit 1 diesel oil transfer pumps to the Unit 2 diesel oil storage tanks. The licensee entered the issue into the corrective action program. The performance deficiency was determined to be more than minor because it adversely affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone objective which of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, a postulated tornado missile could fail the unprotected section of piping, rendering the intertie unable to complete its intended function, thereby reducing the licensees capability to mitigate a design basis tornado event. The team determined the finding to be of very low safety significance (Green) because it did not involve the total loss of any safety function, nor was it identified by the licensee through probabilistic risk assessment, Individual Plant Evaluation of External Events (IPEEE), or similar analysis that would have contributed to external event initiated core damage accident sequences. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000389/FIN-2016001-01 | 2016Q1 | Green | H.8 | Self-revealing | Unauthorized Entry into a High Radiation Area | A self-revealing, NCV of TS 6.12.1.b occurred when a worker entered a high radiation area (HRA) without being made knowledgeable of dose rates in the area prior to entry. Specifically, on November 10, 2015, a worker performing a plant surveillance under radiation work permit (RWP) 15-004, Clearance Tags, Surveillances and Inspections, climbed into overhead in the Unit 2 Pipe Penetration room and received an electronic dosimeter (ED) dose rate alarm. The licensee entered this issue into the CAP as AR 02090225 and took immediate corrective actions which included restricting the operators access to the radiological control area (RCA), performing followup surveys and convening a human performance review board to examine causal factors for the purpose of determining corrective actions. This PD was determined to be more than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of Human Performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Workers permitted entry into HRAs with inadequate knowledge of current radiological conditions could receive unintended occupational exposures. The finding was evaluated using the Occupational Radiation Safety SDP. The finding was not related to as low as reasonably achievable (ALARA) planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. Therefore, the inspectors determined the finding to be of very low safety significance (Green). The inspectors noted that the operator responded properly to the ED dose rate alarm thereby limiting his potential for unintended exposure. This finding involved the cross cutting aspect of (H8) procedure adherence because the individual understood the RWP requirements but failed to comply with them. |
| 05000335/FIN-2016008-04 | 2016Q1 | Green | NRC identified | Failure to Identify Degraded Condition of Unit 1 Electrical Equipment Room Supply Fan Gravity Dampers | The NRC identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify a condition adverse to quality, which prevented the Unit 1 electrical equipment room (EER) supply fan dampers from performing their safety-related function to close. The licensee entered the issue into their corrective action program and implemented compensatory measures to prevent reverse flow of air through the degraded dampers in the event of a failure of their supply fan. This compensatory measure will remain in place until the licensee is able to replace both gravity dampers. This performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inability of the gravity dampers to close upon failure of one of the supply fans would result in room temperatures above the design temperature of 104oF. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000335/FIN-2016403-01 | 2016Q1 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000335/FIN-2016008-05 | 2016Q1 | Green | NRC identified | Failure to Verify the Adequacy of Design of Unit 1 Electrical Equipment Room Ventilation System | The NRC identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of the Unit 1 electrical equipment room (EER) ventilation system design when performing a design calculation. The licensee entered the issue into the corrective action program and plans to re-balance flow rates in the EERs or revise the equipment qualification temperatures for equipment located in the EERs. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the re-analysis of the ventilation system resulted in a reduction in temperature margin, which could impact the reliability and capability of emergency electrical equipment in the EERs. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000335/FIN-2016008-03 | 2016Q1 | Green | NRC identified | Inadequate Testing of 125VDC MCCBs | The NRC identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to perform testing for safety-related 125 volts direct current (VDC) molded case circuit breakers (MCCBs) to detect deterioration. The licensee entered the issue into the corrective action program and plans to make changes to the procedure to ensure deterioration of the safety-related 125VDC MCCBs is adequately detected. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, cycling the breakers multiple times before electro-mechanical testing could mask degradation of the circuit breakers and thus decrease the reliability of the breakers to perform their safety function when called upon. The team determined the finding to be of very low safety significance (Green), because it was not a deficiency affecting the design or qualification of a structure, system, or component which did not maintain its functionality; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for greater than its Technical Specification (TS) allowed outage time or two separate safety systems out-of-service for greater than its TS allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. | |
| 05000335/FIN-2015004-07 | 2015Q4 | Green | Licensee-identified | Licensee-Identified Violation | Contrary to TS 6.8.1, Procedures and Programs, the licensee failed to implement the written procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978, regarding PM of safety-related equipment. Specifically, Regulatory Guide 1.33 Appendix A Section 9.b states, in part, that system parts that have a specific lifetime should be replaced. The licensee implements this guidance in Regulatory Guide 1.33 by following the PM program, ER-AA-204, Preventive Maintenance Program Strategy, Revision 5, which details how PM should be developed and implemented for safety-related equipment. Section 3.2.9 of this procedure states, in part, that to ensure inclusion of vendor technical information, vendor maintenance recommendations should be included in PM bases and frequency requirements. The ESI-EMD owners group recommends a 10-year life for EDG speed switches based on electrolytic capacitor life expectancy. However, there is no evidence that the licensee considered vendor recommendations regarding the periodicity of EDG speed switch replacement when implementing its PM on the EDG. As a result, the existing PM for the speed switches was inadequate and led to the 1A EDG being rendered inoperable when the speed switch failed to function properly during manual local start of the EDG. This violation was associated with the Mitigating Systems Cornerstone and was determined to be of very low safety significance (Green) in accordance with Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, because the finding did not result in a loss of system function or represent an actual loss of function of at least a single train for greater than its TS allowed outage time. The licensee entered this violation into its CAP as AR 2053060. | |
| 05000335/FIN-2015004-01 | 2015Q4 | Green | H.8 | NRC identified | NRC Biennial Written Examinations Did Not Meet Qualitative Standards | An NRC-identified finding related to 10 CFR 55.59, Requalification, was identified based on a determination that greater than 20 percent of the 2014 biennial written exam question sampled for review were flawed. The finding did not involve a violation of NRC requirements. The inspectors determined that the finding was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding adversely affected the quality and level of difficulty of biennial written examinations, which potentially impacted the facilitys ability to appropriately evaluate licensed operators. The risk importance of this issue was evaluated using IMC 0609, Appendix l, Licensed Operator Requalification Significance Determination Process (SDP). The qualitative standards used by the inspectors were defined in TR-AA-220-1004, Licensed Operator Continuing Training Annual Operating and Biennial Written Exams. Because more than 20 percent, but less than 40 percent, of the questions reviewed were flawed, Blocks 4 and 5 of Appendix I characterized the finding as having very low safety significance (Green). A review of the cross-cutting aspects was performed and no associated cross-cutting aspect was identified. |
| 05000335/FIN-2015004-02 | 2015Q4 | Green | NRC identified | Non-willful Compromise of a Remedial Examination Required by 10 CFR 55.59 Affected the Equitable and Consistent Administration of the Exam | An NRC-identified severity level IV (SLIV) NCV of 10 CFR 55.49, Integrity of examinations and tests was identified based on a determination that a non-willful compromise of a remedial examination required by 10 CFR 55.59 affected the equitable and consistent administration of the examination. An associated finding of very low safety significance (Green) was also identified based on a determination that a biennial written remedial examination was not prepared and approved in accordance with licensee procedures. The licensees failure to develop and administer a remedial examination in accordance with TR-AA-220-1004, Licensed Operator Continuing Training Annual Operating and Biennial Written Exams, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency caused an incident of exam compromise that affected the equitable and consistent administration of the exam and resulted in a licensed operator being authorized to resume licensed duties prior to the condition being corrected. Additionally, the finding adversely affected the integrity of a biennial written remedial examination, which impacted the facilitys ability to appropriately evaluate a licensed operator. The licensed operator subsequently passed another remedial examination that was one hundred percent different from his original exam and the previous remedial exam. The operator also demonstrated satisfactory performance while performing licensed operator duties and participating in the licensed operator requalification program. The traditional enforcement violation was evaluated using the NRC Enforcement Policy dated January 28, 2013, and revised February 4, 2015. The inspectors determined the violation was SLIV per Section 6.1.d.2 because the associated finding was evaluated by the SDP as having very low safety significance (i.e., Green). The finding was directly related to the cross-cutting aspect of procedure adherence of the cross-cutting area of Human Performance because the training staff did not follow applicable guidance for the preparation and approval of licensed operator biennial written remedial examinations. | |
| 05000335/FIN-2015010-01 | 2015Q4 | NRC identified | Implementation of Commitments and Aging Management Programs | The inspectors identified a URI associated with the implementation status of various commitments and AMPs. Description: The inspectors identified that there were pending actions for various regulatory commitments/AMPs as a result of commitment changes implemented by the licensee after the renewed operating license was issued. The licensee informed the NRC of such changes, and submitted correspondence to the NRC for review and approval. At the time of this inspection, the NRC was still in the process of reviewing the licensees submittals. While the licensee met its commitment to submit the proposed changes to the NRC prior to the PEO, the inspectors were unable to determine whether the licensees implementation of the affected AMPs was consistent with the staffs final position, which will be provided through the issuance of SERs. The affected commitment items, and their respective pending actions, are summarized below. Commitment 1, Condensate Storage Tank Cross-Connect Buried Piping Inspection On May 12, 2015, the licensee informed the NRC of a commitment change based on the as-found configuration of the cross-tie line after excavation. On September 1, 2015, the NRC issued a Request for Additional information, for which the licensee provided responses in letter L2015-258, dated October 6, 2015. At the time of this inspection the NRC was reviewing the licensees response to the Request for Additional Information, and no SER had yet been issued. Commitments 4 and 5, Reactor Vessel Internals Inspection Program As described in the inspection scope section of this report, the licensee submitted several letters to the NRC after the renewed operating license was issued describing the proposed program to manage the aging effects of the reactor vessel internals. At the time of this inspection, the NRC was reviewing the licensees submittals and no final SER had yet been issued. Commitment 6, Small Bore Class 1 Piping Inspection Program On May 11, 2015, the licensee submitted a revision to the previously approved Small Bore Class 1 Piping Inspection Program for NRC review and approval. The revision was related to the use of destructive examinations in lieu of volumetric examinations. At the time of this inspection, the NRC was reviewing the licensees submittal and no final SER had yet been issued. Commitment 20, Environmentally-Assisted Fatigue of the Pressurizer Surge Line On October 29, 2015, the licensee submitted its proposed program for managing environmentally-assisted fatigue of the pressurizer surge line to the NRC. The inspectors noted that the proposal detailed the licensees intent to utilize the ASME BPVC, Section XI ISI Program (UFSAR Section 18.2.2) to manage the recurring inspections, and the associated evaluations for any flaws noted. At the time of this inspection, the NRC was reviewing the licensees submittal and no SER had yet been issued. In addition to the commitment changes under NRC review, the inspectors identified a followup item for Commitment 17, Reactor Vessel Integrity Program. The inspectors noted that the licensee credited fleet procedure ER-AA-110 to meet the regulatory commitment associated with the integration of all four reactor vessel integrity subprograms into a single program document. Fleet procedure ER-AA-110 requires a plant-specific procedure be developed for each site describing the important parameters needed to meet the regulatory requirements specific to that station. The inspectors noted that the plant-specific procedure for Unit 1, procedure ADM 17.38, was still under development with a target completion date of March 1, 2016. Therefore, the inspectors concluded that there still were pending actions associated with the development of the site-specific program, and additional inspection was required to verify that the Reactor Vessel Integrity Program was implemented as intended. The licensee initiated AR 02094578 to enter this item in the CAP. The inspectors determined that it was necessary to open a URI to further review the implementation of the commitments/AMPs, and verify that the commitments were met as approved by the NRC in the final SERs. This issue requires followup inspection, and will be tracked as URI 05000335/2015010-001, Implementation of Commitments and Aging Management Programs. | ||
| 05000335/FIN-2015004-03 | 2015Q4 | Green | P.3 | NRC identified | Inadequate Corrective Actions to Prevent Fouling of the CCW HXs | An NRC-identified NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to implement corrective actions to prevent fouling of the 2B component cooling water (CCW) heat exchanger (HX) that resulted in the number of blocked tubes exceeding the HXs maximum analyzed limit for plugged tubes. The licensees failure to implement adequate corrective actions was a performance deficiency and was within the licensees ability to prevent. Corrective actions included installing temporary equipment to ensure adequate continuous sodium hypochlorite (SH) is injected through the CCW HXs to prevent biological fouling. The licensee entered this issue into the CAP. The performance deficiency was more-than-minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, inadequate SH injection may cause extensive fouling and can lead to a common mode failure of the CCW HXs preventing the required cooling of safety-related structures, systems, and components (SSCs) analyzed heat loads during a design basis accident (DBA). Using Manual Chapter 0609.04, Significance Determination Process Initial Characterization of Findings, Table 2 dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 Mitigating Systems Screening Questions, dated, June 19, 2012, was used to further evaluate this finding. The finding screened as Green because the finding did not represent either an actual loss of function of at least a single train for greater than its Technical Specification (TS) Allowed Outage Time, or two separate safety systems out-of-service (OOS) for greater than its TS Allowed Outage Time. The finding involved the cross-cutting area of the resolution component in Problem Identification and Resolution (PI&R) because the organization did not take effective corrective actions to address issues in a timely manner commensurate with the safety significance of the CCW HX, in that, even after the repeat fouling issue had been identified on the 2B CCW HX, the immediate resolution of inadequate SH injection remained unresolved until the inspectors addressed this issue with plant management. |
| 05000335/FIN-2015004-04 | 2015Q4 | Green | P.3 | NRC identified | Procedural Non-compliances Relating to Installed Scaffold Located Near Safety-related SSCs | An NRC-identified NCV of TS 6.8.1, Procedures and Programs, was identified for the licensees failure to properly implement written procedures covering activities referenced in NRC Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensee routinely failed to complete engineering evaluations to determine the acceptability of scaffolds that did not meet the 2 inch clearance requirement of NextEra Nuclear Fleet Administrative Procedure MA-AA-100-1002, Scaffold Installation, Modification, and Removal Requests. The licensees failure to erect scaffold in compliance with the NextEra Nuclear Fleet Administrative Procedure was a performance deficiency. This issue has been entered into the licensees CAP. The performance deficiency was more-than-minor because it was associated with the Mitigating Systems Cornerstone Attribute of Protection against External Factors, Seismic, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, routinely failing to complete engineering evaluations of scaffold clearance issues could lead to the continued use of inadequately installed scaffolds, ultimately posing a risk of rendering safety-related equipment inoperable during normal and adverse conditions, such as a design basis seismic event. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, Table 2, Cornerstones Affected by Degraded Condition or Programmatic Weakness, dated June 19, 2012, the inspectors determined the finding affected the Mitigating Systems Cornerstone. Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, was used to further evaluate this finding. The finding screened as Green because no was answered to all four screening questions, i.e. the finding did not represent an actual loss of function of any piece of plant equipment for any amount of time. The finding involved the cross-cutting area of PI&R in the aspect of resolution, in that the organization did not take effective corrective actions to address the scaffolding issues in a timely manner, as evidenced by a period of five months in which the inspectors continued to identify non-conformances with erected scaffold. |
| 05000335/FIN-2015004-05 | 2015Q4 | Green | NRC identified | Failure to Verify the Adequacy of the Unit 1 and Unit 2 Steam Generator Tube-to-Tubesheet Welds Design | An NRC-identified, Non-cited Violation of 10 CFR Appendix B, Criterion III, Design Control, was identified for the failure to verify the adequacy of the Unit 1 and Unit 2 replacement steam generators (RSGs) design with respect to the requirements in the American Society of Mechanical Engineers Boiler Pressure Vessel Code (ASME Code), Section III, Article NB-3000, for the primary stress and fatigue analyses of the pressure-retaining tube-to-tubesheet welds. The licensee entered the issue in the corrective action program, and performed the required analyses for the Unit 1 and Unit 2 RSGs to demonstrate that the design met the ASME Code requirements. The inspectors used the guidance in NRC Inspector Manual Chapter (IMC) 0612, Appendix B, Issue Screening, and determined that the performance deficiency was more-than-minor because it was associated with the design control attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective. Specifically, the failure to verify that the required stress and fatigue analyses were performed in accordance with the ASME Code did not support the objective of limiting the likelihood of primary-to-secondary leakage events that could upset plant stability and challenge critical safety functions during shutdown, as well as power operations. The inspectors evaluated this finding using NRC IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, Exhibit 1 Initiating Events Screening Questions. The finding screened as Green because the stress calculations demonstrated that there was no degraded steam generator (SG) tube condition where one tube could not sustain three times the differential pressure across a tube during normal full power, and none of the SGs violated the accident leakage performance criterion. Additionally, the stress calculations demonstrated that the finding did not result in a condition that exceeded the reactor coolant system leak rate for a small loss of coolant accident (LOCA), or affected other systems used to mitigate a LOCA resulting in a total loss of their function (e.g., Interfacing System LOCA). The inspectors determined that no cross-cutting aspect was associated with this finding because the performance deficiency occurred more than 3 years ago, and it was not reflective of present performance. | |
| 05000335/FIN-2015004-06 | 2015Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR55.49, Integrity of examinations and tests, states, Applicants, licensees, and facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. This includes activities related to the preparation and certification of license applications and all activities related to the preparation, administration, and grading of the tests and examinations required by this part. Contrary to the above, on August 18, 2015, the licensee identified that two licensed operators were administered a 2014 biennial requalification comprehensive written examination that contained five repeat questions from other versions of the biennial written examination that the individuals had either prepared or approved. The inspectors determined that the violation was not greater than very low safety significance (Green) because the licensed operators were not actively performing licensed duties in the control room. This issue was entered in the licensees corrective action program as CR 02067887. | |
| 05000335/FIN-2015405-01 | 2015Q3 | Green | H.8 | NRC identified | Security | |
| 05000389/FIN-2015003-03 | 2015Q3 | Severity level IV | NRC identified | Untimely 10 CFR50.72 Notification | The NRC identified an NCV of 10 CFR 50.72(b)(3)(iv)(A) for the licensees failure to notify the NRC within 8 hours of an event that was not part of a preplanned sequence which resulted in a valid actuation of an emergency AC electrical power system. During Unit 2s refueling outage with Unit 2 in Mode 5 and the 2A emergency diesel generator (EDG) properly tagged out of service for pre-planned maintenance, a phase-to-phase fault on the 6.9kV non-segregated bus from the 2A startup transformer (SUT) to the non-safety related 2A1 bus caused the 1A and 2A SUTs supply breakers to open. The safety related 4.16kV 2A3 bus experienced an under voltage condition which generated a valid actuation signal for the 2A EDG. The licensee failed to recognize this event as reportable pursuant to 10 CFR 50.72(b)(3)(iv)(A). The licensee generated corrective actions (AR 2075703) which included restoring compliance within a reasonable period of time after the violation was identified, and training the appropriate personnel to understand why the situation was reportable pursuant to 10 CFR 50.72. The inspectors determined that the failure to report required plant events or conditions to the NRC had the potential to impede or impact the regulatory process. As a result, the NRC dispositioned this violation of 10 CFR 50.72 using the traditional enforcement process instead of the SDP. The inspectors determined that this issue was more than minor because it is similar to a Severity Level IV example provided in Section 6.9 of the NRC Enforcement Policy. Cross-cutting aspects are not assigned to traditional enforcement violations. | |
| 05000335/FIN-2015003-04 | 2015Q3 | Green | H.12 | Self-revealing | Failure to Follow Reactor Protection System Surveillance Procedure Resulting in Reactor Plant Trip | A Green, self-revealing, NCV of TS 6.8.1 was identified for the licensees failure to adequately implement surveillance procedures during reactor protection system (RPS) testing. Specifically, the licensee failed to implement as-written operations surveillance procedure 1-OSP-63.01, RPS Logic Matrix Test, when operators failed to close two trip circuit breakers (TCBs) prior to proceeding to the next section of the procedure. This resulted in an unplanned automatic reactor trip when a second pair of TCBs were opened. Corrective actions completed for this event included a human performance review that was conducted by the shift manager, operations director and plant general manager, initially implementing around the clock management oversite, and revising the RPS logic matrix test procedure to change it from a reader/doer procedure to a procedure with more concurrent verification steps. The licensee entered this issue into their corrective action program as AR 2065821. The licensees failure to follow procedure 1-OSP-63.01, RPS Logic Matrix Test, as written is a performance deficiency. This performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone and it adversely affected the associated cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions and resulted in an actual plant trip. The inspectors evaluated the risk of this finding using IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined that the finding was of very low safety significance because it did not result in both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The finding involved the cross-cutting area of human performance, with an aspect of avoiding complacency (H.12), in that the licensee failed to ensure that personnel effectively used human performance tools during the logic matrix test to ensure procedure steps were completed as required. |
| 05000335/FIN-2015003-02 | 2015Q3 | NRC identified | Partial Loss of Unit 1 and Unit 2 Offsite Power Due to Unit 2 6.9 kV Non-Segregated Bus Fault | The inspectors identified an unresolved item associated with the partial loss of offsite power as a result of a fault on the 2A1 6.9 kV non-safety bus. On September 17, 2015, a fault of the 2A1 6.9 kV bus connected to the 2A SUT resulted in the loss of power to both the 1A and 2A SUTs. 1A SUT was impacted since it shared a common power supply from the switchyard with the 2A SUT. The 2A1 6.9kV bus is of a bus bar design. The bus is made up of flat copper bars that are bolted together with all three phases contained in a metal enclosure. The phases are supported within the enclosure and insulated from each other using ceramic insulator plates that maintain the spacing between the phases and with the enclosure. Each bar is insulated between the bolted connections with Noryl insulation. Rubber insulating boots cover the bolted connections. The licensees inspection of the 6.9 kV bus determined that the fault occurred at a location where the bus transitions from a vertical to a horizontal orientation. The three insulating boots for this bolted transition were found lying on top of the ceramic insulators between the phases below in the vertical run. The boots had a coating of dust and corrosion products that had flaked off the enclosure. At the close of this inspection period, the licensees root cause evaluation and complete inspection of the 2A1 6.9 kV bus was in progress. The licensee entered this issue in the CAP as AR 2074774. This is an unresolved item pending review of the licensees root cause evaluation to determine whether or not a performance deficiency exists. The NRC will track this issue as an URI. | ||
| 05000335/FIN-2015003-05 | 2015Q3 | Severity level Enforcement Discretion | NRC identified | Unit 2 Shutdown Due to Through Wall Crack and Leak in the 2B2 Safety Injection Tank Discharge Pipe | On March 30, 2015 the operators reviewed the Unit 2 control room logs and identified increased leakage from the 2B2 SIT. On April 11, 2015, the 2B2 SIT was declared inoperable due to a through wall leak identified on the 12-inch diameter Class 2 piping of the discharge header. The licensee determined that the pipe failed due to a legacy support design from construction, which led to higher levels of stress in the supports weld. The licensee concluded through metallurgical analysis that the pipe flaw propagated through wall due to high cycle fatigue. Prior to the through-wall leak being identified, there were no indications that a flaw existed within the pipe support weld. Additionally, there were no examinations required to be performed on the support that would have recognized a flaw within the support weld. As a result, the inspectors concluded that there was no performance deficiency associated with the pipe failure. The inspectors utilized available risk-informed tools to assess the safety significance of the 2B2 SIT inoperability. Based on the fact that the through-wall leak did not preclude the 2B2 SIT from performing its design basis function while inoperable, the inspectors concluded this event was of very low safety significance. St. Lucie Unit 2 TS limiting condition for operation 3.5.1, Safety Injection Tanks (SIT), requires each RCS safety injection tank to be operable in plant operating Mode 1 through Mode 3. With one SIT inoperable, the inoperable SIT must be returned to operable status within 24 hours or Unit 2 placed in hot standby within the next six hours and hot shutdown within the following six hours. Contrary to the above, Unit 2 operated for approximately 13 days from March 30, 2015 to April 12, 2015, with the 2B2 SIT inoperable due to a through-wall leak identified on the 12-inch diameter class 2 piping of the discharge header. Although a violation of the TS occurred, the violation was not attributable to an equipment failure that was avoidable by reasonable licensee quality assurance measures or management controls. Therefore, the TS 3.5.1 violation was not associated with a licensee performance deficiency. The inspectors concluded that the violation would normally be characterized as a Severity Level IV violation based on its very low safety significance. The NRC exercised enforcement discretion in Enforcement Action (EA)-14-047, in accordance with Section 2.2.4.d and 3.5 of the Enforcement Policy because the violation was not associated with a licensee performance deficiency; therefore, it will not be considered in the assessment process or the NRCs Action Matrix. This issue was documented in the licensees corrective action program as AR 2039830. Licensee corrective actions included: replacing the leaking pipe spool piece with the through wall flaw (line I-12-SI-459), modifying the supports for line SI- 459, removing support SI-4203-44, revising procedure STD-C-010, Piping and Support Analysis Requirements St. Lucie Units 1 and 2, to include more detail related to weld attachments to specifically address avoiding extended lugs which develop a bending movement, and incorporating considerations associated with using weld attachments in an environment which involves cyclic loading. This LER is closed. | |
| 05000335/FIN-2015003-01 | 2015Q3 | Green | H.9 | NRC identified | Unsecured Utility Cart With An Unrestrained Operating Pedestal Fan Near Safety-related ECCS Equipment | An NRC-identified, NCV of Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for the licensees failure to implement written procedures covering activities referenced in NRC Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensee failed to follow procedural requirements to properly secure a pedestal fan positioned on a wheeled cart to the extent required to prevent a potential for adverse interaction with safety-related systems, structures or components (SSCs) during a design basis seismic event. Failure to control equipment located near safety-related SSCs to prevent the equipment from interacting with safety-related SSCs during a design basis seismic event was a performance deficiency. Immediate corrective actions included removing the cart and fan assembly from the area and entering this issue into the corrective action program. The performance deficiency was more than minor because the issue was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factors (seismic) and affected the cornerstone objective of ensuring the availability, reliability, and capability of safety-related SSCs to respond to initiating events to prevent undesirable consequences. Specifically, during a design basis seismic event the unsecured cart and unrestrained fan could have damaged the emergency core cooling system low and high pressure safety injection flowrate transmitters causing control room operators to have a loss of safety injection flowrate indication and a small amount of system leakage during accident mitigation. Using Manual Chapter 0609.04, Significance Determination Process Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the Mitigating Systems Cornerstone. Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 - Mitigating Systems Screening Questions dated, June 19, 2012, was used to further evaluate this finding. The finding screened as Green because the inspectors answered No to all four screening questions. The finding involved the cross-cutting aspect in the area of human performance associated with training because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values to ensure temporarily placed equipment located near safety-related SSCs was adequately secured to prevent interaction during a seismic event. |
| 05000335/FIN-2015002-04 | 2015Q2 | Green | Licensee-identified | Licensee-Identified Violation | The St. Lucie Unit 1 Technical Specification 6.8.1(a) states, in part, that the licensee shall establish, implement, and maintain the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Rev. 2, 1978. Section 9(a) of Appendix A to Regulatory Guide 1.33, Rev.2, states, in part, that maintenance that can affect the quality of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above requirements, on April 12, 2015, the licensee did not implement adequate maintenance instructions that were appropriate to the circumstances as specified by WO 40296976 to ensure that the 1C AFW pump was correctly aligned and returned to service. Specifically, the work order instructions required Attachment 10 of procedure 1-PMM-09.04, Auxiliary Feedwater Turbine Mechanical and Electrical Over speed Trip Tests, to be completed as part of the pump restoration. Attachment 10 of procedure 1-PMM-09.04 included a step to position valve V08385 to the open position, and this step was not completed. The licensee entered this issue into the CAP as AR 02042311. The failure to adequately implement the work instructions in WO 40296976 requiring completion of Attachment 10 of procedure 1-PMM-09.04, to ensure the valve was correctly aligned was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance (Green) because the inspector answered No to all of the associated Mitigating Systems screening questions within IMC 0609, Attachment 4, Initial Characterization of Findings. Because this violation was of very low safety significance and was entered CAP, this violation is being treated as a NCV, consistent with Section 2.3.2 o f the NRC Enforcement Policy. | |
| 05000335/FIN-2015002-01 | 2015Q2 | Green | P.5 | NRC identified | Failure to Assess Potential Gaseous Effluents Released from Containment Equipment Hatch Openings during a Loss of Negative Pressure | The inspectors identified a Green non-cited violation of Technical Specification 6.8.1 for the failure to implement procedures for the monitoring, evaluating, and reporting of gaseous effluents in accordance with the methodology in the Off-Site Dose Calculation Manual. Specifically, there was no program in place to assess potential effluent releases from containment equipment hatch openings during periods when negative pressure was lost. The licensee took immediate corrective actions including placement of a low-volume air sampler near the Unit 1 Reactor Containment Building equipment hatch, and entered the issue into their corrective action program as AR 02037629. The performance deficiency was more than minor because it was associated with the Public Radiation Safety cornerstone attribute of Programs and Processes and adversely affects the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The finding was assessed using the Public Radiation Safety Significance Determination Process. Based on the fact that routine (i.e. nonaccident) effluents released from an equipment hatch are unlikely to contribute significantly to public dose, this finding does not represent a substantial failure to implement the effluent program and was determined to be of very low safety significance (Green). This finding has a crosscutting aspect of Operating Experience (P.5) because the licensee failed to recognize the applicability of regulatory issues experienced by other plants regarding equipment hatch monitoring. |
| 05000389/FIN-2015002-05 | 2015Q2 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 6.12.1 requires an area with dose rates greater than 100 millirem per hour (mrem/hr) at 30 centimeters (cm) to be barricaded and conspicuously posted as an HRA. Contrary to this, on January 31, 2015, the licensee identified dose rates in excess of 100 mrem/hr at 30 cm on a five gallon bucket containing drain hoses in a Radiation Area within the U2 Pipe Tunnel, which was not barricaded or posted as a High Radiation Area (HRA). A survey of the bucket identified dose rates of up to 120 mrem/hr at 30 cm. Immediate corrective actions included relocating the bucket to a locked location in a designated HRA. This condition was documented in AR 02022248. This violation was evaluated using the guidance in IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and was determined to be of very low safety significance (Green) because there was no substantial potential for overexposure and the licensees ability to assess dose was not compromised. Because this violation was of very low safety significance and was entered CAP, this violation is being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. | |
| 05000389/FIN-2015002-03 | 2015Q2 | Severity level IV | NRC identified | Problem with LER Reporting | The NRC identified multiple non-cited violations of regulatory requirements that it has decided to group into an example of a problem associated with the licensees reporting program. This problem includes violations of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to address all the applicable reporting criteria and 10 CFR 50.9, Completeness and Accuracy of Information, for the licensees failure to submit complete and accurate information to the Commission, as part of Licensee Event Report (LER) 050000389/2014-001 dated September 22, 2014 . These violations were material to the NRC because the failure to include the appropriate reporting criteria and provide complete and accurate information had the potential to impede or impact the regulatory process and, therefore, is subject to traditional enforcement as described in the NRC Enforcement Policy. The inspectors used the examples provided in Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, of the NRC Enforcement Policy, and concluded that this problem was appropriately categorized as Severity Level (SL) IV. The licensee placed these issues into their corrective action program as AR 02021204 and has submitted a revised LER. | |
| 05000389/FIN-2015002-02 | 2015Q2 | Green | P.2 | NRC identified | Failure to Comply with Technical Specification 3.0.3 | The NRC identified a non-cited violation of Technical Specification (TS) 3.0.3 for the licensees failure to take the required actions to shut down the plant in a timely manner. The licensees failure to perform an adequate operability evaluation in accordance with the requirements of EN-AA-203-1001, Operability Determinations / Functional Assessments, was a performance deficiency. Specifically, the licensee failed to identify in an Immediate Operability Determination that through-wall leakage on the ASME Class 1 pipe riser for vent valve V3811 rendered both Emergency Core Coolig Systems (ECCS) subsystems inoperable, requiring entry into TS LCO 3.0.3 and performance of the applicable action statements. The licensee entered this into their corrective action program as AR 02021204. The performance deficiency was more than minor because it was associated with the equipment reliability attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding was associated with the mitigating systems cornerstone and required a detailed risk evaluation because the finding represented a loss of function on the high pressure safety injection system. A detailed risk evaluation determined the significance of the finding was Green. The inspectors determined the finding was related to the crosscutting aspect of Evaluation (P.2) of the Problem Identification and Resolution area because the licensees failure to thoroughly evaluate the issue commensurate with its safety significance led to the licensee failing to perform an appropriate operability evaluation. |