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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000306/FIN-2018003-04 | 2018Q3 | Prairie Island | Failure to Promptly Identify and Correct 21 125 VDC Battery Lid Conditions Adverse to Quality | The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, as of February 15, 2018, for the licensees failure to promptly identify and correct conditions adverse to quality associated with the 21 125 VDC battery system. |
| 05000282/FIN-2018003-01 | 2018Q3 | Prairie Island | Failure to Repair a D2 EDG Jacket Water Leak per the Leak Management Process | The inspectors identified a finding of very low safety significance (Green) as of July 18, 2018, for the licensees failure to repair a D2 EDG jacket water leak per the Leak Management Process. |
| 05000456/FIN-2018003-01 | 2018Q3 | Braidwood | Inadequate Detail in Maintenance Procedure for Emergency Diesel Generator 2-Year Inspection Contributed to 1A Emergency Diesel GeneratorFuel Rack Binding | A self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to include adequate detail within their maintenance procedures to enable proper lubrication of the emergency diesel generator (EDG) fuel rack control linkage. Specifically, the preventative maintenance template for the fuel rack control linkage required that the manual fuel trip lever and associated linkage be lubricated every 2 years. However, the licensees implementing 2year maintenance procedure failed to include specific instructions to disassemble the lever assembly for lubrication. This lack of lubrication contributed to the mechanical binding of the emergency diesel generator fuel rack and failure of the 1A EDG during surveillance testing on April 22,2018. |
| 05000255/FIN-2018003-01 | 2018Q3 | Palisades | Wire Not Landed on Safety Injection Initiation Relay Circuit | The inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018 |
| 05000282/FIN-2018003-02 | 2018Q3 | Prairie Island | Failure to Maintain a Preventative Maintenance Strategy for 12 and 22 Cooling Water Pump Diesel Engines | The inspectors identified a finding of very low safety significance (Green) and associated NCV of Prairie Island Technical Specification 5.4.1, Procedures, as of August 9, 2018, for the licensees failure to maintain a preventative maintenance strategy for sacrificial zinc anode plugs on the jacket water system for the 12 and 22 cooling water pump diesel engines (DDCLPs). |
| 05000456/FIN-2018003-02 | 2018Q3 | Braidwood | Minor Violation | All Braidwood Station EDG governors were replaced during the late 1990s. During design testing, the licensee noted that the historical EDG frequency response had changed slightly due to installation of new electronic governors. Prior to these governor replacements, EDG frequency was always above 57 hertz (Hz) during load sequencing. However, with the newly installed electronic governors, 1A and 2A EDG frequency was observed to dip below the 57 Hz under frequency relay setpoint following start of the 1A and 2A motor-driven AF pumps. (Note that because the 1B and 2B AF pumps are diesel-driven, there is no corresponding impact on the 1B or 2B EDGs.) As a result, an external 2-second time delay, provided by an Agastat time delay relay, was incorporated into the under frequency trip logic for the 1A and 2A EDGs to provide an additional margin for frequency recovery following motor-driven AF pump load starts. The Braidwood governor modification was installed in 1998, with the external time delay added to the 1A and 2A EDGs as part of the design changes to prevent inadvertent actuations of the under frequency logic.During the licensees investigation into the issue discussed in the subject LER, it was identified that the external Agastat time delay was installed incorrectly on the 1A EDG. Specifically, the original trip logic wiring had not been properly removed, which permitted the actuation of the under frequency trip after the original 0.5 second internal time delay through the bypassing of the additional 2.0 second external time delay. The wiring error was introduced during the original modification installation in October 1998. Screening: The inspectors determined that the error was of minor safety significance. Absent the mechanical binding of the manual fuel trip lever and associated linkage, as discussed in NCV 05000456/201800301 in this report, the 1A EDG had performed reliably and satisfactorily during surveillance testing prior to the Unit 1 refueling outage testing in April of 2018. Additionally, the inspectors determined that the error, having occurred some 20 years ago, was not indicative of current licensee performance.Violation: This failure to comply with the requirements of 10 CFR Part 50, Appendix B, Criterion III , Design Control, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000282/FIN-2018003-03 | 2018Q3 | Prairie Island | Failure to Promptly Identify Degradation of the 122 DDCLP FOST Vent Piping | The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, as of November 28, 2017, for the licensees failure to promptly identify a condition adverse to quality associated with 122 DDCLP FOST vent piping. |
| 05000341/FIN-2018002-03 | 2018Q2 | Fermi | Failure to Adequately Evaluate the Operability of Emergency Diesel Generator11 | A finding of very low safety significance was self-revealed for the licensees failure to adequately evaluate the operability of a condition adverse to quality identified on Emergency Diesel Generator (EDG) 11. Specifically, a lube oil leak was evaluated as having no impact to the operation of the emergency diesel generator. However, during the next surveillance run of EDG 11, the engine had to be shut down and declared inoperable due to the lube oil leak degrading during operation. |
| 05000341/FIN-2018002-02 | 2018Q2 | Fermi | Inadequate Preventative Maintenance in Residual Heat Removal Service Water System Outlet Flow Control Valves Results in Lower Bonnet (Backseat) Bushing Failure | A self-revealed Green finding and associated non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) Part 50, Appendix Criterion V, Instructions, Procedures, and Drawings were identified for failure to ensure activities affecting quality were prescribed in a manner consistent with the circumstances to the residual heat removal service water system(RHRSW). Specifically, preventative maintenance procedure M681 failed to establish an appropriate interval and guidance for periodic valve internals inspections on the Division 2 RHRSW system outlet flow control valve to prevent significant degradation from galvanic corrosion given known internal and external operating experience |
| 05000341/FIN-2018002-04 | 2018Q2 | Fermi | Failure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal Service Water Outlet Flow Control Valve | A self-revealed TBD finding and an associated apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and Technical Specification 3.7.1 Residual Heat Removal Service Water (RHRSW) System, were identified for failure to identify a condition adverse to quality while performing corrective maintenance on Division 2 RHRSW outlet flow control valve E1150F068B prior to returning the Division 2 RHRSW system to service. Specifically, troubleshooting and associated post maintenance testing failed to identify and correct a failed anti-rotation key which resulted in an inoperable Division 2 RHRSW system for longer than its Technical Specification 3.7.1 allowed outage time. |
| 05000341/FIN-2018002-01 | 2018Q2 | Fermi | Failure to Document a Condition Assessment Resolution Document for Reactor Recirculation Motor-Generator Set A Brush Gear Sparking | A self-revealed Green finding was identified for failure to document a Condition Assessment Resolution Document (CARD) for 5-inch rooster tail sparking on reactor recirculation motor-generator set A brush gear, which ultimately resulted in a manual recirculation pump A trip and plant transient. |
| 05000263/FIN-2017004-02 | 2017Q4 | Monticello | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix R, Section III.G.2.a, Fire Protection of Safe Shutdown Capability. Specifically, the licensee identified that the structural steel located in the plant administrative building (PAB) basement supporting the cable spreading room (CSR) floor did not have a 3hour fire rating as required by 10 CFR 50, Appendix R, Section III.G.2.a. Title 10 CFR 50, Appendix R, Section III.G.2.a, requires, in part, that where separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3hour rating is provided, structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier. Contrary to the above, since 1982, the licensee failed to protect the structural steel supporting the fire barrier between the cable spreading room and fire area IV. This failure was identified by the licensee on August 4, 2016 during an Appendix R self-assessment and addressed in CAP 1530637. The licensee issued LER 201600201 in response to this Appendix R non-compliance and implemented the immediate corrective action (compensatory measure) of an hourly fire watch in the affected fire area. The licensee conducted an exposed steel fire simulation and evaluation to understand the performance of the unprotected steel in the event of a fire in the PAB. The inspectors reviewed the licensees simulation and evaluation. This finding was determined to be more-than minor because the performance deficiency was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the structural steel supports the fire barrier between the CSR and the PAB basement and a failure to protect the structural steel from fire damage would degrade the fire barrier separating the CSR and PAB. This could result in a fire in the PAB spreading to the CSR due to the degraded fire barrier resulting in an evacuation of the control room. The only means for operators to shutdown the reactor using the ASDS panel would require travel through the PAB fire area where a fire event is occurring. Therefore, this finding impacted the safe shutdown capability of the plant. After review of the licensees exposed steel fire simulation and evaluation, the finding was determined to be of very low safety significance (Green) because the licensee demonstrated that the unprotected structural steel would provide at least one hour of fire endurance rating under a fire event in the PAB. |
| 05000455/FIN-2017004-01 | 2017Q4 | Byron | Fire Barrier Impaired without Authorization | A finding of very low safety significance and an associated NCV of Technical Specification 5.4.1.c, Procedures, was self-revealed when an Operations department supervisor identified that a fire door separating two rooms containing safety-related equipment was impaired and did not meet the requirements specified in fire protection program procedures. Specifically, on October 5, 2017, a fire door was left unattended and unable to latch due to the presence of tape over the door latch assembly. The supervisor promptly removed the tape to restore the fire doors functionality and documented the as-found condition in IR 04059911, Fire Door 0DSD474 Improperly Impaired Tape Over Latch. This issue was determined to be of more than minor significance because it was associated with the Initiating Events Cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as having very low safety significance (Green) using IMC 0609, Appendix F, Fire Protection Significance Determination Process, Question 1.4.3A, since the fire finding category was determined to be Fire Containment, due to the door not being able to latch, and the combustion loading on both sides of the door was determined to result in less than the 1.5 hour threshold. The finding affected the cross-cutting area of Human Performance in the aspect of Avoiding Complacency (H.12) because the individual that impaired the door did not recognize the inherent risk in their actions and use error reduction tools to mitigate that risk. |
| 05000263/FIN-2017004-01 | 2017Q4 | Monticello | Failure to Maintain Radiation Exposure ALARA | A finding of very low safety significance (Green) was self-revealed due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees radiological work planning and work control program. Specifically, the licensee failed to properly incorporate ALARA strategies, insights while planning, and executing work activities during the 1R28 refueling outage. The Reactor Water Cleanup (RWCU) Inlet Outboard Isolation Valve MO2398 was scheduled for replacement during the outage. The initial dose estimate for this activity was 4.5 person-rem. However, 13.776 actual person-rem of dose was received. This issue was caused by poor radiological planning and work execution of this task. The licensee entered this issue into their Corrective Action Program (CAP) item 1558234. The finding was more than minor because it was associated with the program and process attribute of the Occupation Radiation Safety Cornerstone. Additionally, this issue affected the cornerstone objective of ensuring the adequate protection of the workers health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, the finding is very similar to IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, Example 6.i. This example provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance (Green) because Monticello Nuclear Generating Plants current 3year rolling average collective is 64.637 person-rem (20142016). This is less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. This finding had a cross-cutting aspect in the area of Human Performance, related to the cross-cutting aspect of Work Management, in that the outage plan did not adequately plan, control and execute work activities to ensure the RWCU Inlet Outboard Isolation Valve MO2398 replacement remained ALARA. (H.5) |
| 05000255/FIN-2017003-04 | 2017Q3 | Palisades | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and an associated NCV o f TS 5.7.2, which requires, in part, that each entryway into High Radiation Areas ( HRAs) with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation source shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry. Contrary to the above, on May 4, 2017, the licensee failed to lock or continuously guard an entryway into a HRA with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation source. Specifically, an entryway was left unguarded when the individual assigned to guard the entryway left the area prior to another guard being stationed. This issue was identified by a radiation protection technician who immediately stationed another guard. This issue was entered into the licensees CAP as CR PL 2017 02160. The failure to continuously guard the HRA entryway was a performance deficiency that was within the licensees ability to foresee and should have been prevented. The performance deficiency was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring the adequate protect ion of worker health and safety from exposure to radiation. The finding was determined to be of very low safety significance (Green) because it did not involve as -low -as-reasonably -achievable planning or work controls, there was no overexposure or substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. |
| 05000255/FIN-2017003-03 | 2017Q3 | Palisades | 12 Diesel Generator Trip During Maintenance Resulting in Additional Unavailability of the 12 DG | A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, was self -revealed on March 31, 2017, when the 12 Diesel Generator ( DG ) tripped during performance of monthly TS surveillance procedure MO 7A 2, Emergency Diesel Generator 1 2. Specifically, during conduct of the monthly surveillance procedure, restoration activities associated with maintenance of breaker 152 213, 1 2 DG to Bus 1D, were being performed. When maintenance personnel closed the trip cutouts for the Z -phase of the 1 2 DG differential overcurrent relay, an unbalanced current flow into the differential relay resulted in relay actuation. This actuation resulted in a trip of the output breaker and subsequently the 1 2 DG. The trip caused a delay in the TS surveillance activities and resulted in the extended unavailability and inoperability of the 1 2 DG. The licensee entered this issue into their corrective action program (CAP) as condition report (CR) CR PLP 2017 01291. Corrective actions included retesting the 1 2 DG and updating the work instructions associated with the differential overcurrent relays to include caution statements that opening or closing trip cutouts for the relays while the output breaker s from the DGs to the associated buses were closed could cause the differential relay s to actuate and trip the DG . The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating System s cornerstone attribute of Procedure 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. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, since the inspectors answered No to all screening questions. The finding had a cross- cutting aspect in the area of Human Performance, in the Work Management aspect , for the licensees failure to identify and manage risk commensurate to the work (H.5). |
| 05000255/FIN-2017003-02 | 2017Q3 | Palisades | Cause of 422/RPS Breaker Failure to Open | Introduction: The inspectors identified an URI associated with the failure mechanism of the 42 -2/RPS control rod clutch breaker failure to open. Specifically, at the end of the inspection period the licensee was working to understand the cause of the breaker failure and determine the actions required to address the failure mechanism. Description : On May 17, 2017, the licensee conducted a shutdown to complete emergent repairs to a leaking seal identified on control rod drive mechanism 40. In accordance with GOP 8, Power Reduction and Plant Shutdown to Mode 2 or Mode 3 525 F, the operators depressed the reactor trip pushbutton from the EC 06, reactor protection system panel. When the pushbutton was depressed, the reactor did not trip as expected. The operators successfully tripped the reactor using the reactor trip pushbutton on the EC 02, primary process and reactor controls console. The licensee identified that the 42 1/RPS breaker tripped as expected when the reactor trip pushbutton on the EC 06 panel was depressed, however, the 42 2/RPS breaker did not trip as expected. This resulted in the reactor trip not occurring as expected when the reactor trip pushbutton on the EC 06 panel was depressed as both breakers a re required to open to result in a reactor trip. The licensee performed troubleshooting activities to determine the cause of the 42 2/RPS breaker failure. The direct cause of the breaker failure was found to be the 42 2/RPS breaker undervoltage release mechanism failing to provide enough downward force to fully depress the trip plunger. This resulted in a physical failure of the breaker to open. At the end of the inspection period, the cause of this physical failure mode was unknown. The licensees equipment failure evaluation identified that it could be age- related degradation or a physical degradation of the breaker. As a corrective action, a failure analysis of the breaker was planned. Once the failure analysis i s complete, the licensee plans to re-assess the failure mechanism and determine any additional corrective actions that are required to address the issue. This item is considered unresolved, pending the inspectors review of the failure analysis and any changes made to the equipment failure evaluation, to determine if this issue constitutes a performance deficiency and/or violation of NRC requirements. (URI 05000255/2017003 02, Cause of 42 2/Reactor Protection System Breaker Failure to Open) |
| 05000255/FIN-2017003-01 | 2017Q3 | Palisades | Left Train Emergency Diesel Generator Load Sequencer Failure | Introduction: The inspectors identified an Unresolved Item ( URI ) associated with the failure of the left train emergency DG load sequencer to run its program. Since this sequencer is required for left train DG operability, this condition resulted in an unanticipated entry into a TS shutdown action statement. The cause of this failure is currently unknown, pending the results of a vendor evaluation of a failed load sequencer component. Description : On August 3, 2017, the control room received alarm EK 1145, Sequencer Trouble, unexpectedly. The operators identified that the indication lights were not lit on the left channel load sequencer, MC -34L101; declared the associated DG inoperable; and entered the appropriate TS action statement. The failed sequencer was removed and replaced with a new module that was satisfactorily post -maintenance tested and the left train EDG was subsequently declared operable on August 4, 2017. The failed sequencer was sent to an on -site lab for further troubleshooting. No obvious visual signs of failure were identified and the electrolytic capacitors in the module all tested satisfactorily. The module was then bench tested using a test program, which identified that although it would power up, no program would run. The licensee completed an equipment failure evaluation to review the bench test data, along with information collected in the failure modes analysis, and determined that the direct cause of the failure was a memory fault within the sequencer module that caused the sequencer to lock -up and not run its program. A fault in the memory module, memory processing interface circuitry, or the executive module could have caused the sequencer to lock up. At the end of the inspection period, further examination by t he vendor was required and in progress to determine the exact initiating point of the fault. In addition to replacing the failed sequencer, the licensees immediate corrective actions included inspecting the right train load sequencer and completing the quarterly surveillance test to ensure proper operation; the results of which were satisfactory. A plant operating experience review was conducted and did not identify any prior memory failures on the load sequencers. Once the vendors evaluation is complete, the licensee plans to re-assess the failure mechanism and any additional corrective actions required. This item is considered unresolved, pending the inspectors review of the vendor analysis and any changes made to the equipment failure evaluation, to determine if this issue constitutes a performance deficiency and/or violation of NRC requirements. (URI 05000255/2017003 01, Left Train Emergency Diesel Generator Load Sequencer Failure ) |
| 05000454/FIN-2017002-01 | 2017Q2 | Byron | Failure to Verify Computer Software during a Transformer Replacement Modification | Green . A finding of very low safety significance was self -revealed on March 28, 2017, when operators rapidly reduced generator load in response to a loss of forced cooling for the newly installed Unit 1 East main power transformer ( 1E MPT ) and an indicated rapid rise in transformer winding hotspot temperature caused by vendor data entry errors in the monitoring system software . The process detailed in CC -AA- 256- 101, Software Quality Assurance Process for Plant Digital Instrumentation and Control Systems and Components, to verify and validate the software/firmware during updates was not implemented after the vendor made changes to the digital software during the modification process. The issue was entered into the licensees corrective action program (CAP) and corrective actions included replacement of the cooling group supply breaker, correction of the software errors, and revision of the alarm response procedure and supporting documentation. The inspectors concluded that the issue was more than minor because it adversely impacted the Design Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during plant operations. Specifically, rapid power changes or load reject could challenge operating safety limits. In this event, the rapid rise in the calculated winding hotspot indications and subsequent operator actions to rapidly reduce load over 300 megawatts electric ( MWe ) was the result of two software errors : (1) an incorrect Current Turns (CT) Ratio and (2) the incorrect configuration of the MPT cooling groups in series within the software. The inspectors utilized Exhibit 1, Initiating Events Screening Questions of IMC 0609, Significance Determination Process, Appendix A, dated June 19, 2012, to conclude that the finding was Green, or of very low safety significance, because the event did not cause a reactor trip and the event did not affect any mitigation equipment. A cross -cutting aspect in the Challenge the Unknown element of the Human Performance Are a (IMC 0310 H.11) was assigned because the engineering group based the risk evaluation on the vendor input that the scope of the change was limited. The flawed assumption that the vendor input was correct without verification resulted in a failure to manage the risk prior to implementation through the verification/validation of the software/firmware. |
| 05000255/FIN-2017002-01 | 2017Q2 | Palisades | Inadequate Protection from Tornado Missiles Identified Due to Non- Conforming Design Conditions | A finding and an associated violation of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified based upon the lack of adequate tornado missile protection to the safety -related equipment listed above. The finding was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado- generated missile non -compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15 002, Revision 1, Enforcement Discretion for Tornado- Generated Missile Protection N on- Compliance, and can be found in ADAMS Accession No. ML16355A286. Because this finding and violation was identified during the discretionary period covered by Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado Missile Protection Non-Compliance and because the licensee, prior to the expiration of the associated LCO, took initial compensatory measures that provided additional protection such that the likelihood of tonado missile effects were lessoned, followed by more comprehensive compensatory measures that w ere completed within approximately 60 days of issue discovery , and has final corrective actions planned, the NRC is exercising enforcement discretion by not issuing an enforcement action, as discussed in Section 1R15.2 of this report. |
| 05000454/FIN-2017001-01 | 2017Q1 | Byron | Licensee-Identified Violation | On March 11, 2017 , with Unit 1 shutdown and in a refueling outage, pipefitters as signed to cut out and replace service water valve 1WS413 discovered that piping was blocked upstream of the valve and the work scope was appropriately changed to remove the blocked piping. Taking action they believed was allowed by the work instructions, the pipefitters opened a pipe union and removed the pipe. They then set the removed section containing valve 1WS023C on a nearby tripod to continue work. A system engineer performing a walkdown in the area identified that the removed valve had a clearance (danger) tag on it and immediately stopped work and contacted the operations department. Technical Specification 5.4.1 requires , in part , that written procedures be established, implemented and maintained covering the procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. One administrative procedure recommended in Appendix A is , Equipment Control ( e.g. locking and tagging). OP AA 109 101, Clearance and Tagging, accomplished the locking and tagging requirement for Byron Station. Section 5.2, Danger Tags, established standards for implementation of the tagging process. Step 5.2.2 stated , A component with a Danger Tag attached to it shall not be physically removed from the system. Contrary to the requirements stated above, a component with a danger tag attached was physically removed from the system on March 11, 2017. Specifically, pipefitters disconnected a pipe union and removed associated service water piping from the system that contained valve 1WS023C which had a clearance (danger) tag attached. The licensee immediately verified that the cooler the piping served was out -of-service on both the supply and return sides with a clearance boundary in place and drained so that the workers were not exposed to a pressurized sourc e. The workers immediately acknowledged their error stating they did not see the tag because they were focused on the demolition activities. The issue was entered into the licensees CAP as IR 03984215 , and the maintenance organization conducted a stand down to reinforce the station standards for compliance with the clearance procedure. The inspectors determined that this issue was more than minor because the performance deficiency adversely impacted the Configuration Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown operations. The inspectors determined the issue was of very low safety significance , or Green by answering No to all screening questions in IMC 0609, Appendix G, Shutdown Operations Significant Determination Process, Exhibit 2, Initiating Events Screening Questions. |
| 05000341/FIN-2017001-03 | 2017Q1 | Fermi | Licensee-Identified Violation | TS 3.3.5.1, ECCS Instrumentation, states the ECCS instrumentation for each function in Table 3.3.5.1 1 shall be operable. As specified in Table 3.3.5.1 1, Function 3b, HPCI System High Drywell Pressure (4 channels) and Function 3f, HPCI System Manual Initiation (1 channel) are required to be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. TS 3.3.5.1, Required Action A.1 states with one or more channel(s) inoperable, immediately enter the condition referenced in Table 3.3.5.1 1 for the channel. Table 3.3. 5.1 1, Function 3b, references Condition B for inoperable HPCI System High Drywell Pressure channels. Required Action B.2 states declare the HPCI system inoperable within 1 hour from discovery of loss of HPCI initiation capability and Required Action B.3 states place the affected channel(s) in trip within 24 hours. Table 3.3.5.1 1, Function 3f, references Condition C for an inoperable HPCI System Manual Initiation channel. Required Action C.2 states restore the channel to operable status within 24 hours . If the required actions and associated completion 42 times of Condition B or C are not met, Required Action G.1 states immediately declare the associated supported feature (i.e., HPCI system) inoperable. TS 3.5.1, ECCS Operating, states, in part, each ECCS injection subsystem shall be operable in Modes 1, 2, and 3, except HPCI is not required to be operable with reactor steam dome pressure less than or equal to 150 psig. With the HPCI system inoperable, Required Action E.1 states immediately verify by administrative means RCIC system is operable and Required Action E.2 states restore HPCI system to operable status in 14 days. If the required actions and associated completion times of Condition E are not met, Required Action I.1 states be in Mode 3 in 12 hours. LCO 3.0.4.b is not applicable to HPCI. TS 3.3.5.2, RCIC System Instrumentation, states the RCIC instrumentation for each function in Table 3.3.5.2 1 shall be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. As specified in Table 3.3.5.2 1, Function 4, RCIC System Manual Initiation (one channel per valve) is required to be operable. TS 3.3.5.2, Condition A states with one or more channels inoperable, immediately enter the condit ion referenced in Table 3.3.5. 21 for the channel. Table 3.3.5.2 1, Function 4, references Condition C for an inoperable RCIC System Manual Initiation channel. Required Action C.1 states restore the channel to operable status within 24 hours. If the required actions and associated c ompletion times of Condition C are not met, Required Action E.1 states immediately declare the RCIC system inoperable. TS 3.5.3, RCIC System, states the RCIC system shall be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. With the RCIC system inoperable, Required Action A.1 states immediately verify by administrative means HPCI system is operable and Required Action A.2 states restore RCIC system to operable status in 14 days. If the required actions and associated completion times of Condition A are not met, Required Action B.1 states be in Mode 3 in 12 hours. LCO 3.0.4.b is not applicable to RCIC. TS 3.0.4, Limiting Condition for Operation (LCO) Applicability, Paragraph (a) states, in part, when a LCO is not met , entry into an operational mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation in the operational mode or other specified condition in the applicability for an unli mited period of time. This specification shall not prevent changes in modes or other specified conditions in the applicability that are part of a shutdown of the unit. Contrary to the above: 1. On six occasions (February 10, 2014, April 16, 2014, March 19, 2015, September 13, 2015, May 3, 2016, and November 7, 2016 ), the licensee entered Mode 3 following plant shutdowns without declaring the HPCI system instrumentation functions of high drywell pressur e and manual initiation inoperable and entering LCO 3.3.5.1. During the shutdowns, Fermi 2 was in Mode 3 for up to fifteen hours with reactor steam dome pressure greater than 43 150 psig without the licensee satisfying TS 3.3.5.1, Required Actions A.1, B.2, and G.1. This is a violation of TS 3.3.5.1. With HPCI inoperable as specified by TS 3.3.5.1, Required Actions B.2 and G.1, the licensee did not satisfy TS 3.5.1, Required Action E.1. This is a violation of TS 3.5.1. 2. On six occasions (February 10, 2014, A pril 16, 2014, March 19, 2015, September 13, 2015, May 3, 2016, and November 7, 2016 ), the licensee entered Mode 3 following plant shutdowns without declaring the RCIC system instrumentation function of manual initiation inoperable and entering LCO 3.3.5.2 . During the shutdowns, Fermi 2 was in Mode 3 for up to fifteen hours with reactor steam dome pressure greater than 150 psig without the licensee satisfying TS 3.3.5.2, Required Action A.1. This is a violation of TS 3.3.5.2. 3. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without declaring the HPCI system instrumentation functions of high drywell pressure and manual initiation inoperable and entering LCO 3.3.5.1. For up to nineteen hours during this time, the licensee did not satisfy TS 3.3.5.1, Required Actions A.1, B.2, and G.1. This is a violation of TS 3.3.5.1. With HPCI in operable as specified by TS 3.3.5.1, Required Actions B.2 and G.1, the licensee did not satisfy TS 3.5.1, Required Action E.1. This is a violation of TS 3.5.1. 4. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without declaring the RCIC system instrumentation function of manual initiation inoperable and entering LCO 3.3.5.2. For up to nineteen hours during this time, the licensee did not satisfy TS 3.3.5.2, Required Action A.1. This is a violation of TS 3.3.5.2. 5. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without meeting the LCOs of TS 3.3.5.1 and TS 3.3.5.2 for HPCI and RCIC systems instrumentation functions of high drywell pressure (HPCI only) and m anual initiation ( both HPCI and RCIC) . This is a violation of TS 3.0.4. This violation was entered into the licensees corrective action program as CARD 16 26153. The violation was determined to be of very low safety significance (Green) during a detailed Significance Determination Process review since the CDF was determined to be less than 1.0E -7/year. |
| 05000341/FIN-2017001-02 | 2017Q1 | Fermi | Failure to Maintain Adequate SLC Storage Tank Boron Concentration | Green . A finding of very low safety significance with an associated Non- Cited Violation of TS 3.1.7, Standby Liquid Control (SLC) System, was self -revealed when the licensee measured the boron concentration in the SLC storage tank and discovered the concentration was below the minimum requirement of 8.5 percent. Specifically, the licensee failed to adequately monitor and identify a decreasing trend in SLC storage tank sodium pentaborate concentration concurrent with known dilution of the SLC storage tank during pump and valve testing. The licensee entered this violation into its corrective action program for evaluation and identifi cation of appropriate corrective actions and restored the SLC sodium pentaborate concentration to within TS limits. The finding was of more than minor safety significance 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, a lower than allowable sodium pentaborate concentration affected the SLC systems ability to shut down the reactor during a design basis event. The finding was determined to be a licensee performance deficiency of very low safety significance during a detailed Significance Determination Process review since the delta core damage frequency ( CDF ) was determined to be less than 1.0E 6/year. The inspectors concluded this finding affected the cross -cutting area of human performance and the cross -cutting aspect of resources. Specifically, the licensee failed to ensure equipment and procedures were adequate to support nuclear safety . Th is issue would have been avoided if the system monitoring plan was trending tank level via a pressure indicator . Also, chemistry had no administrative limits in their procedure to add boron prior to the minimum TS limit was reached and the system engineer was not a reviewer on the routine surveillance procedure and was not trending the concentration as a backup. (IMC 0310, H.1 ) |
| 05000341/FIN-2017001-01 | 2017Q1 | Fermi | Inadequate Work Instructions for Maintenance on EDG 14 | Green . A finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations ( 10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self -revealed when plant operators discovered a thick white smoke plume coming from the emergency diesel generator (EDG) 14 engine exhaust manifold during surveillance testing. Consequently, operators shut down the engine and removed it from service. The licensee failed to have work instructions for maintenance on the safety -related EDG appropriate to ensure insulation blankets on the engines exhaust manifold were replaced with insulation blankets conforming to the approved engineering design. The licensee entered this violation into its corrective action program for evaluation and identification of appropriate corrective actions. The licensee replaced the insulation blankets with insulation blank ets conforming to the approved engineering design. The finding was of more than minor safety significance because it was related to the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective o f ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operators shutdown the engine after discovering a thick white smoke plume coming from the engines exhaust manifold , which resulted in unplanned inoperability and unavailability of this onsite emergency power source. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time nor did it represent a loss of function of a non -TS train designated as high safety significant in accordance with the licensees Maintenance Rule Program . The inspectors concluded this finding affected the cross -cutting area of human performance and the cross- cutting aspect of documentation. Plant activities are governed by comprehensive, high -quality, programs, processes and procedures. Design documentation, procedures, and work pack ages are complete, thorough, accurate, and current. In this case, the licensees process for implementing and maintaining engineering configuration control of the newly designed EDG exhaust manifold insulation blankets was inadequate because 3 it did not follow the licensees formal engineering configuration management process. (IMC 0310, H.7) |
| 05000341/FIN-2016004-05 | 2016Q4 | Fermi | Failure to Lock an Area Meeting Locked High Radiation Area Conditions | A finding of very low safety significance with an associated NCV of TS 5.7.2, High Radiation Area, was self-revealed when a locked high radiation area (LHRA) was found to be unlocked. The licensee immediately locked the LHRA and performed follow-up surveys. Subsequent actions included providing additional training for radiation protection technicians. This issue was entered into the licensees CAP as CARD 16-28186. The inspectors determined the performance deficiency was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, not locking LHRAs could lead to inadvertent worker entry into high dose rate areas without knowledge of the radiological conditions. The finding was determined to be of very low safety significance because it did not involve as-low-as-reasonably-achievable planning for work controls, there was no overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors determined the finding affected the cross-cutting area of human performance and the cross-cutting aspect of avoid complacency because individuals did not plan for the possibility of mistakes and implement appropriate error reduction tools. Specifically, the radiation protection technician did not ensure a lock verification was performed on the padlock as required by station procedures (IMC 0310, H.12). |
| 05000341/FIN-2016004-03 | 2016Q4 | Fermi | Failure to Satisfy 10 CFR 50.73 Reporting Requirements for Loss of LOP Instrumentation and AC Electrical Power Safety Functions | The inspectors identified a Severity Level IV NCV of the NRCs reporting requirements in 10 CFR 50.73(a)(1), Licensee Event Report System. The licensee failed to submit a required Licensee Event Report (LER) within 60 days after discovery on September 16, 2016, of an operation or condition which was prohibited by the plants TSs and an event or condition that could have prevented the fulfillment of the safety function to remove residual heat and mitigate the consequences of an accident. The inspectors concluded the licensee failed to satisfy the applicable regulatory reporting requirements due to unwarranted delay in evaluating conditions from the event with respect to compliance with the TSs and reporting requirements. The licensee subsequently submitted LER 05000341/2016-009-00, Emergency Diesel Generator Inoperable Due to Open Circuit on Loss of Power Instrumentation, on December 20, 2016, to report the event. The licensee entered this issue into its CAP as CARD 16-30164. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined the performance deficiency was of minor significance based on No answers to the more-than-minor screening questions. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding. |
| 05000341/FIN-2016004-01 | 2016Q4 | Fermi | Inadequate Work Instructions for Maintenance on Flexible Couplings for EDGs | A finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when plant operators discovered an oil leak coming from a flexible coupling upstream of the emergency diesel generator (EDG) 12 lube oil heater during surveillance testing. The licensee failed to have work instructions for maintenance on safety-related EDGs appropriate to the circumstances to ensure flexible coupling fasteners were correctly torqued as specified by the manufacturer to prevent leakage. The licensee entered this violation into its corrective action program (CAP) as Condition Assessment Resolution Document (CARD) 16-25666 and replaced the leaking flexible coupling. This performance deficiency was of more than minor safety significance because it was related to 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 EDG 12 flexible coupling oil leak resulted in unplanned inoperability and unavailability of this onsite emergency power source. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time nor did it represent a loss of function of a non-TS train designated as high safety significant in accordance with the licensees Maintenance Rule Program. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified. |
| 05000341/FIN-2016004-02 | 2016Q4 | Fermi | Failure to Correctly Interpret and Implement TS Requirements for LOP Instrumentation and AC Electrical Power Functions | The inspectors identified a finding of very low safety significance with an associated NCV of TS 3.3.8.1, Loss of Power (LOP) Instrumentation, and TS 3.8.1, AC (Alternating Current) Sources Operating. The licensee failed to satisfy applicable action requirements for inoperable loss of voltage and degraded voltage instrument channels, inoperable EDGs, and an inoperable offsite power circuit when power was lost to the station transformer 64 auto voltage tap changer and one-half of the instrument channels for engineered safety features bus 64C due to failure of line side potential transformer fuses on April 24, 2016. The licensee entered this performance deficiency into its CAP as CARDs 16-23392, 16-25194 and 16-28120. As an immediate corrective actions the licensee established an expectation to enter Limiting Condition for Operation (LCO) 3.3.8.1 when any of the LOP instrumentation channels are tripped. Other corrective actions included additional training for licensed operators. This performance deficiency was of more than minor safety significance because a failure to correctly implement TS LCO requirements has the potential to lead to a more significant safety concern if left uncorrected. Specifically, a failure to declare an LCO not met, enter the applicable condition(s), and follow the applicable actions could reasonably result in operations outside of established safety margins or analyses. The finding was determined to be of very low safety significance during a detailed Significance Determination Process review since the delta core damage frequency (CDF) was determined to be less than 1.0E-6/year. The inspectors concluded this finding affected the cross-cutting area of human performance and the cross-cutting aspect of training. Specifically, licensed operators failed to correctly apply the TS LCO requirements for inoperable LOP instrument channels and inoperable AC power sources due to lack of knowledge and unfamiliarity with the equipment conditions they faced during the event (IMC 0310, H.9). |
| 05000341/FIN-2016004-04 | 2016Q4 | Fermi | Inadequate Testing of SGTS Filters | The inspectors identified a finding of very low safety significance with an associated NCV of TS 5.5.7, Ventilation Filter Testing Program. The licensee failed to perform testing of the standby gas treatment system (SGTS) high-efficiency particulate air (HEPA) filters that demonstrated a penetration and system bypass of less than 0.05 percent. The licensee entered this violation into its CAP as CARD 1628812. The licensee declared the Division 1 SGTS subsystem inoperable until testing was performed satisfactorily and evaluated the extent of condition on the control room filtration system. This performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute for the control room and auxiliary building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not adequately testing the SGTS HEPA filters, the ability of the SGTS to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured. The finding was determined to be of very low safety significance because it involved only a degradation of the radiological barrier function provided by the SGTS. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified. |
| 05000341/FIN-2016004-06 | 2016Q4 | Fermi | Licensee-Identified Violation | Title 10 of the CFR, Section 50.72(a)(1)(ii), requires, in part, that the licensee shall notify the NRC Operations Center via the Emergency Notification System of those non-emergency events specified in Paragraph (b) that occurred within three years of the date of discovery. 10 CFR 50.72(b)(3) requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the applicable conditions. 10 CFR 50.72(b)(3)(v)(C) requires, in part, that the licensee report any event or condition, that at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. 10 CFR 50.73(a)(1) requires, in part, that the licensee submit an LER for any event of the type described in this paragraph within 60 days after the discovery of the event. 10 CFR 50.73(a)(2)(v)(C) requires, in part, that the licensee report any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above: 1. Between September 1, 2013 and September 30, 2016, the licensee failed to notify the NRC Operations Center via the Emergency Notification System of numerous non-emergency events specified in Paragraph (b) within eight hours of the events. These events involved the loss of safety function of the secondary containment when secondary containment pressure exceeded the TS limit due to known effects of high winds. 2. The licensee failed to submit required LERs within 60 days after the discovery of numerous events between September 1, 2013 and September 30, 2016. These events involved the loss of safety function of the secondary containment when secondary containment pressure exceeded the TS limit due to known effects of high winds. Violations of 10 CFR 50.72 and 10 CFR 50.73 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to make reports to the NRC as required by 10 CFR 50.72(a)(1)(ii) and 10 CFR 50.73(a)(1). The licensee entered this violation into its CAP as CARD 16-27023. Title 10 of the CFR, Section 20.1501, requires, in part, that each licensee shall make, or cause to be made, surveys of areas that may be necessary for the licensee to comply with the regulations in this part and are reasonable for the circumstances to evaluate the magnitude and extend of radiation levels and the potential radiological hazards of the radiation levels. 10 CFR 20.1902(b) states that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION, HIGH RADIATION AREA or DANGER, HIGH RADIATION AREA. Contrary to the above, on August 17, 2016, the licensee failed to conduct reasonable surveys to evaluate radiation levels to ensure compliance with the posting requirements of 10 CFR 20.1902(b) during activities known to cause changes in radiation levels. Specifically, the licensee failed to ensure surveys were performed while draining the annulus of the multi-purpose canister, which is an evolution known to change radiological conditions. An unposted high radiation area was identified several hours later when radiation protection personnel entered the area to perform surveys to ensure compliance with the containers Certificate of Compliance. This violation was entered into the licensees CAP as CARD 16-26586. The finding was assessed in accordance with IMC 0609, Appendix C, Occupational Radiation Safety SDP and determined to be of very-low safety significance because it did not involve as-low-as-reasonably-achievable planning or work controls, there was no overexposure nor substantial potential for an overexposure, and the ability to assess dose was not compromised |
| 05000341/FIN-2016003-02 | 2016Q3 | Fermi | Fluctuating Background Effect on Accident Range Noble Gas Monitor | The inspectors identified an Unresolved Item (URI) associated with the licensees ability, during an accident, to accurately quantify radioactive releases, potentially having an adverse impact on the licensees ability to effectively implement its Emergency Plan. Description: During a walkdown of the facility and discussions with licensee staff, the inspectors identified that the SGT system accident range monitor (AXM) noble gas detector did not utilize the fluctuating background subtraction feature of the unit and only subtracted a fixed background rate. The inspectors questioned this set-up because the physical location of the monitor was in close proximity to the SGT system filtration system, which could significantly change dose rates in the area during an accident. The function of the AXM is to assess radioactivity being released from the plant during accident conditions. These readings are used for various purposes, including accident classification and off-site dose assessment, both of which can affect the protective action recommendations made by the licensee. On September 29, 2016, the licensee provided the inspectors an assessment of the potential impact of fluctuating background from the filtration system on the AXM noble gas detector. This URI remains under review by the inspectors to determine if there is an impact to the AXM noble gas detector that could affect the licensees ability to implement its Emergency Plan and whether any violation of regulatory requirements occurred. (URI 05000341/201600302, Fluctuating Background Effect on Accident Range Noble Gas Monitor) |
| 05000454/FIN-2016003-01 | 2016Q3 | Byron | DOST Flood Barrier Door Left Open | A finding of very low safety significance and an associated NCV of 10 CFR 50 Appendix B, Criterion V, "Instructions, Procedures and Drawings," was self-revealed on September 14, 2016, when a station employee discovered that the flood barrier door for the Unit 1 Train B (1B) diesel oil storage tank (DOST) was open and unattended for three hours and six minutes. The watertight door was installed to protect the DOST fuel oil transfer pumps from the effects of a postulated failure of a circulating water expansion joint at the condenser water boxes in the turbine building, and the open door rendered the 1B diesel generator inoperable. An operator was dispatched to assess the door and, after finding no mechanical issue with the door, closed the door to restore operability to the 1B diesel generator. The issue was entered into the licensees Corrective Action Program (CAP) as IR 02699674. The inspectors determined that the issue was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to close and dog the 1B DOST door impacted the availability of the 1B diesel generator during postulated events. The finding was determined to be of very low safety significance, or Green, in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Appendix A, The Significance Determination Process (SDP) For Findings at Power, because the inspectors answered the Exhibit 2 Mitigating Systems Screening Question B as Yes. The inspectors determined that the finding involved the degradation of equipment specifically designed to mitigate a flooding event and used Exhibit 4 of the same Appendix to evaluate the significance. The inspectors determined that with the flood door open, this single condition during a turbine building flood event would degrade two trains of a multi-train system. Specifically, the turbine building flood would impact the diesel fuel transfer pumps for both Unit 1 emergency diesel generators. Therefore, a Detailed Risk Evaluation was performed by a Senior Risk Analyst who concluded that the change in core damage frequency (CDF) associated with the finding was 4.6E10/year and since the total estimated CDF was less than 1.0E7/year, the issue screened as having very low safety significance (i.e., Green) using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, for large early release frequency (LERF). The inspectors assigned a cross-cutting aspect in the Avoiding Complacency element of the Human Performance Area (IMC 0310 H.12) to this finding because an individual accessing the room through the doorway failed to challenge the door to ensure proper closure in a manner that would have revealed the door was not properly latched. |
| 05000454/FIN-2016003-03 | 2016Q3 | Byron | Failure to Properly Block and Brace a Radioactive Shipment for Transport | A finding of very low safety significance and an associated NCV of 10 CFR 71.5(a) and 49 CFR 171.1(b)(12) was self-revealed when the licensee failed to properly block and brace a Radioactive Waste (Radwaste) Shipment that was shipped to a waste processing facility for disposal. The failure to properly block and brace the Radwaste Shipment caused a breach of the shipping package while in transit to the waste processing facility. When the shipment breach was discovered at the waste processing facility, contamination surveys were immediately conducted and it was determined that no loss of content had occurred during transportation. The surveys also determined that radiation dose limits from the package were below NRC and Department of Transportation (DOT) limits. The waste processing facility notified the licensee of the breach during transport and the licensee entered the event into their CAP as IR 02665985. The inspectors determined that the issue was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and adversely impacted the cornerstone objective of ensuring adequate protection to public health and safety from exposure to radiation from routine civilian nuclear operations. Specifically, the breach of the transportation package by its content could lead to the inadvertent spread of radioactive contamination to the public domain if conditions had been slightly altered. The finding was determined to be of very low safety significance, or Green, in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, because the finding did not involve: (1) a radioactive shipment above radiation limits; (2) a certificate of compliance issue; (3) the failure to make emergency notifications; or (4) a low-level burial issue. A breach of the transportation package occurred during transit. However, the shipment contained less than a Type A quantity of material (LSA II shipment), and there was no loss of package contents or radioactive contamination. The inspectors assigned a cross-cutting aspect in the Resources element of the Human Performance Area (IMC 0310 H.1) to this finding due to inadequate procedures. |
| 05000455/FIN-2016003-02 | 2016Q3 | Byron | Failure to Use Alteration Log Resulted in Fuel Oil Leak | A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1.a, Written Procedures, was self-revealed on August 24, 2016, when a fuel oil leak of approximately one-eighth gallon per minute was identified coming from a tubing connection after the Unit 2 Train B (2B) DG was started for routine surveillance testing. Technicians replaced a fuel oil relay during the previous shift and did not use the procedurally required tools to track alterations made to each individual input line as required by MAAA716100, Maintenance Alteration Process. The issue was entered into the licensees CAP as IR 02707888. As part of their corrective actions, the leak was promptly repaired by tightening the fitting after the diesel generator was shut down; and the technicians reviewed human performance error prevention techniques, including proper use of the Maintenance Alterations Log, with supervisors. The inspectors determined that the issue was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to tighten all fittings during a maintenance activity resulted in a substantial fuel oil leak that could have resulted in a fire or could have impacted the availability of the diesel generator if the tubing had loosened further or become disconnected during a design basis event. The finding was determined to be of very low safety significance, or Green, in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Appendix A, The Significance Determination Process (SDP) For Findings at Power, because the inspectors answered Exhibit 2 Mitigating Systems Screening Question A.1 as Yes since the diesel generator remained operable and functional until the fitting was repaired. The inspectors assigned a cross-cutting aspect in the Avoiding Complacency element of the Human Performance Area (IMC 0310 H.12) to this finding because judicious implementation of human performance error prevention tools could have prevented the failure to properly tighten the fitting, even if the Alterations Log was not used as required. |
| 05000341/FIN-2016002-03 | 2016Q2 | Fermi | Loss of Power Instrumentation TS 3.3.8.1 Applicability Following Bus 64C Potential Transformer Fuse Failures | Loss of Power Instrumentation TS 3.3.8.1 Applicability Following Bus 64C Potential Transformer Fuse Failures Introduction. The inspectors opened an Unresolved Item to further evaluate the applicability of TS 3.3.8.1, Loss of Power (LOP) Instrumentation, following the failure of potential transformer fuses that caused half of the bus 64C LOP and degraded voltage relays to de-energize. Description. On April 24, 2016, a loss of output from the line to neutral potential transformers occurred on 4160 volt AC busses 64A and 64C. This resulted in the loss of bus indications, loss of automatic control of station transformer 64 load tap changer, and actuation of half of the LOP and degraded voltage relaying for safety-related bus 64C. The licensee subsequently discovered all six of the primary side fuses to the potential transformers had blown. Although the actual cause for the blown fuses was not conclusively determined, the most likely cause was attributed to an intermittent low energy transient on the secondary side of station transformer 64 or a transient on the 120 kilovolt electrical grid supplying the transformer. The inspectors noted the licensee did not consider the de-energized LOP and degraded voltage instrument channels to be inoperable, and therefore, did not enter the applicable action requirements of TS 3.3.8.1. The licensee had concluded the affected LOP and degraded voltage instrument channels remained operable since their safety function was believed to have been satisfied while they were de-energized and tripped. The inspectors raised several questions with the licensee concerning the operability of the affected LOP and degraded voltage instrument channels. The questions included whether the potential transformers were part of the LOP and degraded voltage instrumentation described in TS 3.3.8.1 and whether applicable surveillance requirements had been satisfied for the instrumentation prior to and during the event. The licensee entered this issue into its corrective action program as CARD 1625194 for further evaluation. This issue of concern is considered an Unresolved Item pending additional review by the inspectors to determine whether the licensee had correctly applied the TS limitations and satisfied applicable regulatory reporting requirements (URI 05000341/201600203, Loss of Power Instrumentation TS 3.3.8.1 Applicability Following Bus 64C Potential Transformer Fuse Failures). |
| 05000341/FIN-2016001-10 | 2016Q1 | Fermi | Failure to Satisfy 10 CFR 50.72 and 10 CFR 50.73 Reporting Requirements for Primary Containment Isolation Valve Actuations | The inspectors identified a Severity Level IV NCV of 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1), Licensee Event Report (LER) System. Specifically, the licensee failed to make a required 8-hour non-emergency notification call to the NRC Operations Center and also failed to submit a required within 60 days after discovery of a condition that resulted in the valid actuation of containment isolation signals affecting containment isolation valves in more than one system on September 13, 2015, and September 14, 2015 (two separate occurrences). Subsequently, the licensee made an 8-hour notification call on February 27, 2016 to the NRC Operations Center via the Emergency Notification System to report the events (Event Notice 51391, third update). The licensee entered this issue into its corrective action program to evaluate the cause for its failure to satisfy the reporting requirements and to identify appropriate corrective actions. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined the performance deficiency was of minor significance based on No answers to the more-than-minor screening questions. However, in accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.72(a)(1)(ii) and 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding. |
| 05000341/FIN-2016001-05 | 2016Q1 | Fermi | Failure to Satisfy 10 CFR 50.73 Reporting Requirements for a Condition Prohibited by the Plants Technical Specifications | The inspectors identified a Severity Level IV NCV of 10 CFR 50.73(a)(1), Licensee Event Report (LER) System, for the licensees failure to submit a required LER within 60 days after the discovery of an event on July 28, 2015, that was reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by the plants Technical Specifications. The condition involved the licensees failure to complete required actions for an inoperable ultimate heat sink reservoir and for both emergency diesel generators in one division inoperable within the allowed completion times. The licensee subsequently submitted LER 05000341/2015-009-00, Condition Prohibited by Technical Specification Due to Missed Entry into LCO (Limiting Condition for Operation) Condition, on March 31, 2016, to report the event. The licensee entered this issue into its corrective action program to evaluate the cause for its failure to satisfy the reporting requirements and to identify appropriate corrective actions. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was of minor significance based on No answers to the more-than-minor screening questions. However, in accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding. |
| 05000341/FIN-2016001-03 | 2016Q1 | Fermi | Failure to Satisfy 10 CFR 50.72 and 10 CFR 50.73 Reporting Requirements for Loss of RPS Trip Safety Functions | The inspectors identified a Severity Level IV NCV of the 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1), Licensee Event Report (LER) System. Specifically, the licensee failed to make a required 8-hour non-emergency notification call to the NRC Operations Center after discovery of a condition that could have prevented the fulfillment of the safety function to shut down the reactor on February 21, 2015, and on January 6, 2016 (two separate occurrences). In addition, the licensee failed to submit a required LER within 60 days after discovery of the event on February 21, 2015. Subsequently, the licensee made an 8-hour notification call on February 25, 2016 to the NRC Operations Center via the Emergency Notification System to report the two events (Event Notices 51755 and 51756). On March 2, 2016, the licensee updated Event Notices 51755 and 51756 to include an additional reporting criterion. The licensee submitted LER 05000341/2015-008-00, Turbine Stop Valve Closure and Turbine Control Valve Fast Closure Reactor Protection System Functions Considered Inoperable Due to Open Turbine Bypass Valve, on March 29, 2016, to report the February 2015 event. The licensee entered this issue into its corrective action program to evaluate the cause for its failure to satisfy the reporting requirements and to identify appropriate corrective actions. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was of minor significance based on No answers to the more-than-minor screening questions. However, in accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.72(a)(1)(ii) and 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding. |
| 05000341/FIN-2015301-01 | 2015Q3 | Fermi | Inadequate Examination Security on a Simulator Reset | Title 10, Code of Federal Regulations (CFR), Part 55.49, Integrity of examinations and tests, states, in part, that a licensee 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. Contrary to this, the licensee failed to clean a marked-up hard card prior to a job performance measure (JPM) during administration of the operating tests. To correct this issue, the licensee initiated corrective action CARD number 15-26003. Additionally, a replacement JPM was administered to the applicant affected by the compromised JPM. The failure of the licensees staff to ensure that previously used examination materials were not available for an applicant during the initial examination administration was a performance deficiency. The performance deficiency was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV noncited violation because it involved a non-willful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. |
| 05000255/FIN-2014008-10 | 2014Q4 | Palisades | Lack of Analysis for Electrical Containment Penetration Protection | The inspectors identified an unresolved item (URI) regarding lack of an analysis to demonstrate that circuit breakers and fuses provide adequate protection against short circuits and overloads for electrical containment penetrations, as discussed in Regulatory Guide 1.63. Resolution of this issue will be based on clarification of Palisades licensing basis by NRC staff. As part of the review of power supplies to components inside containment, the inspectors requested to review the analysis that demonstrates protection of the electrical penetrations against short circuits and overloads. The licensee responded that such an analysis does not exist, and stated their position that it is not required by their design and licensing bases. Electrical protection of containment penetrations was the subject of the Palisades Systematic Evaluation Program (SEP) Topic VIII-4. A letter from Dennis M. Crutchfield, NRC, to David P. Hoffman, Consumers Power Company, SEP Topic VIII-4, Electrical Penetrations of Reactor Containment, dated March 26, 1981, (ADAMS Accession No. ML8104080152) included an enclosure entitled Position on Protection of Containment Electrical Penetrations against Failures Caused by Fault and Overload Currents for SEP Plants. This position document states: ...the staff requires compliance with the recommendations of Regulatory Guide 1.63 or an acceptable alternative method. For each containment electrical penetration, the protective systems provide primary and backup protection devices to prevent a single failure in conjunction with a circuit overload from impairing containment integrity. The licensee responded in a letter from Robert A. Vincent, Consumers Power Company, to Dennis M. Crutchfield, NRC, SEP Topic VIII-4, Electrical Penetrations of Reactor Containment, dated June 15, 1981, (ADAMS Accession No. ML8106180170), in which they stated: The secondary (backup) interrupt devices (...) would fail to trip prior to the penetration reaching its limiting temperature of 302 C with the postulated combination of faults and failure of the primary interrupters. The licensee committed to perform more detailed evaluations of the capabilities of the protective devices, as well as An evaluation of the adequacy of the Palisades Plant overcurrent protection surveillance testing program. In the following subsequent letters, the licensee reported on the progress of their further evaluations: Letter from Robert A. Vincent, Consumers Power Company, to Dennis M. Crutchfield, NRC, SEP Topic VIII-4, Electrical Penetrations of the Reactor Containment, dated November 16, 1981, (ADAMS Accession No. ML8111200805); Letter from Kerry A. Toner, Consumers Power Company, to Dennis M. Crutchfield, NRC, SEP Topic VIII-4, Electrical Penetrations of the Reactor Containment, dated October 12, 1982, (ADAMS Accession No. ML8210190459); and letter from Kerry A. Toner, Consumers Power Company, to Dennis M. Crutchfield, NRC, SEP Topic VIII-4, Status Update of Program to Evaluate the Adequacy of Penetration Protection from Overload and Short-Circuit Conditions, dated February 11, 1983, (ADAMS Accession No. ML8302240273). The NRC issued their Integrated Plant Safety Assessment Report on this SEP Topic in Letter from Thomas V. Wambach, NRC, to David J. VandeWalle, Consumers Power Company, Integrated Plant Safety Assessment Report (IPSAR) Section 4.26, Electrical Penetrations of Reactor Containment Palisades Plant, dated June 10, 1983, (ADAMS Accession No. ML8306160396). This IPSAR stated, The staff has evaluated this issue for other plants... and concluded that no further action was required for these plants. Based upon the information contained in the licensees letters dated June 15, 1981, October 12, 1982, and February 11, 1983, the staff concludes that the design of the Palisades electrical penetrations are similar to other SEP plants, that the probability of electrical failure is low and that any leakage path due to penetration failure would be small. Therefore, we consider this issue to have been completed satisfactorily and further action by the licensee is not required. This conclusion was reiterated in NUREG 0820, Supplement 1, Integrated Plant Safety Assessment, Systematic Evaluation Program, Palisades Plant, Final Report, dated November 1983, (ADAMS Accession No. ML8311290133). The inspectors requested that the licensee discuss their lack of an analysis for electrical penetration protection in light of Palisades FSAR Section 8.5.1.2, which states: 10 CFR Part 50, General Design Criterion 50, as implemented by Regulatory Guide 1.63 and IEEE Standard 317-1972, requires that electrical penetrations be designed so that the containment structure can accommodate, without exceeding the design leakage rate, the calculated pressure, temperature and other environmental conditions resulting from any Loss of Coolant Accident (LOCA). The licensee responded by initiating corrective action CR-PLP-2014-04450, which states the licensees position that Palisades is not committed to the electrical circuit protection requirements of Regulatory Guide 1.63. Due to complexity of establishing the appropriate design and licensing bases for this issue, the concern will be resolved using the NRCs Task Interface Agreement (TIA) process. Pending resolution, this item will be tracked as an unresolved item. |