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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 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 |
| 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. |
| 05000454/FIN-2018003-01 | 2018Q3 | Byron | Minor Violation | On June 14, 2018, the licensee performed IST surveillance 2BOSR 5.5.8.DO1, Test of the Diesel Oil Transfer System, on the 2A diesel oil transfer pump. On June 19, 2018, the inspectors noted that an issue concerning the calibration of the Flexim ultrasonic flow meter used during the test had not been documented in the licensees Corrective Action Program (CAP). Specifically, the calibration sticker on the flow meter used during the surveillance test indicated that the instrument was calibrated to a 5 percent accuracy when the ASME OM Code required an instrument accuracy of 2 percent. The inspectors discussed the issue with licensee management. The licensee subsequently confirmed that the instrument calibration did not meet ASME OM Code requirements and entered this issue into their CAP.Title 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, requires that measures be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specific periods to maintain accuracy within necessary limits. Licensee procedure ERAA321, Administrative Requirements for Inservice Testing,Section 4.10.3, states, in part, that instrument accuracy and range requirements are specified in the applicable ASME Code Edition/Addenda. ASME OM Code Paragraph ISTB-3510, General, states, in part, that instrument accuracy shall be within the limits of Table ISTB-35101, Required Instrument Accuracy. Table ISTB35101 states that the required instrument accuracy for determining flow rate is 2 percent. Screening: The failure to implement programmatic controls that ensured measurement and test equipment was calibrated to the accuracy requirements of the ASME OM Code was a performance deficiency. The instruments used in IST surveillances were later re-certified to meet the required 2 percent accuracy in the ASME OM Code with no required adjustments. As a result, the performance deficiency was determined to be minor because the inspectors answered No to all of the more-than-minor screening criteria in IMC 0612, Appendix B. The licensee generated Issue Report (IR) 04149294 to document this issue in their CAP. This issue was also incorporated into a corrective action program evaluation (CAPE) report evaluating an adverse trend identified with ASME test performance at the site (AR 04154533). Violation: The failure to comply with 10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, constituted a minor violation that was not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 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. |
| 05000454/FIN-2018002-03 | 2018Q2 | Byron | Minor Violation | Minor Violation: The inspectors identified multiple instances of a failure to perform inservice testing in accordance with written procedures appropriate for the circumstances during this inspection period: 1. On March 30, 2018, the licensee performed 1BOSR 5.5.8.DO2, Test of the Diesel Oil Transfer System, and declared the 1B diesel oil transfer pump inoperable due to flow results being low out of specification. Subsequently, the licensee determined that the instrument setup was incorrect in that an incorrect value was entered into the flow meter for pipe diameter. The licensee declared the surveillance invalid and scheduled a time to re-perform the activity. Acceptable system flow rates were achieved a week later when the correct pipe diameter was used for the instrument setup. 2. On April 26, 2018, while observing the licensee perform 2BOSR 5.5.8.CS.52C, Comprehensive Inservice Testing (IST) Requirements for Containment Spray Pump 1CS01PB, the inspectors noted that the pump suction pressure and discharge pressure test gauges were not installed as described in the Precautions and Limitations section of the procedure. After the inspectors asked how the installed configuration satisfied the procedure requirement, the licensee suspended the test to obtain clarification. After some deliberation between engineers and operators attempting to identify the correct instrument location, the test data was recorded with the instruments at different locations for data gathering and comparison. The licensee verified that pump performance had sufficient margin, including the introduced error, to remain operable and available to perform its safety-related function as expected.3. On May 1, 2018, while observing the licensee perform 2BOSR 5.5.8.SX.51C, Comprehensive Inservice Testing (IST) Requirements for the Essential Service Water (SX) Pump 2SX01PA and Unit 2 SX Pumps Discharge Check Valves, the inspectors noted that operators were not taking data from the ultrasonic flow meter in accordance with the procedure. Specifically, the instrument was not set up to indicate time and flow so that an average flow could be determined as required by a Note in the procedure. Instead the operators were recording instantaneous flowrate. When the inspector asked for clarification and the operators and technicians deferred to their supervisors, the licensee suspended the test to obtain clarification. The test was performed again after the instrument was set up correctly and operators were briefed on how to obtain the correct data.Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions. Contrary to the above, for the diesel fuel oil transfer pump surveillance, 1BOSR 5.5.8.DO2, the procedure listed an incorrect pipe diameter value that was subsequently entered into the flow meter resulting in unacceptable test results; for the containment spray pump surveillance, 2BOSR 5.5.8.CS.52C, the licensee potentially introduced an unaccounted for error in the surveillance test method by not setting up test equipment in accordance with the procedure; and for the SX surveillance, 2BOSR 5.5.8.SX.51C, the licensee introduced a potential error in the surveillance test by not determining an average flow rate as discussed in the procedure Note.Screening: The failure to perform inservice testing in accordance with written procedures appropriate for the circumstances was a performance deficiencyin each of the listed 11 examples. The performance deficiency was determined to be minor in each case because the inspectors answered No to all of the more-than-minor screening questions in IMC 0612, Appendix B. The licensee generated the following issue reports (IRs) to document these issues:AR 04121539, Ultrasonic Flow Measurement Installation IssueAR 04122295, PCR (procedure change request) 1/2BOSR 5.5.8.DO1 AR 04131201, Engineering Clarification Needed on ASME Precaution AR 04133585, NRC ID: Potential Concerns With Execution of 2A SX Pump Surveillance Violation: These failures to comply with 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, constituted minor violations that are not subject to enforcement action in accordance with the NRCs Enforcement Policy |
| 05000454/FIN-2018002-02 | 2018Q2 | Byron | Licensee-Identified Violation | A violation of very low safety significance was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.Licensee procedure ERAA321, Administrative Requirements for Inservice Testing, stated in Step 4.10.5, that acceptance criteria are established using the reference values and the applicable ASME (American Society of Mechanical Engineers) Code. Paragraph ISTA3160, Test and Examination Procedures, of the ASME Operation and Maintenance of Nuclear Power Plants (OM) Code required in part that, Tests and examinations shall be performed in accordance with written procedures. The procedures shall contain the Owner-specified reference values and acceptance criteria. Paragraph ISTA9230, Inservice Test and Examination Results, of the ASME OM Code required, in part, that The results of tests and examinations shall be documented and shall include the following: comparison with allowable ranges of test and examination values, and analysis deviations and requirements for corrective action.Contrary to the above, from July 1, 2016, to May 30, 2018, the licensees procedures did not clearly document acceptance range, alert range, and required action values for the diesel oil (DO) transfer pump IST surveillance tests in accordance with the ASME OM Code. This resulted in several instances where the pump being tested did not meet IST criteria, but no action was taken. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedural Quality 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 clearly identify the acceptance criteria, alert range and required action ranges resulted an in organizational failure to declare the pumps inoperable and to perform required analysis of the equipments condition. The inspectors assessed the significance of the finding using SDP Appendix A and concluded the issue was of very low safety significance (i.e., Green).Corrective Action References: (1) AR 04142617, Acceptance Criteria Not Clearly Listed in DO Pump Procedures, and (2) AR 04142370, DO Pump Test Packages are Not Routed to the IST Coordinator. |
| 05000456/FIN-2018002-03 | 2018Q2 | Braidwood | Inadequate Test Activity Coordination Results in Unintended Valve Actuation and Reactor Coolant System Pressure Drop | A self-revealed finding of very low safety significance (i.e., Green) and an associated NCV of Technical Specification 5.4, Procedures, was identified for the licensees failure to have properly coordinated testing activities associated with redundant Unit 1 pressurizer pressure instruments in accordance with the stations procedural requirements governing such testing. Specifically, during the licensees 20th Unit 1 refueling outage, on April 23, 2018, redundant pressurizer pressure instrumentation channels were inadvertently subjected to simultaneous testing activities. This resulted in the coincidence logic for both of the units pressurizer power-operated relief valves (PORVs) being satisfied and the PORVs opening to depressurize the RCS from approximately 345 pounds per square inch gauge (psig) to approximately 320 psig |
| 05000456/FIN-2018002-02 | 2018Q2 | Braidwood | Work Instruction Error Results in Reactor Coolant System Pressure Transient | A self-revealed finding of very low safety significance (i.e., Green) was identified due to the licensees failure to follow work instructions while performing a digital upgrade to plant control systems. Specifically, while performing maintenance on the volume control tank (VCT) level transmitter on April 10, 2018, maintenance personnel failed to properly track the steps being performed while simultaneously working on multiple packages. This resulted in the Unit 1 reactor coolant system (RCS) experiencing a pressure transient and the actuation of a VCT relief valve. |
| 05000457/FIN-2018002-01 | 2018Q2 | Braidwood | Inadequate Detail in Maintenance Work Instructions Resulted in Failed Gearbox Oil Cooler Head Gasket and Inoperable 2B Auxiliary Feedwater Pump | 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 have adequate detail within their maintenance work instructions to enable proper reassembly of the 2B auxiliary feedwater (AF)pump gearbox oil cooler. Specifically, during the licensees 19th Unit 2 refueling outage in April 2017, the gearbox oil cooler closure head was reassembled following scheduled maintenance using an excessive amount of room temperature vulcanizing silicone (RTV) on the joint and an insufficient amount of torque on the closure head bolting. As a result, on March 16, 2018, the closure head joint failed causing several hours of unplanned inoperability and unavailability for the 2B AFPump. |
| 05000455/FIN-2018002-01 | 2018Q2 | Byron | Overspeed Trip of 2B Auxiliary Feedwater Pump During Surveillance | A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed when the 2B diesel-driven auxiliary feedwater (AF) pump tripped on overspeed during a quarterly inservice test (IST). Specifically, operators with portable instrumentation used an erroneous speed value to adjust pump speed beyond the range specified in the procedure resulting in a pump overspeed trip, entry into a 72-hour technical specification (TS) required action statement, and unplanned pump unavailability with an associated change in Unit 2 risk from green to yellow. |
| 05000255/FIN-2018010-01 | 2018Q1 | Palisades | Licensee-Identified Violation | Violation: Title 10 of theCode of Federal Regulations (CFR) Part 50.55a(g)(4), Inservice Inspection Standards Requirement for Operating Plants, requires that, throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the 2006 edition through 2008 addenda of the ASME Boiler and Pressure Vessel Code. This edition of the AMSE Code requires that a VT3 visual examination of supports other than piping supports be performed once every 10year inservice inspection (ISI) interval. Contrary to the above, since the beginning of plant operation, the safety-related CCW and SW pump lateral supports (classified as ASME Code Section XI Class 3) had never been included in the ISI program and therefore had never had the required VT3 examination performed during each 10year ISI interval. Corrective actions included incorporating the supports into the ISI program, scheduling the inspections as required, and validating that the supports were still capable of performing their safety function and that the CCW and SW systems remained operable.Significance/Severity Level: The inspectors determined that the failure to perform ASME Code Section XI required inspections of the CCW and SW pump lateral supports was a performance deficiency. The inspectors determined the performance deficiency was more than minor because it adversely affected the Design Control attribute of the Mitigating Systems 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, the failure to periodically inspect the pump lateral supports could result in the failure to identify a nonfunctional support that could increase the risk of a pump failure.The inspectors assessed the significance of the finding using Appendix A of the SDP. The finding was determined to be of very low safety significance (Green) because although it was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), the SSC remained operable. Corrective Action Reference: CRPLP201705784, OE Review Identified Palisades Failure to Inspect ASME Class 3 Pump Supports for SW and CCW Pumps, 1/26/2018 Safety Conscious Work Environment Observations Based on interviews with plant staff and reviews ofthe latest safety culture survey results to assess the safety conscious work environment on site, the team determined that, in general, plant personnel appeared willing to raise nuclear safety concerns through at least one of the several means available. Most of those interviewed had an adequate knowledge of the CAP process and would initiate a CR, or work with someone who would do so on their behalf, if they knew of a safety concern. A weakness was identified in plant personnel knowledge ofhow to use the electronic CR system. Specifically, there were some personnel who were not familiar with how to generate a CR or how to track the resolution of a CR. Personnel also expressed an overall frustration with feedback provided on a CR; either with difficulties in being able to see how something was resolved or with not being able to understand the decision-making process for the resolution of issues.Most individuals expressed a willingness to raise safety concerns without fear of retaliation and all employees knew the importance of having a strong safety conscious work environment. There were some instances where the free flow of information or a willingness to raise concerns through an individuals direct line of supervision were hampered due to the perception that supervision was not receptive to receiving the concern or addressing the issue. In some cases, this presented an uncomfortable work environment for the affected individuals. However, when presented with this situation, all individuals knew of other supervisors that they could bring their concerns to or other avenues to use to address anissue. All plant personnel were aware of the Employee Concerns Program (ECP), knew who the ECP coordinator was, and most were willing to use it as an avenue to raise concerns, if desired. However, some individuals believed that the ECP lacked the appropriate level of confidentiality to effectively address concerns. |
| 05000255/FIN-2018001-03 | 2018Q1 | Palisades | Licensee-Identified Violation | A violation of very low safety significance (Green) was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a Non-Cited Violation consistent with Section 2.3.2 of the Enforcement Policy. Enforcement:Violation: Technical Specification 3.7.6 requires that the combined useable volume of the Condensate Storage Tank (CST) and Primary Makeup Storage Tank (T81) shall be greater or equal than 100,000 gallons. LCO 3.7.6, Condition A states that if the useable volume is not within this limit then A.1 Verify OPERABILITY of backup water supplies in 4 hours andA.2 Restore condensate volume to within limit in 7 days. Condition B states that if the Required Action and associated Completion Time is not met then B.1 Be in MODE 3 in 6 hours and B.2 Be in MODE 4 without reliance on steam generators for heat removal in 30 hours. Contrary to the above, on December 7, 2017 and March 3, 2016, the licensee failed to enter and comply with the actions required by LCO 3.7.6 Condition A and Condition B when Primary Makeup Tank Makeup Control Valve CV2008 could not be fully opened, resulting in a combined useable volume of the CST and T81 of less than 100,000 gallons.Significance/Severity Level: The inspectors answered No to all the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, because even though the CST and T81 volume were considered inoperable by the TS requirements, there was not a loss of safety function because credited backup water sources were available and operable.Therefore, the finding screened as Green.Corrective Action References: The licensee entered these issues into their CAP as CRPLP20175589, CRPLP20175554, CRPLP20175551, and CRPLP20161116 |
| 05000255/FIN-2018001-02 | 2018Q1 | Palisades | Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance | On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliances that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The EGM applies specifically to a structure, system, or component (SSC) that is determined to be inoperable for tornado-generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Palisades, the EGM provided for enforcement discretion of up to 3 years from the original date of issuance of the EGM. On December 7, 2017, and as supplemented on January 18, 2018, Palisades submitted a request to the NRC to extend the enforcement discretion from June 10, 2018 to June 10, 2020 (ML17341A415 and ML18018A328, respectively). By letter dated February 16, 2018, the NRC granted the request to extend enforcement discretion until June 10, 2020 (ML18046A675). The EGM permitted NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provide additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within about 60 days of issue discovery. In accordance with the EGM, the comprehensive compensatory measures are toremain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Palisades was licensed prior to issuance of Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC). Specifically, GDC 2, Design Bases for Protection Against Natural Phenomena, and GDC 4, Environmental and Dynamic Effects Design Basis, discuss how SSCs important to safety shall be designed to protect against natural phenomena, such as tornadoes and shall be adequately protected against the dynamic effects of tornadoes, including protection against missiles. Palisades site-specific licensing bases compliance with GDC 2 and GDC 4 are described in the Updated Final Safety Analysis Report (UFSAR) Sections 5.1.2.2 and 5.1.2.4. Palisades protection of SSCs against tornado-generated missiles is also discussed in UFSAR Section 5.5, Missile Protection. On January 31, 2018, the licensee initiated condition report (CR) CRPLP201800556, which identified a nonconforming condition in the Palisades licensing basis. Specifically, the surge line from the component cooling water (CCW) surge tank to the CCW suction line was identified to be potentially vulnerable to a tornado missile through a doorway. The licensee previously identified a CCW system-related vulnerability on March 29, 2017. The March 29, 2017 CCW vulnerability and five additional vulnerabilities of other SSCs, which all received enforcement discretion, are documented in NRC Inspection Report 05000255/2017002 (ML17220A349). The licensee assessed this new vulnerability and concluded that previously established compensatory measures for the CCW system were adequate and no additional comprehensive compensatory actions were required. Therefore, the licensee declared the SSC operable, but nonconforming because no additional compensatory measures designed to reduce the likelihood of tornado-generated missile effects were required and the previously implemented compensatory measures were still in place. Corrective Action: The licensee documented the condition of the SSC in the CAP and documented the SSC as operable but nonconforming.Corrective Action Reference: CRPLP201800556 Enforcement: Violation: Enforcement discretion was applied to the required shutdown actions of the following Technical Specification (TS) Limiting Conditions for Operation (LCOs): TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); andTS 3.7.7, Component Cooling Water (CCW) System.Severity/Significance: The subject of this enforcement discretion associated with tornado missile protection deficiencies 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 EGM 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance (ML16355A286). 11 Basis for Discretion:The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented more comprehensive compensatory actions to resolve the nonconforming conditions within the required 60 days. These comprehensive measures were to remain in place until permanent repairs were completed, which for Palisades were required to be completed by June 10, 2020, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed.The disposition of this enforcement discretion closes LER 05000255/201700101, Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions. |
| 05000255/FIN-2018001-01 | 2018Q1 | Palisades | Failure to Maintain an Appropriate Documented Work Instruction for Reassembly of Primary Makeup Tank Makeup Control Valve CV2008 | A self-revealed Green finding and an associated NCV of Technical Specification 5.4.1, Procedures, was identified for the licensees failure to have an adequate maintenance work instruction for the reassembly of Primary Makeup Tank Makeup Control Valve CV2008. Specifically, because a previous CV2008 maintenance activity failed to properly set the height of the CV2008 jam nuts, the valve guide key fell out of place and in December 2017, CV2008 was unable to be manually stroked during surveillance testing |
| 05000255/FIN-2017004-01 | 2017Q4 | Palisades | Improperly Connected M&TE Leads to Unexpected AFU Fan Trip | A finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to follow step 5.4.4.b of Technical Specification surveillance procedure RT85DA, Control Room Emergency Ventilation Filtration Testing A Train. Specifically, the licensee failed to properly connect maintenance and test equipment (M&TE) across flow transmitter test taps which caused V26A, the air filter unit (AFU) VF26A fan, to stop 17 seconds after operators started the fan from the control room. The licensee entered this issue into their Corrective Action Program (CAP) as condition report (CR) CRPLP201705234. Corrective actions included coaching the vendor on ensuring M&TE is properly connected to plant equipment and ensuring suitable field oversight of the vendor during re-performance of the surveillance.The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Barrier Integrity cornerstone attribute of Human Performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. 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 3, because the inspectors answered "No" to all screening questions. The finding had across-cutting aspect in the area of Human Performance, in the Field Presence aspect, for the failure to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel (H.2). |
| 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. |
| 05000456/FIN-2017008-05 | 2017Q4 | Braidwood | Inaccurate Analysis of Record | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to maintain an accurate and up- to-date analysis of record for a postulated HELB in th e MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and performed an operability evaluation. The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not resul t in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000456/FIN-2017008-04 | 2017Q4 | Braidwood | Untimely Corrective Action for Secondary Missiles | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with t he 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Sc reening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a los s of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000456/FIN-2017008-03 | 2017Q4 | Braidwood | Failure to Properly Correct Errors in Design Analysis for Main Steam Line Break in Main Steam Tunnel | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly correct errors in the design analysis for a main steam line break in the main steam tunnel. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and completed an operability evaluation. The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000456/FIN-2017008-02 | 2017Q4 | Braidwood | Inadequate Blow Out Panel Design Control | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee originall y designed the MSSV blow out panels in a manner that prevented them 3 from functioning properly. The licensee entered this issue into their CAP as A R 4075641 and corrected the design issue in March of 2009 . The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, E xhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000456/FIN-2017008-01 | 2017Q4 | Braidwood | Failure to Prevent Secondary Missiles Following a Postulated HELB | The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design basis for the main steam safety valve (MSSV) room maintenance hatches was maintained. Specifically, the high energy line break ( HELB) analysis performed for the MSSV rooms and steam tunnels prior to initial construction concluded that no secondary missiles were generated as a result of a HELB although maintenance hatches in the ceiling of the MSSV rooms were identified to become secondary missiles following a HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their corrective action program (CAP) as AR 4075641 and performed an operability evaluation. The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC (Structure, System, and Component) , does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross- cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000454/FIN-2017010-05 | 2017Q4 | Byron | Inaccurate Analysis of Record | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to maintain an accurate and up- to-date analysis of record for a postulated HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000454/FIN-2017010-04 | 2017Q4 | Byron | Untimely Corrective Action for Secondary Missiles | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000454/FIN-2017010-03 | 2017Q4 | Byron | Failure to Properly Correct Errors in Design Analysis for Main Steam Line Break in Main Steam Tunnel | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075608 and performed an operability evaluation. The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000454/FIN-2017010-02 | 2017Q4 | Byron | Inadequate Blow Out Panel Design Control | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee originally designed the MSSV blow out panels in a manner that prevented them 3 from functioning properly. The licensee entered this issue into their CAP as A R 4075608 and corrected the design issue in March of 2009 . The finding 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 (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000454/FIN-2017010-01 | 2017Q4 | Byron | Failure to Prevent Secondary Missiles Following a Postulated HELB | The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regualtions (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design basis for the main steam safety valve (MSSV) room maintenance hatches was maintained. Specifically, the high energy line break ( HELB) analysis performed for the MSSV rooms and steam tunnels prior to initial construction concluded that no secondary missiles were generated as a result of a HELB although maintenance hatches in the ceiling of the MSSV rooms were identified to become secondary missiles following a HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their corrective action program (CAP) as AR 4075608 and performed an operability evaluation. The finding 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 system s that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC (Structure, System, and Component) , does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross- cutting aspect was assigned to this finding as it was not reflective of current performance. |
| 05000454/FIN-2017007-02 | 2017Q3 | Byron | Failure to Promptly Identify Degraded Reactor Containment Fan Cooler Circuitry | The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly identify a condition adverse to quality resulting in a safety -related system becoming inoperable. Specifically, from May 5, 2017, to August 4, 2017, the licensee failed to trend available surveillance data in a timely manner and did not identify a degraded condition in the 1A reactor containment fan cooler (RCFC) time delay circuitry prior to the system becoming inoperable. The licensee entered this issue into their CAP as AR 04039037 and AR 04045767, replaced the failed relay, and planned to update the RCFC system monitoring plan to note abnormal changes in time delay relay actuation times and improve coordination between engineering and operations to reduce the time it takes engineering to obtain RCFC surveillance data for trending after surveillances are completed. The inspectors determined that the failure to promptly identify a condition adverse to quality associated with the time delay relay circuitry in the 1A RCFC was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify a degraded condition in the time delay circuitry associated with the 1A RCFC resulted i n a missed opportunity for the licensee to evaluate the cause and initiate prompt actions to respond to the degraded condition prior to the failure. The inspectors answered No to questions A.1 through A.4 of IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions ; therefore, the finding screened as 4 having very low safety significance. This finding affected the Cross -Cutting area of Problem Identification and Resolution in the aspect of Trending because information was available that indicated a degraded condition in the 1A RCFC time delay relay circuitry for three months prior its failure in August , but was not identified and evaluated by the licensee prior to failure (P.4) . |
| 05000454/FIN-2017007-01 | 2017Q3 | Byron | Fai lure to Perform Maintenance in Accordance with Performance Centered Maintenance Template | The inspectors identified a finding of very low safety significance and an associated NCV of TS 5.4.1, Procedures, when licensee personnel failed to perform maintenance in accordance with written procedures as required by Regulatory Guide 1.33. Specifically, from February 3, 2014, through August 25, 2017, the licensee failed to develop and execute work instructions of sufficient scope to accomplish the 3 preventive maintenance to replace flexible hoses on the essential service water (SX) makeup pumps and the diesel driven auxiliary feedwater (AFW) pumps and did not have a technical justification for a deviation from the Exelon Corporate Performance Centered Maintenance (PCM) template. The licensee entered this issue into their CAP as Action Request (AR) 03961955, AR 03971962, and AR 04045769 and planned to replace the flexible hoses at the next available opportunity. The inspectors determined that failure to perform maintenance in accordance with written procedures as required by TS 5.4.1, Procedures, and Regulatory Guide 1.33 was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance 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, failing to replace flexible hoses on the SX makeup pumps and the Unit 1 and Unit 2 diesel -driven AFW pumps at a pre - established frequency could allow hose degradation to remain unidentified and lead to the unplanned inoperability of these safety-related systems. Since the finding is a deficiency affecting the design or qualification of mitigating systems, structures and components (SSC s) and the SSC s remained operable and functional, the finding screened as having very low safety significance. This finding affected the C ross -Cutting area of Human Performance in the aspect of Work Management because the licensee failed to perform required maintenance in accordance with their associated maintenance strategy as well as the corporate PCM template (H.5) . |
| 05000456/FIN-2017003-01 | 2017Q3 | Braidwood | Failure to Implement Adequate Radiological Controls for Treated Liquid Radioactive Effluents Containing Tritium | The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1406(c), when the licensee failed to conduct operations to minimize the introduction of residual radioactivity onto the site. Specifically, the licensee failed to identify and evaluate the environmental risk and control work practices with a credible mechanism to prevent spills and leaks from reaching groundwater at the circulating water blowdown (CWBD) area, a radiologically unrestricted area in the licensees owner controlled area. Specifically, tritium contaminated sump water was intermittently pumped to the environs. The licensee documented this finding in their corrective action program (CAP) as Issue Report (IR) 4020644. The failure to conduct operations and control work practices with a credible mechanism to prevent spills and leaks to reach groundwater and minimize residual radioactivity onto the site represented a licensee performance deficiency. The performance deficiency was of more than minor significance because it was associated with the Program and Process attribute of the Public Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring the 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. In accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because the issue involved a radioactive effluent release, but did not: (1) represent a substantial failure to implement the radioactive effluent release program; or (2) result in public exposure that exceeded the dose values in Appendix I to 10 CFR Part 50 and/or 10 CFR 20.1301(e) limits. The inspectors determined that this finding had a cross-cutting component in the area of Human Performance, in the aspect of Challenging the Unknown, because licensee personnel did not stop when faced with uncertain conditions or evaluate and manage risk before proceeding. |
| 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-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. |
| 05000456/FIN-2017002-03 | 2017Q2 | Braidwood | Failure to Adequately Implement and Maintain the Radiological Environmental Monitoring Program by Collecting Representative Samples from the Principal Food Pathways | Green. A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix I, Section IV(B), were identified by the inspectors for the licensees failure to establish an appropriate surveillance and monitoring program in order to provide data on measurable levels of radiation and radioactive materials in the environment to evaluate the relationship between quantities of radioactive material released in effluents and resultant radiation doses to individuals from principal pathways of exposure. This was an NRC-identified finding for the failure to implement and maintain the licensees radiological environmental monitoring program (REMP) by collecting representative samples from the highest deposition coefficient (D/Q) quadrant locations during annual REMP sampling and collections of food products in 2015. On May 25, 2016, during a review of the stations annual radiological environmental operating report for 2015, the inspectors noted that the licensee documented missed samples in three out of four quadrants where the principal food pathways were grown within the 10 kilometers from the station and missed milk samples. The licensee captured this issue in their CAP as IR 4002540. Licensee corrective actions included, but were not limited to, revising the applicable REMP procedures and investigating the possibility of growing the principal food pathways on the licensees owner controlled area or other approved licensee property within the 10 kilometer site radius. The performance deficiency was determined to be more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of the public from radiation. Specifically, the licensee failed to implement effective sample collection from sample locations for food products from three of the major quadrants during annual REMP sampling and collections in 2015. The licensees Offsite Dose Calculation Manual (ODCM), as written, did not meet 10 CFR Part 50, Appendix I, which requires the licensee to establish and provide data on measurable levels of radiation and radioactive materials in the site environs. The finding was determined to be of very low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, because it only involved the licensees REMP. The inspectors determined that this finding had a 4 cross-cutting component in the area of human performance, change management aspect, because the licensee did not use a systematic process for evaluating and implementing changes in their REMP sampling and collection program. (H.3) |
| 05000456/FIN-2017002-02 | 2017Q2 | Braidwood | Failure to Adequately Implement Technical Specification Surveillance Frequency Requirements into Implementing Procedures | Green. A finding of very low safety significance and an associated NCV of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified by the inspectors for the licensees failure to have appropriate implementing procedures for TS SR 3.9.3.2. Specifically, procedure BwIS NR203, Post Accident Neutron Monitoring System Discriminator Adjustment, 3 did not provide for determining and checking the discriminator voltage for the system at an 18-month frequency, as specified by TS SR 3.9.3.2. The licensee entered this issue into their CAP as IR 4010147 with an action to revise the surveillance frequency to every 18 months for each channel. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective. The finding screened as having very low safety significance (Green) because it did not result in the loss of operability or functionality of any SSC. The licensee performed a review of the records associated with the last three years of operation and did not find any instances in which the post-accident neutron monitors (PANMs) were used to satisfy TS 3.9.3, Nuclear Instrumentation, requirements. No cross-cutting aspect was associated with this finding because it was confirmed not to be reflective of current licensee performance due to the age of the performance deficiency. |
| 05000456/FIN-2017002-01 | 2017Q2 | Braidwood | Failure to Adequately Implement Surveillance Frequency Program for the Deferral of a Technical Specification Surveillance | Green. A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.5.19.b, Surveillance Frequency Program, were identified by the inspectors for the licensees failure to implement the requirements contained in the surveillance frequency control program when making a change to the specified frequency of TS Surveillance Requirement (SR) 3.3.1.11. On May 3, 2017, the licensee improperly deferred a TS required surveillance through the preventive maintenance deferral process due to a belief that it was a preventive maintenance activity and not an activity supporting a TS SR. The licensee entered this issue into their corrective action program (CAP) as Issue Report (IR) 4009050 with an action to re-establish the surveillance at an 18-month frequency and to perform it before the end of the Unit 2 refueling outage (RFO) A2R19. The performance deficiency was determined to be more than minor because if left uncorrected it could lead to a more significant safety concern. The finding screened as being of very low safety significance (Green) because it did not result in the loss of operability or functionality of any system, structure, or component (SSC). The inspectors determined that this finding had a cross-cutting component in the area of human performance, work management aspect, because the licensee failed to utilize a work process that included proper coordination with different groups or job activities. Specifically, licensee personnel conducting the deferral did not coordinate the activity with personnel in either the operations or regulatory assurance departments. Knowledgeable personnel in either of these station organizations could have identified that the wrong process for deferral was being utilized. (H.5) |
| 05000293/FIN-2016011-07 | 2017Q1 | Pilgrim | Failure to Report Condition Prohibited by Technical Specifications and a Safety System Functional Failure | The NRC team identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, associated with Entergys failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. Specifically, on September 28, 2016, Entergy identified the A emergency diesel generator was inoperable. The NRC team determined that the condition was prohibited by technical specifications and the inoperability of the A emergency diesel generator existed for a period of time longer than allowed by Technical Specification 3.5.F, Core and Containment Cooling Systems. This was also reportable as a safety system functional failure. Entergy entered this issue into the corrective action program as CR-PNP-2016-09552. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC team evaluated the performance deficiency using traditional enforcement. The violation was evaluated using Section 2.3.11 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. In accordance with Section 6.9.d, Example 9, of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, the NRC team did not assign a cross-cutting aspect to this violation, in accordance with IMC 0612, Appendix B. |
| 05000293/FIN-2016011-13 | 2017Q1 | Pilgrim | Licensee-Identified Violation | 10 CFR 50.54(q)(2) requires, in part, that the licensee follow and maintain the effectiveness of an emergency plan to meet the planning standard of 10 CFR 50.47(b)(4). Specifically, the licensee was to maintain the necessary equipment to support the effectiveness of EALs. Contrary to these requirements, PNPS identified in CR-PNP-2016-01491 that on three past occasions (March 15 through August 8, 2012; September 4 through October 14, 2012; and June 4 through June 14, 2015) both trains of the H2O2 monitors and the Post-Accident Sampling System were unavailable to ensure the effectiveness of EAL 24, Deflagration concentrations exist inside PC, for the potential loss of the containment barrier within the Fission Product Barrier category of the EALs. This issue meets the criteria for very low safety significance (Green) because, due to other EALs, an appropriate emergency declaration could have been made in an accurate and timely manner. |
| 05000293/FIN-2016011-12 | 2017Q1 | Pilgrim | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, and shall be accomplished in accordance with those structures, procedures, and drawings. Entergy procedure EN-DC-148, Vendor Manuals and Vendor Re-Contact Process, Revision 6, requires, in part, that the station update vendor manuals every three years. Contrary to this, in July 2016, PNPS determined through a self-assessment that they had 13 vendor manuals that had not been evaluated for changes within 3 years. The NRC team determined that this finding did not affect the design or qualification of a mitigating structure, system or component; did not represent a loss of a system and/or function; did not result in loss of a train or two safety systems greater than any technical specification allowed outage time; did not result from an actual loss of safety function; and did not involve loss of any external event mitigating system. Consequently, the NRC team determined that this performance deficiency screened as having very low safety significance (Green). PNPS documented this issue in their corrective action program as CR-PNP-2016-05115. |
| 05000293/FIN-2016011-11 | 2017Q1 | Pilgrim | Failure to Adequately Develop and Implement Targeted Performance Improvement Plans | The NRC team identified a Green finding because Entergy did not adequately develop and implement a CAPR of a root cause related to a Category A CR, as required by Entergy Procedure EN-LI-102, Corrective Action Program. Specifically, Entergy did not adequately develop and implement the Targeted Performance Improvement Plans, which were designated as a CAPR for the root cause for the Nuclear Safety Culture Fundamental Problem. Entergy documented this issue in the corrective action program for further evaluation as CR-PNP-2017-00406. The performance deficiency was more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, inadequate implementation of the Targeted Performance Improvement Plans could result in recurrence of a culture in which leaders are not holding themselves and their subordinates accountable to high standards of performance, resulting in continuing performance issues at the station. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Resources, Change Management, because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. In this case, PNPS leaders did not apply sufficient rigor in development and implementation of the Targeted Performance Improvement Plans such that they would be an adequate method to drive and sustain positive changes in the stations safety culture (H.3). |
| 05000293/FIN-2016011-10 | 2017Q1 | Pilgrim | Failure to Promptly Correct a Condition Adverse to Quality for the Residual Heat Removal System | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not take timely corrective action for a previously identified condition adverse to quality. Specifically, Entergy failed to adequately resolve, through repair or adequate evaluation, gasket leakage on the B residual heat removal heat exchanger, which resulted in continued degradation and leakage for the heat exchanger gasket. Entergy did not consider this leakage as a degraded condition, with the potential to impact both the operability of the residual heat removal system, and PNPSs licensing basis with regards to leakage of a closed loop system outside of containment. After the NRC team raised the issue, Entergy performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. Entergy entered this issue into their corrective action program as CR-PNP-2016-09725. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct identified gasket leakage resulted in continued degradation and leakage of the heat exchanger gasket. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in Human Performance, Conservative Bias, because Entergy failed to use decision making practices that emphasize prudent choices over those that are simply allowable (H.14). |
| 05000293/FIN-2016011-09 | 2017Q1 | Pilgrim | Ineffective Corrective Actions to Address Conditions Adverse to Quality Regarding Components in Contact with or Close Proximity to the Drywell Liner | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with Entergys failure to correct a condition adverse to quality affecting safety-related equipment. Specifically, during a previous NRC inspection in August 2016, inspectors identified numerous locations in the drywell where non-seismic equipment was either in contact, or close proximity, with the drywell liner and had caused damage. Entergy initiated CRs and performed an operability evaluation for the identified issues. However, following a review of these CRs, the NRC team determined that Entergy failed to take corrective actions to address the condition adverse to quality. Entergy entered this issue into the corrective action program as CR-PNP-2016-09346 and CR-PNP-2016-09377 to perform an extent of condition review, secure the loose grating that had caused damage to the liner, and evaluate the need for a clearance criteria between components such as floor grating and support structures and the containment liner. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the NRC team determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the engineering evaluation of the degraded condition identified by the inspectors did not thoroughly evaluate the containment liner issues to ensure that resolutions address causes and extents of condition commensurate with their safety significance (P.2). |
| 05000293/FIN-2016011-08 | 2017Q1 | Pilgrim | Failure to Adequately Monitor the Performance of Maintenance Rule Scoped Components | The NRC team identified a Green non-cited violation of 10 CFR 50.65(a)(2), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, Entergy did not demonstrate that the performance of 18 maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, and did not establish goals and monitoring in accordance with 10 CFR 50.65(a)(1). Entergys immediate corrective action was to initiate a CR to evaluate moving the affected systems to 10 CFR 50.65(a)(1) monitoring requirements. Entergy entered this issue in the corrective action program as CR-PNP-2017-00401. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to demonstrate that the performance of the 18 maintenance rule scoped components was being effectively controlled through the performance of appropriate preventive maintenance which adversely impacts the reliability of those systems. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, in that Entergy failed to thoroughly evaluate and ensure that resolution of the identified issue, maintenance not being performed on maintenance rule scoped components, included reclassifying the components as necessary. Specifically, Entergy failed to demonstrate that the performance of Maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, or through performance goals and monitoring. (P.2). |
| 05000293/FIN-2016011-05 | 2017Q1 | Pilgrim | Failure to Establish Corrective Actions to Address Scope of Procedure Quality Issues | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy implemented inadequate corrective actions to address the procedure quality issues identified in CR-PNP-2016-02058. Specifically, Entergy inappropriately limited their corrective actions to those procedures that increased integrated risk above normal, and did not include other types of safety-related procedures that did not meet their procedure quality standards and resulted in procedure quality being a problem area. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00400. The performance deficiency was more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Entergy limited corrective actions to procedures that increased integrated risk above normal or trip sensitive and failed to include other procedures associated with safety-related components that reflected the broader population reviewed during the collective evaluation. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that this finding had a cross-cutting aspect related to Human Performance, Resources, because the leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, based on available resources, Entergy chose to limit the scope of safety-related procedures being revised to only those that resulted in high integrated risk or were trip sensitive (H.1). |
| 05000293/FIN-2016011-04 | 2017Q1 | Pilgrim | Programmatic Issue with Implementation of the Operability Determination Process | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the NRC team identified a programmatic issue because in some cases, Entergy did not enter the operability determination process when appropriate, and, when the process was entered, did not adequately document the basis for operability, in accordance with Procedure ENOP-104, Operability Determination Process, Revision 11. In each of the examples discussed, though the basis for operability was not adequate, all components were determined to be operable upon further evaluation. Entergy entered this issue into their corrective action program as CR-PNP-2017-00626. The performance deficiency was more than minor because if left uncorrected, could lead to a more significant safety issue. Specifically, the failure to enter and document a basis for operability could lead to not recognizing inoperable safety-related equipment, and place the reactor at a higher risk of core damage in a design basis accident. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Teamwork. Specifically, the operations and engineering departments did not demonstrate a strong sense of collaboration and cooperation with respect to holding each other accountable when performing operability determinations to ensure nuclear safety is maintained (H.4). |
| 05000293/FIN-2016011-03 | 2017Q1 | Pilgrim | Failure to Issue Appropriate Corrective Actions to Preclude Repetition for the Causes of the September 2016 Scram | The NRC team identified a Green finding because Entergy did not issue appropriate CAPRs in accordance with Entergy procedure EN-LI-102, Corrective Action Process, Revision 28. Specifically, Entergy did not issue adequate CAPRs associated with Root Cause 1 of the feedwater regulating valve failure in September 2016 that resulted in a manual scram. As a result of the NRC teams questions, Entergy issued procedure 1.13.2, Vendor and Technical Information Reviews, Revision 0, as continuous use to ensure that planners will always have the checklist in-hand when planning work to ensure that appropriate vendor technical information is always included in applicable work instructions. Entergy entered the NRC teams concerns in the corrective action program as CR-PNP-2017-00687 and CR-PNP-2017-00936. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a significant condition adverse to quality, loss of control of feedwater regulating valve 642A and a manual scram. The NRC team evaluated the finding using Exhibit 1, Initiating Events Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8). |
| 05000293/FIN-2016011-01 | 2017Q1 | Pilgrim | Failure to Identify All Root Causes of a Significant Condition Adverse to Quality | The NRC team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not adequately determine all root causes associated with a significant condition adverse to quality related to the failure to identify, evaluate, and correct the A SRVs failure to open upon manual actuation during a plant cooldown on February 9, 2013. Specifically, Entergy did not establish adequate measures to assure that the cause of a significant condition adverse to quality, inadequate shift manager operability determination rigor and its associated causes, were adequately determined and corrective action taken to preclude repetition. Entergys immediate corrective actions included planning to conduct operations management face-to-face conversations with shift manager qualified individuals to reinforce the shift managers responsibility for operability and functionality determination accuracy and rigor. Entergy entered this issue into the corrective action program as CRPNP-2017-00363 and CR-PNP-2017-00828. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a failure to identify, evaluate, and correct an SRVs failure to open or a similar significant condition adverse to quality. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, Entergy incorrectly assumed that CR-PNP-2013-00825 contained inadequate information to determine that the A SRV had not opened, and this assumption ultimately impacted the root cause results documented in CR-PNP-2016-01621 (H.12). |
| 05000293/FIN-2016011-06 | 2017Q1 | Pilgrim | Design Change Not Appropriately Reviewed by Entergy | The NRC team identified a preliminary greater than Green finding and apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with Entergys failure to ensure that design changes were subject to design control measures commensurate with those applied to the original design and were approved by the designated responsible organization. Specifically, Entergy received a new style right angle drive for the A emergency diesel generator radiator blower fan from a vendor but failed to adequately review the differences in the design of the drives to identify potential new failure mechanisms for the part or the need for related preventive measures. Entergy entered this issue into the corrective action program as CR-PNP-2016-07443. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team screened the finding for safety significance and determined that a detailed risk evaluation was required based on the A emergency diesel generator being inoperable for greater than the technical specification allowed outage time. Region I senior reactor analysts performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk important sequences were dominated by external fire risk. Specifically, a postulated fire in the B 4 kilovolt (KV) switchgear room with a consequential loss of the unit auxiliary generator power supply, non-recoverable loss of off-site power (LOOP) to both safety buses A5 and A6, loss of the B emergency diesel generator with the conditional failure of the A emergency diesel generator, along with the loss of bus A8 feed (from the shutdown transformer or station blackout (SBO) diesel generator) to safety buses A5 and A6. The internal event risk was dominated by weather related LOOPs, failure of the A emergency diesel generator, with failure of the B emergency diesel generator and SBO diesel generator to run, along with failure to recover offsite power or the emergency diesel generators. See Attachment 1, A Emergency Diesel Generator Cooling Water System Degradation Detailed Risk Evaluation, for a detailed review of the quantitative criteria considered in the preliminary risk determination. The NRC team did not assign a cross-cutting aspect to this finding because the performance deficiency occurred in May 2000. Entergys program has undergone changes since May 2000, and the NRC team did not identify any recent examples of this performance deficiency. Other aspects of Entergys performance related to this issue are further discussed in Sections 5.10.3 and 6.3.4. |
| 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. |
| 05000255/FIN-2017001-01 | 2017Q1 | Palisades | Licensee-Identified Violation | The licensee Identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix R, Section III.G.2, which requires, in part, that where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shut down conditions are located within the same fire area outside of primary containment, one means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to the above, as of October 1, 2010, the licensee failed to ensure that one of the redundant trains was free of fire damage in areas where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment. Specifically, the licensee failed to analyze a fire scenario in the 1C switch gear room, screen- house room, and component cooling water pump room that could potentially damage the control cable before the load cable, and therefor e result in the loss of safety -related 2400 volt alternating current (VAC) bus 1C and/or 1D, with subsequent loss of equipment credited for Appendix R compliance to support safe shutdown in the event of such a fire. The licensees failure to analyze an Appendix R fire scenario for the three fire areas described above w as a performance deficiency . 21 The performance deficiency was more- than- minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) 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). The finding was determined to be of very low safety significance (Green) because it did not impact the licensees ability to reach hot shutdown because operator manual actions would have allowed operators to shut down the plant following a fire. The licensee identified this issue during the transition to NFPA 805, entered the issue into their CAP as CR PLP 2010 04255, and implemented compensatory measures, including fire watches. The violation was not willful and routine licensee efforts, such as normal surveillance or quality assurance activities, were not likely to have previously identified the violation due to the specific sequence of fire cable damage required for such an Appendix R fire scenario. As a result, the inspectors concluded that the violation met all four criteria for exercising enforcement discretion established by Section 9.1 of the NRCs Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues; therefore, the NRC is exercising enforcement discretion to not cite this violation |
| 05000293/FIN-2016011-02 | 2017Q1 | Pilgrim | Failure to Establish Corrective Actions to Preclude Repetition of a Significant Condition Adverse to Quality | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not implement CAPRs for a significant condition adverse to quality identified in root cause evaluation CR-PNP-2016-00716, Implementation of the Corrective Action Program, Revision 2. Specifically, the team identified that CAPRs for Entergys continued weaknesses in the implementation of the corrective action program were inadequate. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00053, CR-PNP-2017-00410, and CR-PNP-2017-01134. The performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to preclude repetition of this significant condition adverse to quality could result in continuing weaknesses in implementation of the corrective action program, which was designated as a fundamental problem, and thus a contributing factor for PNPS Column 4 performance. Additionally, weaknesses with corrective action program implementation could result in equipment issues where operability is not maintained. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8). |
| 05000255/FIN-2016004-03 | 2016Q4 | Palisades | Failure to Translate Design Analysis Stack-up Configuration into Specifications, Drawings, Procedures, and Instructions | Green. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to establish measures to assure that the applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to provide instructions in procedures to construct the spent fuel dry cask loading stack-up, in the safety-related auxiliary building, in the configuration that had been analyzed for in the stack-up seismic design basis calculation. In addition, the licensee failed to provide instructions in revised procedures to construct the stack-up without certain gaps as 4 specified in the stack-up seismic design basis document. The licensee documented these issues in their CAP as CRPLP201600646, CRPLP201601308, CRPLP201601558, CRPLP201604497, and CRPLP201604826; revised the stack-up seismic analysis to address the identified issues; and translated the analyzed stack-up design configuration into stack-up installation procedures prior to performing stack-up operations with spent nuclear fuel in the multi-purpose canister. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Design Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a stack-up configuration that did not ensure stack-up dynamic stability or Auxiliary Building structural integrity to maintain radiological barrier functionality during a design basis seismic event. The finding screened as having very low safety significance (Green) because it did not result in the loss of operability or functionality of the Auxiliary Building. The finding had a cross-cutting aspect of Field Presence in the Human Performance cross-cutting area, because licensee senior managers failed to ensure effective supervisory and management oversight of contractor activities related to the seismic analysis and installation of the stack-up configuration (H.2). |