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05000397/FIN-2018003-0330 September 2018 23:59:59ColumbiaFailure to Adequately Control Work Hours for Covered PersonnelThe inspectors identified a Green, non-cited violation of 10 CFR 26.205 associated with the licensees failure to adequately schedule and control work hours for personnel subject to work hour controls. Specifically, the licensee failed to appropriately schedule and control work hours for at least three Chemistry Technicians who were providing covered work as the designated Emergency Response Organization (ERO) Duty Chemistry Technician as defined by the Columbia Generating Station Emergency Plan.
05000373/FIN-2018003-0630 September 2018 23:59:59LaSallePotential Failure to Inspect Containment Post-Tensioned Tendons per Code Requirements and to Follow Corrective Action Program ProcessVertical and horizontal post tensioned tendons, along with reinforcing steel, are required to maintain structural integrity of the primary containment. There are a total of 120 vertical post tensioned tendons along the periphery of the primary containment wall, including 60 Group C tendons and 30 each of Groups A and B. Section 5.5.6 of the Technical Specifications describes the Inservice Inspection (ISI) program for post tensioning tendons and states that the Tendon Surveillance Program shall be in accordance with ASME Section XI, Subsection IWL as required by 10 CFR 50.55a. One Group B tendon (V213B) on Unit 1 was inspected in 1999 and according to the inspection records, water was identified on all components of the tendon. No presence of water is one of the acceptance criteria per Subsection IWL of the ASME Section XI. No condition report was found for this adverse condition. Subsequently, a condition involving degraded vertical tendons was identified during inspections in 2003 and documented in AR 157920. The degradation consisted of broken wires. The Root Cause Report (RCR) for this condition noted that 11 Group A tendons were found degraded and water induced corrosion was the root cause for tendon degradation. The evaluation concluded that five Group A tendons and all 30 Group B tendons in each unit were not susceptible to water intrusion because they were protected by welded covers. These tendons with welded covers were also determined to be inaccessible, and therefore exempt from future inspections requirements in accordance with provisions of ASME Section XI, IWL. The RCR did not address the condition of water found during the Group B tendon inspection in 1999. Additionally, to verify this assumption of welded covers providing protection from water intrusion, a corrective action was generated to inspect one Group A and one Group B inaccessible tendons during the next outage. Pertaining to inaccessible tendons, the inspectors noted the following requirements of 10 CFR 50.55a(b)(2)(viii)(E): Concrete containment examinations: Fifth provision. For Class CC applications, the applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or the result in degradation to such inaccessible areas. For each inaccessible area identified, the applicant or licensee must provide the following in the ISI Summary Report required by IWA6000: (1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation; and (3) A description of necessary corrective actions. After the licensee identified degraded group A tendon locations, to comply with the provision of 10 CFR 50.55a, the licensee documented in its 90 day post outage ISI reports information on the degraded A tendons in 2004 and 2005 for units 1 and 2, respectively. This information included an assumption that the extent of degradation did not apply to the Group B tendon locations because of a welded cover at locations that precluded entry of water. Additionally, a corrective action, CA 15792033, was generated to inspect one Group A and one Group B tendon during the next refueling outage to verify this assumption. The corrective action was closed without inspection of any Group B tendon based on a management decision following satisfactory inspection of a Group A tendon in 2006. The licensees decision failed to take into account the fact that the most recent inspection of a Group B tendon showed presence of water on tendon components and also that the welded closure details were different for tendons in the two groups. Subsequently, the licensee identified a concern regarding inadequate closure of this corrective action during its reviews for the license renewal application in 2014. Specifically, the licensee wrote AR 1658189 to document that due to the differences in the welded cover designs, the results of the Group A tendon inspection may not be applicable to Group B tendons. Therefore the critical assumption regarding the adequacy of Group B tendon covers remained unverified. In particular, the Group B tendon cover used dissimilar metal welds and water was found inside the cover during the most recent inspection. The licensee identified actions to perform inspections on two of the Group B tendons on each unit in addition to inspecting the tendon V213B where water was initially found. These actions were categorized as action tracking items (ACITs), items that do not represent conditions adverse to quality. Since water was found on all tendon components during the last inspection of a Group B tendon, and water induced corrosion was found to be the root cause of many tendon failures, the assumption in the RCR that the welded covers would prevent water intrusion needed to be validated through inspections. This unresolved item remains open pending additional inspector review of the issue with respect to regulatory requirements. Planned Closure Action: Inspectors will seek additional information from the licensee and the NRC will perform internal review to evaluate compliance with NRC regulations.
05000374/FIN-2018003-0730 September 2018 23:59:59LaSallePotential Failure to Promptly Correct the Unit 2 Primary Containment Wall Cavity Leakage Condition and to Follow Corrective Action Program ProcessCondition description in AR 2420888 indicated that leakage through the Unit 2 primary containment wall has been a longstanding open issue. The leak was initially identified in 1998 when water leakage was noticed on the external side of the primary containment wall. The leakage was approximately 2025 drops per minute at the primary location and multiple areas near the 180 degree azimuth at construction joints on elevations 813 and 795. Another minor leak was noticed at a similar location near the 0 degrees azimuth. The condition was documented in AR 2269. The source of water leakage was determined to be a weld on a 2 fuel pool cooling drain line and work order 98109950 was initiated to repair the weld. The work was not scheduled and the work order was eventually cancelled. In 2010, the leakage was documented again in AR 1086083. A technical evaluation documented as ATI14709531847 in 2014 concluded that there was no adverse impact on structural adequacy of the containment. The technical evaluation stated that the leakage was to be repaired in the upcoming outage through work order 855785. Action request 2420888 was written in December 2014 to re-enter the condition in the CAP. It recommended corrective actions for liner ultrasound testing every other refueling outage, completion of weld repair, and performance of a technical evaluation for structural impact on the concrete, reinforcing steel, tendons, and liner. The technical evaluation assignment was closed to the evaluation documented under ATI14709531847 discussed above. The corrective action assignment for the weld repair was closed to a work order which has not been completed to-date. Based on the inspectors review, the licensee has deferred the actions to correct this condition identified in 1998. The inspectors question whether the continuous leakage could lead to deterioration of the concrete, corrosion of the reinforcement, or degradation of post tensioned tendons if it enters the tendon sheath or trumpet area; and therefore a condition adverse to quality. This unresolved item remains open pending additional inspector review of the issue with respect to regulatory requirements. Planned Closure Action: Inspectors will seek additional information from the licensee and the NRC will perform internal reviews to evaluate compliance with NRC regulations
05000373/FIN-2018003-0230 September 2018 23:59:59LaSalleFailure to Establish an Appropriate Inservice Testing ProcedureThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe procedures that were appropriate to the circumstances, for activities affecting quality, that included appropriate quantitative or qualitative acceptance criteria for determining that important activities had been satisfactorily accomplished. Specifically, the CSCS bypass line isolation valve IST procedure did not contain acceptance criteria to verify the necessary valve obturator movement.
05000373/FIN-2018201-0130 September 2018 23:59:59LaSalleSecurity
05000373/FIN-2018003-0830 September 2018 23:59:59LaSalleFailure to Implement Engineering Change Results in Reactor Coolant Boundary LeakageThe inspectors documented a self-revealed finding of very low safety significance (Green) and associated NCVs of TS 5.4.1 Procedures, and TS 3.4.5 for the failure to implement EC 354539 to perform the final piping weld for the 1B33F067B bonnet vent line in the field, resulting in pressure boundary leakage when the weld failed at power.
05000373/FIN-2018003-0330 September 2018 23:59:59LaSalleFailure to Establish Goals to Monitor Steam Tunnel Check DampersIntroduction: The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.65(a)(1) for the licensees failure to establish goals to monitor the performance of steam tunnel check dampers. Specifically, the licensees goals for functional failure and condition monitoring could always be satisfied given a two years monitoring period with only one testing opportunity.
05000397/FIN-2018003-0130 September 2018 23:59:59ColumbiaFailure to Follow Radiologically Controlled Area ProceduresThe inspectors reviewed a self-revealed Green, non-cited violation of Technical Specification 5.4.1(a) when the licensee failed to implement radiation control procedures. On June 1, 2017, a supplemental health physics technician (HPT) entered a posted locked high radiation area without a functioning electronic dosimeter (ED). Although the area was posted as a locked high radiation area (LHRA), there were no measured dose rates in excess of 1 rem per hour during this entry. The HPT logged on to Radiation Work Permit (RWP) 30003852 and entered the radiologically controlled area (RCA) to cover a job to add additional shielding in the travelling in-core probe (TIP) Mezzanine room. The HPT entered the RCA through the HP swing gate near the RCA exit point, in order to obtain survey instruments for the job coverage. The HPT proceeded to the dress out area and then to the TIP Mezzanine room, where he entered with a survey meter. After about 10 minutes in the room, the HPT looked at his ED and noticed that it was in pause mode (i.e., not functioning). The HPT informed the worker he was covering, and they both left the LHRA. During the RWP logging process, there was an error when the HPTs ED was being programmed that went unnoticed. As a result, the HPT was signed-on to the RWP, but the ED was not programmed and active. Because the HPT used the HP swing gate at the RCA exit rather than the normal access point with electronic turnstiles that verify ED function, this errant condition was not identified. The swing gate used was intended for HPTs assigned to assist workers with contamination alarms at the RCA exit, not as an RCA entry point to perform work or cover a job. Licensee Procedure GEN-RPP-04, Entry Into, Conduct In, and Exit From Radiologically Controlled Areas, Section 4.13 Dosimetry and Log-in, paragraph (e), requires workers to ensure that electronic dosimetry is on immediately before entering the RCA. The HPT neither used the electronic turnstiles nor checked to see if the ED was on prior to entering the RCA.Additionally, licensee Procedure 11.2.7.3 High Radiation Area, Locked High Radiation Area, and Very High Radiation Area Controls, Section 3.2.4 Coverage and Monitoring of Work, paragraph (d), describes conducting a peer-check prior to LHRA entries, by the job coverage HPT, to verify that workers are wearing an active ED (i.e., not in pause mode) in the appropriate location on the body. The job coverage HPT checked to see that workers had an ED appropriately placed, but did not check the ED setpoints or if the ED was active.Multiple barriers that could have prevented this situation from occurring were either ineffective or not used. Had the error reduction/prevention measures been used, the ED programming error during RWP log on would have been identified.Corrective Action(s): An immediate corrective action, in addition to the HPT being restricted from the RCA, was a stand down conducted with radiation protection personnel about this incident and coaching on use of the procedures related to the verification of dosimetry and peer-checking prior to entry into LHRAs. Corrective Action Reference: AR 00366701
05000373/FIN-2018003-0430 September 2018 23:59:59LaSalleFailure to Manage the Increase in Risk During a Battery Charger Capacity TesThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.65(a)(4) for the failure to manage risk when the licensee failed to adhere to procedure WCAA101, Revision 28, On-line Work Control Process. Specifically, procedural requirements regarding a dedicated operator for manual restoration actions and written instructions to credit the availability of the A RHRSW pump during the battery charger testing were not met.
05000410/FIN-2018003-0130 September 2018 23:59:59Nine Mile PointFailure to Ensure that Thermal Power is Less Than or Equal to the Licensed Power LimitThe inspectors identified a Green finding and associated non-cited violation (NCV) of the NMPNS Unit 2 Operating License (NPF-69), Condition 2.C(1), Maximum Power Level, when Exelon did not ensure that thermal power was less than or equal to the licensed power limit of 3988 megawatts-thermal (MWth). Specifically, on multiple occurrences between May 22, 2018 and October 19, 2018, licensed operators in the main control room did not appropriately monitor and control 2-hour average thermal power at or below the licensed power limit. The inspectors determined the 2-hour average thermal power exceeded the licensed power limit outside of normal steady-state fluctuations, and did not take timely, effective corrective action to reduce thermal power below the licensed power limit when the 2-hour average was found to exceed the licensed power limit
05000397/FIN-2018003-0230 September 2018 23:59:59ColumbiaFailure to Control Workers in a High Radiation Area (>1.0 rem per hour)The inspectors reviewed a self-revealed Green, non-cited violation of Technical Specification (TS) 5.7.2(b) and (e) when the licensee failed to control worker activities in a locked high radiation area in accordance with the requirements of the RWP and failed to determine radiological conditions in the work area prior to the start of work.
05000373/FIN-2018003-0530 September 2018 23:59:59LaSalleFailure to Translate Fuel Oil Relief Valve Setting into Design Drawing of Record.The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to accurately translate the Division III EDG fuel oil relief valve set point from the design drawing of record, VPF341110, to the fuel oil pressure operator rounds alert value in the Division III EDG operating procedures.
05000410/FIN-2018003-0230 September 2018 23:59:59Nine Mile PointMinor ViolationDuring the review of Licensee Event Report (LER) 05000220/2017-002-01, Manual Reactor Scram Due to Presesure Oscillations, the inspectors identified a minor violation of 10 CFR 50.9, Completeness and accuracy of information. The LER was found to be inaccurate. Specifically, the LER timeline contained inaccurancies regarding the time operators entered a special operating procedure and did not include an actuation of high-pressure coolant injection (HPCI). The timeline stated at 2:10 AM operators entered the special operating procedure for Pressure Regulator Malfunction, due to reactor pressure oscillations of 2-3 psig. At 2:27 AM operators inserted a manual scram of the reactor due to pressure oscillations exceeding procedural limits. This information was confirmed by a review of the operational logs for March 20, 2017. During OI Investigation 1-2018-002, it was determined that this entry was not accurate and although an exact time could not be established is was estimated to have been at 2:20 AM vice 2:10 AM. Additionally the timeline did not include a mention that at 2:16 AM unexpected turbine trip signal was received and HPCI was initiated due to a tagging error. Operators reset HPCI at 2:18 AM and restored main feedwater flow to restore Reactor Vessel water level. A sixty day telephone notification instead of a written licensee event report was conducted for this invalid initiation of HPCI was completed on May, 11, 2017, as EN 52747 as allowed by 10 CFR 50.73(a)(2)(iv). Screening: Violations involving the submittal of inaccrurate or incomplete information are evaluated under Traditional Enforecement because they impact the NRCs regulatory process. Accordingly, the inspectors evlauted this issue against the example violations in Section 6.9 of the NRC Enforcement Policy. Inspectors concluded that the violation is of minor safety significance because the inaccurate information did not change the NRCs review of the licensee event report. Enforcement: 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 22, 2015, Entergy provided information to the Commission that was not complete and accurate in all material respects. In the licensee event report, Exelon documented incorrect information that resulted in the NRC launching a substation further inquiry (OI investigation), but did not substantiate that licensed operators deliberately failed to follow a Technical Specifications required procedure. Exelon identified the inaccuracy and entered the issue into the corrective action program (IR 04091110) on January 7, 2018, and submitted LER 05000220/2017-002-01 on August 18, 2018, revising the timeline to show operators entering N1-SOP-31.2 at 2:20 AM vice 2:10 AM. The disposition of this violation closed Licensee Event Report 05000220/2017-002-01
05000373/FIN-2018003-0130 September 2018 23:59:59LaSalleFailure to Establish Heat Exchanger Inspection Procedures Appropriate for the CircumstancesThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions Procedures, and Drawings, for the licensees failure to ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances. Specifically, the licensee failed to ensure that procedure ERAA3401002 appropriately accounted for partially blocked HX tubes identified during HX inspections.
05000373/FIN-2018002-0130 June 2018 23:59:59LaSalleFailure to Implement a Preventative Maintenance Strategy for Residual Heat Removal Service Water Pump Shorting RelaysA self-revealed Green finding of very low safety significance was identified for the licensees failure to implement a preventative maintenance (PM) strategy for the residual heat removal service water (RHRSW) pump shorting relays in accordance with procedure MAAA716210, Performance Centered Maintenance (PCM) Process, Revision 11. Specifically, a PCM template was issued in 2002 that required periodic as-found testing and calibration for control and timing relays, but a maintenance strategy was never implemented. As a result, one of the normally closed contacts on the Unit 1 D RHRSW pump shorting relay developed a high contact resistance and prevented the Unit 1 D RHRSW pump from starting.
05000410/FIN-2018002-0130 June 2018 23:59:59Nine Mile PointFailure to Ensure Proper Control of the Standby Gas Treatment System Damper Valve, 2GTS*V2000B, Within Procedures, Materials, and Design Control MeasuresThe inspectors identified a Green finding and associated NCVof 10 CFRPart 50, Appendix B, Criterion III, Design Control, when Exelon failed to ensure proper control of the SGTS damper valve 2GTS*V2000B within procedures, materials, and design control measures. Specifically, on April 15, 2018 operators attempted to run B SGTS for containment purge; however, no flow was observed and the system was secured. Operators discovered the 2GTS*V2000B closed due to the failure of the operating mechanism to maintain control of the valve position.
05000373/FIN-2018002-0230 June 2018 23:59:59LaSalleFailure to Follow Procedure and Perform Database Revision Review RequirementsThe inspectors identified a Green finding of very low safety significance for the licensees failure to follow procedure NSWPWM03, Predefine Database Revisions, Revision 0, for retiring procedure LESGM108, Inspection of 480V Motor Control Center Equipment, that performed bus bar inspection on Division 3 motor control centers. Specifically, instead of completing NSWPMW03, step 6.5, Database Revision Review Requirements, to retire the bus bar inspections for Division 3 motor control centers, the licensee retired the procedure based solely on having previously retiring the bus bar inspections for Division 1 and Division 2 in 2002,and did not performthe required review.
05000397/FIN-2018002-0130 June 2018 23:59:59ColumbiaFailure to Maintain Configuration Control in the Diesel Generator 2 Diesel Cooling Water SystemThe inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to perform maintenance in accordance with written procedures appropriate to the circumstances. Specifically, on April 9, 2018, the licensee inadvertently bumped and partially opened a diesel cooling water valve, DCW-V-8B2, while operating a nearby demineralized water valve, DW-V-14B2, as part of work activities under Work Request (WR) 29127677, and rendered diesel generator 2 inoperable and unavailable.
05000220/FIN-2018002-0230 June 2018 23:59:59Nine Mile PointInadequate Procedure Causes Water Hammer Condition Resulting in Isolation and Inoperability of the 12 Train of the Emergency Condenser SystemThe inspectors identified a Green finding and associated NCVof 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, when Exelon did not provide appropriate quantitative or qualitative criteria and guidance to operators in procedure N1- OP- 13 Emergency Cooling System to return an emergency condenser loop to service without inducing a water hammer condition which caused operators to re-isolate the emergency condenser loop and declare it inoperable
05000373/FIN-2018002-0430 June 2018 23:59:59LaSalleMinor Violation - Follow-up of Events and Notices of Enforcement Discretion

Minor Violation: For S/RV 2B21F013L, serial number N63790050012 (hereafter referred to as S/RV 12), the licensee completed a work group evaluation as documented in AR 03975216ACIT No. 3 to investigate the cause for two S/RVs that failed a set pressure lift test out of specification low. For ACIT No. 3, the licensee staff incorporated a vendor letter that documented the results of the S/RV vendors review of the S/RV 12 condition and which recorded an out of tolerance spring condition. It stated that The spring was measured and rate tested. The free height was found to be below the minimum original equipment manufacturer specified tolerance. The licensees vendor subsequently replaced the nonconforming spring with a new spring. In prior vendor correspondence with the licensee (reference E-mail dated June 24, 2015), the vendor stated that Typically we contribute a low as-found lift to an out-of-tolerance spring rate or free height dimension. Therefore, the nonconforming spring free height dimension may have caused the low as-found lift setpoint failure for this valve and as such was relevant (e.g. material) to the determination of a failure cause that was reported in LER 05000374/201700400 and 01. However, the licensee failed to identify this during their cause investigation and erroneously reported in LER 05000374/201700400 and 01 that The vendor reported for both valves that all the spring tolerances were within the acceptance limits. The licensee documented this violation in AR 04134591, Potential Minor Violation for Unit 2 LER 20170401. The licensee also submitted a revision to the LER as LER 05000374/201700402

Screening: The significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which could impede the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors determined that this issue was a Severity Level IV violation based on Example 6.9.d.10 in the NRC Enforcement Policy which states, A failure to identify all applicable reporting codes on a Licensee Event Report that may impact the completeness or accuracy of other information (e.g. performance indicator data) submitted to the NRC. In accordance with the Section 2.2.1.c of the NRC enforcement policy, the severity level of a violation involving the failure to make a required report to the NRC will depend on the significance of and the circumstances surrounding the matter that should have been reported. The NRC had not relied on information in this LER report to make a regulatory decision, and the inspector answered no to each of the more than minor screening questions in Appendix B of IMC 0612 for the issue of concern. Therefore, the NRC determined this was a minor violation because it was associated with a minor performance deficiency. Violation: Failure to comply with 10 CFR 50.9 Completeness and accuracy of information and accurately report the nonconforming S/RV 12 spring tolerance in LER 05000374/201700400 and 01 to the NRC constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000373/FIN-2018010-0130 June 2018 23:59:59LaSalleFailure to Translate Reactor Building Superstructure Design BasisInspectors identified a Green finding and associated Non-Cited Violation of Title 10of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, for licensees failure to assure that applicable Updated Final Safety Analysis Report described design basis for the Reactor Building (RB) superstructure were correctly translated to field documents was a performance deficiency. Specifically, Updated Final Safety Analysis Report Tables 3.8-9 and 3.8-11 define the design basis load combinations and the corresponding design stress limits applicable to the RB superstructure. Design calculation L-003415 evaluates these load combinations and applies RB overhead crane lifting limitations which ensures these design basis are met. The licensee failed to translate these limitations into specifications, drawings, procedures, or instructions which would ensure the specified stress limits for RB design basis load combinations would not be exceeded while operating the RB overhead crane.
05000397/FIN-2018401-0231 March 2018 23:59:59ColumbiaSecurity
05000220/FIN-2018001-0131 March 2018 23:59:59Nine Mile PointPotential Failure to Submit an 8-Hour Event Notification for a Valid Actuation of HPCOn March 18, 2018,at 1:18 a.m., during the Unit 1maintenance outage while the unit was in cold shutdown, operators received multiple low level alarms on the GEMAC 11 and 12 level indications. Operators responded by adjusting reactor water cleanup reject flow and the feedwater minimum flow control valve to raise reactor water level. Upon the operators making the adjustment to reactor water level, the feedwater low flow control valve was slow to respond, but eventually opened more rapidly, and the increased flow from feedwater resulted in a rapid rise in reactor water level. At 1:28 a.m., indicated reactor water level rose to the 95-inch trip setpoint for the 11 and 12 Yarway level indication instruments, resulting in a turbine trip and HPCI initiation signal. The HPCI pumps were tagged out and thus did not inject, and the turbine was offline for the shutdown. The 11 and 12 Yarway level indication instruments provide reactor protection system logic inputs for reactor vessel water level; however, the Yarway level indication instruments are not density compensated. Therefore, under cold shutdown conditions, actual reactor vessel water level was lower than indicated water level on the 11 and 12 Yarways. During cold shutdown conditions, the GEMAC level instruments, which are calibrated to cold shutdown conditions, provide an accurate indication of actual reactor vessel water level. The GEMAC instruments both indicated well below the trip setpoint of 95 inches (indicated ~72 inches) when the turbine trip and HPCI initiation signal were received. Exelon determined that this event was not reportable under 10 CFR 50.72.Title 10 CFR 50.72(b)(3)(iv)(A) states, Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. (B) The systems to which the requirements of paragraph (b)(3)(iv)(A) of this section apply are: 10 (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system. Planned Closure Action(s): The inspectors requested the 10 CFR 50.72 subject matter experts at the Office of Nuclear Reactor Regulation (NRR) and Office of General Council (OGC) to review whether this was a valid actuation and thus reportable. The inspectors are opening an unresolved item (URI) to determine if a performance deficiency exists.Licensee Action(s): Licensee entered the concern into their corrective action program, and communicated with NRC Region I and NRR Staff. Exelons position is that the event was not reportable. Corrective Action Reference:IR 04116336 NRC Tracking Number: 05000220/2018001-01
05000410/FIN-2018001-0231 March 2018 23:59:59Nine Mile PointPotential Inadequate 50.59 Evaluation for TS 3.3.1.1 Bases ChangeOn February 23, 2018, Exelon personnel performed a 50.59 Screening for a change to Unit 2 TS Bases 3.3.1.1, Reactor Protection System (RPS) Instrumentation, for MSIV and TSV surveillance testing. Exelon personnel performed this activity to address operating experience associated with the use of a test box that prevents a scram signal during RPS surveillance testing for TS 3.3.1.1 Function 5 MSIV Closure and Function 8 TSV Closure. TS Bases B 3.3.1.1, C.1, Revision 1 was revised to state, in part, For Function 5 (MSIV Closure), this would require both trip systems to have at least one channel associated with the MSIVs for each main steam line in one Trip Logic Channel (not necessarily the same main steam lines for both trip systems), Operable or in trip (or the associated trip system in trip). For Function 8 (Turbine Stop Valve Closure), this would require both trip systems to have the channels for one Trip Logic Channel, Operable or in trip (or the associated trip system in trip).The inspectors questioned whether the change to TS Bases B 3.3.1.1 resulted in a change to the implementation of TS 3.3.1.1. A licensee is permitted to make changes to their Technical Specification Bases documents without NRC review and approval. However, in certain cases, such as a change to the Technical Specification Bases that would change how the associated Technical Specification is applied, NRC review and approval would be required.Planned Closure Action(s): The inspectors sent written questions to request assistance from NRR to determine whether this change to the TS Bases reasonably would have required NRC review and approval. The inspectors are opening a URI to determine if this is violation of 10 CFR 50.59 and if it is more than minor. Licensee Action: Documented NRCs concern as AR 04055602. Exelons position is the change would not affect how TS 3.3.1.1, or its note, is applied and therefore NRC review was not required.Corrective Action Reference: AR04055602 NRC Tracking Number: 05000410/2018001-02
05000397/FIN-2018401-0131 March 2018 23:59:59ColumbiaSecurity
05000373/FIN-2018001-0131 March 2018 23:59:59LaSallePost-Maintenance Testing Failed to Demonstrate Testable Check Valve FunctionA self-revealed Green finding of very low safety significance and an associated Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion XI, Test Control, was documented by the inspectors for the licensees failure to perform post-maintenance testing that would demonstrate that structures, systems and components (SSCs) would perform satisfactorily in service. Specifically, following maintenance on the Unit 2 B residual heat removal (RHR) shutdown cooling (SDC) return testable check valve, 2E12F050B, and the Unit 1 A RHR SDC return testable check valve, 1E12F050A, the post maintenance test performed failed to identify that they would not open fully when in service, resulting in the valves being unable to pass full flow during SDC mode of RHR operation.
05000373/FIN-2018001-0331 March 2018 23:59:59LaSalleEnforcement Action (EA) 18035: Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection NoncomplianceOn 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 noncompliance 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 an 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. The EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed 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 provided 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 approximately 60 days of issue discovery. The comprehensive measures should remain 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. Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants (GDC), Criterion 4, Environmental and Dynamic Effects Design Basis, states in part that SSCs important to safety shall be adequately protected against dynamic effects including missiles. On February 15, 2018, during evaluation of protection for Technical Specifications (TS) equipment from the damaging effects of tornado generated missiles, LaSalle County Station identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from tornado generated missiles. Specifically, tornado generated missiles could strike the components supporting the operation of Control Room (VC) and Auxiliary Electric Room (VE) ventilation. This could result in inoperable VC/VE systems, which provide a protected environment for occupants to control the unit following an uncontrolled release of radioactivity, hazardous chemicals, or smoke if a tornado were to occur. In addition, the Unit 2 Division 2 motor control center (MCC) 236X1 was affected, which impacted various loads on Unit 2 including the Unit 2 standby gas treatment, Unit 2 Division 2 post LOCA system, B main control room area filtration system supply and exhaust fan, reactor building Division 2 isolation damper control logic, Unit 2 Division 2 battery room exhaust fan and Unit 2 24/48 Volt battery rooms exhaust fans. This would result in a loss of power to components and systems rendering them inoperable. The condition was reported to the NRC in Event Notice (EN) 53213 as an unanalyzed condition and potential loss of safety function. Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS Limiting Conditions for Operation (LCOs) in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of the implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. Initial (immediate) compensatory measures were established by an operations standing order that included: Procedures were verified to be put in place, with associated current training, for performing actions in response to a tornado. Procedures were verified to be put in place, with associated current training, for actions to be taken if a tornado watch is issued for the area. Procedures were verified to be put in place, with associated current training, for actions to be taken if a tornado warning is issued for the area. Verification that training was up to date for individuals responsible for implementing preparation and response procedures; and Established a heightened station awareness and preparedness level relative to identified tornado missile vulnerabilities. The comprehensive (60 day) compensatory measures were established by incorporating the standing order actions and adding additional detail to operating procedure LOATORN001, High Winds/Tornado, Revision 22, for completing additional inspections and restoration actions on equipment vulnerable to tornado missile damage. Corrective Action Program References: AR 4104401; AR 4104391; AR 4104393; AR 4104396; AR 4104397. Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.7.4, Control Room Area Filtration (CRAF) System; TS 3.7.5, Control Room Area Ventilation Air Conditioning (AC); TS 3.6.4.2, Secondary Containment Isolation Valves (SCIVs); TS 3.6.4.3; Standby Gas Treatment (SGT) System; and TS 3.8.7, Distribution SystemsOperating. 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 Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS, Accession No. ML16355A286. 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 actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions were to remain in place until permanent repairs were completed, which for LaSalle were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed
05000373/FIN-2018001-0231 March 2018 23:59:59LaSalleFailure to Update Throttle Valve Position in Accordance with Station ProceduresThe inspectors identified a Green finding of very low safety significance and an associated Non-Cited Violation (NCV) of LaSalle Technical Specifications 5.4.1, Procedures, for the licensees failure to implement station procedures recommended in Regulatory Guide 1.33, Appendix A, Section 9. Specifically, on two separate occasions while performing a flow balance on the Unit 1 A diesel generator (DG) cooling water system, procedural errors resulted in the licensee failing to update the throttle valve position to be used during manual backwash of the Unit 1 A DG cooling water strainer with the correct position.
05000397/FIN-2018001-0131 March 2018 23:59:59ColumbiaFailure to Follow Procedure Leads to Loss of Secondary ContainmentThe inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to perform maintenance in accordance with documented instructions appropriate to the circumstances. Specifically, on September 12, 2017, the failure to verify power sources per Work Order 02072924 caused an electrical transient that caused the reactor building exhaust valve and supply valve to lose power and close, resulting in a loss of secondary containment
05000220/FIN-2017004-0231 December 2017 23:59:59Nine Mile PointInadequate Fill and VentProcedure for Control Room Chiller Results in Unplanned LCO EntryAn NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for Exelons failure to ensure that activities affecting quality were prescribed in a manner appropriate to the circumstances for the Unit 1 control room chiller system. Specifically, Exelon procedure N1-OP-49, Control Room Ventilation System, Revision 03800, Section H.5, Venting of Control Room Chiller Circulating Water Pump 11 and 12 Discharge Piping, led personnel to inadequately fill and vent the 12 control room chiller during system restoration from maintenance, while in a single chiller lineup. As a result, on October 15, 2017, control room chiller 12 tripped on low flow, and due to a prior trip of 11A control room chiller compressor, an unplanned 7-day LCO in accordance with TS 3.4.5.e, Control Room Air Treatment System, was entered due to an insufficient number of available chiller compressors to provide adequate control room cooling. Exelon entered this issue into the CAP as IR 04090200. Corrective actions included generating a procedure change to correct N1-OP-49 Section H.5, which provides instruction for filling and venting when in a single chiller lineup This finding is more than minor because it is associated with the procedure quality 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, Exelon failed to prescribe an adequate fill and vent procedure for the Unit 1 control room chillers which led to a trip of the 12 chiller on low flow while troubleshooting of chiller compressor 11A was on-going, resulting in an unplanned TS LCO entry. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The performance deficiency did not represent a degradation of the radiological barrier function provided for the control room. Additionally, the performance deficiency did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. Therefore, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because between 2014 and 2017 the inspectors noted over 20 issue reports documenting issues affecting reliability of the control room chiller system. Exelon failed to thoroughly evaluate the issues associated with the chillers to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Exelon failed to effectively evaluate previous chiller trips and to prevent additional trips of the chiller system such as the one that occurred on October 15, 2017. (P.2) (Section 1R12.b.2)
05000373/FIN-2017004-0431 December 2017 23:59:59LaSalleFailure of Offsite Power Backfeed Procedure to be Appropriate to the Circumstances Caused Unit 1 ScramA finding of very low safety significance and an associated NCV of LaSalle Technical Specification (TS) 5.4.1, Procedures, occurred on February 13, 2017, for the stations failure to maintain instructions of a type appropriate to the circumstances for energizing offsite electrical systems during a Unit 2 backfeed evolution (an activity affecting quality per Regulatory Guide 1.33). Specifically, the steps of backfeed procedure, LOPAP01, Revision 35, led to a Unit 1 scram because the prescribed switchyard configuration left both units connected to the 345 kilovolt (kv) ring bus, leaving the operating unit susceptible to the large in-rush current induced by the backfeed energization of the Unit 2 main power transformer. As a corrective action from Action Request (AR) 03973724, the licensee revised the backfeed procedure to eliminate the tie between the units on the ring bus when main power transformers are energized. This performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events Cornerstone, and adversely affected the Cornerstone objective of limiting the likelihood of events that upset plant stability because it resulted in a Unit 1 Scram. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, issued June 19, 2012, the inspectors determined that this finding was of very low safety significance because, although the performance deficiency caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the scram to a stable shutdown condition. The inspectors determined there was no cross-cutting aspect because the performance deficiency was not indicative of licensees current performance since the design modification occurred greater than 3 years before the event. This inspection report will also bring to closure the associated Licensee Event Report, (LER) 05000373/201700300.
05000220/FIN-2017004-0131 December 2017 23:59:59Nine Mile PointMain Control Room Annunciators 10 CFR 50.65(a)(2) Demonstration Not MetAn NRC-identified Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65 (a)(2), was identified because Exelon did not adequately demonstrate that the performance of the Unit 1 main control room (MCR) annunciators was effectively controlled through performance of appropriate preventive maintenance. Specifically, Exelon did not identify and properly account for functional failures of the MCR annunciators in June 2017, and therefore did not recognize that the annunciator system exceeded its performance criteria and required a Maintenance Rule (a)(1) evaluation. On December 7, 2017, Exelon entered the issue into their CAP as IR 04081698 and performed a review of the events identified by the inspectors that were applicable to the maintenance rule annunciator system. Corrective actions included Exelon determining that the events were functional failures, and initiated an (a)(1) evaluation based on the MCR annunciator system functional failures exceeding the designated performance criteria of an allowable one functional failure per 24 months.This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, following the two failures of the main control annunciator panel in June 2017, Exelon did not identify the failures as functional failures, and consequently, did not establish goals and monitoring criteria in accordance with 10 CFR 50.65(a)(1). In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system, or component (SSC), did not represent a loss of system and/or function, did not involve an actual loss of a function of at least a single train or two separate safety systems for a greater time than allowed by technical specifications (TS), and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. T his finding has a cross-cutting aspect in the area of Human Performance, Consistent Process, in that Exelon failed to use a consistent, systematic approach to make decisions. Specifically, Exelon did not ensure their review process for issues entered into the CAP was effectively implemented to ensure proper evaluations for all applicable maintenance rule systems affected by a n SSC failure. (H.13)
05000373/FIN-2017004-0331 December 2017 23:59:59LaSallePrimary Containment Structure, Suppression Pool Columns, Downcomer Vent and Downcomer Vent Bracing Did Not Meet Seismic Category I RequirementsA 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 failure to ensure the adequacy of the design for the primary containment, suppression pool columns, downcomer and downcomer vent bracing. Specifically, the inspectors identified three representative examples where the licensee failed to perform adequate design calculations resulting in the design not being in conformance with Seismic Category I requirements as defined in Updated Final Safety Analysis Report (UFSAR) Sections 3.8.1.4.1, 3.8.1.5 and 3.8.6. The licensee documented these violation examples in ARs 4070065, 4074674 and 4070067 and initiated actions to restore compliance. 4 The inspectors determined the licensees failure to perform adequate evaluations to demonstrate Seismic Category I compliance for the primary containment structure, suppression pool columns, downcomer vents and downcomer vent bracing was contrary to the design control measures per 10 CFR Part 50, Appendix B, requirements and was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and adversely 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. The inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, for the Barrier Integrity Cornerstone (r\eactor containment). The inspector answered no to the Barrier Integrity questions for reactor containment. The finding screened as having very low safety significance (Green). The inspectors determined there was no cross-cutting aspect associated with this finding because the deficiency was a legacy design calculational issue and, therefore, was not indicative of licensees current performance.
05000373/FIN-2017004-0231 December 2017 23:59:59LaSalleFailure to Establish Brazing Repair Procedures with Appropriate Acceptance CriteriaA finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to establish instructions with acceptance criteria that were appropriate to the circumstances for the brazing repair of the Unit Common Division I diesel generator (DG) starting air system. Specifically, through worker skill of the craft, the use of a heat sink device was relied upon to ensure that the adjacent joint of a brazed connection did not cross a temperature threshold that could have melted or otherwise unacceptably weakened the filler material; however, the procedure used did not contain any quantitative acceptance criteria for the adjacent joint temperature to determine that this important activity had been satisfactorily accomplished. The finding was considered more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, without quantitative acceptance criteria for temperature of the adjacent joints in close proximity of a brazed connection it is possible that joints could be reheated to near the solidus temperature of the filler material, resulting in joint weakening and potential failure. The licensee entered the issue into its CAP as AR 04090775. Corrective actions included revising procedures associated with brazing repairs to include a temperature value as a quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished and to address the physical condition of the adjacent joint by verifying its conditions under work order (WO) 4702099 performance. The inspectors determined that the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016. Because the finding impacted the Mitigating Systems Cornerstone the inspectors screened the finding through IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012. The finding screened as very low safety significance (Green) because it did not result in the loss of operability or functionality; thus, the inspectors answered No to all of the mitigating system screening questions. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, under the aspect of Work Management. Specifically, WO 4702099 designated DG air start system repair activities as non-code when an American Society of Mechanical Engineers (ASME) code brazing procedure specification, (BPS) 107107BR Revision 0, was being used to satisfy the standard of record, the diesel engine manufacturers standards. (H.5)
05000410/FIN-2017004-0431 December 2017 23:59:59Nine Mile PointIneffective Correction Action Results in Failure of Instrument Air SystemThe inspectors documented a self-revealing Green finding (FIN) of CNG-CA-1.01-1000, Corrective Action Program, Revision 01100, because Nine Mile Point Nuclear Station (NMPNS) failed to implement corrective actions at NMPNS Unit 2 to remove and replace all un-annealed red brass piping for the instrument air system during the April 2008 refueling outage. Specifically, on July 13, 2017, Unit 2 experienced a rupture of un-annealed red brass instrument air pipe which resulted in a feedwater pump trip and a reactor recirculation pump runback to 49 percent. Exelons corrective actions for the July 13, 2017 failure of un-annealed red brass instrument air piping included wrapping the instrument air piping with a material that both supports the piping and prevents potential stress corrosion cracking. Exelon has developed work orders to replace the piping in the upcoming outage in spring 2018. Exelon also improved staff training for accountability and work checking to verify that generated work orders are completed and closed out. Exelon entered this issue into the corrective action program (CAP) as issue report (IR) 04031685, and performed a corrective action program evaluation (CAPE). This finding is more than minor because it is associated with the design control attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, NMPNS staff failed to complete corrective actions to replace Unit 2 un-annealed red brass instrument air piping, which was susceptible to stress corrosion cracking, resulting in a feedwater pump trip and a reactor recirculation runback to 49 percent on July 13, 2017. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued on October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012. The inspectors determined that the finding was of very low safety significance (Green) because it did not result in the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event and affected mitigation equipment. The inspectors determined that this finding did not have a cross-cutting aspect because the performance deficiency occurred greater than 3 years ago; therefore, it is not considered to be indicative of current plant performance.
05000373/FIN-2017004-0131 December 2017 23:59:59LaSalleComplete versus Truncated Shifts on Proficiency WatchesThe inspectors identified an unresolved item (URI) related to the adequacy of the shifts for proficiency watches stood by specific reactor operators (ROs). Clarification was requested for whether the 8-hour proficiency watches stood by only these specific ROs, should be considered complete or truncated watches, which may not meet the requirements of 10 CFR 55.53(e). Description: Title 10 CFR 55.53(e) states, in part: To maintain active status, the licensee shall actively perform the functions of an operator or senior operator on a minimum of seven 8hour or five 12hour shifts per calendar quarter. In NUREG 1021, Revision 11, ES-605 further explains that: In accordance with 10 CFR 55.53(e), to maintain an active status, licensed operators are required to maintain their proficiency by actively performing the functions of an operator or senior operator on at least seven 8hour or five 12hour shifts per calendar quarter. This requirement may be completed with a combination of complete 8 and 12hour shifts (in a position appropriately credited for watch-standing proficiency as discussed below) at sites having a mixed-shift schedule, and watches shall not be truncated when the operator satisfies the minimum quarterly requirement (56 hours). Overtime may be credited if the overtime work is in a position appropriately credited for watch-standing proficiency. As documented in AR 04070501, dated November 3, 2017, it has been LaSalle Stations practice to use an individuals normal shift work hours to determine the length of his/her proficiency watch. While the operating shift crews were assigned to 12hour shifts, those licensed ROs assigned to other staff positions at LaSalle normally worked 8 hours per day. LaSalle refers to these individuals as Administrative ROs. Thus, when 14 LaSalles Administrative ROs stood their proficiency watches, they stood 8hour watches, and turned over to another operator to complete the normal 12hour operating shift. As stated in this AR, 8hour shifts minimized the overtime costs to maintain active licenses for these individuals. The Operator Licensing and Training Branch was requested via Regional Office Interaction ROI1725, Clarification of Complete vs. Truncated Shift for Proficiency Watches, because Administrative ROs stood 8hour proficiency watches, while all other operators stood 12hour shifts. Clarification is needed from the Operator Licensing and Training Branch and the Office of the General Counsel to determine if the current practice meets the requirements of 10 CFR 55.53(e) to maintain an operating license in an active status. (URI 050000373/201700401; 050000374/201700401, Complete versus Truncated Shifts on Proficiency Watches)
05000374/FIN-2017010-0131 December 2017 23:59:59LaSalleFailure To Ensure Fire Door Was Engaged And PinnedThe inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation of License Condition 2.C.15 for Unit 2, for the licensees failure to ensure all fire rated assemblies (i.e., fire doors) were operable. Specifically, during a plant walk down, the inspectors found Fire Door 282 inoperable. The lower pin of the stationary part of the double door was not engaged, because the pin was broken.The licensee entered the issue into their Corrective Action Program and as an immediate action, declared the door inoperable, established hourly fire watch, and subsequently installed a new pin.The inspectors determined that the performance deficiency was more-than-minor because the finding was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the issue screened as having very-low safety significance (Green) by answering Yes to Question 1.4.3.A of IMC 0609, Appendix F, Attachment 1 based on no combustible within 10 feet of Door 282 on the 5A4 side and one pin should still provide sufficient defense-in-depth for several hours before buckling or moving out of the frame. The finding had a cross-cutting aspect in the Procedure Adherence component of the Human Performance cross-cutting area. Specifically, the licensee failed to follow procedural guidance to thoroughly verify that fire doors were pinned when challenging the doors. (H.8)
05000410/FIN-2017004-0331 December 2017 23:59:59Nine Mile PointInadequate Operability Determination forImpairedInternal Flood BarrierAn NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when Exelon failed to perform an adequate operability determination in accordance with OP-AA-108-115, Operability Determinations, Revision 20, upon identification of Unit 2 degraded internal flood barriers that support operability of emergency core cooling system (ECCS) equipment. Specifically, from November 21 until December 10, 2017, Exelon failed to properly evaluate the excavation of internal flood barriers and concluded there was a reasonable expectation for operability of the supported ECCS systems. Exelon entered this issue into the CAP as IR 04082686. Corrective actions included conducting a detailed evaluation of operability for the supported safety-related systems, additional training associated with TS 3.0.9, including a focus on the need for risk assessments when entering TS 3.0.9, and a procedure change to CC-AA-201, Plant Barrier Control Program, and CC-NM-201-1001, Plant Barrier Control Program Implementation, which is the NMPNS specific procedure to address the vulnerabilities associated with impairing multiple required barriers. This finding is more than minor because it is associated with the human performance 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. Specifically, from November 21 until December 10, 2017, Exelon failed to adequately evaluate the operability of a degraded internal flooding barrier and the potential impact on operability of the supported ECCS system equipment. The inspectors identified that the internal flood barrier was excavated such that there was not sufficient material to ensure adequate flood protection, and resulted in a reasonable doubt for the operability of the supported ECCS systems. This finding is also similar to example 3.j and 3.k of IMC 0612 Appendix E, Examples of Minor Issues, issued August 11, 2009, because the condition identified by the inspectors resulted in a reasonable doubt for the operability of the ECCS supported systems and additional analysis was necessary to verify operability. 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, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to vulnerability to external initiating events. This finding has a cross-cutting aspect in the area of Human Performance, Work Management, because Exelon failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. As a result, Exelon personnel failed to recognize that work activities that impaired internal flood barriers on both Division I and II low pressure ECCS pump rooms were executed simultaneously, which led to an unplanned entry into TS Limiting Condition for Operation (LCO) 3.0.9. (H.5)
05000397/FIN-2017003-0130 September 2017 23:59:59ColumbiaInadequate High Pressure Core Spray Fill and Vent ProcedureThe inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, for the licensees failure to have a high pressure core spray system fill and vent procedure appropriate to the circumstances. The licensee entered this issue into the corrective action program as Action Request 368872. The failure to have a high pressure core spray system fill and vent procedure appropriate to the circumstances was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the equipment performance at tribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Procedure SOP- HPCS -FILL, HPCS Fill and Vent, Revision 11, was not appropriate to the circumstances in that it did not ensure the high pressure core spray instrumentation lines were clear of voids. As a result, air remained in the instrumentation lines , and the high pressure core spray minimum flow instrument, HPCS -FIS -6, was degraded. The inspector s performed the initial significance determination using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding did not represent an actual loss of function of one or more non- technical specification trains of equipment designated as high safety -significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding had a cross -cutting aspect in the area of human performance, avoid complacency, in that the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12).
05000373/FIN-2017003-0130 September 2017 23:59:59LaSalleInadequate Maintenance Rule Monitoring of the Low Pressure Core Spray SystemThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulation(CFR) 50.65(a)(1) for the failure to monitor the performance of the Unit 1 low pressure core spray (LPCS) system against licensee-established goals. Specifically, the licensee did not identify and properly account for a maintenance rule functional failure (MRFF) of the Unit 1 LPCS min-flow valve differential pressure switch, which demonstrated that performance of the Unit 1 LPCS system was not being controlled in accordance with the maintenance rule. The Licensees immediate corrective actions included entering this issue into their corrective action program (CAP), re-evaluating and classifying the LPCS min-flow valve differential pressure switch failure as a MRFF, and entering the system into (a)(1) status. This finding was entered into the licensees CAP as action request (AR) 4029999.The performance deficiency was determined to be more-than-minor in accordance with IMC 0612 Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not properly classify the May 17, 2017, failure of the LPCS min-flow valve differential pressure switch as a MRFF. When properly classified, this failure caused the maintenance rule performance criteria for the LPCS system to be exceededcausing the system to receive additional remedial station attention. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 16, 2016, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that this maintenance rule program-based finding is of very low safety significance (Green) since it was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, it did not represent the loss of a system and/or function, it did not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its technical specifications allowed outage time, and it did not represent an actual loss of a non-technical specification equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of Problem Identification and Resolution in the aspect of Evaluation, where the organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the Licensee failed to thoroughly evaluate the failure of the Unit 1 LPCS min-flow valve differential pressure switch on May 17, 2017 (P.2).
05000397/FIN-2017003-0230 September 2017 23:59:59ColumbiaFailure to Report Unplanned Valid Reactor Protection System ActuationThe inspectors identified a Severity Level IV, non- cited violation of 10 CFR 50.72(b)(3)(iv)(A) for the licensees failure to submit an event notification to the NRC 3 within 8 hours of occurrence of an unplanned valid reactor protection system actuation of the reactor protection system. Specifically, the licensee did not report are actor protection system Level 3 scram actuation when reactor vessel level dropped below +13 inches until prompted by the inspectors. The licensee subsequently restored compliance and reported the event in accordance with 10 CFR 50.72(b)(3)(iv)(A) on August 24, 2017, as an update to Emergency Notification System Report 52918 and entered the issue into their corrective action program as Action Request 370529 . The licensees failure to submit the event notification was a violation that impacted the regulator y process and warrants treatment using traditional enforcement . In accordance with Section 2.2.4 and the example in Section 6.9.d.9 of the NRC Enforcement Policy, dated November 1, 2016, the violation was determined to be a Severity Level IV violation. Traditional enforcement violations are not assessed for cross- cutting aspects.
05000373/FIN-2017008-0130 September 2017 23:59:59LaSalleFailure to Correctly Evaluate/Justify Post-Accident Operability Qualification for the Reliance Motor 1(2)VY03C RHR Pumps RoomCooling FanThe inspectors identified a finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.49, Paragraph (f)(4), for the licensees failure to provide adequate analysis in combination with partial type test data to qualify an Environmental Qualification (EQ) component. Specifically, EQ-LS068 failed to provide adequate analysis to justify the Post-Accident Operability Qualification for the Reliance Electric motor utilized for 1(2)VY03C. The EQ Binder incorrectly relied on test values that was strictly performed for thermal aging (for normal plant conditions) to justify a Post-Accident Qualification. The licensee captured the inspectors concern into their Corrective Action Program (CAP) as Action Request (AR) 04030532.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, as an immediate corrective action, the licensee performed a preliminary assessments and concluded that the motors could be EQ qualified for the environmental conditions for which they could be exposed. The finding was associated with a cross-cutting aspect in the area of Human Performance, Design Margin. (H.6)
05000397/FIN-2017010-0130 September 2017 23:59:59ColumbiaFailure to Transfer Byproduct Material to a Disposal Facility in Accordance with the Terms of the Facilitys LicenseThe inspectors reviewed a self-revealed non-cited violation of 10 CFR 30.41(b)(5) for the failure to transfer byproduct material to an authorized waste disposal facility in accordance with the terms of the facilitys license. Specifically, License Condition No. 21.C of the US Ecology license requires that all radwaste shall be packaged in such a manner that waste containers received at the facility do not show an increase in the external radiation levels as recorded on the manifest, within instrument tolerances. On July 20, 2017, Columbia Generating Station personnel transferred byproduct material to US Ecology for disposal (Shipment 17-51). The disposal facilitys surveys identified that the dose rate on contact with the waste liner was 1100 millirem per hour, whereas the manifest for this shipment recorded a dose rate of 12 millirem per hour. The licensees corrective actions included providing a corrected shipment manifest to US Ecology and proposed enhancements to the Columbia Generating Station procedures for shipping. This issue was documented in the licensees corrective acti on program as Action Request AR 00369215. The failure to transfer byproduct material to a low-level radwaste disposal facility in accordance with the facilitys license was a performance deficiency. The performance deficiency was more than minor because it was associated with the program and process attribute of the Public Radiation Safety Cornerstone and adversely affected the associated cornerstone objective to ensure 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. Using NRC Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined that the finding had very low safety significance (Green) because it was a low-level burial ground nonconformance; however, it was not Class C waste or greater and the waste did conform to the waste characteristics of 10 CFR 61.56. The finding has a cross-cutting aspect in the area of Human Performance, Resources, because licensee leaders failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety (H.1).
05000373/FIN-2017008-0230 September 2017 23:59:59LaSalleFailure to have Adequate Justification for Extending the life of Lubricant used in EQ Motor BearingsThe inspectors identified a finding of very-low safety significance and an associated NCV of 10 CFR Part 50.49, Paragraph (j), Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, for the licensees failure to have adequate justification for extending the service-life for grease used in the bearing for EQ motors installed in harsh environment. Specifically, the licensee extended the bearing grease qualified service life for several EQ motors installed in Zone H4A, H6 and H5A from 31.5, 20.5 and 19.5 years respectively to 60 years based on incorrect assumptions. The justification for 60 years extension incorrectly assumed that the calculated service-life was based on continuous operation of the motor. The licensee captured the inspectors concern into their CAP as AR 04030538. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, as an immediate corrective action, the licensee performed a preliminary evaluation that concluded that the grease remained qualified based on test data which showed that the grease consistency remained within acceptable range during the thermal age test. The finding did not have a cross-cutting aspect associated with it because it was not representative of current performance.
05000397/FIN-2017010-0230 September 2017 23:59:59ColumbiaFailure to Control a High Radiation Area with Dose Rates Greater Than 1000 Millirem Per Hour at 30 CentimetersThe inspectors identified a non-cited violation of Technical Specification 5.7.2 for the failure to control a high radiation area with dose rates greater than 1000 millirem per hour at 30 centimeters. Specifically, equipment boxes placed adjacent to high radiation area barrier fencing in the reactor building 471 elevation, which created a natural ladder into the area, resulted in an uncontrolled entryway to a high radiation area with dose rates greater than 2500 millirem per hour. Once informed, the licensee immediately removed the equipment boxes from the barrier and added signage that restricted the placement of any items adjacent to the fencing forming the high radiation area barrier. This issue was documented in the licensees corrective action program as Action Request AR 00355646. The failure to properly control and barricade an entryway to a locked high radiation area in the reactor building, 471' elevation, was a performance deficiency. The performance deficiency was more than minor because it was associated with the program and process (exposure control) attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material. Using NRC Inspection Manual Chapter 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding had very low safety significance (Green) because the finding was not an as low as reasonably achievable planning or work control issue, there was no overexposure or potential for an overexposure, and the licensee's ability to assess dose was not compromised. The finding had a cross- cutting aspect in the area of Human Performance, Field Presence, because leaders were not commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations, resulting in a lack of oversight of work activities, to include contractors and supplemental personnel (H.2).
05000374/FIN-2017002-0230 June 2017 23:59:59LaSalleFailure to Implement a Preventive Maintenance Strategy for Main Generator AuxiliariesGreen. A self-revealed finding of very low safety significance was identified for the failure to implement a preventive maintenance strategy for main generator auxiliaries in accordance with MAAA716210. Specifically, a performance centered maintenance template was issued in 2004 that required 10 year inspections for stator cooling heat exchanger isolation valves, but the maintenance strategy was never implemented. As a result, 2GCY08 had a stem-to-disc separation that ultimately led to a manual reactor scram on January 23, 2017. As part of the corrective actions, the licensee shifted to the standby stator cooling heat exchanger and restarted the reactor on January 25, 2017. The performance deficiency was documented in the licensees corrective action program (CAP). The performance deficiency was more than minor because it was associated with the Initiating Events Cornerstone attribute of equipment performance 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. Specifically, the performance deficiency resulted in a reactor scram. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding was of very low safety significance because although the performance deficiency caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the scram to a stable shutdown condition. Although the performance deficiency occurred in 2005, the licensee performed a vulnerability review of the stator cooling system in 2015 that did not identify 2GCY08 as critical. Therefore, the inspectors determined that the finding represented present performance. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensee failed to plan and execute preventive maintenance for valve 2GCY08
05000397/FIN-2017002-0130 June 2017 23:59:59ColumbiaMechanism Operated Cell Switch FailureGreen . The inspectors reviewed a self -revealed finding for the licensees failure to follow plant Procedure SWP -CA P-01, Corrective Action Program, that ensures corrective actions are timely. As a corrective action for failures associated with mechanism operated cell switches for nonsafety 4160 VAC circuit breakers in 2013 and 2015, the licensee assigned modifications to the mechanism operated cell switches but failed to implement t hem in a timely manner. Consequently, on July 20, 2016, circuit breaker E -CB -S/3 mechanism operated cell switches failed to change state resulting in a loss of a main feed pump and an unplanned runback to 70 percent reactor power. As corrective action, the licensee declared the startup transformer inoperable, modified the mechanism operated cell assembly for circuit breaker E -CB -S/3 to remove one switch, and performed post -maintenance testing. The licensee also initiated Action Request 352504 to perform an apparent cause review and address long -term corrective actions. The failure to follow plant Procedure SWP -CAP -01, Corrective Action Program, that ensures corrective actions are timely was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Initiating Event Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the loss of major loads on E -SM -3 upset plant stability by causing a loss of feed and reactor runback transient. The inspector performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, the licensee remained at power and maintained diverse feed and condensate pumps. This finding had a cross -cutting aspect in the area of human performance, consistent process, in that the licensee failed to use a systematic approach to make decisions including incorporating risk insights. Specifically, circuit breaker E -CB -S/3 is utilized at least monthly 3 for emergency diesel generator surveillance testing and a failure could render the startup transformer inoperable. The mechanism operated cell assembly modification, recommended in 2013 and assigned for action in 2015, was not planned or scheduled as a work order at the time of the failure in 2016 (H.13).
05000373/FIN-2017009-0130 June 2017 23:59:59LaSalleAnchor Darling Double Disc Gate Valve 1E22-F004 and 2E22-F004 Pressed-FitCollar Related 10 CFR Part 50, Appendix B, Criterion III ViolationIn 1990, the licensee had reviewed and accepted the vendors weak link analyses that provided the upper torque and thrust limits for all safety-related ADDDGV in service at the station. This analysis documentedthat the 1E22-F004 and 2E22-F004 valve stems were the weak link valve components in the closing direction (i.e.,provided enough closing thrust, thevalve stems would be the firstcomponent to becomenonfunctional).Therefore, theclosed thrust limit forthe 1E22-F004 and 2E22-F004 valves was approximately 260,000 lbf. The licensee had set up the valves ina manner that would ensure that the valveswould have enough torque and thrust tooperate under design basis conditions while staying below the maximum weak link limits. Maintenance and test records showed that thelicensee consistently verifiedthat these two valves were setup and maintained within this design window. Typical as-found and as-left closed thrust limits ranged from approximately between200,000240,000 lbf.As described in the licensees failure analysis report and as discussed above, the licensee identified that the pressed-fitcollar could relax its pre-load when operating the valve well within the established maximum closed thrust limitations. The licensees failure analysis report estimated that approximately 130,000 lbf was necessary to shift the collar up and relax the pre-load. Therefore,theteam concluded that the licensees weak link analysis was inadequate based upon the 2E22-F004 valve failure and associated failure analysiswhich determined that the pressed-fitcollar was a weaker component as compared to the valve stem. The team did not identify an associated performance deficiencyfor the inadequate weak link analysis. This determination was based upon the weak link analysis originating from the vendor in 1990, licensees review of that analysis, and latent design issue that had not been previously identified within the industry until recently identified by the licensee.Additionally, the team did not identify a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. This determination was based, in part, that correcting the unknown stem collar pre-torque issueafter receiving the 10 CFR Part 21 Flowserve notification would not necessarily have identified and corrected the non-conforming inadequate weak link design control issue. Enforcement: Title10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,inpart that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2, and as specified in the license application, for those structures, systems, and components to which this appendix apply are correctly translated into specifications, drawings, procedures, and instructions.Contrary to the above, since original plant construction, the licensee failed to ensure thatapplicable design basismaximumclosed thrust and torque valuesfor the safety-related Unit 1 and Unit 2 HPCS injection valves (1E22-F004, 2E22-F004)werecorrectly translatedinto specifications. Specifically, it was identified that the stem-to-wedgepre-torque credited within the design could relax by applying closed direction torque and thrust well within the specified design limitbecause that limit was based uponthe wrong weak link component. The loss of the stem-to-wedgepre-torque could subsequently break the wedge pin and result in stem-to-wedgethread degradation ultimately leading to valve failure.The NRC determined that issue was a Severity Level III Violation based upon Section6.1(c)(2) of the Enforcement Policy. Specifically, a system that is part of the primary success path and which functions or actuates to mitigate a design base accident or transient that either assumes the failure of or presents a challenge to the integrity of the fission product barrier not being able to perform its licensing basis safety function because it is not fully qualified.The NRC exercised enforcement discretion in accordance with Sections 3.10 of the Enforcement Policy and Section 3 of Part1 of the Enforcement Manual. Enforcement Policy Section 3.10 states that the NRC may exercise discretion for violations of NRC requirements by reactor licensees for which there are no associated performance deficiencies. This violation was entered into the Corrective Action Programas Issue Report3972901 and has been corrected by replacing the 1E22-F004 and 2E22-F004 valve stems with integral collars.
05000373/FIN-2017002-0130 June 2017 23:59:59LaSalleFailure to Implement A Preventive Maintenance Strategy for 1B RHR Low Pressure Permissive Pressure SwitchGreen. An NRCidentified finding of very low safety significance was identified for the failure to implement a preventive maintenance strategy for the 1B residual heat removal injection valve low pressure permissive switch in accordance with procedure ERAA2001001, Equipment Classification, Revision 3. The switch failed and was replaced on February 18, 2017. The performance deficiency was documented in the licensees CAP. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating System 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 performance deficiency resulted in the inoperability of an emergency core cooling system train of equipment. 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, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance because all the screening questions associated with IMC 0609, Appendix A, Exhibit 2, were answered No. The switch was replaced and returned to service within 24 hours of when it was initially identified as a problem. This finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current licensee performance.
05000397/FIN-2017002-0230 June 2017 23:59:59ColumbiaFailure to Conduct Adequate Surveys of Spent Filters Moved from the Spent Fuel PoolGreen . The inspectors reviewed a self -revealed, non- cited violation of 10 CFR 20.1501 resulting from the licensee's failure to conduct radiation surveys necessary to establish appropriate controls to support movement of spent filters from the spent fuel pool to a shipping cask. This issue was entered into the licensee's corrective action program as Action Requests 356390 and 358265. The licensees failure to perform surveys necessary to establish appropriate controls to support the movement of filters from the spent fuel pool to a shipping cask was a performance deficiency. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and adversely affected the associated cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. Specifically, the inadequate radiation surveys resulted in inadequate controls being implemented causing unplanned and unintended personnel dos e. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green), because it did not involve: (1) ALARA planning and controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. The finding had a cross- cutting aspect in the area of human performance, associated with work management, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees organization and work processes failed to include the identification and management of radiological risk commensurate with the spent fuel pool filter project and the need for strict coordination with different groups or job activities (H.5).