Semantic search
Jump to navigation
Jump to search
| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000373/FIN-2018201-01 | 2018Q3 | Green | NRC identified | Security | ||
| 05000373/FIN-2018412-01 | 2018Q3 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000373/FIN-2018003-01 | 2018Q3 | Green | H.12 | NRC identified | Failure to Establish Heat Exchanger Inspection Procedures Appropriate for the Circumstances | The 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-2018003-02 | 2018Q3 | Green | NRC identified | Failure to Establish an Appropriate Inservice Testing Procedure | The 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-2018003-03 | 2018Q3 | Green | NRC identified | Failure to Establish Goals to Monitor Steam Tunnel Check Dampers | Introduction: 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. | |
| 05000373/FIN-2018003-04 | 2018Q3 | Green | H.13 | NRC identified | Failure to Manage the Increase in Risk During a Battery Charger Capacity Tes | The 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. |
| 05000373/FIN-2018003-05 | 2018Q3 | Green | NRC identified | Failure 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. | |
| 05000373/FIN-2018003-06 | 2018Q3 | NRC identified | Potential Failure to Inspect Containment Post-Tensioned Tendons per Code Requirements and to Follow Corrective Action Program Process | Vertical 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-2018003-07 | 2018Q3 | NRC identified | Potential Failure to Promptly Correct the Unit 2 Primary Containment Wall Cavity Leakage Condition and to Follow Corrective Action Program Process | Condition 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-2018003-08 | 2018Q3 | Green | H.4 | NRC identified | Failure to Implement Engineering Change Results in Reactor Coolant Boundary Leakage | The 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-2018002-04 | 2018Q2 | Severity level Minor | NRC identified | Minor 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-2018002-03 | 2018Q2 | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification LCO 3.4.4 (applicable for Modes 1, 2 and 3) states: The safety function of 12 safety relief valves (S/RVs) shall be OPERABLE, and Action Statement A states that One or more required S/RVs inoperableA.1 be in mode 3 in 12 hours and A.2 be in Mode 4 in 36 hours. Technical Specification SR 3.4.4.1 states that Verify the safety function lift setpoints of the required S/RVs are as follows Number of S/RVs Setpoint (psig 2 1205 36. 3 1195 35. 2 1185 35. 4 1175 35. 2 1150 34. Contrary to the above, during portions of previous Unit 1 and 2 operating cycles from 2012 through January of 2017, two main steam S/RVs did not meet these lift pressure setpoint requirements. Specifically S/RV 2B21F013C lifted at 1131 psig instead of from 1139.8 to 1210.2 psig and S/RV 2B21F013L lifted at 1130 psig instead of from 1159.2 to 1230.8 psig (reference: Licensee Event Report 05000374/201700400; 01, Two Main Safety Relief Valves Failed Inservice Lift Inspection Pressure Test. Significance/Severity: This licensee identified finding affected the Initiating Events Cornerstone and was screened in accordance with Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power. The two affected SRVs lifted low outside of their setpoint band, which was conservative with respect to maintaining the reactor coolant system overpressure protection safety function of these valves. Therefore, the inspectors determined that this finding is of very low safety significance (Green) because after a reasonable assessment of degradation, the finding would not have resulted in exceeding the reactor coolant system leak rate for a small LOCA and did not affect other systems used to mitigate a loss-of-coolant accident. Corrective Action Reference: AR 3974669 | |
| 05000373/FIN-2018002-02 | 2018Q2 | Green | NRC identified | Failure to Follow Procedure and Perform Database Revision Review Requirements | The 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. | |
| 05000373/FIN-2018002-01 | 2018Q2 | Green | Self-revealing | Failure to Implement a Preventative Maintenance Strategy for Residual Heat Removal Service Water Pump Shorting Relays | A 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. | |
| 05000373/FIN-2018010-01 | 2018Q2 | Green | H.12 | NRC identified | Failure to Translate Reactor Building Superstructure Design Basis | Inspectors 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. |
| 05000373/FIN-2018001-03 | 2018Q1 | Severity level Enforcement Discretion | NRC identified | Enforcement Action (EA) 18035: Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance | On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of 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-2018001-01 | 2018Q1 | Green | Self-revealing | Post-Maintenance Testing Failed to Demonstrate Testable Check Valve Function | A 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-2018001-02 | 2018Q1 | Green | NRC identified | Failure to Update Throttle Valve Position in Accordance with Station Procedures | The 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. | |
| 05000373/FIN-2017004-01 | 2017Q4 | NRC identified | Complete versus Truncated Shifts on Proficiency Watches | The 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-2017010-01 | 2017Q4 | Green | H.8 | NRC identified | Failure To Ensure Fire Door Was Engaged And Pinned | The 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) |
| 05000373/FIN-2017004-02 | 2017Q4 | Green | H.5 | NRC identified | Failure to Establish Brazing Repair Procedures with Appropriate Acceptance Criteria | A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was 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) |
| 05000373/FIN-2017004-03 | 2017Q4 | Green | NRC identified | Primary Containment Structure, Suppression Pool Columns, Downcomer Vent and Downcomer Vent Bracing Did Not Meet Seismic Category I Requirements | A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the 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-2017004-04 | 2017Q4 | Green | NRC identified | Failure of Offsite Power Backfeed Procedure to be Appropriate to the Circumstances Caused Unit 1 Scram | A 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. | |
| 05000373/FIN-2017003-01 | 2017Q3 | Green | P.2 | NRC identified | Inadequate Maintenance Rule Monitoring of the Low Pressure Core Spray System | The 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). |
| 05000373/FIN-2017008-02 | 2017Q3 | Green | NRC identified | Failure to have Adequate Justification for Extending the life of Lubricant used in EQ Motor Bearings | The 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. | |
| 05000373/FIN-2017008-01 | 2017Q3 | Green | H.6 | NRC identified | Failure to Correctly Evaluate/Justify Post-Accident Operability Qualification for the Reliance Motor 1(2)VY03C RHR Pumps RoomCooling Fan | The 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) |
| 05000373/FIN-2017009-01 | 2017Q2 | Severity level Enforcement Discretion | NRC identified | Anchor Darling Double Disc Gate Valve 1E22-F004 and 2E22-F004 Pressed-FitCollar Related 10 CFR Part 50, Appendix B, Criterion III Violation | In 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. | |
| 05000374/FIN-2017002-02 | 2017Q2 | Green | H.6 | Self-revealing | Failure to Implement a Preventive Maintenance Strategy for Main Generator Auxiliaries | Green. 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 |
| 05000373/FIN-2017002-01 | 2017Q2 | Green | NRC identified | Failure to Implement A Preventive Maintenance Strategy for 1B RHR Low Pressure Permissive Pressure Switch | Green. 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. | |
| 05000373/FIN-2017001-03 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and drawings, requires, in part, that activities affecting quality be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to this, on February 6, 2017, the licensee failed to accomplish an activity affecting quality in accordance with licensee procedure, CC AA 201, Revision 11, Plant Barrier Control Program. Specifically, the licensee failed to implement compensatory actions required by the Plant Barrier Control Program which resulted in multiple doors being impaired at the same time such that safety -related equipment in the Unit 2 Division II switchgear room and Unit 2 749 Auxiliary Building were declared inoperable. The licensee documented the issue in their CAP as Action Request (AR ) 3972830. The inspectors determined that this issue was of very low safety significance because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component ( SSC ); (2) did not represent a loss of system and/or function; (3) did not represent the actual loss of safety 35 function of at least a single train for greater than its technical specification ( TS ) allowed outage time; (4) did not represent an actual loss of one or more non TS trains of equipment during shutdown designated as risk significant for greater than 24 hours; and (5) did not degrade a functional auto- isolation of residual heat removal ( RHR) . | |
| 05000373/FIN-2017001-02 | 2017Q1 | Green | P.3 | Self-revealing | Failure to Perform Preventive Maintenance Resulted in Stem -to-Disc Separation of Safety -Related Valve | Green . A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self -revealed for the licensees failure to ensure that activities affecting quality were prescribed in a manner appropriate to the circumstances for the Unit 2, Division 3 , diesel generator (DG) system . Specifically, the licensees processes for the control and administration of preventive maintenance (ER AA 200/WC AA 120) failed to ensure that safety -related valve, 2E22 F319, the 2B DG cooling water strainer backwash valve, was replaced or refurbished at a frequency that would prevent corrosion- related stem -to-disc separation. The licensee entered this issue int o the ir CAP as AR 1122320. Corrective actions planned and completed included replacement of the 2E22 F319 valve with a stainless steel design and performing an apparent cause evaluation of the degraded condition. The performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) . Specifically, the failure to perform preventive maintenance on the 2E22 F319 valve resulted in a degraded condition which adversely affected the reliability of the high pressure core spray system to respond to an initiating event. The inspectors evaluated the finding using the significance determination process in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, dated June 19, 2012. The inspectors reviewed the Mitigating Systems screening questions in Exhibit 2 and answered No to question A.1, If the finding is deficiency affecting the design or qualification of a mitigating SSC (structure, system, or component) , does the SSC maintain its operability or functionality. The inspectors answered Yes to question A.2, Does the finding represent a loss of system and/or function; therefore, a detailed risk evaluation was required. The detailed risk evaluation determined that the finding screened as having very low safety significance (Green). This finding had a cross -cutting aspect in the area of Problem Identification and Resolution, because t he organization failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P .3). |
| 05000373/FIN-2017001-01 | 2017Q1 | Green | H.5 | NRC identified | nadequate Controls for ASME Code VT 3 Internal Examination of Pumps and Valves | Green . The inspectors identified a finding of very-low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (CFR) , Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee failed to establish a procedure that ensured the American Society of Mechanical Engineers (ASME) Code VT 3 examination of the internal surface of valves or pumps occurred in the as -found condition (e.g., prior to repairs). Consequently, the licensee repaired internal damage to the 2B33 F067B valve prior to the Code VT 3 examination which potentially resulted in an ineffective VT 3 examination. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 3972620, initiated actions to complete another VT 3 examination of valve 2B33 F067A or valve 2B33 F067B during the current outage and was evaluating additional controls for scheduling VT 3 internal examinations of pumps and valves. The performance deficiency was determined to be more- than- minor because it affected the Initiating Events cornerstone attribute of equipment performance and adversely affected the cornerstone objective to l imit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, if left uncorrected, this finding would lead to a more significant safety concern because it increa sed the likelihood of an operational challenge to the plant caused by a recirculation system line break initiated from undetected service -induced defects left in service inside pumps or valves as a result of ineffective VT 3 examinations. The finding was screened in accordance with Inspection Manual Chapter 0609, Appendix A, and the inspectors answered No to the applicable Phase 1 Initiating Events Screening question because the finding did not result in a reactor trip and/or loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Therefore, this finding was determined to have very- low safety significance (Green) . The finding had a cross -cutting aspect of Work Management in the Human Performance cross -cutting area because licensee managers failed to establish an adequate process of planning, controlling, and executing 3 work activities such that nuclear safety is the overriding priority as evidenced by the lack of appropriately controls for scheduling the VT 3 internal examination of the 2B33 F067B valve (H.5) |
| 05000373/FIN-2016004-01 | 2016Q4 | Green | H.7 | NRC identified | Failure to Perform Required Monthly Fire Extinguisher Inspections per National Fire Protection Association Code | The inspectors identified a finding of very low safety significance with an associated NCV of the LaSalle County Station Unit 1 and Unit 2 operating licenses, NFP11, Section 2.C.(25), Fire Protection Program, and NFP18, Section 2.C.(15), Fire Protection Program, respectively, for the licensees failure to meet the inspection requirements of National Fire Protection Association (NFPA) 101975 for portable fire extinguishers. Specifically, from October 10, 2011, to January 9, 2017, the licensee failed to perform inspections on portable fire extinguishers in high radiation areas on the required monthly frequency, including some fire extinguishers that were in place in case of a fire in safety-related areas, such as outside emergency core cooling system pump rooms. The licensee entered this issue into the Corrective Actions Program (CAP) as Action Request (AR) 02739987. Licensees corrective actions include completion of an evaluation which provided a technical justification for a deviation from the monthly inspection requirements of NFPA10. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors, including fire, and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, had a fire occurred in one of the affected fire zones containing safety-related mitigation equipment (e.g., RHR pump room) and a licensee responder attempted to use an extinguisher that may not be functional due to an unknown degraded condition allowed to exist because monthly checks were not performed, the fire could progress further and render the mitigating system inoperable. The finding was screened as very low safety significance (Green) because the fire finding was associated with portable fire extinguishers not used for hot work fire watches. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, accurate and up-to-date documentation. Specifically, the licensee failed to ensure that procedures governing the monthly inspection of portable fire extinguishers contained accurate information regarding the use of a deviation from NFPA10. (IMC 0310, H.7) |
| 05000373/FIN-2016004-03 | 2016Q4 | Green | Self-revealing | Failure to Perform Preventive Maintenance Resulting in Two Subsequent Unit 1 RCIC Turbine Trips During Surveillance Testing | A finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to perform preventive maintenance on the Unit 1 reactor core isolation cooling (RCIC) electronic governor-remote (EGR) actuator. Specifically, from June 4, 1993, to November 17, 2016, the licensees processes for the control and administration of preventive maintenance failed to ensure that the Unit 1 RCIC EGR actuator was replaced or refurbished on an interval that would prevent internal fouling of the EGR actuator from adversely affecting governor performance. As a result, contaminates and degradation accumulated in the EGR actuator from January 16, 2004, to November 17, 2016, ultimately causing the RCIC turbine to trip during quarterly surveillance testing on October 18, 2016, and again on November 17, 2016. The licensee entered this issue into the CAP as ARs 02729757 and 02742254. Corrective actions planned and completed included replacing the Unit 1 and Unit 2 RCIC EGRs and performing a root cause evaluation of the degraded condition. The performance deficiency was more than minor, and thus a finding because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to perform preventive maintenance on the Unit 1 RCIC EGR resulted in a degraded condition which adversely affected the reliability of the system to respond to an initiating event. A detailed risk evaluation determined that the finding screened as having very low safety significance (Green). This finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current licensee performance. | |
| 05000373/FIN-2016004-04 | 2016Q4 | Green | H.2 | NRC identified | Block Wall Evaluations Not Consistent with As-Built Condition | The inspectors identified a finding of very-low safety significance when licensee personnel failed to ensure that the design inputs used in block wall evaluations for determination of their seismic capacities were consistent with the conditions as built or as specified in the design documents. Specifically, the material properties and wall configurations used in the analyses were not consistent with the as-built conditions. The evaluations were a part of the licensees response to the NRC Request for Information Pursuant to 10 CFR, Part 50.54(f) regarding Recommendation 2.1: Seismic, of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident. The performance deficiency did not impact the operability or functionality of the walls and was captured in the licensees CAP under ARs 2712569, 2711669, 2711877, 2710850, 2711337 and 2711875, with actions to revise the affected calculations. The performance deficiency was more than minor, and thus a finding because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors 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 licensees failure to determine and use correct design inputs adversely impacted the evaluations of block walls required for protection of the components attached to or located in proximity of the walls, and needed to support implementation of the diverse and flexible coping strategies. The finding screened as having very-low safety significance (Green) because the finding did not result in the loss of operability or functionality of any affected structures, systems, and components. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of field presence where senior managers ensure supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, the licensee failed to provide supervisory and management oversight for the activities of the contractor performing the block wall evaluations. (IMC 0310, H.2) |
| 05000373/FIN-2016004-02 | 2016Q4 | Green | P.5 | NRC identified | Failure to Provide Sufficient Guidance for the Successful Troubleshooting of Safety-Related Equipment | The inspectors identified a finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to provide procedural guidance of a type appropriate to the circumstances. Specifically, licensee procedure MAAA716004, Conduct of Troubleshooting, Revision 13, did not prescribe appropriate quantitative or qualitative acceptance criteria for determining whether a failed component existed in the 1VY03C control circuit (a safety-related component) using the simple troubleshooting methods outlined by the procedure. The licensee entered this issue into the CAP as ARs 02680921 and 02722425. Corrective actions included revision of the MAAA716004 procedure to include instructions that drove more thorough troubleshooting activities, as recommended by an internal fleet assessment documented in AR 02516457. The performance deficiency was more than minor, and thus a finding because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee concluded troubleshooting on fan 1VY03C without correcting the degraded condition which adversely affected the reliability of the fan to automatically start in response to an initiating event. The finding screened as of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating system, structure and component (SSC), did not represent a loss of system or function, did not represent an actual loss of function of a train for greater than the TS allowed outage time and did not represent a loss of function of a nonTS train of equipment. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee failed to ensure evaluation and implementation of internal operating experience in a timely manner after a fleet wide issue concerning less than adequate troubleshooting was entered in the CAP. (IMC 0310 P.5) |
| 05000373/FIN-2016007-02 | 2016Q2 | Green | NRC identified | Failure to Ensure that Both Feed Supply Breakers for Swing DG Components Were Closed During Normal Plant Operation | The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to have the capability to verify the supply breakers of both reactor units feeding the swing diesel generator (DG) components were closed during normal plant operation. Specifically, the circuit design and procedures for the swing DG room fan, fuel oil transfer pump, and fuel storage tank room exhaust fan did not ensure the detection of the condition where one of these feeder breakers was tripped in the open position during normal plant operation. The licensee captured this issue in their CAP as AR 02668759 and created a special log to monitor the associated breakers once per day. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of system and/or function, 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 (TS) allowable outage time, and represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant for greater than 24 hours. Specifically, a historical review did not find an example where the swing DG was non-functional for a period greater than the applicable TS allowable outage time as a result of this finding during the last year. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the mean to detect an opened breaker associated with the affected loads was established more than 3 years ago. | |
| 05000373/FIN-2016007-03 | 2016Q2 | Green | NRC identified | Inadequate Procedures for Containment Debris Management | The team 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 failure to establish procedures that were appropriate to manage containment debris consistent with the emergency core cooling system strainer debris loading design basis and supporting design information. Specifically, the procedures did not contain instructions for evaluating containment debris sources consistent with the associated analyses and other design documents. The licensee captured the team concerns in their CAP as AR 02663076 and AR 02656299. The immediate corrective actions included an operability evaluation that reasonably determined all of the affected emergency core cooling system strainers remained operable. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that 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, the licensee performed an operability review and reasonably determined that only a portion of the unqualified coatings would be available for transport to the strainers and this quantity was bounded by the associated design basis analysis. In addition, this review reasonably determined that sufficient analytical margin existed to accommodate the quantities of the other debris types found during recent inspections. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the associated procedures were established more than 3 years ago. | |
| 05000373/FIN-2016002-02 | 2016Q2 | Green | H.1 | NRC identified | Failure to Implement and Maintain Written Procedures Regarding Respirator Fit Testing | A finding of very-low safety significance and an associated NCV of 10 CFR 20.1703 was identified by the inspectors on May 11, 2016, for the licensees failure to implement and maintain written procedures regarding respirator fit testing. The issue was entered into the licensees CAP as AR 2668632. Corrective actions included invalidating the results for the observed test, removing the qualification from the technician that performed the tests, reaffirmed the procedure requirements with all technicians through a read and sign process, and requested several changes to the fit test procedure RPAA444, Controlled Negative Pressure (CNP) Fit Testing to improve alignment to requirements in 29 CFR 1910.134, Appendix A, Fit Testing Procedures (Mandatory). The inspectors determined that not consistently performing fit tests in accordance the methods described in 29 CFR 1910.134, Appendix A, was a performance deficiency, the failure of which was reasonably within the licensees ability to foresee and prevent. This performance deficiency was determined to be more than minor, because it was associated with program and process attribute of the Occupational Radiation Safety cornerstone and affected its objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the respirator fit testing was being used to certify respirator protection factors of workers which were relied upon to provide protection of workers and any discrepancy affected the licensee's ability to control and limit radiation exposures from airborne sources. The inspectors determined that the finding had a very-low safety significance (Green) because the finding did not involve: (1) as-low-as-is-reasonably-achievable planning and controls, (2) a radiological overexposure, (3) a substantial potential for an overexposure, or (4) a compromised ability to assess dose. This finding had a cross-cutting aspect in the area of Human Performance, Resources. Specifically, the licensee did not ensure that personnel equipment, procedures, and other resources were available and adequate to support nuclear safety (IMC 0310 H.1). |
| 05000373/FIN-2016002-01 | 2016Q2 | Green | H.6 | NRC identified | Partial-Length Rods Exceeded Burnup Limit in Design Basis Method of Analysis | A finding of very low safety significance and a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (CFR) 50.59, Changes, Tests, and Experiments, was identified by the inspectors for the licensees failure to provide a written evaluation which provided the basis for the determination that exceeding the peak burnup limit of 62 gigawatt days per metric ton of uranium (GWd/MTU) for fuel did not require a license amendment. Specifically, the licensee failed to provide a basis supporting the application of an alternate burnup limit, which exceeded their current licensing basis limits to the radiological consequence analysis or alternate source term (AST) analysis. The licensee entered this issue into the corrective action program (CAP) as Action Request (AR) 02537519 and AR 2537664. Regarding corrective actions, the effected fuel bundles in Unit 1 were removed from the vessel following the last refueling outage in the first quarter of 2016. As for Unit 2, the licensee planned to restore compliance prior to the burnup exceeding the 62 GWd/MTU limit. The performance deficiency was determined to be more than minor because the inspectors could not reasonably determine that the activity of operating beyond limits documented in the UFSAR would not have required prior NRC approval. The inspectors answered No to all of the Barrier Integrity screening questions. Therefore, this issue screens as having very low safety significance (Green). Because violations of 10 CFR 50.59 potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the finding screened as having very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance, Design Margins because the licensee did not operate and maintain equipment within design margins (IMC 0310 H.6). |
| 05000373/FIN-2016007-04 | 2016Q2 | Green | P.3 | NRC identified | Alternate Shutdown Procedures Failed to Ensure RCIC MOVs Supply Breakers Were Closed | The team identified a finding of very-low safety significance (Green) and associated NCV of the LaSalle County Station Operating License for the failure to ensure that procedures were in effect to implement the alternate shutdown capability. Specifically, the abnormal operating procedures (AOPs) established to respond to a fire at the main control room did not include instructions for verifying that supply breakers for three reactor core isolation cooling motor-operated valves (MOVs) were closed to ensure they could be operated from the remote shutdown panel. Fire-induced failures could result in tripping MOV power supply breakers prior to tripping the MOV control power fuses. The licensee captured the team concerns in their CAP as AR 02668854 and established compensatory actions to reset the affected breakers, if required The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external events (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 finding screened as of very-low safety significance (Green) because it was assigned a low degradation factor. Specifically, the procedural deficiencies could be compensated by operator experience/familiarity and the fact that the AOPs included steps to verify other breakers at the same locations were closed would likely prompt operators to close the remaining breakers. The team determined that this finding had a cross cutting aspect in the area of problem identification and resolution because the licensee failed to take effective corrective actions for a similar issue identified in 2014. Specifically, the resolution of this issue included actions to revise the affected AOPs to include verifying all the reactor core isolation cooling MOVs supplied breakers were closed. However, the licensee failed to include all of the MOVs in the revised AOPs. (P.3) |
| 05000373/FIN-2016007-01 | 2016Q2 | Green | NRC identified | Failure to Monitor the Fouling Conditions of the CSCS Equipment Area Coolers | The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to monitor the fouling conditions of the core standby cooling system (CSCS) equipment area coolers. Specifically, the licensee did not develop performance test procedures to assess the fouling conditions of the safety-related CSCS equipment area coolers and did not have acceptance criteria that delineate when to remove accumulations. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02665463 and established a standing order for operations to impose more restrictive service water temperature limits to reasonably assure the operability of the affected coolers until long term corrective actions were implemented to restore compliance. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The 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, the licensee reviewed actual service water temperature values measured during the last 12 months, performed an operability evaluation, and concluded that the historical temperatures did not exceed the operability limits established by the operability evaluation. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance. Specifically, the test program for the CSCS equipment area coolers was developed in the decade of 1990s. | |
| 05000373/FIN-2016001-02 | 2016Q1 | NRC identified | Partial Length Rods Exceeded Burnup Limit in Design Basis Method of Analysis | (Opened) Unresolved Item 05000373; 05000374/201600102: Partial Length Rods Exceeded Burnup Limit in Design Basis Method of Analysis Introduction: The inspectors identified an URI related to the licensees use of partial length fuel rods beyond the burnup limit specified in their design basis method of analysis. The inspectors could not resolve the issue of concern during the inspection period due to the need for additional information. Description: In action request (AR) 01647125, Exelon-corporate identified a concern with the potential excessive exposure in partial length rods for LaSalle Unit 1 Cycle 16. This issue also affects the core design of the current Unit 2 cycle. Subsequently, AR 02537519 was written to document the condition as it relates to the Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000. Footnote 11 of the RG, provides an applicability limitation of 62,000 megawatt days per metric ton of uranium for non-loss-of-coolant accident gap release fractions specified in Table 3 of the RG. This RG is the licensees NRC-approved method of analysis for alternate source term as described in the LaSalle UFSAR, and is therefore a part of the licensees design and licensing basis. With this proposed change to the application of RG 1.183, the licensee performed a 50.59 evaluation to review the potential impact of partial length rods operating above 62,000 megawatt days per metric ton of uranium burnup to determine if prior NRC approval was necessary to implement the change. The inspectors have reviewed the licensees 50.59 evaluation, FCP 397411, Revision 1, and calculation, L003067, Revision 2C, and have identified an issue of concern. Specifically, the licensee concluded that prior NRC review and approval were not needed to operate its fuel in a manner that deviated from the limitations delineated within the NRC-approved methodology of RG 1.183 in their current licensing basis. The inspectors are opening this URI because more information/guidance is needed from the NRC Headquarters Office of Nuclear Reactor Regulation to determine if this issue of concern represents a violation of regulatory requirements. (URI 05000373; 05000374/201600102: Partial Length Rods Exceeded Burnup Limit in Design Basis Method of Analysis). | ||
| 05000373/FIN-2016403-01 | 2016Q1 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000373/FIN-2016001-01 | 2016Q1 | NRC identified | Adequacy of Changes to Pool Swell Analysis | (Opened) Unresolved Item 05000373; 05000374/201600101: Adequacy of Changes to Pool Swell Analysis Introduction: The inspectors identified an unresolved item (URI) related to the licensees changes to the assumptions in their design basis method of analysis associated with the pool swell calculation of record. The inspectors could not resolve the issue of concern during the inspection period due to the need for additional information. Description: While reviewing the recent revision to operability evaluation 12003, the inspectors identified an issue of concern regarding the licensees changes to the assumptions of the design basis calculation of record for the loss of coolant accident suppression pool swell analysis. This operability evaluation assessed the effects of a previous error identified by the licensee in the design calculation and incorporated additional changes in the design assumptions which resulted in the recapture of significant amount of margin in the analysis. Specifically, the licensee changed the initial blowdown characteristics from all air to an air/steam mixture, which improved the margin of the analysis. The inspectors are evaluating the changes against the guidance of IMC 0326, Operability Determinations & Functionality Assessments for Conditions Adverse to Quality or Safety. Additionally, the inspectors are reviewing whether or not regulatory relief was/is required to be sought by the licensee from the NRC for the ASME Code requirement per the guidance in IMC 0326. The inspectors are opening this URI because more information/guidance is needed from the NRC Headquarters Office of Nuclear Reactor Regulation to determine if this issue of concern represents a violation of regulatory requirements. (URI 05000373; 05000374/201600101: Adequacy of Changes to Pool Swell Analysis). | ||
| 05000374/FIN-2016001-03 | 2016Q1 | Green | H.5 | Self-revealing | Failure to Maintain Appropriate Work Instructions Led to Lost Parts in the Reactor Vessel | A finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed for the licensees failure to verify zero differential pressure across the jet pump plug seals prior to plug removal, an activity affecting quality, in a manner that was appropriate to the circumstances regarding timeliness of the removal. The verification was required by steps 6.13.1 and 6.12 of work orders (WO) 174735903 and 180438305, respectively. The licensee entered this issue into their CAP as action requests 2466339 and 2508333. Corrective actions planned and completed include performed additional analysis and testing of jet pump plug tooling, revised procedures/work instructions, and planned upgrades to the jet pump plug tooling to increase the margins associated with the forces required to displace the seal from the plug. The performance deficiency was determined to be more than minor because if left uncorrected, it had the potential to become a more significant safety concern. Specifically, the robust physical characteristics of the plugs were such that, if unrecovered and unmitigated, coolant flow through certain peripheral fuel assembly orifices could have become blocked by the plugs and potentially led to fuel melt. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Under Exhibit 4, Barrier Integrity Screening Questions, the inspectors answered No to all of the screening questions. Therefore, this issue screened as having very low safety significance (Green). This finding had a cross-cutting aspect in the area of human performance, work management because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priorityas evidenced by the in-field staff verifying zero differential pressure, but then delaying plug removal due to conflicting activities (e.g., shift turnover). As a result, plug removal was later recommenced without re-verifying that conditions had not changed in the intervening period (IMC 0310, H.5). |
| 05000373/FIN-2015004-02 | 2015Q4 | Green | H.6 | NRC identified | Failure to Ensure that Painting Instructions were Appropriate to Preclude Challenging the Operability of Standby Gas Treatment and Control Room Ventilation Charcoal Filters | The inspectors identified a finding of very low safety significance (Green) and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the licensees failure to have instructions or procedures that were appropriate to the circumstances for activities affecting quality. Specifically, procedure LAP9001, LaSalle In-Plant Painting, Revision 22, did not contain instructions or limitations to safeguard against the potential overloading of the charcoal absorber beds of the safety-related standby gas treatment (SBGT) system or the control room ventilation/auxiliary electrical equipment room (VC/VE) due to the volatile organic compounds (VOC) present in painting products (e.g., paint, primer, thinner, etc.). The performance deficiency was determined to be more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to limit the quantity or type of paint used within the ventilation boundaries of the safety-related SBGT or VC/VE emergency filtration systems could have caused those systems to be unable to perform their safety function in the presence of uncontrolled quantities of VOC. In accordance with IMC 0609, Appendix H, Containment Integrity Significance Determination Process, the inspectors determined the finding to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Design Margins, because design margins were not carefully guarded with special attention placed on safety-related equipment. Specifically, licensee staff failed to recognize the importance of understanding the VOC loading limitations of the SBGT and VC/VE systems with respect to operability, given the large scale of the painting activities throughout the plant (H.6). |
| 05000373/FIN-2015004-04 | 2015Q4 | Green | H.4 | Self-revealing | Entry into an Area with Unknown Dose Rates | The inspectors reviewed a finding of very low safety significance (Green) with an associated NCV of TS 5.7.1, which was self-revealed when a worker received a dose rate alarm from an electronic dosimeter when he entered an area with an unknown dose rate. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening. Specifically, the performance deficiency impacted the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of workers health and safety from exposure to radiation, in that, the unauthorized entry into an area where the dose rates were unknown removed a barrier intended to prevent the worker from receiving unexpected dose. The inspectors concluded that this activity was within the licensees ability to foresee and should have been prevented. This finding was not subject to traditional enforcement since the incident did not result in a significant safety consequence, did not impact the NRCs ability to perform its regulatory function, and was not willful. The finding was assessed using the Occupational Radiation Safety Significance Determination Process, and was determined to be of very low safety significance (Green) because the problem was not an ALARA planning issue, there were no overexposures nor substantial potential for overexposures, and the licensees ability to assess dose was not compromised. The inspectors concluded that the issue involved a cross-cutting component in the human performance area of teamwork due to communication issues that were reported by the licensee during the pre-job brief for the job (H.4). |
| 05000373/FIN-2015004-03 | 2015Q4 | Green | P.6 | NRC identified | Failure to Follow Procedure Associated with Sealed Source Inventory and Leak Testing | The inspectors identified a finding of very low safety significance (Green), and an associated NCV of Technical Specification (TS) requirements for the failure to perform leak tests required by station procedures. The inspectors identified multiple discrepancies with the records that are required to demonstrate that sealed radioactive sources were leak tested to prevent the spread of radioactive contamination. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening." Specifically, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern, in that, the failure to ensure that the sources are free of external contamination could spread radioactive contamination, including alpha contamination that is not readily detected by personnel monitoring equipment, and result in increased exposure to radiation. The inspectors concluded that this activity was within the licensees ability to foresee and should have been prevented. This finding was not subject to traditional enforcement since the incident did not result in a significant safety consequence, did not impact the NRCs ability to perform its regulatory function, and was not willful. The finding was assessed using the Occupational Radiation Safety Significance Determination Process, and was determined to be of very low safety significance (Green) because the problem was not an as-low-as-reasonably-achievable (ALARA) planning issue, there were no overexposures nor substantial potential for overexposures, and the licensees ability to assess dose was not compromised. The inspectors determined that the finding involved a cross-cutting component in the area of problem identification and resolution. Specifically, the licensee did not conduct self-critical and objective assessment of the program and practice (P.6). |
| 05000373/FIN-2015004-01 | 2015Q4 | Green | P.2 | NRC identified | Failure to Perform Required Monthly Fire Extinguisher Inspections per National Fire Protection Association Code | The inspectors identified a finding of very low safety significance (Green) and an associated NCV of LaSalle Units 1 and 2 operating licenses, NFP11 section 2.C.(25) and NFP18 section 2.C.(15), respectively, for failing to ensure that the inspection requirements of National Fire Protection Association (NFPA) 10 for portable fire extinguishers were satisfied. Specifically, on two separate occasions, the licensee failed to perform the required monthly inspection on all applicable portable fire extinguishers in the reactor building due to a deficiency in station procedure, LMSFP21, Monthly Inspection of Portable Fire Extinguishers. The licensee entered this issue into the corrective action program (CAP) as action requests (ARs) 02574270, 02574457, and 02604244. The failure to meet the inspection requirements of NFPA10 for portable fire extinguishers was a performance deficiency. The performance deficiency was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of protection against external factors, including fire, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this performance deficiency could have led to the failure of a fire extinguisher to perform when called upon by station personnel or the fire brigade. The inspectors determined the finding was of very low safety significance (Green) in accordance with IMC 0609 Appendix F, Fire Protection Significance Determination Process. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee failed to initially evaluate the issue thoroughly in order to determine the root cause and extent of condition to prevent subsequent inspections from being missed after the issue was brought to their attention by the NRC inspectors (P.2). |