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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000254/FIN-2018010-01 | 2018Q3 | Quad Cities | Minor Violation | On March 13, 2013, the licensee initiated AR 1487225 to document and evaluate an installed weld nonconformance that attached core spray keep fill line 114342LX to pipe support M983FH1. The licensees immediate evaluation of the nonconforming weld documented that the pipe support would perform its function of restraining the pipe for all loading conditions. Although the inspectors concluded the immediate evaluation provided a reasonable expectation the pipe support would perform its function of restraining the pipe, the inspectors noted the licensee did not provide a more detailed evaluation of the nonconformance using acceptance guidance per procedure OPAA108115, Operability Determinations (CM-1). As a result of the inspectors inquiry, the licensee initiated AR 417429, performed a more detailed operability evaluation in EC 625648, and concluded both the piping and pipe support would be able to meet design allowable stress limits with the nonconforming weld configuration. The inspectors reviewed the current design calculation for pipe support M983FH1, analysis 27.0200.1053.019.031 and performed a field walkdown of the installed piping and pipe support configuration for core spray keep fill line 114342LX to verify the adequacy of information provided used in EC 625648. In addition, licensee procedure OPAA108115 also required corrective action at the next opportunity, normally the next refueling outage, with a provision for deferral with proper documented justification. As a result of inspector inquiry regarding the timeliness of the corrective action, the licensee initiated AR 4177210 which documented the nonconformance repair was deferred from Q1R23 (March 2015) and Q1R24 (March 2017) without documented justification. The licensee plans to repair the nonconforming weld in the upcoming Q1R25 (March 2019) outage. The inspectors determined that this is a minor violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, related to the licensees evaluation and timeliness of corrective action for a safety-related pipe support nonconformance. Screening: The inspectors used Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009 and determined that timeliness of corrective action and the lack of a detailed operability evaluation were minor issues. Specifically, the inspectors compared the weld nonconformance to a calculation error in Example 3a of Appendix E and concluded the issue was minor because licensee EC 625648 provided reasonable justification the nonconforming weld configuration will meet design allowable stress for all loading conditions without modification. Violation: This failure to comply with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000373/FIN-2018003-08 | 2018Q3 | LaSalle | 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. |
| 05000374/FIN-2018003-07 | 2018Q3 | LaSalle | 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-2018003-06 | 2018Q3 | LaSalle | 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. |
| 05000373/FIN-2018003-05 | 2018Q3 | LaSalle | 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-2018003-04 | 2018Q3 | LaSalle | 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. |
| 05000341/FIN-2018003-03 | 2018Q3 | Fermi | Failure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal Service Water Outlet Flow Control Valve | A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and TS 3.7.1 Residual Heat Removal Service Water (RHRSW) System, were self-revealed for the licensees failure to identify a condition adverse to quality on the Division 2 RHRSW outlet flow control valve E1150F068B. Specifically, troubleshooting and the associated post maintenance testing failed to identify and correct a failed anti-rotation key which resulted in an inoperable Division 2 RHRSW system for longer than its TS 3.7.1 allowed outage time. |
| 05000373/FIN-2018003-03 | 2018Q3 | LaSalle | 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. |
| 05000341/FIN-2018003-02 | 2018Q3 | Fermi | Failure to Ensure Electrolytic Capacitors Installed in the Plant Did Not Have Expired Shelf Lives | A finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components was self-revealed when the reactor water cleanup system inlet flow square root converter failed, resulting in a failure of the reactor water cleanup (RWCU) differential flow instrument and loss of automatic isolation function of the RWCU isolation valves. Specifically, electrolytic capacitors were installed in the RWCU system logic that had expired shelf lives, resulting in failures of the automatic isolation function of the RWCU system. |
| 05000373/FIN-2018003-02 | 2018Q3 | LaSalle | 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. |
| 05000341/FIN-2018003-01 | 2018Q3 | Fermi | Failure to Apply Torque Values Described in Maintenance Procedure for Flexible Couplings on Emergency Diesel Generator 12 | A finding of very low safety significance with an associated non-cited violation of Technical Specification 5.4.1.a was self-revealed when plant operators discovered a pencil-thick lube oil leak coming from a flexible coupling on emergency diesel generator 12 during planned surveillance testing. Specifically, a lube oil leak developed when the flexible coupling located between the engine driven lube oil pump and the lube oil filter failed due to improper torque applied to the coupling On April 20, 2018, the licensee was performing a routine slow start surveillance of emergency diesel generator 12 (EDG12), when plant operators noted a pencil-thick lube oil leak from the flexible coupling fastener located between the engine driven lube oil pump and the lube oil filter with the engine running in idle. Plant operators subsequently shut down the engine, discontinued the surveillance, and EDG12 was declared inoperable. The licensee performed an investigation and found the flexible coupling fastener was torqued to 120 in/lbs. Maintenance procedure 35.307.008, Emergency Diesel Generator Engine General Maintenance, Enclosure X, Revision 44 required a torque value of 240260 in/lbs for the size of piping the fastener was on. The coupling was last disturbed in 2011, and the maintenance procedure at that time did not contain information regarding torque values for flexible couplings. A similar flexible coupling fastener failed in 2016 due to inadequate work instructions for torqueing flexible couplings (NCV 05000341/201600401, ADAMS Accession Number ML17030A328), and corrective actions were developed to use the vendor recommended values that had already been added to the maintenance procedure as Enclosure X in 2014. However, the corrective actions did not require all flexible couplings to be checked to ensure they were appropriately torqued. Opportunities existed for the licensee to ensure these flexible couplings were properly torqued according to vendor recommendations, either through scheduled maintenance online or during refueling and forced outages. Therefore, on April 20, 2018, another flexible coupling that was not checked as an extent of condition failed due to an under torqued condition. |
| 05000373/FIN-2018003-01 | 2018Q3 | LaSalle | 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. |
| 05000440/FIN-2018003-01 | 2018Q3 | Perry | Application of ASME Code Case N5133 for the Emergency Service Water Piping Degradations | The inspectors identified an Unresolved Item concerning the Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, and ASME Code requirements for the ESW piping systems with regards to the licensees application of ASME Code Case N5133, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping, Section XI, Division 1. Updated Safety Analysis Report (USAR) Section 9.2.1 describes that the function of ESW system is to provide a reliable source of water to safety-related components required for normal and emergency reactor operation. USAR Table 3.21, Equipment Classification, delineates that the ESW piping system is safety-related and designed in accordance with the requirements of ASME Section III, Subsection ND (Class 3). The regulation in 10 CFR 50.55a(g) requires, in part, that Class 3 components and their supports meet the requirements of ASME Section XI of the ASME Boiler and Pressure Vessel (BPV) Code or equivalent quality standards. The ASME also publishes Code Cases, which provide alternatives to existing Code requirements. The NRC Regulatory Guide (RG) 1.147 identifies that Code Case N5133 provides acceptable alternatives to applicable parts of Section XI, provided it is used with any identified conditions or limitations. Code Case N5133, Section 2(d) requires that a flaw evaluation shall be performed to determine the conditions for flaw acceptance. Section 3 provides accepted methods for conducting the required analysis. In addition, Section 3 requires, in part, that nonplanar flaws shall be evaluated in accordance with the requirements in 3.2. Additionally, Section 5 requires that an augmented volumetric examination or physical measurement to assess degradation of the affected system shall be performed as follows: (a) From an engineering evaluation, the most susceptible locations shall be identified. A sample size of at least five of the most susceptible and accessible locations, or, if fewer than five, all susceptible and accessible locations shall be examined within 30 days of detecting the flaw. (b) When a flaw is detected, an additional sample of the same size as defined in 5(a) shall be examined. (c) This process shall be repeated within 15 days for each successive sample, until no significant flaw is detected or until 100 percent of susceptible and accessible locations have been examined. On June 13, 2018, a through-wall leakage on the 20 ESW piping was identified in CR 201805504. As a result, the licensee invoked the Code Case to evaluate this flaw and permit the degraded ESW piping system to remain in service for a limited period without repair/replacement. The licensees evaluation involved characterization of this flaw as nonplanar, and subsequently, the methodology as described in Section 3.2 of the Code Case was utilized for this nonplanar flaw. Additionally, the licensee identified the five most susceptible and accessible locations in the ESW system and performed examination in accordance with Section 5(a). From the examination of the five additional locations, another localized wall degradation was identified on the 8 ESW pipe elbow on July 10, 2018. The licensee initiated CR 201806205 to document this condition. The licensee characterized this degradation also as a nonplanar flaw, and this degradation represented approximately 80 percent wall loss from its nominal thickness. During the review of the licensee evaluation of this degraded pipe elbow, the inspectors identified that the methodology as described in Section 3.2 of the Code Case had not been utilized. Instead, the licensee elected to use an alternate methodology to evaluate and disposition for its acceptability. Furthermore, the inspectors identified that the licensee essentially redefined the term flaw in the Code Case to reflect the ASME Section XI, IWA9000 definition of the term defect. The ASME Section XI, IWA9000 defines a flaw as an imperfection or unintentional discontinuity that is detectable by nondestructive examination. It also defines a defect as a flaw (imperfection or unintentional discontinuity) of such size, shape, orientation, location, or properties as to be rejectable. With respect to the Code Case, the licensee essentially restricted the criteria for examination scope expansion only to the flaws that were rejectable; therefore, the licensee had not expanded the scope to perform examination of additional locations in accordance with Section 5(b). In essence, two items are to be further evaluated and addressed: (1) whether the use of methodology not described in the Code Case Section 3.2 was appropriate for evaluation of the nonplanar flaw on the 8 ESW pipe elbow, and (2) whether the stopping of scope expansion for examination as required by the Code Case Section 5(b) was appropriate based on the licensees redefining of the term flaw. In response to the inspectors concern, the licensee initiated CR 201808483, NRC ID: Code Case N5133 Interpretation, September 26, 2018. The licensee also plans to perform examination of five additional locations in November of 2018. This represents an item where the inspectors identified Code interpretation issues that resulted in a disagreement with the licensee. This will require additional review to determine whether a violation exists. Therefore, this issue is considered an unresolved item pending completion of inspector review and evaluation and discussion with the Office of Nuclear Reactor Regulation. Licensee Action: The licensee plans to perform examination of five additional locations in November of 2018. Corrective Action Reference: CR 201808483 |
| 05000254/FIN-2018010-02 | 2018Q3 | Quad Cities | Minor Violation | In 2009, due to a control room envelope differential pressure test failure at another nuclear station, the licensee completed an engineering change to develop separate correction factors (uncertainty) for different test methods. The engineering change recommended procedure QCOS 575016, Control Room Envelope DP (Differential Pressure) Surveillance, be revised to perform additional testing using alternate test methods with reduced correction factors. However, the procedure was not revised. In January 2016, during the control room envelope differential pressure test, seven areas failed the acceptance criteria. The licensee utilized the engineering change and performed a temporary procedure change to use a different method with a reduced correction factor for testing. The test was successfully performed and acceptance criteria were met. After the test, the procedure reverted to the old method. A condition report was written regarding this issue and actions were assigned to perform repair of the control room boundary and determine if a more accurate instrument is needed. Although repairs were made to the control room boundary, the licensee has not yet determined if a more accurate instrument is needed. Also, the procedure was not revised to use the alternate methods. The inspectors determined this is a minor violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for not having procedure appropriate to the circumstances. Screening: The inspectors determined this issue is not more than minor because the existing procedure is not incorrect but missing the steps the licensee could take when unsatisfactory results are obtained. Violation: This failure to comply with 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000373/FIN-2018002-04 | 2018Q2 | LaSalle | 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 | LaSalle | 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 | LaSalle | 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 | LaSalle | 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. |
| 05000341/FIN-2018002-01 | 2018Q2 | Fermi | Failure to Document a Condition Assessment Resolution Document for Reactor Recirculation Motor-Generator Set A Brush Gear Sparking | A self-revealed Green finding was identified for failure to document a Condition Assessment Resolution Document (CARD) for 5-inch rooster tail sparking on reactor recirculation motor-generator set A brush gear, which ultimately resulted in a manual recirculation pump A trip and plant transient. |
| 05000341/FIN-2018002-04 | 2018Q2 | Fermi | Failure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal Service Water Outlet Flow Control Valve | A self-revealed TBD finding and an associated apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and Technical Specification 3.7.1 Residual Heat Removal Service Water (RHRSW) System, were identified for failure to identify a condition adverse to quality while performing corrective maintenance on Division 2 RHRSW outlet flow control valve E1150F068B prior to returning the Division 2 RHRSW system to service. Specifically, troubleshooting and associated post maintenance testing failed to identify and correct a failed anti-rotation key which resulted in an inoperable Division 2 RHRSW system for longer than its Technical Specification 3.7.1 allowed outage time. |
| 05000341/FIN-2018002-03 | 2018Q2 | Fermi | Failure to Adequately Evaluate the Operability of Emergency Diesel Generator11 | A finding of very low safety significance was self-revealed for the licensees failure to adequately evaluate the operability of a condition adverse to quality identified on Emergency Diesel Generator (EDG) 11. Specifically, a lube oil leak was evaluated as having no impact to the operation of the emergency diesel generator. However, during the next surveillance run of EDG 11, the engine had to be shut down and declared inoperable due to the lube oil leak degrading during operation. |
| 05000341/FIN-2018002-02 | 2018Q2 | Fermi | Inadequate Preventative Maintenance in Residual Heat Removal Service Water System Outlet Flow Control Valves Results in Lower Bonnet (Backseat) Bushing Failure | A self-revealed Green finding and associated non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) Part 50, Appendix Criterion V, Instructions, Procedures, and Drawings were identified for failure to ensure activities affecting quality were prescribed in a manner consistent with the circumstances to the residual heat removal service water system(RHRSW). Specifically, preventative maintenance procedure M681 failed to establish an appropriate interval and guidance for periodic valve internals inspections on the Division 2 RHRSW system outlet flow control valve to prevent significant degradation from galvanic corrosion given known internal and external operating experience |
| 05000461/FIN-2017007-01 | 2017Q3 | Clinton | Failure to Evaluate Defeating Reactor Core Isolation Cooling System Interlocks and Trips before Adding Them to an Emergency Operating Support Procedure | The inspectors identified a Severity Level IV NCV of Title 10 Code of Federal Regulations (CFR) 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the change to an emergency operating procedure (EOP) support procedure to incorporate three reactor core isolation cooling (RCIC) system interlock defeats and did not provide a basis for the determination that this change would not create a possibility for a malfunction of a structure, system or component (SSC ) important to safety with a different result than any previously evaluated in the updated safety analysis report. The licensee entered this issue into the CAP as action request ( AR ) 04056394 and planned to perform a screening for the procedure change. 3 This performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the change did not ensure RCIC system reliability and availability during and following design basis accidents because it introduced a new failure mode and added reliance on monitoring activities and manual actions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non- technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV significance because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The team determined that this finding had a cross -cutting aspect of Resources in the area of Human Performance because the licensee did not ensure that procedures, and other resources were available and adequate to support nuclear safety. Specifically, the procedure which required a 50.59 screening for changes to EOP support procedures, was not explicit in requiring the screening. (H.1) |
| 05000454/FIN-2017007-02 | 2017Q3 | Byron | Failure to Promptly Identify Degraded Reactor Containment Fan Cooler Circuitry | The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly identify a condition adverse to quality resulting in a safety -related system becoming inoperable. Specifically, from May 5, 2017, to August 4, 2017, the licensee failed to trend available surveillance data in a timely manner and did not identify a degraded condition in the 1A reactor containment fan cooler (RCFC) time delay circuitry prior to the system becoming inoperable. The licensee entered this issue into their CAP as AR 04039037 and AR 04045767, replaced the failed relay, and planned to update the RCFC system monitoring plan to note abnormal changes in time delay relay actuation times and improve coordination between engineering and operations to reduce the time it takes engineering to obtain RCFC surveillance data for trending after surveillances are completed. The inspectors determined that the failure to promptly identify a condition adverse to quality associated with the time delay relay circuitry in the 1A RCFC was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify a degraded condition in the time delay circuitry associated with the 1A RCFC resulted i n a missed opportunity for the licensee to evaluate the cause and initiate prompt actions to respond to the degraded condition prior to the failure. The inspectors answered No to questions A.1 through A.4 of IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions ; therefore, the finding screened as 4 having very low safety significance. This finding affected the Cross -Cutting area of Problem Identification and Resolution in the aspect of Trending because information was available that indicated a degraded condition in the 1A RCFC time delay relay circuitry for three months prior its failure in August , but was not identified and evaluated by the licensee prior to failure (P.4) . |
| 05000454/FIN-2017007-01 | 2017Q3 | Byron | Fai lure to Perform Maintenance in Accordance with Performance Centered Maintenance Template | The inspectors identified a finding of very low safety significance and an associated NCV of TS 5.4.1, Procedures, when licensee personnel failed to perform maintenance in accordance with written procedures as required by Regulatory Guide 1.33. Specifically, from February 3, 2014, through August 25, 2017, the licensee failed to develop and execute work instructions of sufficient scope to accomplish the 3 preventive maintenance to replace flexible hoses on the essential service water (SX) makeup pumps and the diesel driven auxiliary feedwater (AFW) pumps and did not have a technical justification for a deviation from the Exelon Corporate Performance Centered Maintenance (PCM) template. The licensee entered this issue into their CAP as Action Request (AR) 03961955, AR 03971962, and AR 04045769 and planned to replace the flexible hoses at the next available opportunity. The inspectors determined that failure to perform maintenance in accordance with written procedures as required by TS 5.4.1, Procedures, and Regulatory Guide 1.33 was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences . Specifically, failing to replace flexible hoses on the SX makeup pumps and the Unit 1 and Unit 2 diesel -driven AFW pumps at a pre - established frequency could allow hose degradation to remain unidentified and lead to the unplanned inoperability of these safety-related systems. Since the finding is a deficiency affecting the design or qualification of mitigating systems, structures and components (SSC s) and the SSC s remained operable and functional, the finding screened as having very low safety significance. This finding affected the C ross -Cutting area of Human Performance in the aspect of Work Management because the licensee failed to perform required maintenance in accordance with their associated maintenance strategy as well as the corporate PCM template (H.5) . |
| 05000331/FIN-2017007-01 | 2017Q2 | Duane Arnold | Failure to Include Valves in the Inservice Testing (IST) Program | Green . The inspectors identified a finding and an associated non- cited violation of Title 10 of the Code of Federal Regulations (10 CFR ) 50.55a(f)(1) for the licensees failure to scope in multiple check valves of the main steam isolation valve leakage treatment system (LTS) into the Inservice Testing (IST) Program. Specifically, these valves were credited to mitigate the consequences of the main steam isolation valve leakage following a loss of coolant accident but they were not scoped into the IST program. Since the licensee made a commitment to the NRC to put these valves into the IST p rogram as part of License Amendment 207, this issue is also a Deviation in accordance with the NRC Enforcement Policy. The licensee put this issue into the CAP 3 as Action Requests ( AR s) 2193481 and 2193482 and planned to include these v alves in the full IST program . This performance deficiency was m ore than minor because i f left uncorrected, there was a potential to lead to a more significant safety concern. Specifically, th ese valves that were credited to mitigate the consequence of an accident were not tested in accordance with the IST program. T he finding screened as very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment , containment isolation system , and heat removal components , nor did it involve an actual reduct ion in function of hydrogen igniters in the reactor containment . The inspectors determined this finding affected the cross -cutting area of problem identification and resolution in the aspect of evaluation because the licensee justified that the valves be put into the augmented IST program since they were non- code components . In addition, the licensee did not re -scope these components into the IST program when 10 CFR 50.55(f)(1) was changed in 1999. This misconception continued when the licensee discovered several valves of the LTS were not in the IST program scope in 2015 . (P.2) |
| 05000341/FIN-2017007-01 | 2017Q2 | Fermi | Failure to Correct a Design Deficiency that Mis-Quantified Unidentified Leakage | Green . The inspectors identified a finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) , Part 50, Appendix B, Criterion XVI, Corrective Actions , for the licensees failure to correct a design deficiency that mis -quantified unidentified leakage from reactor coolant system (RCS ) pressure boundary . Specifically, in April 2007, the licensee identified that the driver mount drain for the reactor recirculation pump could potentially drain leakage from nearby pipe cracks to the identified leakage collection point. However, the licensee had 3 not correct ed this design deficiency as of the start of this inspection. The licensee documented this issue into the CAP as Condition Assessment Resolution Document (CARD) 17 25489 and developed a night order to direct the operators how to calculate unidentified leakage. The licensee also planned to revise procedure 24.000.02 as an interim measure until the modification was implemented. The inspectors determined that the licensees failure to correct the design deficiency that mis-quantified unidentified leakage is a performance deficiency that is reasonably within the licensees ability to foresee and correct. The inspectors determined that this issue is more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Specifically, leakage that would normally be collected and measured as unidentified leakage could be collected and measured as identified leakage, leading to a potential violation of the TS unidentified leakage rate . Because the finding did not represent a loss of system or function, or represent an actual loss of function of at least a single train for greater than its Technical Specification (TS) Allowed Outage Time, or represent an actual loss of function of one or more non TS trains of equipment designated as high safety -significant in the licensees Maintenance Rule Program, it was screened as very low safety significance. The inspectors did not identify a cross -cutting aspect since the issue originated more than three years ago. |
| 05000440/FIN-2016004-03 | 2016Q4 | Perry | RCS Pressure Boundary Leakage Operation Prohibited by TSs | Green. A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification (TS) 3.4.5, RCS Operational Leakage, was self-revealed when the licensee operated with reactor coolant system (RCS) pressure boundary leakage, as a result of the failure of the weld connecting the root appendage of the vent line on the recirculation loop A discharge valve, between January 19, 2016, and January 24, 2016, which is a condition prohibited by TS. The licensee entered this issue into the Corrective Action Program (CAP) as Condition Report (CR) 201601071 and performed a significant condition adverse to quality root cause evaluation due to a principal safety barrier being seriously degraded, replaced the vent line appendage on the recirculation loop A discharge valve with a more robust pipe and cap, and developed plans to replace ten additional vent and drain line appendages on the reactor recirculation loops prior to the end of the 1R17 refueling outage in 2019. The inspectors determined that the licensees operation with RCS pressure boundary leakage, a condition prohibited by TSs, was a performance deficiency requiring evaluation. The inspectors determined that the finding was more than minor because it adversely impacted the Initiating Events cornerstone attribute of equipment performance-barrier integrity, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined this finding was of very low safety significance because the leak would not have exceeded the RCS leak rate for a small loss-of-coolant accident (LOCA) and would not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function. The inspectors concluded that this finding had no additional cross-cutting aspects than what was discussed in Inspection Report 0500440/2016001. |
| 05000220/FIN-2016004-01 | 2016Q4 | Nine Mile Point | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Technical Specification 6.4.1 for Unit 1 and TS 5.4.1 for Unit 2 state that, Written procedures shall be established, implemented, and maintained covering the following activities: ... d. Fire Protection Program Implementation. Procedure OP-AA-201-003, Fire Drill Performance, implements portions of the Exelon fire protection program, and OP-AA-201-003 states fire drills shall be conducted quarterly for each shift fire brigade. Contrary to the above, Exelon failed to correctly implement its fire brigade training program procedure. Specifically, Exelon failed to conduct fire drills for brigade teams A, C, and D for the third quarter 2015 as required by OP-AA-201-003 because of scheduling conflicts caused by emergent and planned work and station activities. For fire brigade A, the brigade was initially scheduled to perform a third quarter fire drill on July 22, 2015. However, that drill as well as subsequent drills which were scheduled to be performed on August 25, 2015, and September 30, 2015, respectively, were cancelled to facilitate other work activities. For fire brigade C, the third quarter drill which was scheduled for September 8, 2015 was cancelled because of station work load. For fire brigade D, a third quarter drill that was scheduled for September 24, 2015, was also canceled because of other work activities. On October 2, 2015, Exelon documented the missed quarterly drills in the CAP as IR 02564520. During the fourth quarter, crews A, C, and D each conducted a make-up drill as well as a regular quarterly drill. However on October 31, 2015, crew C failed its make-up drill. Crew C was remediated and passed a repeat drill on November 10, 2015. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because if left uncorrected, it could lead to a more significant safety concern. The inspectors decision was informed by examples 3j and 3k in IMC 0612, Appendix E, Examples of Minor Issues. The examples refer to an issue not being minor if significant programmatic deficiencies were identified with the issue that could lead to worse errors if left uncorrected. Specifically, 60 percent of the NMPNS fire brigade teams had missed their quarterly fire drill requirement which is indicative of a programmatic issue and there is a reasonable concern as to the effectiveness of the fire brigade since the required training had not been completed and one crew subsequently failed its next drill. Based on IMC 0609, Attachment 4, Initial Characterization of Findings, findings that involve discrepancies with the fire brigade are directed to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. From IMC 0609 Appendix A Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that a the final significance must be determined using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding was determined to have very low safety significance (Green) because a prior similar violations significance bounded this findings significance. The prior similar violation occurred at NMPNS, which was documented in inspection report 05000220/410/2009006-01 as an NCV. Because this violation was determined to be of very low safety significance and entered into the CAP in IR 02564520, it is being treated as an NCV, consistent with section 2.2.3 of the NRC Enforcement Policy. |
| 05000373/FIN-2016004-01 | 2016Q4 | LaSalle | 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-2016004-02 | 2016Q4 | LaSalle | 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-2016004-03 | 2016Q4 | LaSalle | 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 | LaSalle | 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) |
| 05000440/FIN-2016004-01 | 2016Q4 | Perry | ECC B Heat Exchanger Flow Root Valves Out of Position | Green. A finding of very-low safety significance and associated NCV of TS 5.4.1, Procedures, was self-revealed for the licensees failure to follow valve lineup procedure restoration requirements after an emergency service water (ESW) pump B and valve operability test. Specifically, incorrect valve manipulations of the root valves for 1P42R043B and 1P42R043A flow indicators caused the emergency closed cooling (ECC) heat exchanger B flow to read zero with flow through the heat exchanger. The incorrect flow indication rendered the remote shutdown panel inoperable. The licensee subsequently re-positioned the root valves, 1P42R043B and 1P42R043A, and restored the remote shutdown panel to operable. The licensee entered this issue into the CAP as CR 201612935. The inspectors determined that the performance deficiency for failure to follow procedure was more than minor and thus a finding because it was associated with the Mitigating Systems cornerstone attribute of human performance. The performance deficiency 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 has a cross-cutting aspect in the area of human performance, avoid complacency because the licensee failed to ensure that individuals follow processes, procedures, and work instructions. Specifically the individual performing the surveillance did not utilize all the required human performance tools to prevent the error (H.12). |
| 05000440/FIN-2016004-02 | 2016Q4 | Perry | Modifications to Underdrain and Gravity Discharge System Manhole Covers Without a 10 CFR 50.59 Safety Evaluation | Green-Severity Level IV. The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the installation of grated manhole covers, replacing the rubber gasket, watertight manhole covers for the underdrain and gravity discharge systems and did not provide a basis for the determination that this change would not result in a more than a minimal increase in the likelihood of occurrence of a malfunction of a system structure or component important to safety. The licensee entered this issue into the CAP as CR 201611864 and performed a prompt operability determination to show that the underdrain and gravity drain systems remained functional while the engineering change package was developed to support the change and bring the underdrain and gravity discharge systems into compliance with the design basis. The performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone attribute of protection against external factors and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Per IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Mitigating Systems Screening Questions and determined to be of very low safety significance (Green) because the finding did not cause the underdrain and gravity discharge systems to become inoperable or non-functional. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, procedure adherence, in that individuals did not follow processes, procedures, and work instructions. Specifically, a design engineer authorized the permanent modification to be made without the required 50.59 evaluation being completed (H.8). |
| 05000440/FIN-2016004-04 | 2016Q4 | Perry | Failure to Notify the NRC within Eight Hours of a Non-Emergency Event that Could Have Prevented the Fulfillment of a Safety Function | Severity Level IV. The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.72(b)(3)(v)(A) and (D), for the licensees failure to report to the NRC within eight hours, an event or condition that could have prevented the fulfillment of a safety function. The licensees evaluation of this condition, where both trains of the standby liquid control (SLC) system had been inoperable simultaneously, determined that it was not a reportable event. However, the inspectors determined that as described in NUREG 1022, Event Reporting Guidelines 50.72 and 50.73, Revision 3, Section 3.2.7, the licensee had failed to make a non-emergency eight hour report as required by 10 CFR 50.72(b)(3)(v)(A) and (D). The licensee submitted the eight-hour report on December 30, 2016, and entered this issue into the corrective action program (CAP) as CR 201700098. The failure to make an applicable non-emergency eight-hour event notification report within the required time frame was determined to be a performance deficiency. The inspectors determined that traditional enforcement was applicable to this issue because it impacted the NRC's regulatory process. In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d.9 of the NRC Enforcement Policy, this violation was screened as a Severity Level IV violation that was more than minor. In accordance with IMC 0612, because this violation involved traditional enforcement and does not have an underlying technical violation that would be considered more-than-minor, a cross-cutting aspect was not assigned to this violation. |
| 05000440/FIN-2016002-01 | 2016Q2 | Perry | Failure to Comply with ODCM During Liquid Effluent Discharge | A finding of very low safety significance, and an associated NCV of Technical Specification (TS) 5.5.1 was identified by the NRC inspectors for the failure to follow Offsite Dose Calculation Manual (ODCM) requirements during the execution of a liquid effluent discharge. The license entered this event into their CAP as CR201607572 and the individual was coached regarding procedure compliance. The inspectors determined that the performance deficiency was more than minor because the issue impacted the program and process attribute of the Public Radiation Safety cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, on February 1, 2016, a liquid effluent discharge was performed with the radwaste to essential service water discharge monitor inoperable and without the required independent verification of release rate calculations. The finding was determined to be of very low safety significance (Green) because it was not a failure to implement the Effluent Program, nor did public dose exceed Appendix I or Title 10 of the Code of Federal Regulations (CFR), Part 20.1301(e) criteria. The inspectors concluded that the finding had a cross-cutting aspect in the human performance area of procedure adherence because procedures for this task were not followed (IMC 0310, H.8). |
| 05000440/FIN-2016002-02 | 2016Q2 | Perry | Licensee-Identified Violation | Title 10 of the CFR, Part 50.65(a)(4) states, in part, Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, the licensee identified that it failed to perform the required risk assessment prior to commencing maintenance activities. Specifically, on April 28, 2016, the licensee racked out the L1006 breaker, which was the alternate supply to bus L11 from the Unit 1 startup transformer. The At-The-Controls licensed reactor operator then questioned whether or not an availability log entry was required for this breaker being racked to the disconnect position. At that time, the unit supervisor ran the probabilistic risk assessment (PRA) program for the L1006 breaker being racked to the disconnect position with the plant in its current configuration. The result of the PRA resulted in an increase in risk from Green Risk to Yellow Risk. The Unit Supervisor stopped the evolution and directed the breaker to be racked back in. After an evaluation of the schedule was performed and permission was received from the general plant manager to continue with the plant in Yellow PRA Risk, breaker L1006 was racked out and preventive maintenance proceeded. The finding was determined to be of very low safety significance because the risk deficit incremental core damage probability risk assessment increase of 6.6E10 was less than the 1E6 threshold. The licensee initiated CR201606093 to address this issue. |
| 05000440/FIN-2016002-03 | 2016Q2 | Perry | Licensee-Identified Violation | Technical Specification 5.5.1, Offsite Dose Calculation Manual (ODCM), requires, in part, that radioactive effluents control information be contained within the ODCM. ODCM, Revision 20, requires that the liquid radioactive waste to essential service water radiation monitor have periodic channel functional tests performed. Contrary to the above, on December 13, 2015, it was identified by the licensee that this test, which was required on September 13, 2015, had not been performed. On August 26, 2015, the licensee identified that the monitor would not pass the required functional test. The licensee incorrectly deferred the required channel functional test. This was identified by the operation department staff. The licensee documented this issue in CR201516711, December 13, 2015. The finding was determined to be of very-low safety significance (Green) because it was not a failure to implement the Effluent Program, nor did public dose exceed Appendix I or 10 CFR 20.1301(e) criteria. |
| 05000440/FIN-2015004-02 | 2015Q4 | Perry | Liquid Effluent Calibration | The inspectors identified that the efficiency calibration for the liquid effluent radiation monitor, 0D17K0606, could not be located. The licensee performed a new efficiency determination on the monitor during digital modification upgrade on 2006 for all effluent monitors. According to the licensee, the calculated count rate using a new standard National Institute of Standards and Technology traceable sources indicated a close approximation to the liquid detector count rates data determined during the detector initial (primary) calibration. To date, the licensee was unable to provide the initial calibration paperwork indicating that the calibration count rates for the detector efficiency determinations were correlatable. The inspectors attempted to assess whether the original standard count rates for efficiency determination were correlatable to the initial calibration paperwork; however, this assessment could not be completed within this inspection period. The issue remains under review by the U.S. Nuclear Regulatory Commission (NRC) pending further information from the licensee, and is categorized as an Unresolved Item (URI) pending completion of that NRC review. |
| 05000254/FIN-2015004-01 | 2015Q4 | Quad Cities | EAL Threshold Values Were Not Revised | An Unresolved Item (URI) was identified because additional information is required to determine whether a performance deficiency that is more than minor exists, and if a violation of 10 CFR 50.54(q)(2), which requires a licensee to develop and maintain an emergency plan that meets the requirements of 10 CFR 50.47(b), and 10 CFR Part 50, Appendix E, had occurred. The licensee identified an issue of concern when the Quad Cities General Abnormal procedures (QGAs) were revised with a new value for Minimum Steam Cooling Reactor Pressure Vessel Water Level (MSCRWL) but the associated EALs that use the MSCRWL value as an EAL threshold were not revised. On March 12, 2015, the QGAs were revised with a new value for MSCRWL. However, the site EALs that should use the revised QGA value as an EAL threshold value were not revised. The licensee scheduled the revisions of the QGAs to support implementation of changes that were associated with the diverse and flexible coping strategies (FLEX) implementation and the sites transition to new Optima2 fuel. Both of the changes were scheduled to be implemented in March 2015 during the Quad Cities Unit 1 Refueling Outage as part of a revision package. Because of the new fuel, the MSCRWL value changed from -166 inches to -190 inches. On April 28, 2015, the licensee identified the EALs were not changed to correspond with the new MSCRWL values incorporated in the QGAs. The specific EALs that are affected are MG2 and FG1, which are used to determine if a General Emergency should be declared based on the MSCRWL value. Since the value remained at -166 inches, the licensee concluded that the issue could have potentially caused, under certain conditions, the site to declare a General Emergency earlier than needed and issue an unnecessary Protective Action Recommendation (PAR) to the public. Following identification of the issue, the licensee implemented the appropriate changes to EALs MG2 and FG1 on April 30, 2015. Since there was a discrepancy between the QGAs and the EAL threshold values that could have affected the timely and accurate classification of a General Emergency, a potential performance deficiency exists. However, in order to determine if the performance deficiency is more than minor significance, additional information is needed. The URI was identified pending additional information and inspection follow-up. Specifically, additional information is required to: understand if the discrepancy in the MSCRWL values documented in the QGAs and the EALs would have led to an overclassification of a General Emergency and issuance of an unnecessary PAR; understand if there are events that could be postulated where the -166 inches could be exceeded without reaching the -190 inches; and understand the timeline from when the fuel was transitioned to Optima2 until the discovery of this issue. This information will assist the inspectors to determine if the performance deficiency is more than minor and if a violation of 10 CFR 50.54(q)(2) occurred. |
| 05000254/FIN-2015004-02 | 2015Q4 | Quad Cities | Licensee-Identified Violation | Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established procedure QCOP 650007, Racking in a 4160 V Horizontal Type AMHG or G25 Circuit Breaker, as the implementing procedure for installing fuse blocks in safety-related breakers, an activity affecting quality. Contrary to the above, prior to August 21, 2015, the licensee failed to have a procedure for installing fuse blocks in safety-related breakers. Specifically, QCOP 350007 failed to ensure that fuse blocks for safety-related 4160 Volt breakers were properly installed to ensure the breakers would perform their function. The procedure did not provide the operators guidance to ensure the fuse blocks were fully seated, and on August 21, 2015, following post-maintenance testing for the Unit 1A RHR pump breaker and the system being declared operable, an equipment operator on rounds identified that the breaker closing springs were not charged. The licensee determined that the breaker fuse blocks were not fully seated following breaker maintenance. The licensee captured this issue in the CAP as IR 2550801. The inspectors evaluated the finding under Inspection Manual Chapter 0609, Appendix A, The SDP (Significance Determination Process) for Findings At-Power, issued June 19, 2012. The inspectors answered No to questions A1A4 in Exhibit 2, Mitigating Systems Screening Questions, and determined the finding was of very low safety significance (Green). |
| 05000331/FIN-2015007-04 | 2015Q2 | Duane Arnold | Licensee-Identified Violation | As required, in part, by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Procedure PIAA1041000, Corrective Action, Revision 3, Section 4.10, Step 1, requires, in part, that closure of corrective actions is not permitted until the corrective actions are completed as prescribed or appropriate justification and approval for intent change or cancellation/nonperformance of the corrective action is documented in the condition report. Contrary to the above, on November 24, 2014, a long term corrective action to replace the northwest corner room cooler coils due to copper tube degradation was closed without replacing the cooler coils or obtaining an approval for cancellation of the correction action. Because the cooler was maintained in an operable status and did not initiate a transient based upon the deficient condition, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02016918. Immediate corrective actions included reopening the corrective action assignment and obtaining approval for extending the due date. |
| 05000331/FIN-2015007-02 | 2015Q2 | Duane Arnold | Failure to Correctly Update the Updated Final Safety Analysis Report | The inspectors identified a Severity Level IV NCV of 10 CFR 50.71(e) for failure to assure that the information included in the last update of the updated final safety analysis (UFSAR) report contained the latest information developed. The licensee implemented a change to the UFSAR, in preparation for License Amendment 243 that did not contain the latest information developed. Specifically, Section 5.4.6.1 (page 5.430 of Revision 17) was updated with a note that stated the reactor core isolation cooling system was not safety-related. In fact, the reactor core isolation cooling system had always been designated as safety-related. The licensee entered this issue into the CAP as CR 01974995 and prepared an updated final safety analysis report (UFSAR) change that removed the statement that the reactor core isolation cooling system was not safety-related. The inspectors determined that the update to the UFSAR with incorrect information was a performance deficiency in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on September 7, 2012. The inspectors concluded that traditional enforcement applied because the failure to correctly update the UFSAR impacted the regulatory process. The Enforcement Policy, dated February 4, 2015, Section 6.1.d.3, gave the example that if, a licensee fails to UFSAR as required by 10 CFR 50.71(e) but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures; then this was a Severity Level IV violation. In this case, the UFSAR was updated incorrectly and did not, result in any unacceptable change to the facility or procedures. The inspectors determined this to be a similar example and therefore was more than minor and a Severity Level IV violation. This violation was not associated with a finding that was evaluated by the significance determination process. Therefore, a cross-cutting aspect was not assigned to this traditional enforcement violation. |
| 05000331/FIN-2015007-03 | 2015Q2 | Duane Arnold | Licensee-Identified Violation | As required, in part, by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Procedure PIAA205, Condition Evaluation and Corrective Action, Revision 23, Section 4.10, Step 8, requires, in part, that corrective actions shall not be closed to other existing actions unless the description, scope, condition designation, and intent of the action that will remain open is equivalent to that of the action being closed. Contrary to the above, in December 2013, while performing RCE associated with CR 01884408, the licensee inappropriately credited a long term corrective action developed in an apparent cause evaluation associated with CR 01835557 to accomplish the stated CAPR action. Further, procedure PIAA1041000, Corrective Action, Revision 1, Section 4.11, Step 6, requires, in part, that requests for significance level 1 CAPR due date extensions be documented on PIAA1041000F01, Change Request Form, Revision 1, receive MRC approval for the stated changes, and be attached to the condition report. Contrary to that, on December 5, 2014, the long term corrective action due date was extended 4 months without Management Review Committee approval. The CAPR and long term corrective action were related to the failure of a safety related main steam line temperature instrument. Because the instrument was maintained in an operable status and did not initiate a transient based upon the deficient condition, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02025444 and 02044053. Immediate corrective actions included adding an assignment to correctly restate that the plant modification to replace the temperature indicators as a CAPR in CR 01884408. |
| 05000331/FIN-2015007-01 | 2015Q2 | Duane Arnold | Inappropriate Diesel Generator Maintenance Procedure | The inspectors identified a finding of very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances. Specifically, the licensee implemented GENERAF01001, 1E053A2 (B2) Flange Inspection, Section W, Revision 5, Step 5.1.3.3.b as a corrective action to NCV 05000331/201400902, in order to ensure proper alignment of the 1E053A2 (B2) flange. The procedure was inappropriate for the circumstances because the instructions, as written, in Step 5.1.3.3.b would not result in meeting the acceptance criteria for flange alignment listed in GENERAF01001, 1E053A2 (B2) Flange Inspection, Section W, Revision 5, Attachment 8. The licensee entered this issue into the CAP as condition report (CR) 02041369. The inspectors determined the licensees failure to provide procedures of a type appropriate to the circumstances to assure that for a significant condition adverse to quality, the cause of the condition was determined and corrective actions were taken to preclude repetition was a performance deficiency warranting further review. The inspectors determined that this finding was more than minor because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Misalignment of the flanges could lead to excessive oil leak that rendered the diesel generator inoperable. The inspectors determined the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating system, structure or component and did not result in a loss of operability or functionality. In addition, the finding did not represent a loss of system or function, did not represent an actual loss of function of a least a single train for longer than its technical specification allowed outage time, and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. The inspectors determined this finding had a cross cutting aspect in the area of PI&R, specifically resolution, because licensee personnel failed to take effective corrective actions to ensure that the resolutions address causes and extent of conditions commensurate with their safety significance (P.3). |
| 05000331/FIN-2015001-01 | 2015Q1 | Duane Arnold | Failure to Classify and Declare a Notification of Unusual Event | The inspectors identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) 50.54(q)(2), and 10 CFR 50.47(b)(4) for the failure of the licensee to classify and declare a Notification of Unusual Event. Specifically, on June 30, 2014, the licensee failed to classify and declare a Notification of Unusual Event after a control room instrument peaked at a wind speed that exceeded the Unusual Event Emergency Classification Level threshold for 4 seconds. The licensee entered the issue into the corrective action program (CAP) as condition report (CR) 01975495. Corrective actions included procedure changes to ensure available indications for wind speed are monitored during high wind events. The failure to classify and declare a Notice of Unusual Event when conditions warranted was a performance deficiency. The finding was more than minor because it adversely affected the emergency response organization (ERO) performance attribute of the Emergency Preparedness (EP) cornerstone objective to ensure that licensees are capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Because the finding only involved a failure to declare a Notification of Unusual Event, the finding screened as being of very low safety significance (Green). This finding was associated with the cross-cutting aspect of avoid complacency in the area of Human Performance, because control room operators did not walk-down instrumentation that was available to them in the control room. (H.12) |
| 05000331/FIN-2015001-03 | 2015Q1 | Duane Arnold | Licensee-Identified Violation | Duane Arnold TS 5.4, Procedures, Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, contains, in part under Section 8.b(2)(t), surveillance test procedures for inspection of the reactor coolant system pressure boundary. Contrary to the above, on November 8, 2014, the licensee failed to properly implement surveillance test procedure (STP) 3.10.102, Non Nuclear Heat Class 1 Ten Year System Leakage Pressure Test, Revision 32. Specifically, during the Fall 2014 refueling outage, licensee personnel identified leakage during visual undervessel inspections per STP 3.10.102. Although several CRs were generated to capture the identified leakage locations and approximate leakage rates from control rod drive mechanism (CRDM) flanges, the personnel failed to fully implement STP 3.10.102, Attachment 3 requirements to perform a detailed inspection of the associated CRDM flanges to identify the leakage source and to verify pressure boundary integrity. Had this identification/verification been performed, STP 3.10.1-02, Attachment 3, further required implementation of GMPTEST66, CRD (**-**) Troubleshooting Procedure, Revision 8, for CRDM flange leakage. Because CRs were written, the licensee personnel considered the under-vessel inspection results satisfactory and moved forward in the STP. Upon further review of the completed STP, the licensee identified that required detailed inspections were not performed for the CRDM flange leaks. The licensee entered the issue into the CAP and successfully re-performed STP 3.10.102 after resolving the leakage issues. Because the inspectors answered No to all questions under Exhibit 4 of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02006364. Corrective actions included a revision to STP 3.10.102 to more clearly define under-vessel visual inspection requirements. |
| 05000331/FIN-2015001-04 | 2015Q1 | Duane Arnold | Licensee-Identified Violation | Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Contrary to the above, on September 27 and 29, 2014, the licensee failed to prescribe an instruction appropriate to the circumstances associated with the replacement of shielded cables between the A and C RHRSW pump motors and the associated 4kV supply breakers. Specifically, SPECE512, Cable and Wire Installation, Revision 14, did not ensure that shielded cables be grounded only at the switchgear end, and that the cables be routed back through ground fault (ring) current transformers in the cabinet before being grounded. This resulted in the improper development of work instructions used in the installation of replacement cables for the A and C RHRSW pumps and a resultant non-conforming condition which was discovered by the licensee during an extent of condition review in March of 2015. Because the SSCs maintained operability based on the deficiency affecting the design of the SSCs, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02023605. Immediate corrective actions included a determination of operability (the ground fault protection had no required safety supporting function for the RHRSW pumps and switchgear), equipment configuration control until resolution was taken, re-routing of the affected cables to restore full design, and a revision to SPEC-E512 to clearly describe shielded cable installation requirements. |
| 05000331/FIN-2015001-02 | 2015Q1 | Duane Arnold | Failure to Report Required Monitoring Results to the NRC | The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 20.2206 for the licensees failure to report results of individual radiation exposure monitoring for individuals required to be monitored by 10 CFR 20.1502. Specifically, on or before April 30, 2014, the licensee failed to report results for all individuals requiring monitoring for the calendar year 2013 to the NRCs Radiation Exposure Information and Reporting System (REIRS) database. The issue was entered into the licensees CAP as CR 02028468. Immediate corrective actions included the resubmittal of radiation exposure data to the REIRS database, which included radiation exposure for all individuals that were required to be monitored. The violation of 10 CFR 20.2206 was assessed in accordance with the traditional enforcement path in IMC 0612, Appendix B, Issue Screening. The inspectors determined that traditional enforcement did apply because reporting failures impact the regulatory process. In accordance with the NRC Enforcement Policy, Section 6.9(d)(2), failures to make a timely written report as required by 10 CFR 20.2206 are categorized as SL IV violations. Cross-cutting aspects are not assigned to traditional enforcement violations. |