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	Quarter	Site	Title	Description
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.
05000315/FIN-2018001-02	2018Q1	Cook	Operation of Letdown System Leads to Voiding and Subsequent Relief Valve Lift	The inspectors identified a finding of very low safety significanceand associated Non-Cited Violation of Technical Specification 5.4, Procedures, when the licensee failed to maintain a procedure for operating the letdown system. As a result, a water-hammer occurred which caused a safety-related relief valve to lift, which discharged reactor coolant to the Pressurizer Relief Tank until letdown was isolated
05000315/FIN-2018001-01	2018Q1	Cook	Failure of Unit 1 Turbine Driven Auxiliary Feedwater Pump to Reach Rated Speed	A self-revealed finding of very low safety significance with an associated Non-Cited Violation of Technical Specification 5.4 Procedures, occurred on December 21, 2017, when the Unit 1 Turbine-Driven Auxiliary Feedwater Pump failed to reach rated speed during a surveillance. Procedure 12MHP5021056008, Turbine-Driven Auxiliary Feedwater Pump Governor Valve Maintenance, was not appropriate for the circumstances in that direction was not given to check that the governor valve could fully open following maintenance on the governor valve.
05000461/FIN-2017004-04	2017Q4	Clinton	Failure to Perform an Evaluation in Accordance with 10 CFR 72.48 for Changes Made to the Time-to-Boil Calculation	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 72.48(d)(1), Changes, Tests, and Experiments, for the licensees failure to perform a written evaluation which provides the bases for the determination that changes do not require a Certificate of Compliance amendment pursuant to 10 CFR 72.48(c)(2). Specifically, the licensee accepted Engineering Change Order ECO501825R0 (1), ECO501848R1 (1), and ECO501848R1 (4) on June 20, 2016, to the time-to-boil calculation as described in the HI-STORM FW Final Safety Analysis Report and incorrectly screened out performing an evaluation of those changes in accordance with 10 CFR 72.48. The licensee documented this issue in its CAP as AR 02714091 and AR 04081583. The licensee is performing a 10 CFR 72.48 evaluation for Engineering Change Order ECO501825R0 (1) and ECO501848R1 (4) while planning to revise the acceptance of ECO501848R1(1). The inspectors determined that the violation was of more than minor significance as the inspectors could not reasonably conclude that the above changes did not require prior NRC approval. The violation screened as a Severity Level IV non-cited violation using example 6.1.d.2 of the NRC Enforcement Policy. No cross-cutting aspect was identified since cross-cutting aspects are not assigned to traditional enforcement violations.
05000461/FIN-2017004-03	2017Q4	Clinton	Failure to Identify the Extent of Condition for an Inadequate 10 CFR 50.59 Evaluation	The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to follow a procedure that implemented the Quality Assurance Program requirements. Specifically, the licensee failed to follow procedure PIAA1251003, Corrective Action Program Evaluation Manual, and identify the extent of condition for a lack of proficiency in applying the licensing basis when performing 10 CFR 50.59 evaluations. The licensee documented this issue in their CAP as AR 04075581. The licensee planned an Updated Safety Analysis Report (USAR) Upgrade Project which reportedly would include a review of safety evaluations for USAR changes that dated back to 1986 and determined the scope of this project would be adequate to identify the extent of condition for this issue. The inspectors determined that this issue was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, because the extent of condition review was not adequate, there is a potential for other safety systems to have been adversely affected by a lack of proficiency in applying the licensing basis during safety related system changes. As a result, safety-related systems may not be able to perform intended safety functions as defined in the USAR. This issue would also adversely affect the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was screened against all cornerstones and determined to be of very low safety significance because the finding met each of the applicable screening questions to be characterized as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence, which stated individuals follow processes, procedures, and work instructions. Specifically, when the NRC violation was documented in the CAP previously it was not appropriately classified in accordance with PIAA120 and was also incorrectly closed to an unrelated evaluation. This contributed to the failure to appropriately perform an extent of condition. (H.8)
05000461/FIN-2017004-02	2017Q4	Clinton	Failure to Assess and Manage Risk Associated with the Performance of Control Rod Time Testing	The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, the licensee failed to assess and manage the increase in risk associated with performing control rod scram time testing in Mode 1, prior to performing the activity. As corrective actions, the licensee assessed the increase in risk for performing control rod scram time testing at power and developed a risk mitigation plan that was used to complete the testing. This performance deficiency was determined to be more than minor because the finding was associated with the human performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to assess risk and develop a risk mitigation plan for control rod time testing at power contributed to an automatic reactor scram. Using IMC 0609, Attachment 4, Initial Characterization of Findings, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, issued May 5, 2005, and Appendix M, Significance Determination Process Using Qualitative Criteria dated April 12, 2002, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance. Specifically, the inspectors and the Region III Senior Reactor Analyst (SRA) determined that Appendix K was not directly applicable to this finding because the licensee performs qualitative evaluations of maintenance activities that have the potential to cause a transient rather than quantitative evaluations. The SRA used insights from Appendix K to support a qualitative SDP evaluation using the principles of Appendix M. The SRA determined that the maintenance activity could only result in an uncomplicated reactor transient event and that the increased risk of a transient compared to the baseline risk of the plant was of very low safety significance. The SRA considered the conditional core damage probability of an uncomplicated transient in this evaluation, which was less than 1E6, to conclude that the finding was Green. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. While performing the work order risk screening for completing the control rod scram time testing while the reactor was shut down, the screener identified that a new screening would be needed if the testing was performed at power. However, no holds were placed on the work order to ensure the risk screening was completed. (H.5)
05000461/FIN-2017004-01	2017Q4	Clinton	Reactor Down Power due to Reactor Recirculation Pump Motor Lower Bearing Oil Leak	A self-revealed finding of very low safety significance was identified for the licensees failure to comply with the requirements of station procedure MAAA716004, Compression Fittings Inspection, Installation, Remake and Repair, Revision 3. Specifically, the licensee failed to properly assemble a joint that was part of the reactor recirculation (RR) pump motor B oil level monitoring system that subsequently leaked requiring the plant operators to perform an unplanned power reduction to allow for identification and repairs of the leak. The licensee documented this issue in the corrective action program (CAP) as Action Request (AR) 04029024. As corrective actions, the licensee made repairs to the effected joint, inspected the remaining joints to ensure proper integrity, and filled the lower bearing reservoir. This issue was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the leak on the reactor recirculation pump motor oil level monitoring system would have eventually resulted in the failure of the reactor recirculation pump causing a transient and upsetting plant stability. This finding was determined to be of very low safety significance because the event did not cause a reactor trip. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management, where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the licensee failed to ensure that all personnel installing the compression fittings received an installers brief prior to performing work on the reactor recirculation pump motor oil level monitoring system. (H.5)
05000237/FIN-2017004-01	2017Q4	Dresden	Failure to Follow Procedure,Results in Non-Functional Fire Door	The inspectors identified a finding of very-low safety significance and associated NCV of Technical Specification 5.4.1.c for the licensees failure to implement the established Fire Protection Program procedures which ensure Fire Barrier Integrity. Specifically, the licensee ran an electrical cable through the doorway of an automatically closing fire door. This was contrary to Procedure DFPP 417501, which requires in part that fire doors must not be blocked open by props or any other material in its closing path. The licensee took immediate actions to restore the fire door, by removing the obstruction and entered the issue into their Corrective Action Program (CAP). The inspectors determined that the performance deficiency was more-than-minor because it affected the Mitigating Systems cornerstone objective since the electrical cable could have prevented the fire door from performing its function. The finding was of very-low safety significance per Task 1.4.3A of IMC 0609, Appendix F. Specifically, the total combustible loading on both sides of the affected fire door was representative of a fire duration less than 1.5 hours. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, associated with the Training component, because the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee believed the performance deficiency was caused by the one of the new temporary contractors brought onto the site to work in support of the D2R25 refueling outage. (H.9)
05000456/FIN-2017003-01	2017Q3	Braidwood	Failure to Implement Adequate Radiological Controls for Treated Liquid Radioactive Effluents Containing Tritium	The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1406(c), when the licensee failed to conduct operations to minimize the introduction of residual radioactivity onto the site. Specifically, the licensee failed to identify and evaluate the environmental risk and control work practices with a credible mechanism to prevent spills and leaks from reaching groundwater at the circulating water blowdown (CWBD) area, a radiologically unrestricted area in the licensees owner controlled area. Specifically, tritium contaminated sump water was intermittently pumped to the environs. The licensee documented this finding in their corrective action program (CAP) as Issue Report (IR) 4020644. The failure to conduct operations and control work practices with a credible mechanism to prevent spills and leaks to reach groundwater and minimize residual radioactivity onto the site represented a licensee performance deficiency. The performance deficiency was of more than minor significance because it was associated with the Program and Process attribute of the Public Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring the adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. In accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because the issue involved a radioactive effluent release, but did not: (1) represent a substantial failure to implement the radioactive effluent release program; or (2) result in public exposure that exceeded the dose values in Appendix I to 10 CFR Part 50 and/or 10 CFR 20.1301(e) limits. The inspectors determined that this finding had a cross-cutting component in the area of Human Performance, in the aspect of Challenging the Unknown, because licensee personnel did not stop when faced with uncertain conditions or evaluate and manage risk before proceeding.
05000249/FIN-2017003-01	2017Q3	Dresden	Granted Notice of Enforcement Discretion 173001: LCO 3.1.7 Required Action B.1 per TS 3.1.7, Standby Liquid Control System	Inspection Scope The inspectors reviewed the licensees response to and assessment of a through- wall leak that developed on the Unit 3 SLC A pump discharge piping . Specifically, on September 12, 2017, during a system operational pressure test, licensee personnel observed a through- wall leak from the forged body of a 1.5 stainless steel pipe T in the Unit 3 SLC system. The affected component is a part of the ASME Code Class 2 boundary. Due to the piping being ASME Code Class 2, it was required to be immediately isolated in accordance with Technical Requirements Manual 3.4.a, Structural Integrity. Isolating this piping resulted in both trains of the Unit 3 SLC system becoming inoperable as the leak was unisolable from both pumps. With both trains inoperable, the licensee entered Limiting Condition for Operation ( LCO ) 3.1.7, Required Action B.1 which requires the restoration of at least one train of SLC within 8 hours. 15 The inspectors examined the sites actions to uncover the issue with the Unit 3 SLC system , their actions to address the issue once it was identified, and their compensatory actions associated with the receipt of the Notice of Enforcement Discretion ( NOED ). The inspectors also reviewed licensee documents to verify that information contained in the NOED request was accurate. Inspection activities included gathering additional information regarding the licensees bases for requesting the NOED; examining the sites decision -making process for the issue; reviewing the licensees condition evaluation; observing the licensees compensatory actions; and evaluating the licensees operability determination. To correct this issue and exit the NOED, the licensee completed replacement of the affected Unit 3 pipi ng and connections, satisfactorily tested the replaced components, and declared the Unit 3 SLC system operable. Documents reviewed are listed in the Attachment. This event follow up review constituted one sample as defined in IP 71153 05. b. Findings Introduction : The inspectors opened an unresolved item associated with a potential noncompliance with TS 3.1.7 Required Action B.1 that occurred on September 12, 2017. NOED 17 3001 was granted by the NRC staff agreeing not to enforce compliance with the TS completion time for an additional 35 hours. Description : On September 10, 2017, with the Unit 3 SLC system in standby operation, an equipment operator performing rounds noted sodium pentaborate crystallization build -up under piping insulation. The licensee removed the insulation from the potential leak location, and noted a dry sodium pentaborate stain on the back of a forged piping T on the 1.5 stainless steel discharge line of the A SLC pump. The licensee Shift Manager made an immediate operability determination of operable based on the dry nature of the stain and its location being on a forged body , and not at a connection or weld location. The licensees initial evaluation surmised the stain was historical in nature and was from an adjacent valve packing leak. In the event that further investigation of the stain indicated a through -wall leak, the licensee investigated American Society of Mechanical Engineers ( ASME ) code compliant permanent and temporary repair options, to include the construction of an Engineered Clamp. This method was eventually dismissed as supports required for the clamp would have been impractical based on system configuration. On September 12, 2017, the licensee cleaned the stain off of the piping T and performed a visual inspection for leakage with the system at full operating pressure. During this test, a leak was observed emanating from the body of the piping T. Due to the leak occur ring within the ASME Code Class 2 boundary, the licensee was required to isolate it in accordance with Technical Requirements Manual 3.4.a, Structural Integrity. Isolating this piping resulted in both trains of the Unit 3 SLC system becoming inoperable, and therefore the licensee entered LCO 3.1.7, Required Action B.1, with an 8 hour required action. With a through wall leak discovered and the plant in a short duration shutdown LCO, the licensee implemented a repair plan for a permanent piping replacement and requested a NOED from the NRC to complete repairs prior to entering Required Action C.1 and C.2, which require placing the Unit in Mode 3 (hot shutdown) and Mode 4 (cold shutdown) within 12 and 36 hours , respectively. The NRC granted a NOED for an additional 35 hours at 5:46 p.m. on September 12, 2017. Consistent with NRC policy, the NRC agreed not to enforce 16 compliance with the specific TSs in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine whether there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item. (Unresolved Item 05000249/2017003 01, Granted Notice of Enforcement Discretion 17 3001: LCO 3.1.7 Required Action B.1 per TS 3.1.7, Standby Liquid Control System )
05000237/FIN-2017002-01	2017Q2	Dresden	Failure to Maintain Configuration Control in the Unit 2 Containment Pressure Suppression System	Green . A finding of very low safety significance and associated non- cited violation of Technical Specification ( TS ) 5.4.1, Procedures, was self -revealed on May 26, 2017, for the licensees failure to maintain configuration control in the Unit 2 containment pressure suppression system. Specifically, the licensee failed to maintain the instrument air stop valve to the actuator for the Unit 2 torus vent , air operated valve (AOV) 21601 60, open with the react or mode switch in Run (Mode 1) and reactor power approximately 100 percent rated thermal power (RTP). The inspectors determined that the licensees failure to maintain configuration control of the Unit 2 containment pressure suppression system was contrary to procedures for the emergency depressurization of containment with the reactor in Mode 1 and was a performance deficiency. The inspectors determined that the performance deficiency was more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the mitigating systems cornerstone attribute of configuration control with regards to the plants operating equipment alignment while affecting the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) . The inspectors determined that a Detailed Risk Evaluation was required to be performed based on answering Yes to the Mitigating Systems screening question A.4 in IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2. The result of the detailed risk evaluation was a finding of very low safety significance (Green) . This finding has a cross -cutting aspect of Resolution in the area of Problem Identification and Resolution because the licensee did not implement appropriate robust barriers to prevent bumping of the 2 1601 60SV in response to previous corrective actions 511878 02 and 2414608 16. Specifically, an identical maintenance induced bumping event resulted in the instrument air stop valve to t he Unit 3 torus main vent AOV 3 1601 60 being unintentionally repositioned closed in November 2014. Licensee corrective actions from that event addressed restraining potentially vulnerable valves prior to maintenance activities as well reassessing which ball valves required permanent robust barrier installation. (P.3)
05000237/FIN-2017001-02	2017Q1	Dresden	Secondary Containment Inoperability Due to Lapse in Procedure Use and Adherence	A self-revealed finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.4.1, Procedures, occurred on November 8, 2016, due to the licensees failure to follow procedures designed to ensure secondary containment integrity, when reactor building (RB) pressure relative to the outside environment was less than 0.25 inches water column (in WC) vacuum as required by TS 3.6.4.1, Secondary Containment. Specifically, work group personnel did not communicate to operations regarding degraded sealing surfaces on the RB Equipment Access outer door as required by procedure DAP 1303, Unit 2 Reactor Building Trackway Interlock Door Access Control, therefore when standby gas treatment (SBGT) started as a part of a planned surveillance test, vacuum lowered, rendering secondary containment inoperable. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Barrier Integrity Cornerstone Attribute of Human Performance and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the drop in secondary containment differential pressure to less than 0.25 in WC vacuum, resulted in a loss of secondary containment and failure of its safety function as specified by TS 3.6.4.1 and Updated Final Safety Analysis Report (UFSAR) section 6.2.3. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. The inspectors reviewed the Barrier Integrity Screening Questions in Appendix A, Exhibit 3 and answered Yes to question C.1. As a result, the finding was determined to have very low safety significance (Green). This finding has a cross cutting aspect in the area of Problem Identification and Resolution, Identification, because individuals failed to identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee did not report a condition adverse to quality with regards to degraded seals on the RB equipment access outer door to operations as required by procedure DAP 1303, therefore not ensuring secondary containment integrity.
05000237/FIN-2017001-01	2017Q1	Dresden	Failure to Correct a Condition Adverse to Quality Associated with EDG Single Largest Load Rejection Surveillance Testing	The NRC identified a finding of very low safety significance and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality, originally identified in Issue Report (IR) 2501498, associated with instructions and acceptance criteria in the emergency diesel generator (EDG) surveillance procedures to ensure that the single largest load rejection test bounded the power demand of the largest load in accordance with Technical Specification Surveillance Requirement (TSSR) 3.8.1.10. Specifically, the failure to correct a condition adverse to quality associated with the inadequate performance of TSSR 3.8.1.10 required an operability determination and engineering assessment to ensure continued operability of the sites three EDGs. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events t prevent undesirable consequences (i.e. core damage). The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. The inspectors answered No to all questions in Exhibit 2, Mitigating Systems Screening Questions, Section A: Mitigating SSCs and Functionality. Therefore, the finding was screened as very low safety significance. The inspectors concluded that the cause of the finding involved a cross-cutting component in the area of Human Performance, Documentation, in that the licensee did not create and maintain complete, accurate and up-to-date documentation. Specifically, the licensee utilized surveillance procedures (DOS 660003, 04 and 05) which did not ensure that design post-accident conditions were met during testing. In addition, the licensee created Corrective Action Program (CAP) actions, to make procedure changes to operations surveillance DOS 660012 to establish bounding conditions for TSSR 3.8.1.10, that were never incorporated.
05000237/FIN-2016004-01	2016Q4	Dresden	Failure to Comply With Radiation Work Permit Requirements Resulting In Unplanned Dose Rate Alarms	A finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Technical Specification 5.4.1 was self-revealed when workers violated a radiation work permit (RWP) by entering an area that was outside of the scope of the original RWP brief without obtaining a required appropriate brief, resulting in these workers receiving unplanned electronic dosimeter dose rate alarms. These workers immediately exited the area and reported the event to the radiation protection staff. The licensee entered these issues as two separate events into their CAP as Issue Reports (IR) 02735594 and IR 02735651. The inspectors determined that the performance deficiency was more than minor in accordance with Inspection Manual Chapter 0612, Appendix B, because the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, worker entry into areas beyond the RWP briefing could lead to unintended dose. The finding was determined to be of very-low safety significance (Green) in accordance with Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The inspectors concluded that the cause of the finding involved a cross-cutting component in the human performance area of challenging the unknown because the individual did not stop when faced with an uncertain condition. Risks were not evaluated and managed before proceeding (H.11).
05000237/FIN-2016004-02	2016Q4	Dresden	Licensee-Identified Violation	Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(4) states, A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, between April 2013, and February 2016, the licensee failed to maintain the effectiveness of the emergency plan by failing to maintain the effluent parameters contained in the standard emergency classification and action level scheme. Specifically, the standard emergency classification and action level scheme associated with the radiological effluents at Dresden Nuclear Power Station was not updated to reflect the changes in the X/Q dispersion factor that were made during the April 2013, Offsite Dose Calculation Manual revision. Consequently, the effluent monitor emergency classification and action level thresholds were non-conservative by a factor of 3.8 until this condition was identified and corrected by Dresden Nuclear Power Station in February 2016. The inspectors determined that the finding was of very low significance (Green) in accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Figure 5.41, because the emergency action level classification of an Unusual Event, RU1, would be declared in a degraded manner, not within the required 15 minutes. The emergency action level classification for the Alert, Site Area Emergency, and General Emergency (RA1, RS1, and RG1) would still be capable of being declared in timely manner, within 15 minutes, using alternate conditions within the emergency action level. Because this finding is of very low safety significance, and has been entered into Exelons CAP under IR 02652711, this violation is being treated as a Green NCV consistent with Section 2.3.2 of the NRCs Enforcement Policy.
05000461/FIN-2016003-01	2016Q3	Clinton	Failure to Scope Fuel Building Ventilation Pressure Control into Maintenance Rule	The inspectors identified a finding of very low safety significance and an NCV of 10 CFR 50.65 (b), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to scope a non-safety related structure, system and component (SSC), whose function is used in one or more Emergency Operating Procedures (EOP) and whose failure could cause actuation of a safety-related system, into maintenance rule. Specifically, the licensee failed to scope the non-safety related fuel building pressure control function into their maintenance rule program. The licensee has entered this issue into their corrective action program as AR 02716300. The licensee is scoping the pressure control function of fuel building ventilation into maintenance rule. The inspectors determined that the licensees failure to scope a non-safety related system whose function is used in one or more EOPs and whose failure caused the actuation of a safety-related system into maintenance rule was a performance deficiency. The performance deficiency was determined to be more than minor, in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it affects the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance because the inspectors answered yes to the question does the finding only represent a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, SBGT system (BWR)?. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of avoiding complacency, where individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes because the licensee identified water intrusion of the fuel building pressure sensing line was a longstanding, latent, known problem and failed to recognize and appropriately challenge how the function was scoped into maintenance rule. (H.12)
05000237/FIN-2016003-02	2016Q3	Dresden	Licensee-Identified Violation	The following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Appendix R, Section III.G.3 requires, in part, that an alternative dedicated shutdown capability and its associated circuits, independent of cables, systems, or components in the area, room, or zone under consideration should be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of paragraph G.2 of this section. Compliance with 10 CFR Part 50, Appendix R, Section III.L is considered necessary to satisfy the requirements of 10 CFR Part 50, Appendix R, Section III.G. Section III.L of 10 CFR Part 50, Appendix R, requires implementation of an alternative dedicated shutdown capability as required by Section III.G.3 of 10 CFR Part 50, Appendix R. Section III.L.3 of 10 CFR Part 50, Appendix R, states, in part, that alternative shutdown capability shall be independent of the specific fire area and that procedures shall be in effect to implement this capability. Contrary to the above, from October 15, 2003, until present, the licensee failed to maintain in effect all provisions of 10 CFR Part 50, Appendix R, Section III.G.3 and Section III.L. Specifically, the licensee failed to ensure that systems that were required for alternative shutdown capability were not free of fire effects, therefore, were not independent of the specific fire area. The licensee credits the HPCI system as the alternative to the isolation condenser for hot shutdown. Licensee procedures DSSP 0100C, Hot Shutdown Procedure Path C Revision 27 and DSSP 0100D, Hot Shutdown Procedure Path D Revision 26, inappropriately direct operators to lift leads and install electrical jumpers in order to defeat HPCI suction transfer from the condensate storage tank (CST) to the torus on low CST level or high torus level. Installation of jumpers and lifting leads is considered a repair and is not permissible for systems required to achieve safe hot shutdown. In accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP), Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection SDP. The inspectors determined that the finding impacted the ability to achieve safe shutdown and, assigned the finding to the category of 1.4.5 Post-fire Safe Shutdown using Table 1 in IMC 0609, Appendix F, Attachment 1, Part 1: Fire Protection SDP Phase 1 Worksheet, dated September 20, 2013. The inspectors answered no to Question 1.4.5B, Does the fire finding affect the ability to reach and maintain a stable plant condition within the first 24 hours of a fire event? in Task 1.4.5 of IMC 0609, Appendix F. The repair actions already in place in procedures DSSP 0100C and DSSP 0100D, while not allowed by Appendix R, were determined to be a viable compensatory measure that would allow the plant to reach and maintain a stable hot shutdown condition. Therefore, the inspectors determined that the finding screened as having very low safety significance (Green). This issue was entered into the licensees CAP as IR 2651479.
05000461/FIN-2016003-03	2016Q3	Clinton	Failure to Perform a 50.59 Screening for Changing the Frequency of Exercising the Turbine Bypass Valves	The inspectors identified a Severity Level IV NCV of 10 CFR 50.59 4(d)(1), Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform a written evaluation which provided the bases for determining that changing the turbine bypass valve surveillance testing frequency from every 31 days, as specified in the Updated Safety Analysis Report, to once a year did not require a license amendment. The licensee has entered this issue into their corrective action program as AR 02720163. The licensee is currently evaluating the issue in accordance with their procedure for changes to the facility. The inspectors determined that the licensees failure to perform a written evaluation to provide the basis for the determination that a change to the facility, a change to a procedure, or a change to a test or experiment did not require a license amendment was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. The licensee made a decision to proceed with implementation of a change to the turbine bypass valve surveillance testing frequency after a plant oversight committee review in lieu of following their consistent, systematic process for evaluating changes to the USAR. (H.13)
05000461/FIN-2016003-04	2016Q3	Clinton	Failure to Prevent Recurrence of a Significant Condition Adverse to Quality	The inspectors documented a self-revealed finding of very low safety significance and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to take corrective actions to preclude repetition of a significant condition adverse to quality. After identifying intergranular stress corrosion cracking (IGSCC) on main steam flex hoses as the root cause of reactor coolant system (RCS) pressure boundary leakage in 2007 and concluding the leakage constituted a significant condition adverse to quality, the licensees corrective action to prevent recurrence failed to prevent pressure boundary leakage at the same location in 2016. The licensee entered this issue into their corrective action program as action request (AR) 02670593. The affected hoses were replaced. The remaining hoses will be replaced in the next maintenance outage. The licensee is also developing a design change to address at least one of the three factors that contributes to IGSCC to ensure pressure boundary leakage does not occur at the main steam flex hoses. The inspectors determined that the licensees failure to take corrective actions to preclude repetition for a significant condition adverse to quality was a performance deficiency. Specifically, the licensees corrective action in 2007 to replace main steam flex hoses at 16 year intervals failed to preclude pressure boundary leakage due to IGSCC on the main steam flex hoses in 2016. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not result in exceeding the RCS leak rate for a small loss of coolant accident (LOCA) and did not affect other systems used to mitigate a LOCA resulting in a total loss of their function. No cross cutting aspect was assigned because the performance deficiency was not indicative of current plant performance.
05000461/FIN-2016003-05	2016Q3	Clinton	Spent Fuel Pool Liner Design Not Verified per Code	The inspectors identified a finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure of the licensees design control measures to provide for the verifying or checking of the adequacy of design of the spent fuel pool liner. Specifically, calculations involving the liner had not been verified or checked to ensure the design basis requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 2, were included. The licensee documented this issue in the corrective action program as AR 02690744 and initiated actions to restore compliance. The inspectors determined that the failure of the calculation to include the design basis requirements was contrary to the design control requirements of 10 CFR Part 50 and was a performance deficiency. The performance deficiency was determined to be more than minor because if left uncorrected it could lead to a more significant safety concern if independent spent fuel storage installation loading was conducted. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with IMC 0609, The Significance Determination Process for Findings AtPower, Appendix A, Exhibit 3 Barrier Integrity Screening Questions (Section D). Based on answering No to all the questions in Exhibit 3, Section D, the inspectors determined the finding to be of very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of design margin, where the organization operates and maintains equipment within design margins. Margins are carefully guarded and changed only through a systematic and rigorous process. Specifically, the licensee failed to ensure the SFP liner reflected the intended design margins of the design and licensing basis. (H.6)
05000461/FIN-2016003-06	2016Q3	Clinton	Licensee-Identified Violation	The following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement policy for being dispositioned as an NCV. Clinton Technical Specification 3.3.2.1, Control Rod Block Instrumentation requires the control rod block instrumentation for each Function in Table 3.3.2.11 shall be operable. If one or more rod withdrawal limiter (RWL) channels is inoperable, the required action is to suspend control rod withdrawal. The completion time for the action is immediately. Contrary to the above, on April 13, 2016, both RWL channels became inoperable and the required action to immediately suspend control rod withdrawal was not completed. Specifically, on April 13, 2016, operations lowered reactor power below the control rod withdrawal limiter high power set point to remove the B turbine driven reactor feed pump from service for repairs. Technical Specification Surveillance Requirement 3.3.2.1.2 requires a functional test of the 4-notch control rod withdrawal limit of the RWL within one hour of resetting the high power set point during power reduction if it has not been completed within the previous 92 days. The surveillance was last performed on April 25, 2015, making the RWL channels inoperable below the high power set point. The licensee subsequently withdrew control rods to restore reactor power above the high power set point with the RWL channels inoperable. The licensee discovered the missed surveillance on April 21, 2016, in preparations for lowering reactor power to place the B turbine driven reactor feed pump into service after repairs. The surveillance was successfully performed after lowering power below the RWL high power set point. The licensee entered the issue into the CAP as AR 02659195. The issue was determined to be more than minor because the performance deficiency affected equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the finding did not affect a single reactor protection system trip signal to initiate a reactor scram, did not involve control manipulations that unintentionally added positive reactivity or result in a mismanagement of reactivity by operators.
05000237/FIN-2016003-01	2016Q3	Dresden	Failure to Assess Scope Changes to Corrective Maintenance Activities Affecting Safety-Related Structures, Systems, and Components	A finding of very low safety significance and associated NCV of TS 5.4.1.a, Procedures, was self-revealed for the licensees failure to maintain maintenance procedures appropriate for the circumstances that could affect performance of safety related equipment. Specifically, procedures MAAA716010, Maintenance Planning, Revision 20 and DAP 1518, Work Order Supplemental Information and Lessons Learned, Revision 17 did not ensure that scope revisions in support of corrective maintenance activities performed on high pressure coolant injection (HPCI) piping in 2013 were properly reviewed and evaluated for technical adequacy directly resulting in a through-wall steam leak on the Unit 2 HPCI inlet drain pot drain piping and safety system inoperability in May 2016. Immediate corrective actions included the replacement of the failed piping section, a determination of the extent of condition of susceptible piping to include the scheduling of a replacement work window, and changes to the maintenance planning procedures requiring engineering scope determination and oversight of scope changes for safety related corrective maintenance. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone Attribute of Procedure Quality 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). Specifically, the failure to ensure work planning procedures controlled the process of major revisions to corrective maintenance activities ensuring adequate engineering reviewing and assessment resulted in continued degradation and ultimate failure of the Unit 2 HPCI inlet drain pot drain piping. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, June 19, 2012. The inspectors answered No to all questions in Exhibit 2, Mitigating Systems Screening Questions. Therefore, the finding was screened as very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee failed to thoroughly evaluate corrective maintenance scope changes to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee incorrectly removed scope without engineering evaluation for adequacy from the Unit 2 HPCI inlet drain pot drain line corrective maintenance following a through wall leak in 2012. Piping that was identified as part of the extent of condition of the failure in 2012, was removed from the scope of corrective maintenance activities due to maintenance personnel short falls. This specific piping failed in May of 2016 resulting in the loss of the HPCI system safety function. (P.2)
05000461/FIN-2016003-02	2016Q3	Clinton	Exceeded Technical Specification Allowed Outage Time for Main Turbine Bypass System	The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 3.7.6, Main Turbine Bypass System, for the licensees failure to meet the limiting conditions for operation and complete the associated required actions after making a deficient change to the turbine bypass valve surveillance testing frequency. Specifically, with the main turbine bypass system inoperable and without the Core Operating Limits Report (COLR) limits for thermal power, minimum critical power ratio (MCPR), and linear heat generation rate (LGHR) with the main turbine by pass system inoperable applied, thermal power was not reduced to less than 21.6 percent of rated thermal power within six hours. The licensee entered this issue into their corrective action program as AR 02690657. The licensee restored compliance by applying the COLR limits for reactor thermal power, MCPR and LGHR. The inspectors determined the failure to meet the limiting conditions for operation and complete the associated required actions prior to the end of the specified completion times was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of change management, where leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority because the licensees change management process was not fully utilized by senior management when evaluating and implementing a change to the turbine bypass valve surveillance testing frequency. (H.3)
05000237/FIN-2016002-01	2016Q2	Dresden	Failure to Implement and Maintain Written Procedures Regarding Breathing Air Quality Testing	A finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1703, was an NRC-identified finding for failure to implement and maintain written procedures regarding breathing air quality that resulted in the failure to perform a continuous in-line breathing air quality test during filling of self-contained breathing apparatus (SCBA) cylinders since 2009. Specifically, on May 4, 2016, during an inspection of the licensees air compressor, the inspectors identified that the in-line carbon monoxide (CO) detector located at the compressor highpressure filling station was inoperable since 2009, the procedure does not specify an alternative method of CO monitoring during the filling of the SCBA cylinders. Without specifying an alternative method of monitoring and only relying on the high-temperature safety shut-off, hazardous CO gas could be introduced into the SCBA cylinders, thus degrading the Grade-D air quality, during a compressor malfunction. The licensees corrective actions included but were not limited to revising the applicable procedures, servicing or replacing the CO monitor by the manufacturer, and installing a new air compressor at the facility. The inspectors determined that that the finding was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, in that the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation through the use of SCBAs during an emergency response use by maintaining certified air quality. Specifically, the licensee failed to implement and maintain written procedures regarding an alternative method of monitoring air quality testing to maintain the Grade-D air quality during filling of SCBA cylinders. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an as-low-as-reasonably-achievable planning issue, there was no overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting component in the area of human performance, resources, in that, the license did not ensure the adequacy of the procedure describing the alternate methods of CO monitoring during filling of Grade D air into the SCBA cylinders. (H.1)
05000237/FIN-2016001-01	2016Q1	Dresden	Failure to Maintain Design Control of the 2/3 Emergency Diesel Generator	A finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was self-revealed associated with the licensees failure to assure that the applicable design basis for applicable structures, systems, and components were correctly translated into specifications, procedures, and instructions. Specifically, since initial plant construction the licensee failed to correctly identify the effect a loss of non-safety 2/3 emergency diesel generator (EDG) room ventilation could have on maintaining operability of the 2/3 EDG. On November 6, 2015, during a planned maintenance outage of the normal non-safety related instrument air pneumatic supply and a failure resulting in the depressurization of the back-up non-safety related nitrogen system, the 2/3 EDG ventilation intake and exhaust dampers failed closed making the 2/3 EDG inoperable for approximately 20 minutes on two occasions from the time of discovery of the condition. The licensee incorrectly believed that a loss of the non-safety related instrument air system and its non-safety related back-up nitrogen system would cause the dampers to fail in the conservative open position. This feature was never tested; and therefore the licensee incorrectly believed the non-safety related control systems for the room ventilation system would not adversely affect the safety-related EDGs operability. The performance deficiency was determined to be more than minor, and thus a finding, in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and if left uncorrected could lead to a more significant safety concern. The finding screened as very low safety significance (Green) because 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, dated June 19, 2012. This finding has a cross-cutting aspect in the area of Human Performance, Training, because the licensee did not ensure licensed operations and engineering personnel properly understood the operation and configuration of the 2/3 diesel generator ventilation system under accident conditions and its impact on the safety-related 2/3 EDGs ability to accomplish its design function. Specifically, the licensee incorrectly believed that the 2/3 EDG room ventilation system failed in a conservative manner with a loss of its non-safety related pneumatic supply systems. Corrective Action Program documents and other engineering products up until September 2015 incorrectly state that the 2/3 EDGs operability was not adversely affected by a loss of damper control pneumatics as the dampers were expected to fail open.
05000374/FIN-2016001-03	2016Q1	LaSalle	Failure to Maintain Appropriate Work Instructions Led to Lost Parts in the Reactor Vessel	A finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed for the licensees failure to verify zero differential pressure across the jet pump plug seals prior to plug removal, an activity affecting quality, in a manner that was appropriate to the circumstances regarding timeliness of the removal. The verification was required by steps 6.13.1 and 6.12 of work orders (WO) 174735903 and 180438305, respectively. The licensee entered this issue into their CAP as action requests 2466339 and 2508333. Corrective actions planned and completed include performed additional analysis and testing of jet pump plug tooling, revised procedures/work instructions, and planned upgrades to the jet pump plug tooling to increase the margins associated with the forces required to displace the seal from the plug. The performance deficiency was determined to be more than minor because if left uncorrected, it had the potential to become a more significant safety concern. Specifically, the robust physical characteristics of the plugs were such that, if unrecovered and unmitigated, coolant flow through certain peripheral fuel assembly orifices could have become blocked by the plugs and potentially led to fuel melt. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Under Exhibit 4, Barrier Integrity Screening Questions, the inspectors answered No to all of the screening questions. Therefore, this issue screened as having very low safety significance (Green). This finding had a cross-cutting aspect in the area of human performance, work management because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priorityas evidenced by the in-field staff verifying zero differential pressure, but then delaying plug removal due to conflicting activities (e.g., shift turnover). As a result, plug removal was later recommenced without re-verifying that conditions had not changed in the intervening period (IMC 0310, H.5).
05000373/FIN-2016001-01	2016Q1	LaSalle	Adequacy of Changes to Pool Swell Analysis	(Opened) Unresolved Item 05000373; 05000374/201600101: Adequacy of Changes to Pool Swell Analysis Introduction: The inspectors identified an unresolved item (URI) related to the licensees changes to the assumptions in their design basis method of analysis associated with the pool swell calculation of record. The inspectors could not resolve the issue of concern during the inspection period due to the need for additional information. Description: While reviewing the recent revision to operability evaluation 12003, the inspectors identified an issue of concern regarding the licensees changes to the assumptions of the design basis calculation of record for the loss of coolant accident suppression pool swell analysis. This operability evaluation assessed the effects of a previous error identified by the licensee in the design calculation and incorporated additional changes in the design assumptions which resulted in the recapture of significant amount of margin in the analysis. Specifically, the licensee changed the initial blowdown characteristics from all air to an air/steam mixture, which improved the margin of the analysis. The inspectors are evaluating the changes against the guidance of IMC 0326, Operability Determinations & Functionality Assessments for Conditions Adverse to Quality or Safety. Additionally, the inspectors are reviewing whether or not regulatory relief was/is required to be sought by the licensee from the NRC for the ASME Code requirement per the guidance in IMC 0326. The inspectors are opening this URI because more information/guidance is needed from the NRC Headquarters Office of Nuclear Reactor Regulation to determine if this issue of concern represents a violation of regulatory requirements. (URI 05000373; 05000374/201600101: Adequacy of Changes to Pool Swell Analysis).
05000373/FIN-2016001-02	2016Q1	LaSalle	Partial Length Rods Exceeded Burnup Limit in Design Basis Method of Analysis	(Opened) Unresolved Item 05000373; 05000374/201600102: Partial Length Rods Exceeded Burnup Limit in Design Basis Method of Analysis Introduction: The inspectors identified an URI related to the licensees use of partial length fuel rods beyond the burnup limit specified in their design basis method of analysis. The inspectors could not resolve the issue of concern during the inspection period due to the need for additional information. Description: In action request (AR) 01647125, Exelon-corporate identified a concern with the potential excessive exposure in partial length rods for LaSalle Unit 1 Cycle 16. This issue also affects the core design of the current Unit 2 cycle. Subsequently, AR 02537519 was written to document the condition as it relates to the Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000. Footnote 11 of the RG, provides an applicability limitation of 62,000 megawatt days per metric ton of uranium for non-loss-of-coolant accident gap release fractions specified in Table 3 of the RG. This RG is the licensees NRC-approved method of analysis for alternate source term as described in the LaSalle UFSAR, and is therefore a part of the licensees design and licensing basis. With this proposed change to the application of RG 1.183, the licensee performed a 50.59 evaluation to review the potential impact of partial length rods operating above 62,000 megawatt days per metric ton of uranium burnup to determine if prior NRC approval was necessary to implement the change. The inspectors have reviewed the licensees 50.59 evaluation, FCP 397411, Revision 1, and calculation, L003067, Revision 2C, and have identified an issue of concern. Specifically, the licensee concluded that prior NRC review and approval were not needed to operate its fuel in a manner that deviated from the limitations delineated within the NRC-approved methodology of RG 1.183 in their current licensing basis. The inspectors are opening this URI because more information/guidance is needed from the NRC Headquarters Office of Nuclear Reactor Regulation to determine if this issue of concern represents a violation of regulatory requirements. (URI 05000373; 05000374/201600102: Partial Length Rods Exceeded Burnup Limit in Design Basis Method of Analysis).
05000237/FIN-2015004-01	2015Q4	Dresden	Failure to Maintain Design Control of Secondary Containment Interlock Doors	A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed on September 4, 2015, when the integrity of the Secondary Containment for Units 2 and 3 was not maintained for 39 minutes when interlock features designed to prevent both doors of a Secondary Containment interlock from being simultaneously open prevented the closure of Reactor Building to Turbine Building doors 47 and 48 following simultaneous operation during routine access of the interlock by plant personnel. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity cornerstone attribute of design control, and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance (Green) because the inspectors answered yes to the Barrier Integrity Screening Question C.1, Exhibit 3 of IMC 0609, Appendix A. This finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because the licensee did not use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee failed to implement a modification which addressed a known design deficiency in the 570 foot elevation Secondary Containment interlock in 2013. The licensee reasoned that the interlock was a low traffic area and that it would be unlikely that the doors would be open simultaneously. (H.14)
05000237/FIN-2015003-01	2015Q3	Dresden	Failure to Evaluate the Available Net Positive Suction Head for the Diesel Generator Cooling Water Pumps Following a Dam Failure	The inspectors identified a finding of very low safety significance, and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the adequacy of the diesel generator cooling water (DGCW) pumps design. Specifically, the licensee failed to evaluate the net positive suction head (NPSH) available to the DGCW pumps at the most limiting ultimate heat sink (UHS) level after a postulated dam failure where they are expected to perform a safety function. The licensee entered this finding into their Corrective Action Program (CAP) and, after a review of their DGCW pump NPSH calculation and an evaluation, concluded that the DGCW pumps had adequate NPSH available to them at the most limiting UHS level after a postulated dam failure, and therefore remained operable. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green) because although it affected the design or qualification of the DGCW pumps, it did not result in the loss of operability or functionality of the pumps. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency.
05000237/FIN-2015003-02	2015Q3	Dresden	Failure to Perform Ultimate Heat Sink Surveys in Accordance With Quality Assurance Program	The inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to perform surveys of the UHS in accordance with the Quality Assurance Program. Specifically, the licensee failed to ensure that: (1) the requirements and acceptance limits in the Updated Final Safety Analysis Report (UFSAR) and other applicable design documents for the high points of the intake and discharge canals were incorporated into the UHS test; (2) the evaluation of the bathymetric survey results accounted for the instrument uncertainty of the test equipment; and (3) the UHS bathymetric survey results were evaluated in accordance with the Quality Assurance Program to assure that the UFSAR required UHS volume was satisfied. The licensee entered this finding into their CAP and, after a review of past bathymetric surveys and other information, determined that: (1) the discharge canal high point had not degraded and was not expected to significantly degrade; and (2) the UHS remained operable because the available UHS water volume still remained above the UFSAR required volume. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of procedure quality, and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green) because although it affected the design or qualification of the UHS, it did not result in the loss of operability or functionality of the UHS. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance, Design Margins, because the licensee did not pay special attention to maintaining the safety-related UHS. Specifically, special attention was not placed on guarding the design margin of the UHS, and a test program that did not meet all the quality assurance requirements was accepted to demonstrate the adequacy of the UHS.
05000249/FIN-2015003-03	2015Q3	Dresden	Failure to Post Protected Pathway Signs	The inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to implement all necessary prescribed risk management actions as required by licensee procedure OP-AA-108-117, Protected Equipment Program, during a Unit 3 low pressure coolant injection (LPCI) logic system functional test (LSFT) maintenance window. Specifically, the licensee failed to post protected equipment signs for the Unit 3 2203-28 instrument rack whose unavailability impacts the high pressure coolant injection (HPCI) system and would have taken the unit into a licensee-defined Orange risk condition and Technical Specification (TS) Limiting Condition for Operations (LCO) 3.0.3. The performance deficiency was determined to be more than minor because the licensee failed to implement prescribed risk management actions which if left uncorrected could have become a more significant safety concern. Specifically, operators inadvertently removed the protected equipment postings on instrument rack 2203-28 which could have significantly degraded the key safety function of reactor coolant inventory control if both the HPCI and LPCI systems became unavailable simultaneously and were required to mitigate the consequences of an accident that could result in potential offsite exposure comparable to the 10 CFR Part 100 guidelines. The finding screened as very low safety significance (Green) because the inspectors answered No to all of the Mitigating Systems Screening Questions. The inspectors determined this finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the licensee failed to implement risk management actions in accordance with procedure OP-AA-108-117. Specifically, the licensee inappropriately removed protected equipment postings prior to commencing maintenance activities which required the risk management activity to be in place.
05000249/FIN-2015003-04	2015Q3	Dresden	Licensee-Identified Violation	A violation of TS 3.3.6.1(A.1), Primary Containment Isolation Instrumentation was identified by the licensee during a review of recent operations logs and CAP documents on September 7, 2015 by on-shift operators. On September 5, 2015, operators performing shiftly TS rounds identified that the Unit 3, A main steam line (MSL) steam flow detector, 3-0261-2A, was reading more than 25 pounds per square inch differential (psid) lower than the other three steam flow detectors on the A MSL contrary to Surveillance Requirement 3.3.6.1.1. This issue was entered into the licensees CAP at this time as IR 2551890, where the on-shift Senior Reactor Operator (SRO) recommended performing DIS 0250-01, Dresden Unit 3 Quarterly Main Steam Line High Flow Switch Calibration and incorrectly assessed TS 3.3.6.1 function 1.d Main Steam Line Flow High as being inoperable but not requiring the failed channel to be placed in trip for the failed steam flow switch. The SRO incorrectly applied the operable logic of two required channels on each MSL per trip system and did not recognize that one failed channel would require entry into TS 3.3.6.1(A.1). TS 3.3.6.1(A.1) requires that operators place the failed channel in trip within 24 hours of discovery. Operations on-shift reviews of recent logs and issue reports identified the missed entry into TS 3.3.6.1(A.1) on September 7, 2015 more than 24 hours later. The missed TS entry was entered into the licensees CAP as IR 2552152. Subsequent performance of DIS 0250-01 identified two additional channels outside of TS required trip values for MSL high steam flow. The failure to enter a required TS action statement when declaring A MSL flow detector 3-0261-2A inoperable was considered a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, a separate single failure could have prevented the safety function of isolating the A MSL on a steam line rupture casualty. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered No to the Appendix A, Exhibit 3 Barrier Integrity Screening Questions, therefore, the finding was determined to be of very low safety significance (Green). Licensee corrective actions included disqualifying the SRO from shift duties pending remedial training, performing an apparent cause evaluation concerning the missed TS action statement entry, replacing the failed A MSL flow switch, and calibrating all other channels which were outside the required operating band of DIS 0250-01.
05000237/FIN-2015002-04	2015Q2	Dresden	Failure to Ensure Continued Operability of Unit 2 ERV 2-02033C (2C) Following Implementation of Extended Power Uprate Plant Conditions	A finding preliminarily determined to be of lowto-moderate safety significance, and an associated Apparent Violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control; TS 3.4.3, Safety and Relief Valves; and TS 3.5.1, ECCS Operating, was self-revealed on February 7, 2015, following the discovery that one of the Unit 2 electromatic relief valves (ERVs) would not have performed its intended safety function. Vibration induced wear experienced while operating at extended power uprate (EPU) power levels resulted in the degradation of multiple ERV actuator subcomponents, which rendered the valve inoperable. This finding does not represent an immediate safety concern in that the licensee has replaced all Unit 2 and 3 ERV actuators with a hardened design successfully utilized at the Quad Cities Nuclear Power Station, which has also experienced significant steam line vibrations post EPU. The inspectors determined that the licensees apparent failure to ensure measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of SSCs, in particular ERV 2-0203-3C (2C), was a performance deficiency warranting a significance evaluation. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone attributes of design control and equipment performance, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance. The inspectors determined that this finding has a cross-cutting aspect of Resolution in the area of Problem Identification and Resolution, since it involves the failure to implement effective corrective actions to address issues in a timely manner commensurate with their safety significance. This cross-cutting issue is conditional depending on the outcome of the preliminary White finding. (P.3)
05000237/FIN-2015002-03	2015Q2	Dresden	Reactor Scram Due to Feedwater Level Control System Failure with a Reactor Recirculation Pump Runback	A finding of very-low safety significance (Green) was self-revealed on January 13, 2015, and again on February 6, 2015, when a loss of power to the Unit 2 feedwater level control (FWLC) system resulted in a reactor scram. The loss in power to the Unit 2 FWLC system was determined to be the result of a human performance error during the original installation of the system under Work Order (WO) 97102835, in that two spade-lug connections associated with the systems +5 Vdc power supply were not properly landed resulting in the intermittent losses in power, and reset of the FWLC system. In addition, a dual in-line package switch on a FWLC Input/Output card was improperly positioned which led to an improper anti-cavitation reactor recirculation pump runback during both events. The inspectors determined that the failure to properly land the leads associated with the Unit 2 FWLC system +5 Vdc power supply in accordance with the work instructions in WO 97102835 was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the configuration control attribute of the Initiating Events cornerstone, and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very-low safety significance (Green), because the inspectors answered "No" to the screening question, Did the finding cause a reactor trip AND the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss off condenser, loss of feedwater)? This finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee did not thoroughly evaluate repetitive alarms and a failure of the FWLC system to ensure that resolutions addressed causes and extent of condition prior to restart following the January 13, 2015, FWLC failure and reactor scram. Specifically, licensee analysis of alarms received prior to the January 13, 2015, scram and troubleshooting of the FLWC system failure on January 13, 2015, was overly focused on multi-functional processor cards which happened to be approaching their end of expected life. Activities to investigate loose wiring connections following the January 13, 2015, scram failed to identify the incorrectly landed spade-lug connections for the +5 Vdc power supply. (P.2)
05000237/FIN-2015002-01	2015Q2	Dresden	Failure to Meet Technical Specification Surveillance Requirements Due to Foreign Material Left in the Unit 2 EDG Starting Circuit	A finding of very-low safety significance (Green) was self-revealed on April 21, 2015 while performing TS Surveillance DOS 6600-12, Diesel Generator Tests: Endurance and Margin/Full Load Rejection/ECCS (Emergency Core Cooling System)/Hot Restart, in support of Surveillance Requirement 3.8.1.16 which requires the EDG to achieve rated frequency and voltage conditions within 13 seconds when started less than or equal to five minutes from a previously loaded run, the Unit 2 Emergency Diesel Generator (EDG) failed to complete a hot restart. Licensee troubleshooting identified a degraded pressure switch associated with main bearing lube oil pressure in the start circuit which was taking several minutes to return to a low-pressure condition upon shutting down the EDG. This resulted in a failure of the start circuit relay to be energized upon initiating a start of the EDG, until the pressure switch returned to its appropriate low-pressure state. An internal investigation of the pressure switch identified strips of Teflon tape in the bellows of the pressure switch, which resulted in the pressure switchs sluggish response to lowering lube oil pressure, and a failure to meet the TS hot restart criteria. The inspectors determined that the failure to implement Procedure MAAA716-008, Foreign Material Exclusion Program, and therefore the inability to perform TS Surveillance Requirement 3.8.1.16 was a performance deficiency, and was considered more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone, and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors utilized Attachment 0609.04, Initial Characterization of Findings, and determined that this issue was of very-low safety significance because each question provided in Inspection Manual Chapter (IMC) 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, was answered No. The inspectors concluded that this finding was cross-cutting in the Human Performance, Documentation area, because licensee procedure MA-AA716-008, Foreign Material Exclusion Program, work instructions associated with Work Order 01410972-01, and previous calibrations of pressure switch 2-6641-526 did not include specific instructions and warnings regarding the proper use of Teflon tape with regards to preventing it from becoming foreign material. Other Dresden maintenance procedures, specifically MA-DR-0300-001, Preventive Maintenance of Hydraulic Control Unit, and DEP 0300-16, Rebuilding the Unit 2 (3) ASCO Scram Solenoid Pilot Valves, have specific warnings regarding the proper use and potential for Teflon tape to become foreign material. (H.7)
05000237/FIN-2015002-02	2015Q2	Dresden	Inadvertent Manipulation of a Test Switch at ESF Bus 23-1 During Surveillance Testing Results in the Inoperability of the 2/3 EDG to Unit 2	A finding of very low safety significance (Green), and an associated NCV of TS 5.4.1, Procedures, was self-revealed on May 19, 2015, when the 2/3 EDG was made inoperable to Unit 2 due to the incorrect manipulation of a test switch by operations personnel during a TS required surveillance test. Specifically, while the licensee performed procedure DIS 1500-05, Division I and II Low-Pressure Coolant Injection ECCS Initiation Circuitry Logic System Functional Test, Step 106 ofChecklist B, operations personnel incorrectly opened test switch TS-159SD2/3 at motor control center 23-1 removing the under-voltage trip associated with the feed breaker for the Division I safety-related 4.16 kV engineered safeguards bus, causing the 2/3 EDG to be inoperable to Unit 2. The licensees failure to properly implement steps in the procedure was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the Mitigating Systems Cornerstone Attribute of Configuration Control, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very- low safety significance (Green), because each of the questions provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, were answered No. The finding has a cross-cutting aspect in the area of Human Performance, Field Presence, for failing to ensure senior managers applied the appropriate oversight of infrequently performed and first time work activities. Specifically, the licensee field supervisor or another senior operations manager was not present for the switching activities, which led to the configuration control error. In this instance, the surveillance test is infrequently performed (every 24 months), and the activity, which included using a maintenance procedure vice an operating procedure, was a first time evolution for both equipment perators involved. (H.2)
05000237/FIN-2015002-05	2015Q2	Dresden	Licensee-Identified Violation	The TS 5.5.1 required that the Offsite Dose Calculation Manual (ODCM), and its Radiological Environmental Monitoring Program (REMP) be established, implemented, and maintained. ODCM Radiological Environmental Control No. 12.6.1 defined the surveillance requirements for the REMP. Step E of this section provided requirements for Milk Station D-25 (Control) be sampled within 10 km to 30 km semimonthly as indicated in Table 12.6-1.4.a. Contrary to the requirements, the licensee did not sample the control Milk Station D-25. The missed samples were not identified by the licensee until March 2015. This issue was entered into the licensees CAP as AR-02469852.
05000374/FIN-2015001-05	2015Q1	LaSalle	Loss of Jet Pump Plug Seals during L2R15	During Unit 2s most recent RFO, L2R15, maintenance was performed on normally unisolable sections of the reactor recirculation system. To facilitate this maintenance, temporary seals were installed on the jet pumps inside the reactor vessel; however, three of the individual jet pump plug seals somehow became detached from the plug assemblies and were lost in the reactor. The first seal was lost on February 8, 2015, from the plug at the jet pump number 5 location as documented in AR 02449326. A potential safety concern was later identified by licensee personnel on February 11, as documented in AR 02450949. This AR stated that the seals could potentially block reactor coolant flow to peripheral fuel bundles, potentially leading to TS Safety Limits being exceeded. Two additional seals were lost on February 18 from the plug at the jet pump number 14 location, as documented in AR 02455054. These three seals were not successfully retrieved from the reactor vessel and the unit was restarted on February 26 with the lost seals still in the reactor. Unit 2 was held at approximately 22 percent power until March 15, 2015, when reactor recirculation pump speed was upshifted to reach approximately 50 percent power, and remained at this power level until March 25, when full power was achieved. These restrictions to power operation were to ensure that safety limits were not violated until testing and evaluation was completed to ensure the seals had degraded to a point that they would safely pass through the most limiting fuel support piece orifice. At the time of this report, the licensee was conducting a root cause evaluation to determine the causal factors that led to the loss of the three plugs. This issue is considered a URI pending the inspectors review of that root cause report to determine if a performance deficiency exists.
05000237/FIN-2015001-01	2015Q1	Dresden	10 CFR 20.1701; Failure to Implement Effective Radiological Engineering Controls	A finding of very-low safety significance, and an associated NCV of 10 CFR 20.1701 was self-revealed during work activities associated with the failure to effectively implement planned radiological engineering controls during reactor head reassembly that resulted in personal contaminations and unintended radiological intakes to workers. On November 14, 2014, during the cleaning of the reactor head studs, several workers on the refuel floor were contaminated, and received unplanned and unintended intakes of radioactive material. Corrective actions included revising applicable procedures to improve the engineering and contamination controls during reactor head reassembly. The inspectors determined that that the finding was more than minor in accordance with IMC 0612, in that the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to implement effective radiological engineering and contamination controls during the cleaning of the contaminated reactor head studs resulted in personal contaminations and intakes to several workers. The inspectors concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an as low as reasonably-achievable planning issue, there was neither overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting component in the human performance in that the licensees management did not ensures that effective radiological engineering controls was either managed or coordinated commensurate to the work activities.
05000249/FIN-2014005-02	2014Q4	Dresden	Failure to Ensure Continued Operability of Unit 3 Electromatic Relief Valve 302033E Following Implementation of Extended Power Uprate Plant Conditions	An apparent violation (AV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having a preliminary low to moderate safety significance, was self-revealed on November 6, 2014, following the discovery that one of the Unit 3 electromatic relief valves (ERVs) would not have performed its intended safety function. Increased vibrations experienced while operating at extended power uprate (EPU) power levels resulted in the degradation of multiple ERV actuator components which rendered the valve inoperable. The inspectors determined that the licensee fully implemented the Unit 3 EPU following a main generator rewind in November 2010, but failed to verify that the ERV actuator design was suitable for operation at the continuously increased vibration levels experienced at EPU power levels. This finding does not represent an immediate safety concern in that the licensee has replaced all four Unit 3 ERV actuators with a hardened design successfully utilized at the Quad Cities Generating Station, which also experienced significant steam line vibrations post EPU. The inspectors determined that the licensees failure to ensure the continued operability of the Unit 3 ERVs following the establishment of EPU plant operating conditions was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone attributes of design control and equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel (SERP), using IMC 0609, Appendix A, Significance Determination Process For Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The inspectors determined that this finding has a cross-cutting aspect of operating experience in the area of Problem Identification and Resolution, since it involves the failure to implement relevant internal and external operating experience in a timely manner.
05000249/FIN-2014005-01	2014Q4	Dresden	Unit 3A LPCI Heat Exchanger Supports Returned to Service with Unacceptable Indications	A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR 50.55a(g)(4) was identified by the inspectors for the licensees failure to maintain American Society of Mechanical Engineers (ASME) Code Class 2 components in accordance with ASME Code Section XI requirements. Specifically, the licensee failed to repair or replace the Unit 3A Low Pressure Coolant Injection (LPCI) heat exchanger support welds identified to have unacceptable linear flaws prior to return to service. The inspectors determined that the licensees acceptance of linear flaws in the Unit 3A LPCI heat exchanger supports that are determined to be unacceptable for continued service IAW with the ASME Code Section XI, Article IWC3000 requirements was a performance deficiency (PD). The inspectors determined that the PD was more-thanminor, and a finding, because if the PD remained uncorrected it could lead to a more significant safety concern. Absent NRC identification, the LPCI support welds with unacceptable linear flaws would have remained in service without repair or replacement. This condition could potentially lead to the failure of the Unit 3A LPCI heat exchanger supports, which in turn, could lead to a potential failure of the Unit 3A LPCI heat exchanger. The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of Findings, Table 3Significance Determination Process (SDP) Appendix Router. The inspectors answered No to the question in Section A of Table 3; and therefore, evaluated the finding using the SDP in accordance with IMC 0609, The Significance Determination Process for At-Power Operations, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors answered No to the questions in Exhibit 2 and determined that this finding did not result in a deficiency affecting the structures, systems, and components (LPCI heat exchanger) to maintain its operability or functionality. Therefore, the finding was determined to have very low safety significance. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, training, for the licensees failure to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee staff dispositioned unacceptable flaws in the LPCI heat exchanger supports for continued service using an engineering evaluation because the licensee staff lacked the specific ASME Code knowledge concerning disposition of the unacceptable indications. Therefore, the licensee failed to return the LPCI heat exchanger supports to within ASME Code acceptable flaw limits via repair or replacement prior to return to service.
05000249/FIN-2014005-03	2014Q4	Dresden	Failure to Maintain Configuration Control in the Unit 3 Containment Pressure Suppression System	A finding of very low safety significance and associated non-cited violation of Technical Specification (TS) 5.4.1, Procedures, was self-revealed on November 19, 2014, for the licensees failure to maintain configuration control in the Unit 3 containment pressure suppression system. Specifically, the licensee failed to maintain the instrument air stop valve to the actuator for Unit 3 torus vent 3160160 open with the reactor in the Start-up and Run Mode following refueling outage D3R23. The inspectors determined that the licensees failure to maintain configuration control of the Unit 3 containment pressure suppression system was contrary to procedures for the emergency depressurization of containment as well maintaining TS required atmospheric conditions inside the primary containment with the reactor in Mode 1 and was a performance deficiency. The inspectors determined that the finding was more than minor because it was associated with the barrier integrity cornerstone attribute of configuration control in how containment design parameters are maintained while affecting the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that the finding was of very low safety significance based on answering No to all of the Barrier Integrity screening questions in IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3. The finding has a cross-cutting aspect of conservative bias in the area of Human Performance because the licensee did not implement appropriate robust barriers to prevent bumping of the 3160160SV in response to corrective action 51187802. Specifically, the licensee previously evaluated 3160160SV and non-conservatively determined that this particular valve did not require a seal to prevent inadvertent operation.
05000237/FIN-2014004-03	2014Q3	Dresden	Licensee-Identified Violation	A violation of 10 CFR 50.65(b)(2)(i) was identified by the licensee during a review of systems and components utilized in the emergency operating procedures (EOP) as compared to functions scoped into the sites Maintenance Rule Program. While reviewing emergency operating procedure DEOP 3001, Secondary Containment Control the Site Maintenance Rule Coordinator (SMRC) noted that one of the entry criterion for the procedure included receiving a Reactor Building Floor Drain Sump (RBFDS) Hi-Hi level alarm. The SMRC noted that the RBFDS was scoped into the Maintenance Rule Program, but the Hi-Hi level alarm function was not. This event was entered into the licensees CAP as IR 1698084. The failure to scope into the Maintenance Rule Program non-safety related structures, systems, and components that are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures was considered a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, no maintenance performance criteria would have been established to ensure the reliability of a function serving as an entry criterion for an EOP associated with maintaining containment integrity. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Tables 2 and 3, and Appendix A, The Significance Determination Process (SDP) for Findings At Power, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered No to the Appendix A, Exhibit 3 barrier integrity screening questions, therefore, the finding was determined to be of very low safety significance (Green). Licensee corrective actions included adding the RBFDS alarm function into the Maintenance Rule Program and performing an extent of condition review of all EOPs for SSC not scoped into the Maintenance Rule Program.
05000237/FIN-2014004-02	2014Q3	Dresden	Inadequate Evacuation Time Estimate Submittals	The NRC identified a NCV of 10 CFR 50.54(q)(2) associated with 10 CFR 50.47(b)(10) and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Dresden Nuclear Power Station Emergency Plan as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date. Exelon submitted the Dresden Nuclear Power Station ETE to the NRC on December 12, 2012, prior to the required due date of December 22, 2012. The NRC completeness review found the ETEs to be incomplete due to Exelon fleet common and site-specific deficiencies, thereby preventing Exelon from providing the ETEs to responsible OROs and from updating site-specific protective action strategies as necessary. The NRC discussed its concerns regarding the completeness of the ETE, in a teleconference with Exelon on June 10, 2013, and on September 5, 2013, Exelon resubmitted the ETEs for its sites. The NRC again found the ETEs to be incomplete. The issue is a performance deficiency because it involves a failure to comply with a regulation that was under Exelons control to identify and prevent. The finding is more than minor because it is associated with the emergency preparedness cornerstone attribute of procedure quality and because it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding is of very low safety significance because it was a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The licensee had entered this issue into their corrective action program (CAP) and re-submitted a new revision of the Dresden Nuclear Power Station ETE to the NRC on May 2, 2014, which was found to be complete by the NRC. The cause of the finding is related to the cross-cutting element of Human Performance, Documentation.
05000237/FIN-2014004-01	2014Q3	Dresden	Failure to Perform an Adequate 10 CFR 50.59 Evaluation for Procedure DOP 130002	The inspectors identified a NCV of 10 CFR 50.59, Changes, Tests and Experiments, when, on February 10, 2011, the licensee failed to complete a 10 CFR 50.59 evaluation when they revised procedure DOP 130002 to change the position of Motor Operated Valve (MOV) 213013, Reactor Inlet Isolation, such that the Isolation Condenser (IC) system would not meet its design requirement of removing 84.2E+06 BTUs in 20 minutes when initiated from its minimum Technical Specification(TS) level and maximum TS temperature. The inspectors determined that the licensees failure to identify that the valve position adjustment required a 10 CFR 50.59 evaluation was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. This finding was more than minor because there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, by establishing a new position setting of MOV 213013, the licensee failed to determine that the proposed change would cause isolation condenser tubes to become exposed in the design basis accident such that it adversely affected a Final Safety Analysis Report described design function, which required an evaluation to be performed. In accordance with IMC 0612, Appendix B, Issue Screening, traditional enforcement does apply as the violation impacted the regulatory process. Using IMC 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 was a design or qualification deficiency that did not represent a loss of the system and/or function, did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, and did not result in the actual loss of one or more trains of non-technical specification equipment. Inspectors assessed the violation in accordance with the Enforcement Policy, and determined it to be a Severity Level IV violation because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). This finding has a cross-cutting aspect of Design Margins (IMC 0310, H.6) in the area of human performance, for failing to carefully guard and maintain the IC design requirement of removing 84.2E+06 BTU in 20 minutes.
05000237/FIN-2014003-01	2014Q2	Dresden	Failure to Take Appropriate Corrective Action When a Maintenance Rule Performance Goal for the Standby Coolant System was Not Met	The inspectors identified a finding of very low safety significance and non-cited violation of 10 CFR 50.65(a)(1), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to take corrective actions by performing an (a)(1) determination when the standby coolant supply system preventative maintenance (a)(2) demonstration was failed. Specifically, in November 2013, the standby coolant supply system exceeded its maintenance rule performance criteria when it experienced an additional maintenance preventable functional failure. The licensee failed to appropriately account for this failure in their Maintenance Rule Program and, as a result, the site failed to perform appropriate corrective action, by failing to perform an (a)(1) determination in accordance with Procedures ERAA310, Implementation of the Maintenance Rule, and ERAA-3101005, Maintenance RuleDispositioning Between (a)(1) and (a)(2), Revision 6. Corrective actions taken by the licensee to address this issue included performing a maintenance rule (a)(1) determination and placing the system into (a)(1) status. The issue was entered into the licensees corrective action program as issue report (IR) 1644740, NRC Questions D2R23 Performance of DOS 390001, and IR 1650033, MRule A1 Determination Needed for Missed MRFF Z391. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstones attribute of Equipment Performance and affected the cornerstones objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to identify a functional failure during a periodic (a)(2) demonstration purposed to provide reasonable assurance that the structures, systems, and components (SSCs), the standby coolant injection valve MO 23902, was capable of performing its intended function as specified in licensee emergency operating procedure DEOP 050003, Alternate Water Injection Systems, Revision 22. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 for the Mitigating Systems cornerstone. The inspectors answered Yes to the question Does the finding represent a loss of system and/or function and determined that a Detailed Risk Evaluation was required. The Senior Reactor Analysts (SRAs) evaluated the finding using the Dresden Standardized Plant Analysis Risk (SPAR) model version 8.18 and Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.0.9.0 software. The exposure time for the unavailability of the Standby Coolant Supply Valve 23902 was assumed to be the maximum value of one year. The result was a delta core damage frequency (CDF) of 6.6E8/yr. The dominant sequence was a medium loss of coolant accident initiating event with a failure of suppression pool cooling, a failure of power conversion system recovery, and a failure of late injection. Based on the Detailed Risk Evaluation, the SRAs determined that the finding was of very low safety significance (Green). This finding had a crosscutting aspect in Human Performance, Procedure Adherence, because the licensee failed to appropriately document the failure of a standby coolant supply valve in accordance with periodic test procedure DOS 390001, Standby Coolant Supply Functional Test. (H.8)

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