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| Start date | Site | Identified by | Title | Description | |
|---|---|---|---|---|---|
| 05000331/FIN-2018003-01 | 30 September 2018 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | A violation of very low safety significance (Green)was identified by the licensee and has been entered into the corrective action program. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. System Design Specification APEDA61019, Pressure Integrity of Piping and Equipment Pressure Parts Data Sheet, required in the applicable castings section T1.3.3.b, all accessible surfaces including machine surfaces shall be examined by either the magnetic particle or liquid penetrant method in either the furnished or finished condition. Contrary to the above, in October 2016, measures were not established to assure that applicable design basis requirements as defined in 10 CFR 50.2 were translated into work instructions repairing the B inboard main steam isolation valve, CV 4415, during RFO 25. Specifically, instructions to perform a NDE of machined surfaces following the valve repair were not included in the work package. As a result, the non-destructive examination was not performed prior to placing the valve into service. |
| 05000331/FIN-2018003-02 | 30 September 2018 23:59:59 | Duane Arnold | Self-revealing | Minor Violation | During Mode 1 power operations on July 9, 2018, the licensee had both doors of a secondary containment airlock open simultaneously, and a minor violation of Technical Specification (TS) 3.6.4.1 Secondary Containment was self-revealed. During the time both doors were open, approximately 3 seconds, the allowable penetration opening area was exceeded and rendered the secondary containment inoperable. Technical Specification 3.6.4.1 requires secondary containment to be operable in Modes 1, 2 and 3. Technical Specification Surveillance Requirement 3.6.4.1.2 supports secondary containment operability by verifying that either the outer door(s) or the inner door(s) in each secondary containment access opening are closed. The posted instructions at each secondary containment airlock door stated, ATTENTION Push Button To Be Held In For 2 Seconds Prior To Opening Door, to be of a type appropriate for traversing the containment airlock. Contrary to the above, at approximately 1:34 p.m. on July 9, 2018, while operating in Mode 1 at 97 percent power, two individuals simultaneously traversing through opposite doors of a secondary containment airlock each failed to hold the airlock interlock push button for two seconds prior to opening their respective doors resulting in a momentarily inoperability of secondary containment. Operability was restored upon the immediate closure of one of the two doors. Subsequently, maintenance was unable to recreate the condition and satisfactorily performed Surveillance Test Procedure (STP) 3.6.4.102, Secondary Containment Airlock Verification, and GMPELEC44,Section A5.1,Airlock Door Interlock Checks.The licensee entered this |
| 05000331/FIN-2018002-02 | 30 June 2018 23:59:59 | Duane Arnold | NRC identified | Minor Violation | Minor Violation: On June 19, 2016, while operating at 82 percent power, two secondary containment access airlock doors were opened simultaneously during surveillance testing as part of STP 3.6.4.102, Secondary Containment Airlock Verification. The inspectors determined this event was caused by inadequate procedural guidance which directed the user to attempt to open one airlock door while the other door was already open. During this test, the interlock failed because the permanent magnets had rotated and were misaligned. This failure could have been identified without challenging airlock interlock integrity if the second airlock door wasnt held open. The failure to have adequate procedural guidance for testing the secondary containment airlock doors was a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, which requires licensees to have procedures appropriate to the circumstance when performing safety-related activities. In response to this issue, the licensee immediately closed the airlock doors. In addition, the licensee submitted a TS change request to address the concurrent opening of two secondary containment airlock doors. The licensees corrective action program is tracking the TS change as CR 02034076, Secondary Containment Airlock Doors #225 and 228 Both Opened. Screening: The issue screened as minor because all of the questions associated with a minor issue found in IMC 0612, Appendix B were answered No due to the licensee reestablishing secondary containment operability immediately after the second airlock door opened. In addition, the inspectors considered the failure to have an appropriate procedure was less than a Severity Level IV violation in accordance with the NRCs Enforcement Policy. Violation: The failure to comply with 10 CFR Part 50, Appendix B, Criterion V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes LER 05000331/2016001. |
| 05000331/FIN-2018011-01 | 30 June 2018 23:59:59 | Duane Arnold | NRC identified | Failure to Translate Environmental Qualification Requirements into Maintenance Procedures/Instructions | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures to assure that Environmental Qualification (EQ) requirements for qualified components correctly translated into procedures and instructions. Specifically, the inspectors identified two examples of the licensees failure to ensure that the EQ requirements for O-ring installed in EC290 connector/plug-in cable assemblies were translated into the associated maintenance procedures and instructions(i.e.,EQ Files, warehouses storage requirements). The licensee failed to correctly establish an end-of-life replacement schedule for the O-ring used in the cable assemblies installed in the dry well and failed to establish a 2-year shelf-life for the O-ring stored in the warehouse. |
| 05000331/FIN-2018002-01 | 30 June 2018 23:59:59 | Duane Arnold | NRC identified | Inappropriate Procedural Guidance Resulted in Loss of Scram Function and Failure to Enter Technical Specification Limiting Condition for Operation | The inspectors identified a finding of very low safety significance (Green) and a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have procedures appropriate to the circumstance for testing the main steam isolation valve (MSIV) and turbine stop valve (TSV) closure functions. Specifically, STP 3.3.1.117, MSIV Functional Test, and STP 3.3.1.119, Main Turbine Stop and Combined Intermediate Valves Test, directed the use of a reactor protection system test box which disabled the MSIV and the TSV closure automatic reactor scram functions while testing specific combinations of MSIVs and TSVs and failed to require entry into appropriate Technical Specification Limiting Condition for Operation action statements. |
| 05000331/FIN-2018411-01 | 31 March 2018 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | |
| 05000331/FIN-2018010-02 | 31 March 2018 23:59:59 | Duane Arnold | NRC identified | Failure to Include Operator Action in the Plant Operating Procedure | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited ViolationofTitle10 of the Code of Federal Regulations, Part 50.48(c), and NFPA805, Section 4.2.4.1.6,Operations Guidance. Specifically, during the transition process to NFPA 805, performance-based standard fortheFire Protection Program, the licensee inadvertently removed a required operator action in the control room from plant operating procedure AOP 913, Fire. |
| 05000331/FIN-2018010-01 | 31 March 2018 23:59:59 | Duane Arnold | NRC identified | Failure to have Adequate Pre-Fire Plans | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations, Part50.48(c), and National Fire Protection Association (NFPA)805, Section 3.4.2,Pre-Fire Plans. Specifically, the inspectors identified two examples for the licensees failure to have current and detailed pre-fire plans. The first example for the failure to provide adequate guidance in the pre-fire plans for smoke and heat removal in the event of a fire in switchgear rooms. The second example was for the failure to show the addition of the Flexible Coping Strategiesbattery packsas a potential hazard in the pre-fire plan for the battery roomcorridor. |
| 05000331/FIN-2018001-01 | 31 March 2018 23:59:59 | Duane Arnold | Self-revealing | Failure to Perform Nondestructive Examination of Main Steam Isolation Valve 4415 Following Machining of Valve Bore | On February 28, 2018, while the licensee was planning contingency work order packages for possible MSIV repairs during the upcoming refuel outage (RFO) 26, a self-revealing failure to perform an NDE of MSIV CV 4415 during RFO 25 was identified. During RFO 25, inboard MSIV CV 4415 failed local leak rate testing. The valve was disassembled; the valve bore was machined, and the valve was reassembled. Subsequently, the valve passed local leak rate testing. However, the licensee failed to perform post-machining NDE of the valve bore as required by plant design. The MSIV was purchased and installed in accordance with USAS B 31.1.0. The General Requirements of USAS B31.1.0, Part 107, Valves, states that (a) Valves complying with the standards and specifications listed in Table 126.1 may be used within the specified pressure-temperature ratings, (b) Valves not complying with Paragraph (a) above shall be of a design, or equal to the design, which the manufacturer recommends for the service as stipulated in Paragraph 102.2.2. Purchase Specification General Electric Spec. No. 21A9230, Revision 2, requires in Part 6.2.4, Castings for pressure-containing parts shall be 100 percent examined by radiography and all accessible surfaces, including machined surfaces, shall be examined by either liquid penetrant or magnetic particle methods following heat treatment. Re-examination of repaired areas shall be by the above techniques following heat treatment. Contrary to the above, the licensee failed to perform either liquid penetrant or magnetic particle testing following machining of the valve bore. The inspectors determined that the failure to perform design required NDE was a performance deficiency. However, the inspectors cannot assess whether the performance deficiency is more than minor until they review the NDE results. Planned Closure Actions: The inspectors will review the results of the NDE after it is performed during RFO 26 and determine the significance of the violation. Licensee Action: The licensee will perform an NDE of the MSIV bore during RFO 26. Corrective Action Reference: CR 02251009; Missed NDE During Rebuild Of MSIV CV4415. On January 10, 2018, the inspectors evaluated the licensees response to a B Channel half SCRAM condition. The half SCRAM was caused by the failure of Average Power Range Monitor (APRM) B. The inspectors determined the operators appropriately implemented pertinent procedures and Technical Specification requirements which allowed the operators to bypass APRM B and reset the half SCRAM. The licensee performed simple troubleshooting and determined the power supply for APRM B had failed. The power supply had recently been replaced. The inspectors determined the licensee had replaced the power supply in accordance with NextEra procedures and that the APRM power supply failure was not within the licensees ability to foresee and prevent. The licensee entered this minor issue into their corrective action program as CR 02243984; E/S 265 No Voltage Output. The inspectors discussed the corrective actions and associated evaluations with licensee personnel. |
| 05000331/FIN-2017004-01 | 31 December 2017 23:59:59 | Duane Arnold | Self-revealing | Failure to Maintain Reactor Water Level within Procedurally Required Level Band Results in Reactor Recirculation Pump Runback | The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, due to operations personnel failing to control reactor vessel water level in accordance with Integrated Plant Operating Procedure 2, Startup, Revision 160. Specifically, during a reactor startup, while at 55 percent reactor power with only one reactor feed pump running, the operating crew failed to maintain reactor water level within the procedurally required level band which resulted in a recirculation pump runback to 45 percent speed and an unplanned reactor power decrease from 55 to 43 percent. The licensee responded to the transient and verified that reactor power stabilized at 43 percent without complications, conducted a human performance review, and entered this issue into their corrective action program (CAP) as condition report (CR) 02233094.The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to control reactor water level within the procedurally specified water level band resulted in an unplanned recirculation pump runback and a decrease in reactor power from 55 to 43 percent. The finding was determined to be of very low safety significance because the finding did not cause a reactor trip. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of teamwork, where individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, a reactor operator dialed down the reactor water level control set point without notifying the control room supervisor, briefing the evolution, or obtaining a peer check. (H.4) |
| 05000331/FIN-2017004-02 | 31 December 2017 23:59:59 | Duane Arnold | NRC identified | Failure to Evaluate Site Fire and Explosion Hazards in Accordance with 10 CFR 72.212(b) (6) | The inspectors identified a Severity Level IV NCV of 10 CFR 72.212(b)(6), Conditions of General License Issued under 72.210, for the failure of thelicensee as of June 9, 2003, to determine whether or not reactor site parameters were enveloped by the cask design bases as considered in the Updated Final Safety Analysis Report (UFSAR). Specifically, the licensee failed to evaluate site-specific fire and explosion hazards that were allowed to be near the dry cask storage systems under its Administrative Control Procedure (ACP) 1412.2, Control of Combustibles, Revision 48. The licensee documented this issue in its CAP as CR 02228514 and CR 02228558 and took timely corrective actions.The inspectors determined that the violation was of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues. Example 4k is applicable to this issue in that the lack of evaluation showing that the quantity of combustible and flammable liquids stored near the dry cask storage system were bounded by the design basis in the UFSAR allowed for a credible unanalyzed fire and explosion scenario that could affect the important-to-safety dry cask storage system. The violation screened as a Severity Level IV NCV. Cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000331/FIN-2017007-01 | 30 June 2017 23:59:59 | Duane Arnold | NRC identified | Failure to Include Valves in the Inservice Testing (IST) Program | Green . The inspectors identified a finding and an associated non- cited violation of Title 10 of the Code of Federal Regulations (10 CFR ) 50.55a(f)(1) for the licensees failure to scope in multiple check valves of the main steam isolation valve leakage treatment system (LTS) into the Inservice Testing (IST) Program. Specifically, these valves were credited to mitigate the consequences of the main steam isolation valve leakage following a loss of coolant accident but they were not scoped into the IST program. Since the licensee made a commitment to the NRC to put these valves into the IST p rogram as part of License Amendment 207, this issue is also a Deviation in accordance with the NRC Enforcement Policy. The licensee put this issue into the CAP 3 as Action Requests ( AR s) 2193481 and 2193482 and planned to include these v alves in the full IST program . This performance deficiency was m ore than minor because i f left uncorrected, there was a potential to lead to a more significant safety concern. Specifically, th ese valves that were credited to mitigate the consequence of an accident were not tested in accordance with the IST program. T he finding screened as very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment , containment isolation system , and heat removal components , nor did it involve an actual reduct ion in function of hydrogen igniters in the reactor containment . The inspectors determined this finding affected the cross -cutting area of problem identification and resolution in the aspect of evaluation because the licensee justified that the valves be put into the augmented IST program since they were non- code components . In addition, the licensee did not re -scope these components into the IST program when 10 CFR 50.55(f)(1) was changed in 1999. This misconception continued when the licensee discovered several valves of the LTS were not in the IST program scope in 2015 . (P.2) |
| 05000331/FIN-2017002-01 | 30 June 2017 23:59:59 | Duane Arnold | NRC identified | Failure to Comply with Technical Specification Program Requirements | The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.5.13, Control Building Envelope (CBE) Habitability Program, for the licensees failure to implement all TS required elements in the CBE habitability assessment. Specifically, the assessments were not performed at the required frequency and did not verify that the unfiltered air leakage limits for hazardous chemicals would ensure that the CBE occupants exposure to these hazards were within the assumptions in the licensing basis. The violation was entered into the licensees Corrective Action Program as Condition Report 02211000, NRC Violation-CRE Habitability Program. Corrective actions included re-performing the CBE habitability assessment to determine that the unfiltered air leakage limits for hazardous chemicals ensured that the CBE occupants exposure to these hazards were within the assumptions in the licensing basis as required by TS 5.5.13.e and performing a review and revision of the CBE Habitability Program implementing procedure, Administrative Control Procedure (ACP) 1208.12, to ensure full compliance with TS 5.5.13. The inspectors determined the failure to perform a complete and comprehensive assessment that addressed all CBE Habitability Program requirements was a performance deficiency and was within the licensees ability to foresee and correct. Specifically, the licensee did not address the impact on CBE occupants from data gathered during the performance of offsite chemical surveys. The finding was determined to be more than minor because the finding was associated with the Barrier Integrity cornerstone attribute of procedure quality and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because no degradation of the barrier function of the CBE against smoke or a toxic atmosphere existed. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of documentation, such that the organization creates and maintains complete, accurate and up-to-date documentation. Specifically, the licensee failed to maintain adequate documentation to ensure that TS program requirements were being met. (H.7) |
| 05000331/FIN-2017001-02 | 31 March 2017 23:59:59 | Duane Arnold | NRC identified | Failure to Enter Technical Specification 3.6.1.3 for an Inoperable Outboard Containment Purge Valve | Green: The inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(2)(i) for the licensees failure to enter TS 3.6.1.3 for an inoperable outboard drywell purge valve. Specifically, as a result of a deficient immediate operability determination (IOD), the licensee failed to declare the outboard drywell containment purge valve inoperable when it failed to satisfy surveillance requirement (SR) 3.6.1.3.4, perform leakage rate testing for each primary containment purge valve with resilient seals. The licensee entered this issue into the Corrective Action Program (CAP) as CR 2183505. Corrective actions included licensed operator training to share lessons learned. This performance deficiency is more than minor because it impacted the Barrier Integrity cornerstone attribute of SSCs and barrier performance, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding was screened as very low safety significance because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components, nor did it involve an actual reduction in function of hydrogen igniters in the reactor containment. This finding has a cross-cutting aspect of consistent process in the Human Performance cross-cutting area because the licensee used Tseal pressure and non-accident drywell conditions which are not related to the primary containment isolation valve (PCIV) specified safety function to declare the outboard drywell purge valve operable. (H.13) |
| 05000331/FIN-2017001-01 | 31 March 2017 23:59:59 | Duane Arnold | Self-revealing | Inadequate Maintenance Instructions for Containment Purge Valves Led to Violation of Technical Specification 3.6.1.3 | Green: A Green finding and an associated NCV of Technical Specification (TS) 3.6.1.3 was self-revealed due to the failure to ensure the outboard containment purge valve was operable when the rector was placed in Mode 2 on October 30, 2016. Specifically, the licensee performed maintenance on the outboard containment purge valve during the refueling outage using procedures which failed to contain acceptance criteria to ensure critical dimensions necessary for reliable containment purge valve sealing could be re-established following valve stroking. The licensee entered the issue into the corrective action program as Condition Report (CR) 2181838. Corrective actions included establishing the critical dimensions necessary to actuate the drywell purge valve seals, revising the procedure for containment isolation valve maintenance, developing a new test procedure for containment isolation valve testing and developing just-in-time training to be used during future containment isolation valve maintenance. The inspectors determined that the failure to have the outboard containment purge valve operable while operating in Modes 1 and 2 as required by TS 3.6.1.3 was a performance deficiency. This performance deficiency is more than minor because it is associated with the System, Structure and Component (SSC) and barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (e.g., containment) protect the public from radionuclide releases caused by accidents or events. This finding was of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment, containment isolation system, or heat removal components, nor did it involve an actual reduction in function of hydrogen igniters in the reactor containment. This finding has a cross-cutting aspect in the Human Performance area of Resources, because NextEra personnel did not ensure that procedures were adequate to support nuclear safety. Specifically, maintenance procedures for the drywell purge valves did not contain steps to ensure that a critical dimension for valve operability was re-established following valve stroking. (H.1) |
| 05000331/FIN-2017001-03 | 31 March 2017 23:59:59 | Duane Arnold | NRC identified | Failure to Incorporate Aspects of Vendor Manual into Self-Contained Breathing Apparatus Maintenance Procedure | Green: The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 20.1703, Use of Individual Respiratory Protection Equipment, for the licensees failure to develop and maintain written procedures regarding the maintenance and testing that incorporated all of the visual and functional tests specified by the manufacturer. This issue has been entered into the licensees CAP as CR 02183134. Corrective actions included an evaluation of that the self-contained breathing apparatus (SCBA) available for use would perform as expected. The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the inadequate testing of the SCBA could have resulted in a functional failure during use. The removal of the SCBA in this occurrence could have led the individual to be exposed to radiological airborne hazards and more importantly atmospheres that are immediately dangerous to life and health. The finding was determined to be of very low safety significance because it was not an as-low-as-reasonably-achievable planning issue, there were no overexposures, nor substantial potential for overexposures, and the licensees ability to assess dose was not compromised. The inspectors determined that the cause of the issue had a cross-cutting aspect of evaluation in the problem identification and resolution area. Specifically, the licensee did not thoroughly evaluate the issue to ensure the resolution addressed the extent of conditions from a previously identified discrepancy between the manufacturers manual and licensee procedures. (P.2) |
| 05000331/FIN-2016004-01 | 31 December 2016 23:59:59 | Duane Arnold | NRC identified | Failure to Control Emergency Preparedness Drill Scenario | Green. The inspectors identified a finding of very low safety significance and an NCV of 10 CFR 50.54(q)(2) associated with the failure to maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR 50, Appendix E. Specifically, the licensee failed to control an emergency preparedness table-top drill scenario, per procedure EPAA1011000, to avoid preconditioning Emergency Response Organization (ERO) drill participants. The licensee entered this issue into their Corrective Action Program as CR 02172325. Corrective actions included removing the Drill/Exercise Performance Indicator credit for the drill conducted on November 30, 2016 and from any preconditioned individuals. The licensee also planned to remove the drill scenario from Emergency Planning Department Manual (EPDM) 1024. The inspectors determined that the licensees failure to control an emergency preparedness table-top drill scenario in accordance with EPAA1011000 to avoid preconditioning ERO drill participants was more than minor because it was associated with the ERO performance attribute of the Emergency Preparedness cornerstone and 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 was determined to be of very low safety significance because the finding was a failure to comply with the requirements in planning standard 10 CFR 50.47(b)(14) but did not involve a loss of planning standard function. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of avoid complacency, where individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reductions tools. Specifically, the licensee failed to implement error reduction tools by reviewing the table-top procedure before conducting the drill and failed to plan for the inherent risk of players seeing the same scenario multiple times when the scenario is repeated (H.12). |
| 05000331/FIN-2016004-02 | 31 December 2016 23:59:59 | Duane Arnold | NRC identified | Inadequate Evaluation for a Change to the Low Pressure Coolant Injection Annunciator Response Procedure | Severity Level IV. The inspectors identified a finding of very low safety significance (Green) and an associated Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50.59, Changes, Tests, and Experiments, when licensee personnel failed to perform an adequate written evaluation to demonstrate that a procedure change did not require a license amendment. Specifically, the licensee implemented a change to annunciator response procedure (ARP) 1C03B, Reactor and Containment Cooling and Isolation, that revised low pressure coolant injection (LPCI) system operability determination information which impacted the safety-related function of the LPCI system. The licensee entered the inspectors concerns into their corrective action program as condition report (CR) 02158897. Corrective actions included providing operating crew orders to preclude challenging the low pressure coolant injection systems function, performing a condition evaluation and revising the ARP. The inspectors determined the failure to provide an adequate evaluation that documented the basis for determining the change to ARP 1C03B did not require a license amendment was a performance deficiency. The performance deficiency was determined to be more than minor because it impacted the Mitigating System cornerstone attribute of procedure quality 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. 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 Significance Determination Process (SDP) as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of design margin, where margins are carefully guarded and changed only through a systematic and rigorous process. Specifically, the licensee made a decision to proceed with revising the annunciator response procedure using the results of a condition evaluation (a non-design basis document) in lieu of following their systematic and rigorous process for evaluating changes to the Updated Final Safety Analysis Report (UFSAR) (H.6). |
| 05000331/FIN-2016003-02 | 30 September 2016 23:59:59 | Duane Arnold | NRC identified | Failure to Identify and Evaluate a Condition Adverse to Quality | A finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, was identified by the inspectors for the licensees failure to follow Quality Assurance Program implementing procedure PIAA1041000, Condition Reporting. Specifically, the licensee failed to properly classify a condition report documenting the inappropriate revision of an alarm response procedure as a condition adverse to quality. This issue was subsequently entered into the licensee CAP as CR 2160423. Corrective actions included revising the alarm response procedure and taking action to evaluate the incorrect classification. The inspectors determined that the failure to follow a Quality Assurance Program implementing procedure was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined the finding to be of very low safety significance because it did not represent an actual loss of function for greater than the TS allowed outage time. The finding was associated with the Problem Identification and Resolution cross-cutting aspect of Evaluation because the licensee failed to thoroughly evaluate issues to ensure that resolutions addressed the causes and extent of conditions commensurate with their safety significance. |
| 05000331/FIN-2016003-01 | 30 September 2016 23:59:59 | Duane Arnold | NRC identified | Failure to Satisfy 10 CFR 50.72 and 10 CFR 50.73 Reporting Requirements for a Condition that Could Have Prevented Fulfillment of a Safety Function | The inspectors identified a Severity Level IV NCV of 10 CFR Part 50.72(a)(1) and 10 CFR Part 50.73(a)(1) due to the licensees failure to make a required 8hour non-emergency notification and a 60 day Licensee Event Report to the NRC after discovering a loss of safety function for the reactor core isolation cooling (RCIC) system. The licensee documented this issue in the CAP as CR 02156273 and planned to perform a causal evaluation for the failure to recognize the reportable condition. The inspectors previously evaluated the RCIC systems loss of safety function under the SDP as a finding of very low safety significance (Green) as documented in Section 1R22.b of NRC Integrated Inspection Report 05000331/201600201 (ML16221A619). Violations of the NRCs reporting requirements 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 the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy and determined the violation associated with the failure to report was a Severity Level IV Violation because the previously evaluated loss of safety function was determined to be a Green finding under the SDP. No cross cutting aspect was assigned to this issue due to the issue being a traditional enforcement violation. |
| 05000331/FIN-2016003-03 | 30 September 2016 23:59:59 | Duane Arnold | Self-revealing | Failure to Implement Controls to the Duane Arnold Energy Center Switchyard Resulting in an Unplanned Technical Specification Limiting Condition for Operation 3.8.1 Entry and an Unplanned Risk Change from Green to Yellow | A self-revealed finding of very low safety significance and a non-cited violation (NCV) of Technical Specification (TS) 5.4, Procedures, was self-revealed due to the licensees failure to implement a written procedure recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, the licensee did not implement Administrative Control Procedure 1408.23, Controls to the DAEC (Duane Arnold Energy Center) Switchyard, which led to the loss of one credited offsite power source and an increase in plant risk on August 22, 2016. This issue was entered into the licensee corrective action program (CAP) as Condition Report (CR) 02151255. The licensees corrective actions included correcting the incorrect relay wiring information which led to the loss of the offsite source and revising ACP 1408.23 to define the systematic process that will be used to review modifications, either planned or emergent, made by ITC to the DAEC Switchyard. The inspectors determined the licensees failure to implement a written procedure recommended in Regulatory Guide 1.33 was a performance deficiency. This issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it affected the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the lockout of the T1 transformer which required entry into TS 3.8.1 due to the loss of a required offsite power source. Using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, the inspectors determined the finding to be of very low safety significance because all of the questions in Exhibit 2, Mitigating Systems Screening Questions, were answered no. The finding was associated with the cross-cutting aspect of Work Management because the licensee failed to identify and manage risk and coordinate within different job groups. (H.5) |
| 05000331/FIN-2016002-01 | 30 June 2016 23:59:59 | Duane Arnold | Self-revealing | Failure to Accomplish a Surveillance Test Procedure in Accordance with Instructions Resulting in Safety System Inoperability | A self-revealing finding of very low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to accomplish surveillance test procedure (STP) 3.3.6.1-28, (Reactor Core Isolation Cooling) RCIC Steam Line Flow HI Channel Functional Test. Specifically, on April 28, 2016, licensee personnel placed a relay block on the incorrect relay finger which when the relay was actuated, in accordance with the procedure, caused the steam supply to the RCIC system to isolate which resulted in an unplanned RCIC inoperability. Corrective actions included ceasing the performance of the STP, restoring the RCIC system to an operable status and performing an apparent cause evaluation. The apparent cause evaluation corrective actions included updated and expanded training on the proper implementation of place keeping and error reduction techniques. Blocking the wrong relay contacts was a performance deficiency. The finding was more than minor because it affected the mitigating systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Although the finding constituted a loss of safety function, the finding was determined to be of very low safety significance (Green) because the three hours of system unavailability was less than the Technical Specification allowed outage time. Corrective actions included ceasing the performance of the STP, restoring the RCIC system to an operable status and performing an apparent cause evaluation. The apparent cause evaluation corrective actions included updated and expanded training on the proper implementation of place keeping and error reduction techniques. The finding was associated with the cross-cutting aspect of avoid complacency in the area of human performance because individuals failed to implement appropriate error reduction tools. (H.12) |
| 05000331/FIN-2016407-01 | 30 June 2016 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | |
| 05000331/FIN-2016007-02 | 31 March 2016 23:59:59 | Duane Arnold | NRC identified | Failure to Document 50.59 Evaluation for UFSAR Change Concerning Radiological Dose Consequence Analysis Methodology | The inspectors identified a Severity Level IV, NCV of 10 CFR 50.59, Changes, Tests, and Experiments, having very-low safety significance (Green) for the licensees failure to document the basis for making a change to Updated Final Safety Analysis Report (UFSAR) Table 15.0-2 to allow the use of RADTRAD Version 3.03 for all Chapter 15 Accidents. Specifically, the licensee failed to demonstrate that the change to UFSAR Table 15.0-2 did not constitute a Departure from a Method of Evaluation described in the UFSAR and would have never required prior NRC review and approval. The inspectors determined that the failure to evaluate whether the change to UFSAR Table 15-0.2 constituted a Departure from a Method of Evaluation was contrary to 10 CFR 50.59(d)(1) and was a PD. The PD was determined to be more than minor, and a finding, because if left uncorrected, the PD had the potential to become a more significant safety concern. Specifically, the inspectors could not reasonably determine that use of RADTRAD version 3.03 for all UFSAR Chapter 15 Accidents would not have increased the control room dose value during accidents. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined that finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process. Using Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined that the finding affected the Barrier Integrity cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3 for the Barrier Integrity cornerstone. The inspectors answered Yes to question C.1 in Exhibit 3 Barrier Integrity Screening Questions. Specifically, the inspectors determined the finding only represented a degradation of the radiological barrier function provided for the control room. In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). In accordance with IMC 0612, Power Reactor Inspection Reports, Section 07.03.c, the inspectors did not assign a cross-cutting aspect to this violation because the violation and underlying technical finding was not indicative of current plant performance. |
| 05000331/FIN-2016007-01 | 31 March 2016 23:59:59 | Duane Arnold | NRC identified | Failure to Document Reviews Performed in 50.59 Screen for New Abnormal Operating Procedure | The inspectors identified a finding of very low safety significance (Green) and associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to document the review performed to conclude that a 50.59 evaluation was not required. Specifically, the licensee failed to document the reviews performed to determine that installation of portable electric heaters in battery rooms would not have an adverse effect on the safety related batteries. The inspectors determined that the licensees failure to document the reviews performed to conclude that a 50.59 evaluation was not required was contrary to procedure EN-AA-203-1201, 10 CFR Applicability and 10 CFR 50.59 Screening Reviews, and was a performance deficiency (PD). The PD was determined to be more than minor, and a finding, because if left uncorrected, the PD would become a more significant safety concern. Specifically, installation of portable electric heaters in battery rooms may increase the probability of hydrogen ignition and challenge the ability of safety related batteries to perform their safety function. 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 finding screened as very-low safety significance (i.e. Green) because it did not result in the loss of operability or functionality of any structure, system, or component. Specifically, the licensee did not enter a condition that required the installation of portable electric heaters in the battery room per Procedure AOP 904. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current licensee performance. |
| 05000331/FIN-2015004-01 | 31 December 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Scope Safety Related and Nonsafety Related Breaker Into the Maintenance Rule | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR 50.65(a)(1), 10 CFR 50.65(a)(2), and 10 CFR 50.65(b), due to the licensees failure to scope the load-shed function of safety related and nonsafety related 4160 volt (V) and 480V breakers into the Maintenance Rule. The load-shed function of the breakers was to ensure upon receipt of load-shed signal that the required breakers would separate from the associated essential buses such that the Standby Diesel Generators (SBDGs) could close into the vital buses. The licensee entered the inspectors concerns into the CAP as CR 2065346. Corrective actions included scoping those breakers of concern into the Maintenance Rule program, establishing breaker performance criteria, and performing a review of past breaker failures against the established criteria. The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems Cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) with respect to the SBDGs. Because the finding did not affect the design or qualification of the SBDGs, nor did it represent a loss of a system or function, the finding screened as very low safety significance. This finding was not indicative of licensee performance since the scoping aspects were determined in 1994, which was prior to the rules effective date of July 10, 1996. Therefore, no cross-cutting aspect was assigned to this finding. |
| 05000331/FIN-2015004-04 | 31 December 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Follow Temporary Configuration Control Procedure | The inspectors identified a finding of very low safety significance for the licensees failure to follow procedure EN-AA-205-1102, Temporary Configuration Changes, Revision 6. Specifically, the licensee constructed a shaft housing enclosure on the B condensate pump without applying the rigor provided by the temporary configuration change (TCC) process. This resulted in water intrusion into the B condensate pump lower motor bearing. This finding does not involve enforcement action because no violation of a regulatory requirement was identified. The licensee entered the inspectors concerns into the corrective action program (CAP) as condition report (CR) 2100521. Corrective actions included the performance of an apparent cause evaluation and the creation of a form to document engineering positions with respect to TCC applicability. The inspectors determined that the failure of the licensee to follow procedure EN-AA-205-1102 to document the addition of the B condensate pump shaft housing shield and forced air blower as TCCs was a performance deficiency. The finding was determined to be more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the addition of the shaft housing shield resulted in a very high humidity environment which resulted in water passing through the lower motor shaft seal and entering the lower motor bearing oil reservoir. This resulted in the need for repetitive feeding and bleeding of the lower motor bearing oil reservoir to prevent emulsification of the oil. The feeding and bleeding of the B condensate pump lower motor bearing oil reservoir was an evolution that could have resulted in bearing damage, pump trip, and reactor scram. The finding was determined to be of very low safety significance because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident, cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of or caused an initiating event, and did not affect mitigation equipment. This finding was associated with the cross-cutting aspect of operating experience in the area of problem identification and resolution because the licensee failed to implement relevant internal operating experience in a timely manner. |
| 05000331/FIN-2015004-02 | 31 December 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Declare High Pressure Coolant Injection and Reactor Core Isolation Cooling Inoperable when the High Pressure Coolant Injection and Reactor Core Isolation Cooling Pump Suction Swap Logic was Inoperable | The inspectors identified a finding of very low safety significance, with two examples, and an associated NCV of Technical Specifications (TS) Sections 3.3.5.1, Condition D and 3.3.5.2, Condition D, for failure to initiate required TS action statements 3.3.5.1.D.1 and 3.3.5.2.D.1. Specifically, the licensee failed to declare the high pressure coolant injection (HPCI) and the reactor core isolation cooling (RCIC) systems inoperable when the automatic HPCI/RCIC pump suction swap function on low condensate storage tank (CST) level was revealed to be inoperable during surveillance testing. The licensee entered the inspectors concerns into the CAP as CR 2080489 and replaced the failed time delay relay. The inspectors determined the failure to declare the HPCI/RCIC systems inoperable when the pump suction swap function on low CST level failed during surveillance testing was a performance deficiency because it resulted in the licensees failure to implement TS required actions and the cause was reasonably within the licensees ability to foresee and should have been prevented. The performance deficiency was determined to be more than minor and a finding because if left uncorrected, failing to implement TS required actions reduced the margin of safety and had the potential to lead to significant safety concerns. The finding was determined to be of very low significance because the CST was assumed to contain sufficient inventory for HPCI and RCIC to perform their function for most scenarios. This finding was associated with the cross-cutting aspect of conservative bias in the area of human performance because the licensee failed to use decision-making practices that emphasize prudent choices over those that are simply allowable when the licensee failed to conservatively evaluate unexpected surveillance test results. (H.14) |
| 05000331/FIN-2015004-03 | 31 December 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Satisfy 10 CFR Part 50.73 Reporting Requirements for a Condition Prohibited by Technical Specifications and for a Condition that Could Have Prevented Fulfillment of a Safety Function | The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (CFR), Section 50.73, Licensee Event Report System. Specifically, the licensee failed to submit a required Licensee Event Report within 60 days after the discovery of an event that was reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition that was prohibited by the plants TS and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented fulfillment of a safety function. The licensee documented the inspectors concern into the CAP as CR 2099065. Planned corrective actions included the performance of an apparent cause evaluation for the failure to recognize the reportable condition and to submit a licensee event report. This issue was determined to be more than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the 10 CFR 50.73 in order to perform its regulatory function. The inspectors previously determined in Section 1R15 of this report that the underlying issue (i.e., the failure of the HPCI/RCIC suction swap function as discovered during surveillance requirement testing) was a finding of very low safety significance. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation. No cross-cutting aspect was assigned to this traditional enforcement violation. |
| 05000331/FIN-2015002-02 | 30 June 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Maintain Breathing Air Quality Requirements for SCBAs | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR 20.1703, Use of Individual Respiratory Protection Equipment, for the licensees failure to supply breathing air in accordance with manufacturer requirements, which voided the National Institute for Occupational Safety and Health (NIOSH) approval of their self-contained breathing apparatuses (SCBAs). This issue was entered into the licensees CAP as CR 2056826. Corrective actions included obtaining an air sample to be sent for analysis and ensuring that the required dew point would be maintained in compliance with the manufactures specifications. The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the introduction of excessive moisture to the SCBA could have resulted in the sudden loss of all breathing air to the individual. The removal of the SCBA in this occurrence could have led the individual to be exposed to both radiological and non-radiological airborne hazards. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, but did not identify any similar examples. The finding was assessed using IMC 0609, Appendix C, Occupation Radiation Safety SDP, dated August 19, 2008, and determined to be of very-low safety significance (Green) because it was not an as-low-as-reasonably achievable planning issue, there were no overexposures, nor substantial potential for overexposures, and the licensees ability to assess dose was not compromised. The inspectors determined that the cause of the issue had a cross-cutting aspect of resources in the human performance area. Specifically, the licensee did not ensure that procedures for testing breathing air quality were in compliance with manufacturer requirements, and therefore, the NIOSH approval for the SCBAs was void. (H.1.) |
| 05000331/FIN-2015007-03 | 30 June 2015 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | As required, in part, by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Procedure PIAA205, Condition Evaluation and Corrective Action, Revision 23, Section 4.10, Step 8, requires, in part, that corrective actions shall not be closed to other existing actions unless the description, scope, condition designation, and intent of the action that will remain open is equivalent to that of the action being closed. Contrary to the above, in December 2013, while performing RCE associated with CR 01884408, the licensee inappropriately credited a long term corrective action developed in an apparent cause evaluation associated with CR 01835557 to accomplish the stated CAPR action. Further, procedure PIAA1041000, Corrective Action, Revision 1, Section 4.11, Step 6, requires, in part, that requests for significance level 1 CAPR due date extensions be documented on PIAA1041000F01, Change Request Form, Revision 1, receive MRC approval for the stated changes, and be attached to the condition report. Contrary to that, on December 5, 2014, the long term corrective action due date was extended 4 months without Management Review Committee approval. The CAPR and long term corrective action were related to the failure of a safety related main steam line temperature instrument. Because the instrument was maintained in an operable status and did not initiate a transient based upon the deficient condition, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02025444 and 02044053. Immediate corrective actions included adding an assignment to correctly restate that the plant modification to replace the temperature indicators as a CAPR in CR 01884408. |
| 05000331/FIN-2015007-01 | 30 June 2015 23:59:59 | Duane Arnold | NRC identified | Inappropriate Diesel Generator Maintenance Procedure | The inspectors identified a finding of very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances. Specifically, the licensee implemented GENERAF01001, 1E053A2 (B2) Flange Inspection, Section W, Revision 5, Step 5.1.3.3.b as a corrective action to NCV 05000331/201400902, in order to ensure proper alignment of the 1E053A2 (B2) flange. The procedure was inappropriate for the circumstances because the instructions, as written, in Step 5.1.3.3.b would not result in meeting the acceptance criteria for flange alignment listed in GENERAF01001, 1E053A2 (B2) Flange Inspection, Section W, Revision 5, Attachment 8. The licensee entered this issue into the CAP as condition report (CR) 02041369. The inspectors determined the licensees failure to provide procedures of a type appropriate to the circumstances to assure that for a significant condition adverse to quality, the cause of the condition was determined and corrective actions were taken to preclude repetition was a performance deficiency warranting further review. The inspectors determined that this finding was more than minor because it affected the Mitigating Systems Cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Misalignment of the flanges could lead to excessive oil leak that rendered the diesel generator inoperable. The inspectors determined the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating system, structure or component and did not result in a loss of operability or functionality. In addition, the finding did not represent a loss of system or function, did not represent an actual loss of function of a least a single train for longer than its technical specification allowed outage time, and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance. The inspectors determined this finding had a cross cutting aspect in the area of PI&R, specifically resolution, because licensee personnel failed to take effective corrective actions to ensure that the resolutions address causes and extent of conditions commensurate with their safety significance (P.3). |
| 05000331/FIN-2015007-04 | 30 June 2015 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | As required, in part, by 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Procedure PIAA1041000, Corrective Action, Revision 3, Section 4.10, Step 1, requires, in part, that closure of corrective actions is not permitted until the corrective actions are completed as prescribed or appropriate justification and approval for intent change or cancellation/nonperformance of the corrective action is documented in the condition report. Contrary to the above, on November 24, 2014, a long term corrective action to replace the northwest corner room cooler coils due to copper tube degradation was closed without replacing the cooler coils or obtaining an approval for cancellation of the correction action. Because the cooler was maintained in an operable status and did not initiate a transient based upon the deficient condition, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02016918. Immediate corrective actions included reopening the corrective action assignment and obtaining approval for extending the due date. |
| 05000331/FIN-2015002-01 | 30 June 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Accomplish a Procedure for an Activity Affecting Quality in a Manner Appropriate to the Circumstances | A finding of very low safety significance and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to identify a degraded condition warranting compensatory measures/actions for assuring operability, as required by Section 4.3.12 of procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, on April 29 and April 30, 2015, the licensee failed to accomplish procedure EN-AA-203-1001 in a manner appropriate to the circumstances to assess equipment operability following the identification of an unexpected condition associated with the flow indication controller (FIC) on the B Standby Gas Treatment (SBGT) subsystem. Following completion of the surveillance testing, the B SBGT system was initially declared operable and fully qualified. In response to questions from the inspectors, the licensee subsequently declared the B SBGT system operable but degraded. The licensee entered the issues associated with the FIC into the Corrective Action Program (CAP) as condition report (CR) 02044191 and CR 02044702, and entered the inspectors concerns into the CAP as CR 02052508. The performance deficiency was determined to be more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected, failing to properly assess operability as required by procedure could potentially result in incorrect/untimely operability conclusions, and the failure to take action to correct degraded or deficient conditions as required by the Technical Specifications. Specifically, the performance deficiency resulted in not properly assessing operability of the B SBGT subsystem that resulted in the failure to implement appropriate compensatory measures/actions to assure operability until final corrective actions were taken. The performance deficiency is associated with the Barrier Integrity attribute of systems, structures, and components and Barrier Performance associated with standby gas trains, which adversely affects the cornerstone objective. (To provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.) 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) SDP Appendix Router, and transitioned to IMC 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012. Per Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that because the finding only represented a degradation of the radiological barrier function provided by the SBGT system, the finding screened as very-low safety significance (Green). This finding was associated with the cross-cutting aspect of conservative bias in the area of Human Performance, because the licensee did not use decision-making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensees decision making practices in implementing EN-AA-203-1001 were non-conservative, and failed to identify the FIC-5828B process deviation as a degraded condition warranting compensatory measures/actions which resulted in incorrectly declaring the B SBGT subsystem operable and fully qualified. (H.14) |
| 05000331/FIN-2015007-02 | 30 June 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Correctly Update the Updated Final Safety Analysis Report | The inspectors identified a Severity Level IV NCV of 10 CFR 50.71(e) for failure to assure that the information included in the last update of the updated final safety analysis (UFSAR) report contained the latest information developed. The licensee implemented a change to the UFSAR, in preparation for License Amendment 243 that did not contain the latest information developed. Specifically, Section 5.4.6.1 (page 5.430 of Revision 17) was updated with a note that stated the reactor core isolation cooling system was not safety-related. In fact, the reactor core isolation cooling system had always been designated as safety-related. The licensee entered this issue into the CAP as CR 01974995 and prepared an updated final safety analysis report (UFSAR) change that removed the statement that the reactor core isolation cooling system was not safety-related. The inspectors determined that the update to the UFSAR with incorrect information was a performance deficiency in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on September 7, 2012. The inspectors concluded that traditional enforcement applied because the failure to correctly update the UFSAR impacted the regulatory process. The Enforcement Policy, dated February 4, 2015, Section 6.1.d.3, gave the example that if, a licensee fails to UFSAR as required by 10 CFR 50.71(e) but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures; then this was a Severity Level IV violation. In this case, the UFSAR was updated incorrectly and did not, result in any unacceptable change to the facility or procedures. The inspectors determined this to be a similar example and therefore was more than minor and a Severity Level IV violation. This violation was not associated with a finding that was evaluated by the significance determination process. Therefore, a cross-cutting aspect was not assigned to this traditional enforcement violation. |
| 05000331/FIN-2015403-02 | 31 March 2015 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | |
| 05000331/FIN-2015001-03 | 31 March 2015 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | Duane Arnold TS 5.4, Procedures, Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, contains, in part under Section 8.b(2)(t), surveillance test procedures for inspection of the reactor coolant system pressure boundary. Contrary to the above, on November 8, 2014, the licensee failed to properly implement surveillance test procedure (STP) 3.10.102, Non Nuclear Heat Class 1 Ten Year System Leakage Pressure Test, Revision 32. Specifically, during the Fall 2014 refueling outage, licensee personnel identified leakage during visual undervessel inspections per STP 3.10.102. Although several CRs were generated to capture the identified leakage locations and approximate leakage rates from control rod drive mechanism (CRDM) flanges, the personnel failed to fully implement STP 3.10.102, Attachment 3 requirements to perform a detailed inspection of the associated CRDM flanges to identify the leakage source and to verify pressure boundary integrity. Had this identification/verification been performed, STP 3.10.1-02, Attachment 3, further required implementation of GMPTEST66, CRD (**-**) Troubleshooting Procedure, Revision 8, for CRDM flange leakage. Because CRs were written, the licensee personnel considered the under-vessel inspection results satisfactory and moved forward in the STP. Upon further review of the completed STP, the licensee identified that required detailed inspections were not performed for the CRDM flange leaks. The licensee entered the issue into the CAP and successfully re-performed STP 3.10.102 after resolving the leakage issues. Because the inspectors answered No to all questions under Exhibit 4 of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02006364. Corrective actions included a revision to STP 3.10.102 to more clearly define under-vessel visual inspection requirements. |
| 05000331/FIN-2015001-01 | 31 March 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Classify and Declare a Notification of Unusual Event | The inspectors identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) 50.54(q)(2), and 10 CFR 50.47(b)(4) for the failure of the licensee to classify and declare a Notification of Unusual Event. Specifically, on June 30, 2014, the licensee failed to classify and declare a Notification of Unusual Event after a control room instrument peaked at a wind speed that exceeded the Unusual Event Emergency Classification Level threshold for 4 seconds. The licensee entered the issue into the corrective action program (CAP) as condition report (CR) 01975495. Corrective actions included procedure changes to ensure available indications for wind speed are monitored during high wind events. The failure to classify and declare a Notice of Unusual Event when conditions warranted was a performance deficiency. The finding was more than minor because it adversely affected the emergency response organization (ERO) performance attribute of the Emergency Preparedness (EP) cornerstone objective to ensure that licensees are capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Because the finding only involved a failure to declare a Notification of Unusual Event, the finding screened as being of very low safety significance (Green). This finding was associated with the cross-cutting aspect of avoid complacency in the area of Human Performance, because control room operators did not walk-down instrumentation that was available to them in the control room. (H.12) |
| 05000331/FIN-2014011-02 | 31 March 2015 23:59:59 | Duane Arnold | NRC identified | Inadequate Coating Strainer Debris Loading Calculation | The team identified a finding having very-low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for all unqualified coating debris available for transport to the ECCS strainers. Specifically, the licensee relied on assumptions that were inconsistent with their licensing basis, resulting in a non-conservative unqualified coating debris loading value which was used in the strainer hydraulic calculations. This finding was entered into the licensees CAP to evaluate and resolve, including correction of th affected calculations. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very-low safety significance (Green) because it did not result in the loss of operability or functionality. Specifically, the licensee removed conservatisms from the calculations, added the additional unqualified coating debris loading, and reasonably determined the system remained operable. The unqualified torus coating debris associated with the apparent violation did not affect this finding. The team did not identify a cross-cutting aspect associated with this finding because it was not indicative of current licensee performance due to the age of the performance deficiency. |
| 05000331/FIN-2014011-01 | 31 March 2015 23:59:59 | Duane Arnold | Self-revealing | Failure to Install Torus Coating in Accordance with Established Processes | An apparent violation of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion IX, Control of Special Processes, was self-revealed for the failure to install the torus coating under suitable controlled conditions. Specifically, inadequate quality controls during the application of the torus coating resulted in unqualified torus coating in excess of the debris loading design margin of the emergency core cooling system (ECCS) suction strainers. This finding was entered into the licensees Corrective Action Program (CAP) to evaluate and resolve, including removal of the unqualified torus coating in excess of design margin. The performance deficiency was determined to be more than minor because it wa associated with the Mitigating System Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has been preliminarily determined to be White, a finding of low-to-moderate safety significance, because it affected the function of the low-pressure ECCSs. The team determined that this finding had a cross-cutting aspect in the area of Human Performance because licensee senior managers did not ensure supervisory and management oversight of work activities, including contractors and supplemental personnel. (H.2) |
| 05000331/FIN-2015001-04 | 31 March 2015 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Contrary to the above, on September 27 and 29, 2014, the licensee failed to prescribe an instruction appropriate to the circumstances associated with the replacement of shielded cables between the A and C RHRSW pump motors and the associated 4kV supply breakers. Specifically, SPECE512, Cable and Wire Installation, Revision 14, did not ensure that shielded cables be grounded only at the switchgear end, and that the cables be routed back through ground fault (ring) current transformers in the cabinet before being grounded. This resulted in the improper development of work instructions used in the installation of replacement cables for the A and C RHRSW pumps and a resultant non-conforming condition which was discovered by the licensee during an extent of condition review in March of 2015. Because the SSCs maintained operability based on the deficiency affecting the design of the SSCs, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02023605. Immediate corrective actions included a determination of operability (the ground fault protection had no required safety supporting function for the RHRSW pumps and switchgear), equipment configuration control until resolution was taken, re-routing of the affected cables to restore full design, and a revision to SPEC-E512 to clearly describe shielded cable installation requirements. |
| 05000331/FIN-2015001-02 | 31 March 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Report Required Monitoring Results to the NRC | The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 20.2206 for the licensees failure to report results of individual radiation exposure monitoring for individuals required to be monitored by 10 CFR 20.1502. Specifically, on or before April 30, 2014, the licensee failed to report results for all individuals requiring monitoring for the calendar year 2013 to the NRCs Radiation Exposure Information and Reporting System (REIRS) database. The issue was entered into the licensees CAP as CR 02028468. Immediate corrective actions included the resubmittal of radiation exposure data to the REIRS database, which included radiation exposure for all individuals that were required to be monitored. The violation of 10 CFR 20.2206 was assessed in accordance with the traditional enforcement path in IMC 0612, Appendix B, Issue Screening. The inspectors determined that traditional enforcement did apply because reporting failures impact the regulatory process. In accordance with the NRC Enforcement Policy, Section 6.9(d)(2), failures to make a timely written report as required by 10 CFR 20.2206 are categorized as SL IV violations. Cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000331/FIN-2015008-01 | 31 March 2015 23:59:59 | Duane Arnold | NRC identified | Failure to Identify and Evaluate the Effects of Vessel OverfillScenario | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR) 50.48(c), and National Fire Protection Association Standard 805, Section 2.4.3.2 for the licensees failure to address in the Fire Probabilistic Risk Assessment (PRA) the risk contribution with all potentially risk-significant fire scenarios. Specifically, the licensee did not address potential damage to safety relief valves (SRVs), or the SRV tailpipes as a result from fire induced overfill of the reactor pressure vessel. The licensee entered this issue into their Corrective Action Program to review the multiple spurious operations Expert Panel report, and properly disposition the scenario. The inspectors determined that the performance deficiency was more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Protection against External Factors (i.e., fire), and it affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the missed failure mechanism for the SRVs had the potential to impact the ability to achieve safe and stable conditions. In accordance IMC 0609, Appendix F, Fire Protection SDP, Attachment 1, Step 1.6.1, Screen by Licensee PRA-Based Safety Evaluation, the inspectors were able to use the Licensees PRA to evaluate the safety significance of the finding. The increase in core damage frequency (CDF) as a result of the identified scenario was found to be approximately 2.6E-7 per year; therefore, the inspectors concluded that this finding was of very-low safety significance (Green). This finding did not have a cross-cutting aspect because it was not representative of current licensee performance. |
| 05000331/FIN-2015404-01 | 31 March 2015 23:59:59 | Duane Arnold | NRC identified | Security | |
| 05000331/FIN-2015403-01 | 31 March 2015 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | |
| 05000331/FIN-2014005-07 | 31 December 2014 23:59:59 | Duane Arnold | Self-revealing | Ineffective Radiological Engineering Controls Resulted in Unplanned and Unintended Radiological Intakes to Workers | A finding of very-low-safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Section 20.1701 was self-revealed during work activities associated with the failure to implement effective radiological engineering controls during reactor pressure vessel (RPV) disassembly that resulted in personal contaminations and unplanned and unintended radiological intakes to workers. Specifically, on October 5, 2014, several individuals working on the refuel floor were contaminated and several received small intakes of radioactive material while venting the RPV head. The licensee entered the issue into the Corrective Action Program as condition report 01996216. Corrective actions included revising applicable procedures for RPV flood-up with the RPV vented to atmosphere on the refuel floor. The finding was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected th cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to implement effective radiological engineering controls during RPV disassembly resulted in personal contaminations and low dose intakes to several workers. The inspectors also 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-lowsafety significance because it was not an ALARA planning issue; there was neither overexposure nor a substantial potential for an overexposure; and the licensees ability to assess dose was not compromised. This finding was associated with the crosscutting aspect of operating experience in the area of Problem Identification and Resolution because the licensee did not systematically implement relevant external operating experience in a timely manner. (P.5) |
| 05000331/FIN-2014005-10 | 31 December 2014 23:59:59 | Duane Arnold | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established for the selection and review for suitability of application of materials and parts that are essential to the safety-related functions of structures, systems, and components. Contrary to the above, in October 2012, the licensee failed to properly select an review the suitability of application of several electrical cable splices and terminal strips during the replacement of safety-related electrical cables associated with the A and B SBDGs. Specifically, two modification packages associated the cable replacements during the 2012 RFO 23 did not appropriately evaluate the impacts of the stations environmental qualification (EQ) program and the effects of an internal turbine building flood. Because the SSCs maintained operability based on the deficiency affecting the qualification of the SSCs, the finding screened as very low safety significance (Green). This issue was documented in the licensees CAP as CR 01979556. Immediate corrective actions included a determination of operability (the SBDGs had no specified safety function for the EQ and turbine building flood events in question per the UFSAR), installation of a temporary flood barrier to compensat for the non-conforming condition for the A SBDG, cable splice and terminal strip replacement for the B SBDG, and the performance of a root cause evaluation. |
| 05000331/FIN-2014005-05 | 31 December 2014 23:59:59 | Duane Arnold | NRC identified | Inadequate Containment Isolation Valve Leak Tightness Test Procedure | The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish an adequate procedure for an activity affecting quality for a system penetrating the primary containment pressure boundary. Specifically, Surveillance Test Procedure STP 3.6.1.1-09, Containment Isolation Valve Leak Tightness Test Type C Penetrations TIP (traversing in-core probe) Valves, Revision 4, failed to include leak rate testing instructions for all of the fittings inboard of the outboard TIP valves tested, which constituted part of the primary containment pressure boundary. The licensee entered the issue in their Corrective Action Program as condition report 02003580. As part of their corrective actions, the licensee re-performed a local leakage rate test to verify the fittings were leak tight. The inspectors determined that the finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedure quality and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. As it related to the finding, procedure STP 3.6.1.1-09 lacked adequate instructions to ensure no leakage of a system penetrating the primary containment pressure boundary. The finding was of very low safety significance (Green) because it did not represent an actual open pathway of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding was associated with the cross-cutting aspect of resources in the area of Human Performance because STP 3.6.1.1-09 did not include testing of the fittings inboard of the outboard TIP valve as required. (H.1) |
| 05000331/FIN-2014005-01 | 31 December 2014 23:59:59 | Duane Arnold | NRC identified | Construction Code Used during a Replacement Activity Not Reconciled with the Owner's Requirements | A finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Section 50.55a, Codes and Standards, was identified by the inspectors for the failure to reconcile the construction code and owners requirements when replacing rod hangers associated with the high pressure coolant injection (HPCI) system. The licensee subsequently performed a code reconciliation and concluded the applicable construction code requirements were met. The licensee captured this issue in its Corrective Action Program as condition report 01999594. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of HPCI to respond to initiating events to prevent undesirable consequences. Specifically, the failure to reconcile the construction code and owners requirements when replacing HPCI support rod hangers reduced confidence in the systems capability to meet its mitigating function consistent with its design basis. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. This finding had a cross-cutting aspect of procedure adherence in the area of Human Performance because the licensee failed to follow American Society for Mechanical Engineers Section XI, Administrative Manual for Repair, Replacement, and Modification. (H.8) |
| 05000331/FIN-2014005-02 | 31 December 2014 23:59:59 | Duane Arnold | NRC identified | Liquid Penetrant Testing Procedures Not Qualified for their Full Applicability Range | A finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors for the failure to properly qualify nondestructive testing procedures in accordance with applicable codes. Specifically, liquid penetrant testing procedures were not qualified for their full applicability temperature ranges in accordance with American Society for Mechanical Engineers (ASME) Code, Section V, Nondestructive Examination. The licensee entered this issue into the Corrective Action Program as condition report 01950601 and 01999596. As an immediate corrective action, the licensee reviewed completed liquid penetrant examination records and did not find an example where the procedures were implemented at the unqualified temperature range. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, since the liquid penetrant testing procedures were not qualified for their full applicability temperature ranges, liquid penetrant examinations were not assured to detect flaws in the unqualified temperature ranges. As a consequence, the potential would exist for a rejectable flaw to go undetected affecting the operability of the affected system. This finding affected the Initiating Event, Mitigating System, and Barrier Integrity cornerstones. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the inadequate qualifications were performed more than three years ago and was not confirmed to reflect current performance. |