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05000237/FIN-2018003-022018Q3DresdenLicensee-Identified ViolationViolation: Dresden Technical Requirements Manual (TRM) Control Program (Appendix G of TRM), Section 1.5, Program Implementation, requires that proposed changes to the TRM are screened and reviewed under the 10 CFR 50.59 process in accordance with plant specific procedures. Contrary to the above, in October 2017 Dresden station approved and implemented an extension to the surveillance frequency of DIS 150020, Division I & II Low Pressure Coolant Injection (LPCI) Pumps Suction and Injection Valves Circuitry Logic System Functional Test, on Unit 2 per the Surveillance Frequency Control Program (SFCP) without the required 50.59 review.
05000249/FIN-2018003-012018Q3DresdenFailure to Follow Maintenance Procedures for Assembling Unit 3 HPCI Room Cooler FanA self-revealing, Green non-cited violation (NCV) of Technical Specification (TS) 5.4, Procedures, was identified for the licensees failure to follow maintenance procedures DMP 570004, LPCI and HPCI Room Cooler Maintenance, and DEP 570004, HPCI Room Cooler Fan Preventive Maintenance, when assembling the Unit 3 HPCI room fan. Specifically, on one occasion when maintenance was performed on the fan, technicians installed the cam locking collar in the opposite direction of the fan shaft rotation, and on the other occasion, technicians tensioned the fan belt to the wrong value and misadjusted the alignment of the shaft sheave. Over time, this improper maintenance caused the inboard and outboard fan bearings to wear on the shaft, causing increased vibrations, and eventually leading to HPCI being declared inoperable to emergently work on the fan
05000282/FIN-2018002-012018Q2Prairie IslandResults of ISFSICask Array Dose Calculation Not Incorporated into FSARPrairie Island ISFSI FSAR, as updated, Revision 18, Section A7A.7 evaluates off-site dose rates for an array of ISFSI casks. In this dose rate calculation, explicit modeling credit is given to the earthen berm that surrounds the Prairie Island ISFSI as discussed in Section A7A.7.1. The earthen berm provides radiation shielding for the ISFSI. This calculation allows the licensee to demonstrate, in part, compliance with Title 10 of the Code of Federal Regulations (CFR) 72.104(a) which requires, in part, that, During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ. Calculation TN40HT0502, TN40HT Far Field Shielding Calculations, Revision 0, was performed by the licensee in support of a License Amendment Request (LAR) to modify the Prairie Island ISFSI TN40 cask design (designated as TN40HT casks). The TN40HT LAR was submitted to the NRC by the licensee on March 28, 2008. This dose rate calculation does not credit the earthen berm and, in part, also allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a). The licensee also provided this calculation directly to the NRC in a February 29, 2012, letter in response to a Request for Supplemental Information (RSI) from the NRC associated with the license renewal application for the Prairie Island ISFSI. Although the results from calculation TN40HT0502 for a single cask was incorporated into the Prairie Island ISFSI FSAR, Revision 18, in Tables A7A.22 and A7A.61, the results from TN40HT0502 for an array of casks which, in part, allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a), has not been incorporated into the ISFSI FSAR, Revision 18.Title 10 CFR 72.70, Safety analysis report updating requires, in part, that (a) Each specific licensee for an ISFSI shall update periodically, as provided in paragraphs (b) and (c) of this section, the FSAR to assure that the information included in the report contains the latest information developed (b) Each update shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last update to the FSAR under this section. The update shall include the effects of: (2) All safety analyses and evaluations performed by the licensee in support of approved license amendments.This Unresolved Item is being opened to determine whether or not the licensee is required to update the ISFSI FSAR with the results of calculation TN40HT0502 for an array of casks in accordance with 10 CFR 72.70.Planned Closure Action: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Material Safety and Safeguards to determine whether or not calculation TN40HT0502 is subject to the FSAR updating requirements of 10 CFR 72.70 for the Prairie Island ISFSI.
05000237/FIN-2017004-012017Q4DresdenFailure to Follow Procedure,Results in Non-Functional Fire DoorThe 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)
05000249/FIN-2017003-012017Q3DresdenGranted Notice of Enforcement Discretion 173001: LCO 3.1.7 Required Action B.1 per TS 3.1.7, Standby Liquid Control SystemInspection 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-2017002-012017Q2DresdenFailure to Maintain Configuration Control in the Unit 2 Containment Pressure Suppression SystemGreen . 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)
05000440/FIN-2017001-022017Q1PerryLicensee-Identified ViolationIn part, 10 CFR 20.1703 (c)(5) states, The licensee shall implement and maintain a respiratory protection program that includes Determination by a physician that the individual user is medically fit to use respiratory protection equipment. Contrary to the above, the licensee identified that an individual wore a powered air purifying respirator (PAPR) three times during the period of March 67, 2017 for the purpose of radiological protection without the required medical determination. This was entered into the licensees corrective action program, CR 201702957, Vessel Technician Wore PAPR Three Times without Being Qualified. The significance of this violation was determined in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process dated August 19, 2008. This violation was determined to be of very low safety significance (Green), because this violation was not associated with ALARA Planning or Work Controls, there was no overexposure nor substantial potential for overexposure and the ability to access dose was not compromised.
05000440/FIN-2017001-012017Q1PerryFailure to Implement Procedures for Combating a Loss of Shutdown CoolingGreen. A finding of very-low safety significance and associated NCV of TS 5.4, Procedures, was identified by the inspectors for the failure to implement procedures for combating a loss of shutdown cooling (SDC). Specifically, the licensee failed to implement its procedure for combating a loss of SDC resulting from emergency service water (ESW) inoperability and during high decay heat load. This finding was entered into the licensees Corrective Action Program to perform analyses for various conditions to identify available alternate methods of decay heat removal and provide associated procedural guidance. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as very-low safety significance (Green) because it was a design deficiency that did not impact the operability or Probabilistic Risk Assessment functionality of any mitigating structures, systems, and components. The inspectors did not identify a cross-cutting aspect associated with this finding because it did not reflect current performance due to the age of the performance deficiency
05000237/FIN-2017001-022017Q1DresdenSecondary Containment Inoperability Due to Lapse in Procedure Use and AdherenceA 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-2017001-012017Q1DresdenFailure to Correct a Condition Adverse to Quality Associated with EDG Single Largest Load Rejection Surveillance TestingThe 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-2016004-012016Q4DresdenFailure to Comply With Radiation Work Permit Requirements Resulting In Unplanned Dose Rate AlarmsA 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-2016004-022016Q4DresdenLicensee-Identified ViolationTitle 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.
05000249/FIN-2016010-012016Q4DresdenFailure to Verify the Adequacy of Design for the Unit 3 HPCI AOP Motor Shunt Resistor SettingA self-revealing finding preliminarily determined to be of low to moderate safety significance, and an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was associated with the licensees failure to ensure that the applicable design basis for applicable structures, systems, and components was maintained by the performance of design reviews, through the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, the licensee failed to verify the adequacy of design for the Unit 3 high pressure coolant injection (HPCI) auxiliary oil pump (AOP) motor shunt resistor setting during motor replacement in March of 2002, and then again in March of 2015, eventually resulting in pump failure in June of 2016, and inoperability of the HPCI system. The licensee documented this issue in its corrective action program (CAP) as IR 2686163. The inspectors determined that the licensees failure to verify the adequacy of design for the Unit 3 HPCI AOP motor shunt resistor setting was a performance deficiency, the cause was reasonably within the licensees ability to foresee and correct due to previous events and licensee generated causal determinations regarding the significance of adjusting the shunt field resistors on motor and pump operations, and should have been prevented. The inspectors determined the issue was more than minor because it adversely impacted the Mitigating Systems Cornerstone attribute of Design 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 (i.e. core damage). Specifically, the failure to control the design of the Unit 3 HPCI AOP motor resulted in the degradation and ultimate failure of the pump motor windings, which is a required component for HPCI operation. The inspectors applied IMC 0609, Attachment 4, and IMC 0609, Appendix A, Exhibit 2, Section A, for Mitigating Systems to screen this finding and determined that a detailed risk evaluation was required because the finding represented a loss of system and/or function. Therefore, a coordinated effort between inspection staff and regional Senior Reactor Analyst (SRA) was required to perform an appropriate risk evaluation for the degraded condition that resulted from the finding. The SRA used the Dresden Standardized Plant Analysis Risk (SPAR) model, version 8.24 for the detailed risk evaluation. This evaluation concluded that the exposure time for the HPCI system was 1 year. The total delta core damage frequency (CDF) for the 1 year exposure period was 6.9E6/year, which is a finding of low to moderate safety significance (White). HPCI is an important high pressure injection system that is used to mitigate internal events, internal flooding, and internal fire events at Dresden. The inspectors determined the contributing cause that provided the most insight into the performance deficiency was associated with the crosscutting area of Human Performance, Design Margins because the licensee failed to operate and maintain equipment within design margins, in that margins are carefully guarded and changed only through a systematic and rigorous process with special attention placed on maintaining fission product barriers, defense-in-depth, and safety-related equipment (H.6). Specifically, the licensee failed to verify the adequacy of design for the Unit 3 HPCI AOP motor shunt resistor setting during motor replacement in March of 2002 and then again in March of 2015.
05000237/FIN-2016003-022016Q3DresdenLicensee-Identified ViolationThe 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.
05000237/FIN-2016003-012016Q3DresdenFailure to Assess Scope Changes to Corrective Maintenance Activities Affecting Safety-Related Structures, Systems, and ComponentsA 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)
05000255/FIN-2016003-012016Q3PalisadesFailure to Appropriately Select and Review for Suitability of Application the Control Switch and Circuit Design of the Engineered Safeguards Room Cooler FansA self-revealed finding of very low safety significance and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, was identified for the failure to appropriately select and review for suitability of application the control switch and circuit design of the engineered safeguards room cooler fans. Specifically, on July 27, 2016, when the licensee was conducting troubleshooting activities for the tripping of engineered safeguards room cooler fan V27B, it was revealed that the control switch design was break before make and as the hand switch was transitioned from one position to the next, the supply voltage and the motor became out of phase and caused an overcurrent trip of the breaker. This resulted in an unplanned entry into a 72 hour limiting condition for operation (LCO) for the right train of the emergency core cooling system (ECCS). In the apparent cause evaluation (ACE) for this issue, the licensee determined that the contributing cause had not previously addressed this particular failure mode (i.e. the control switch and circuit design) when similar overcurrent events occurred in the past. Prior corrective actions included adding guidance to system operating procedures to pause between hand switch movements and replacing other components within those systems. These actions were not successful in eliminating this failure mode. The licensee documented the issue in their CAP, planned to revise the control circuit and switch design, and added specific procedural steps on how to operate these fans until the design change was implemented. The finding was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Reliability and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, as a result of the overcurrent trip of its breaker, V27B was declared non-functional and unavailable and the equipment in the room it cooled was declared inoperable, which included the A high pressure safety injection (HPSI) pump and the A containment spray (CS) pump. This led to an unplanned entry into a 72 hour LCO for the right train of ECCS. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution and was related to the cross-cutting component of Evaluation, which required that the licensee thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. As discussed above, in the ACE for this issue the licensee determined that the corrective actions associated with the identified contributing cause following similar overcurrent events that occurred in the past had not addressed or been successful in eliminating this failure mode (PI.2).
05000255/FIN-2016003-022016Q3PalisadesHourly Fire Tour DiscrepanciesThe inspectors identified an unresolved item (URI) related to discrepancies found during fire tour daily log sheet and corresponding badge record reviews. Specifically, the NRC is in the process of reviewing the licensees evaluation of the root and contributing causes of the issue, as well as the corrective actions to prevent recurrence. Also, the NRC will verify that the licensees actions taken to address the issue are sustainable. On May 24 and 25, 2016, while the inspectors were observing a maintenance activity on a service water pump in the screenhouse, they noted that hourly fire tours were not being conducted consistently by security personnel. The inspectors requested plant room badging records and copies of the hourly fire tour daily log sheets from the licensee for hourly fire tours completed on May 24 and 25, 2016. The inspectors identified that some areas on the fire tour log sheets were annotated as complete, yet there were no corresponding badge records for these areas. The inspectors requested additional fire tour daily log sheets and badge records for May 31 and June 1, 2016 for an extent of condition review. Additional issues were identified with the fire tour log sheets not corresponding with badge records for certain plant areas required to be covered by the hourly fire tours. On June 8, 2016, the inspectors discussed these discrepancies with the licensee. The licensee entered this issue into the CAP and promptly began an extent of condition review of the fire tour daily log sheets and plant room badging records for the period of March 1, 2016 through June 8, 2016. The condition report included actions to conduct a root cause evaluation to determine the root and contributing causes of the discrepancies identified in the fire tour and badging records and formulating corrective actions to prevent recurrence. The licensees immediate interim corrective actions included direct supervisor observation of all hourly fire tours being conducted, newly formatted fire tour log sheets with additional detail added, and re-training of personnel conducting the tours on the requirements and expectations for completion of the activity. Pending NRC review of the licensees evaluation of the issue, subsequent corrective actions to prevent recurrence, and verification that the actions are sustainable, this issue is unresolved.
05000237/FIN-2016002-012016Q2DresdenFailure to Implement and Maintain Written Procedures Regarding Breathing Air Quality TestingA 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-2016001-012016Q1DresdenFailure to Maintain Design Control of the 2/3 Emergency Diesel GeneratorA 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.
05000237/FIN-2015004-012015Q4DresdenFailure to Maintain Design Control of Secondary Containment Interlock DoorsA 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)
05000440/FIN-2015010-012015Q4PerryUnqualified Radiation Protection ManagerThe inspectors identified a finding of very low safety significance, and an associated violation of Technical Specification (TS) 5.3.1 when an unqualified individual was designated and performed the duties of the Radiation Protection Manager since early 2015. Specifically, the individual did not have the required experience and background necessary to provide sound judgement for safe and successful operation of the plant. This designation occurred after an April 29, 2015 report documented an internal review by the licensees Fleet Oversight group that concluded that the candidate did not meet qualifications of TS 5.3.1. The NRC determined that this violation did not meet the criteria to be treated as a Non-Cited Violation because this issue was not documented in the licensees Corrective Action Program. In addition, the licensees staff communicated to the inspector that no violation of TS had taken place The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612 because it was associated with the human performance 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, in that the lack of experience and background necessary to provide sound judgement for the Radiation Protection Program affects the licensees ability to control and limit radiation exposures. The finding was determined to be of very low safety significance (Green) 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 an overexposure nor a 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 aspect in the area of Human Performance, change management, because the licensee did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000461/FIN-2015004-012015Q4ClintonFailure to Update the Final Safety Analysis Report (FSAR) Hydrogen Water Chemistry SystemThe inspectors identified a Severity Level IV Violation of 10 CFR 50.71(e), Periodic Update of the FSAR (Final Safety Analysis Report), for the licensees failure to update the FSAR after installing a hydrogen water chemistry system into the plant to reduce rates of intergranular stress corrosion cracking in recirculation system piping and reactor vessel internals. Specifically, the licensee did not update Section 5.4.15, Hydrogen Water Chemistry System, of the FSAR to include a design basis and description of the process and the system used to periodically injection noble metals. The licensee entered this issue into the corrective action program as AR 02594259 and planned to revise the FSAR to include a design basis and description of the process and the system used to periodically injection noble metals. The inspectors determined that the failure to update the FSAR in accordance with 10 CFR 50.71(e), Periodic Update of the FSAR, with the design basis and description of the process and the system used to periodically injection noble metals was a performance deficiency warranting a significance evaluation. The inspectors reviewed this issue in accordance with NRC IMC 0612 and the NRC Enforcement Manual. Violations of 10 CFR 50.71(e) 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.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because cross cutting aspects are not assigned to traditional enforcement only violations.
05000461/FIN-2015004-022015Q4ClintonFailure to Perform Activities Affecting Quality in Accordance with Prescribed ProceduresThe inspectors identified a finding of very low safety significance for the failure to ensure that activities were accomplished in accordance with prescribed procedures as required by station procedure HU-AA-104-101 Procedure Use and Adherence. Specifically, the inspectors identified two examples where the licensee failed to adhere to prescribed station procedures when performing activities in the plant. The licensee placed both issues in their corrective action program as Action Request (AR) 02600726 and addressed the nonconformances created by the failure to follow the procedures. The licensee planned to perform an apparent cause evaluation to determine why there was an adverse trend related to procedure adherence. The inspectors determined that the failure to perform activities in accordance with prescribed procedures as required by station procedure HU-AA-104-101, Procedure Use and Adherence, 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, if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, by not performing activities in accordance with a procedure the licensee could manipulate equipment, challenge the operators, and cause unexpected transients. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) 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 cause a reactor trip or the loss of mitigation equipment, and it did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which stated, Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with unknown conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to station equipment.
05000461/FIN-2015004-032015Q4ClintonFailure to Follow Station Procedures for Plant ActivitiesThe inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of Title 10 Code of Federal Regulations (CFR), Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings. Specifically, the inspectors identified two examples where the licensee failed to perform activities affecting quality in accordance with prescribed procedures. The licensee entered this issue into their corrective action program as AR 02600726 and planned to perform an apparent cause evaluation to address the trend. Separate action requests were also written and immediate corrective actions were taken for each identified example to address the nonconformances created by the failure to follow procedures. The inspectors determined that the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings as required by 10 CFR 50, Appendix B, Criterion V, 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, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities affecting quality in accordance with a procedure the licensee could manipulate equipment and challenge the operators by causing unexpected transients or impact safety-related equipment. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, issued May 9, 2014, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the finding did not represent a loss of system safety function, it did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, it did not represent an actual loss of safety function of one or more non-TS trains of equipment during shutdown for equipment designated as risk significant for greater than 24 hours, and it did not degrade a functional auto-isolation of residual heat removal on low reactor vessel level. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which states, Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with uncertain conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to safety-related equipment and activities.
05000263/FIN-2015003-012015Q3MonticelloInadequate Evaluation of Refueling Floor Structural Steel BeamsThe inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, on September 3, 2008, licensee personnel failed to verify the adequacy of design when they failed to use correct section properties in their calculation of stresses on structural steel beams supporting the refueling floor for the increased spent fuel cask loading. Reevaluation of the beams using correct methodology resulted in the conclusion that the beams would not meet the design basis stress limits. Immediate corrective actions for this issue included initiation of a CAP, performance of a functionality assessment which concluded that the refueling floor remained functional but non-conforming, and creating compensatory measures which limited the refueling floor live load in the cask loading area (CAP 1492837). The inspectors determined that the licensees calculational methodology was contrary to the standard engineering principles applicable for determination of stresses in structural members, which resulted in a failure to meet Criterion III, Design Control, and was a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612 because it was associated with the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers (reactor building) protect the public from radionuclide releases caused by accidents or events. Additionally, More than Minor Example 3.j of IMC 0612, Appendix E, Examples of Minor Issues, was used to inform the more than minor screening. The inspectors used IMC 0609, SDP, Attachment 4, Initial Characterization of Findings, and Appendix A of IMC 0609 to screen this finding. The inspectors answered No to questions C.1 and C.2 in Exhibit 3, Barrier Integrity Screening Questions. As a result, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000255/FIN-2015003-022015Q3PalisadesFailure to Establish, Implement, and Maintain the Offsite Dose Calculation ManualA finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.5.1, Offsite Dose Calculation Manual, was identified for the failure to establish, implement, and maintain the Offsite Dose Calculation Manual (ODCM) relative to dose calculation parameters. Specifically, the licensee failed to modify the parameters used in public radiation calculations when changes in the use of unrestricted areas were identified. As a result, the quarterly and annual doses that were calculated every 31 days, as required by the ODCM, were incorrect and non-conservative. In addition to entering this issue into their CAP as CRPLP20152972, the licensee recalculated the dose using the correct calculation parameters. The performance deficiency was determined to be more than minor 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. The finding was determined to be of very low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, because the issue did not represent a significant deficiency in evaluating a planned or unplanned effluent release since the resulting dose was not grossly underestimated. The finding had a cross-cutting aspect of Training in the Human Performance cross-cutting area because the licensee did not ensure adequate knowledge transfer to maintain a knowledgeable, technically competent workforce.
05000254/FIN-2015003-022015Q3Quad CitiesFailure to Establish Adequate Procedure to Preclude Unacceptable Preconditioning of the SBGT SystemA finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to establish a procedure appropriate to the circumstances that precluded unacceptable preconditioning of the standby gas treatment (SBGT) system during surveillance testing. The licensee performed an evaluation and concluded the SBGT system was operable and planned additional testing on the relay timing function. Other corrective actions included revising the applicable procedures such that unacceptable preconditioning would not occur. The licensee captured this issue in their CAP as IR 2524699. The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Procedure Quality and 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. Specifically, the inadequate procedure had the potential to mask the ability of the SBGT system to initiate in time to prevent ex-filtration of radioactive gases during a design basis accident. The finding was determined to be of very low safety significance because it represented a degradation of the radiological barrier function for the SBGT system. This finding had a cross-cutting aspect of questioning attitude in the area of human performance because the licensee did not recognize the possibility of mistakes, latent problems, or inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to recognize that performing the steps in the specified sequence could unacceptably precondition the time-delay relay for the SBGT system and mask the ability of the system to perform its function (H.12).
05000316/FIN-2015003-022015Q3CookFailure of Steam Dump Valves Results in Plant TripA finding of very low safety significance (Green) was self-revealed on April 23, 2015, when two condenser steam dump valves failed open during startu following the Unit 2 refueling outage. In response to the failure, the licensee manually tripped the Unit 2 reactor. Contrary to the requirements of PMP5040MOD007, Engineering Modifications, the design of the new valves that were installed was not compatible with the steam dump system. This finding does not involve enforcement action because no violation of a regulatory requirement was identified. The licensee replaced three steam dump valves on Unit 2 with a new design during the spring refueling outage. Shortly following reactor startup, two of the new valves failed open after being placed in service. The resulting temperature transient required operators to manually trip the reactor to comply with Technical Specification (TS) requirements for minimum temperature while critical. Design work and planning to perform the modifications failed to meet timeliness milestones prior to the outage. Contrary to the modification procedure for these circumstances, the change was not considered fast-track, therefore, additional risk assessments and management oversight were not provided. As a result, the operational impact of the new design was not fully realized. The steam dump system can be subject to significant amounts of condensate. The new valves trapped some of the condensate. This, along with a different plug design, caused a backpressure of sufficient force to cause the valves to fail open when steam was admitted. The licensee stabilized the plant following the trip, replaced two valves with the old design, isolated the other via a temporary modification, and returned the unit to service. The issue was also entered into the Corrective Action Program (CAP) as Action Request (AR) 20155825. The issue was more than minor because it adversely affected the Design Control attribute of the Initiating Events Cornerstone, whose objective is to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the inadequate design caused the new valve to fail open, which resulted in a manual reactor trip. Utilizing IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, effective July 1, 2012, the inspectors determined the finding was Green, or very low safety significance, by answering no to the Transient Initiators question in Exhibit 1. Specifically, while the transient caused a plant trip, all mitigation equipment remained available to respond to the trip. The inspectors determined the finding had an associated cross-cutting aspect in the Human Performance area, namely, H.8, Procedure Adherence. The licensee failed to follow the requirements of the modification procedure, which would have prompted a more thorough review of the modification.
05000315/FIN-2015003-052015Q3CookPRA Model ErrorsDuring review of MSPI, the inspectors identified an URI associated with the adequacy of the Probable Risk Assessment (PRA) model of record used for MSPI. Specifically, an error in the PRA model resulted in extremely high hours for system unavailability. The same error impacted the PRA for NFPA805 and internal events. Description: While reviewing MSPI, the inspectors identified an URI regarding the licensees PRA model. While reviewing margin reports as part of the inspection, the inspectors noted that available margin for unavailability for all the SSCs covered by MSPIs was extremely high. The SSC with the least amount of available margin had about 250,000 hours of unavailability before the MSPI would cross the threshold from Green to White. The inspectors discussed the issue with licensee and learned that an error in the PRA model of record caused the high numbers. AR 20143184 documents the issue. The licensee told the inspectors that during NFPA 805 reviews, the NRC had inquired about the models treatment of test and maintenance. In response, the licensee identified the error, and corrected the model for NFPA 805. The licensee has also updated the model for on line risk management. Because of additional controls on the model used for MSPI, the corrected model did not become the PRA model of record until September 2015. The inspectors inquired if the licensee had performed any evaluation to validate that information previously submitted to the NRC using the flawed model resulted in a masked greater-than-green MSPI result. The licensee stated they believed all the prior MSPI submittals would remain green; however, the licensee also stated that the NRC endorsed NEI guidance did not require licensees to resubmit MSPI data if the model changed. In particular, the licensee noted that NEI 9902 guidance regarding model revisions requires the model to be in effect for the entire quarter. A revised PRA model becomes effective on the first day of the quarter following approval of the revised model as the PRA model of record. Thus, a revised model would not impact previously submitted MSPI data. In addition, the inspectors learned that some of the licensee conditions for NFPA 805 addressed the adequacy of the PRA model and imposed requirements for conducting peer review of the PRA model. During the inspection period, the inspectors could no determine if the licensee satisfied the associated license conditions. The inspectors noted that NEI 9902 also requires the PRA model to be technically accurate with requirements included in appendix G. In addition, NEI 9902 also requires licensees to correct data errors for previously submitted data. In this case, the inspectors, during the inspection period, could not determine: 1) If the flawed model met the Appendix G requirements for technical adequacy; 2) if the model does not meet appendix G, would the data be considered in error and in need of correction 3) impacts of the flawed PRA model on other documents submitted to the NRC; and 4) impact of the error on license condition for NFPA 805 Pending resolution of the above items, this issue is considered a URI. (URI 05000315/2015003-05; 05000316/2015003-05; PRA Model Errors)
05000315/FIN-2015003-032015Q3CookDeletion of Hot Shutdown Panel ProceduresThe inspectors identified an Unresolved Item (URI) related to deletion of procedures used to operate the HSD. The UFSAR and TS bases describe the HSD and its use; therefore procedures to operate the panel should have remained in place. Licensing actions, including NFPA 805 conversion and transition to improved TSs complicate the current license bases requirements for the HSD. Description: In 2003, the licensee determined that the HSDs were not required under appendix R since local instrumentation panels had been installed. The licensee prepared a 50.59 screen that inappropriately concluded that the procedures could be deleted without assessing the deletion using a full evaluation. The licensee deleted the procedures but failed to address the discussion of the HSDs in the UFSAR and TS bases. Subsequent to deletion of the procedures, the licensee received approval to convert their TSs from custom TSs to improved TSs. The revised TS still discussed the HSDs; however, reference to specific instruments were moved from the TS to the TS bases. In addition to the conversion to improved TSs, the licensee also converted fire protection from appendix R to NFPA 805 via the license amendment process. This revision recognized that the local panels would be credited for achieving and maintaining safe shutdown from outside the control room. However, the HSD satisfies draft GDC 11, which is part of the current licensing basis, and states the license must be able to shutdown the reactor and maintain it in a safe condition if access to the control room is lost due to fire or other cause. In 2009, the licensee recognized the UFSAR still substantively discussed use of the HSDs despite the deletion of procedures for them, and entered this issue into the CAP; however, the CAP did not result in substantive changes to the UFSAR and also failed to recognize the improper screen performed in 2003. In reviewing this issue, the inspectors recognized that the issue involved multiple changes to the license bases and that multiple violations of NRC requirements might exist. Because of the interactions between various licensing actions and requirements, this issue will remain a URI pending better understanding of potential violations and the current license bases for the HSD. As part of the inspection, the inspector reviewed the requirements of TS 3.3.4, Remote shutdown Monitoring Instrumentation. This TS addresses five indication functions on the HSDs and the licensee continues to perform surveillances on these instruments. Therefore, instrumentation remains operable. In addition, the licensee has entered the condition into the CAP and developed new procedures to operate the HSD.
05000263/FIN-2015003-042015Q3MonticelloDrywell to Torus Vacuum Breaker Past OperabilityDuring the cycle preceding the 2015 refueling outage, two evaluations associated with torus to drywell vacuum breaker operation were developed due to issues identified in the first quarter 2014. These included: CAP 1417977, Failure of drywell-torus vacuum breaker to close, which identified an occasion of dual indication during Procedure 0143 procedure. A second occurrence was observed several days later and was documented in CAP 1418471, AO-2382A Torus-to-DW vacuum breaker closed indication anomaly. CAP 1420318, DW-Torus vacuum breaker work performed with inadequate PMT, identified the PMT following shaft sealing component (O-ring) replacement during the 2013 outage was not performed as planned. The licensee evaluations for these CAP conditions concluded the Drywell to Torus vacuum breakers were operable. However, neither evaluation specifically considered the effect of an interference between the vacuum breaker test lever and vacuum breaker test actuator stem. Since this specific mechanism was not addressed in these two evaluations, past operability of the torus to drywell vacuum breakers was questioned. As a result, the licensee established a past operability evaluation be conducted via CAPs 1479198 and 1478212. The licensee completed its past operability evaluation on June 26, 2015. After review, the inspectors conveyed a number of questions to the licensees engineering staff in regard to the past operability evaluation. Although the licensee provided responses for the majority of these questions during the remainder inspection quarter, the licensee had requested external input in regard to one of the inspectors questions. Specifically, inspectors questioned whether it was possible for the bottom of the lever arm to be at an elevation above the top of the actuator stem at valve disc full open and if so, could the valve test lever arm have come to rest on top of the actuator stem, potentially impacting the ability of the vacuum breaker valve to close. Upon the close of this inspection period, that input had not yet been finalized and made available to the inspectors. As a result, this issue was considered to be an unresolved item pending a review of the licensees response and past operability for CAPs 1479198 and 1478212, including and the licensee response to open inspector questions.
05000263/FIN-2015003-052015Q3MonticelloFailure to Provide Complete and Accurate Information in LER 05000263/2015-002-00The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9 due to the licensees failure to provide information to the NRC that was complete and accurate in all material respects in accordance with the NRCs reporting requirements in 10 CFR 50.73(a)(1), Licensee Event Report (LER) System. Specifically, on June 29, 2015, the licensee failed to include an accurate assessment of the safety consequences and implications of a loss of shutdown cooling event when they issued LER 05000263/2015-002-00. This LER included an inaccurate assessment of safety implications, stating that engineering calculations show a potential worst case maximum temperature of 115 degrees Fahrenheit (F). The inspectors identified that engineering models actually showed potential worst case temperatures of 25-26 degrees F higher, which could have challenged or exceeded fuel pool cooling design specifications. Corrective actions included issuance of a revision to LER 2015-002-00 which contained the correct engineering modeling results and associated discussion of safety implications. The licensee entered this issue into its CAP (CAP 1484633). This issue was of more than minor significance under the Traditional Enforcement Process because the NRC relies on licensees to identify and correctly report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., loss of shutdown cooling) was evaluated separately and determined to be a finding of very low safety significance as documented in the 2015 2nd Quarter Integrated Inspection Report (05000263/2015002-01). In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was of more than minor concern with relatively inappreciable potential safety significance and is related to a finding that was determined to be a more than minor issue. Consistent with Example 6.9.d.1, this represented an example where the licensee submitted inaccurate information in a required report, which resulted in expansion of the scope of the next regularly scheduled inspection and required LER revision. Because there was no finding evaluated with this violation, the inspectors did not assign a cross-cutting aspect to this issue.
05000315/FIN-2015003-042015Q3CookChanges to Minimum 60-Minute Emergency Responder Staffing Without Prior ApprovalThe inspectors identified a finding of very-low safety significance with an associated Severity Level IV (SL-IV) NCV of Title 10, Code of Federa Regulations (CFR) 50.54(q)(3) and 10 CFR 50.54(q)(4) related to a staffing change in the licensees Emergency Plan that reduced the effectiveness of the Plan, which was made without prior NRC approval. Specifically, in March 2004, the licensee made changes to wording in the Donald C. Cook Emergency Plan that allowed two Radiation Protection (RP) Technician positions to be augmented by staff that were not qualified RP Technicians. This issue was placed in the licensees CAP and was corrected by revising the Emergency Plan to the approved augmented staffing minimum. The finding was of more than minor significance because it was associated with the Emergency Preparedness Cornerstone attribute of Procedure Quality, and affected the cornerstone objective of ensuring the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, a failure to evaluate changes to the Emergency Plan as required by 10 CFR 50.54(q)(3) resulted in unacceptable changes made to the plan that decreased its effectiveness without prior NRC approval as required by 10 CFR 50.54(q)(4) and reduced the licensees capability to perform an emergency planning function in the event of a radiological emergency. The finding was of very low safety significance because it was a failure to comply that did not result in a loss of the planning standard function. In accordance with Section 6.6.d of the NRC Enforcemen Policy, this violation was categorized as SL-IV because it involved the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the Emergency Plan decreases. The inspectors concluded that because the performance deficiency involved a change to the licensees Emergency Plan in March 2004, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified. (Section 1EP4.b.1) Violations of very low safety or security significance or SL-IV that were identified by th licensee have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees CAP. These violations and CAP tracking numbers are listed in Section 4OA7 of this report.
05000263/FIN-2015003-022015Q3MonticelloFailure to Perform High Radiation Area Portable Fire Extinguisher SurveillancesThe inspectors identified a finding of very low safety significance and an NCV of Technical Specification (TS) 5.4.1.d when the licensee failed to implement procedures associated with Fire Protection Program Implementation, to ensure that required refueling outage surveillances were performed for fire extinguishers located in high radiation areas (HRAs). Specifically, between March 2007 and May 2015, the licensee failed to implement steps 9 and 10 of 1123, Portable Fire Extinguishers, which required weighing and verifying adequate hydrostatic testing of the fire extinguishers in HRAs on a refueling outage frequency. Corrective actions included surveillance process changes and evaluation of the current status of the high radiation area fire extinguishers which resulted in the determination that outside of the surveillance process, a separate work activity had exchanged all the affected extinguishers with ones that were current on their surveillances in May 2015. This issue was entered into the licensees Corrective Action Program (CAP) 1484257 The inspectors determined that the failure to implement HRA fire extinguisher surveillances was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and 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). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Work Management aspect because of the failure to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority and the failure to identify the need for coordination with different groups or job activities
05000263/FIN-2015003-032015Q3MonticelloFailure to Identify Safe Shutdown Equipment Impacts in Fire Strategy ProceduresThe inspectors identified a finding of very low safety significance and an NCV of TS 5.4.1.d when the licensee failed to maintain procedures associated with Fire Protection Program Implementation, consistent with the Updated Safety Analysis Report (USAR), to ensure that fire strategy procedures accurately indicated safe shutdown (SSD) equipment. Specifically, on June 25, 2015, the licensee failed to maintain A.3-12-C, Condenser Room Fire Strategy, to ensure SSD equipment was appropriately identified. In this case, fire strategy A.3-12-C failed to identify any SSD equipment in the room, despite the fact that SSD cabling ran through the room and was included in the USAR Fire Hazards Analysis. Corrective actions included performance of an extent of condition review which identified 40 other fire strategies where safe shutdown cabling was not identified, and initiation of procedure changes to include the appropriate SSD equipment. This issue was entered into the licensees CAP (CAP 1484142). The inspectors determined that the failure to maintain fire strategy procedures to ensure that SSD equipment was identified was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and 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). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Self-Assessment aspect because of the licensees failure to conduct self-critical and objective assessments of its programs and practices.
05000255/FIN-2015003-012015Q3PalisadesFailure to Justify Continued Service of Safety-Related Electrolytic Capacitors Installed Beyond Their Service LifeAn NRC-identified finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, was identified for the failure to justify continued service of safety-related electrolytic capacitors that were installed beyond their recommended service life associated with the safety-related containment floor level indicating transmitters (LITs). Specifically, on June 21, 2015, containment floor LIT LIT0446B and LIT0446A did not satisfy the acceptance criteria of the technical specification surveillance monthly channel checks and LIT0446B was declared inoperable. Further troubleshooting identified a failure of the electrolytic capacitor within the transmitters converter module and that this failure was most likely due to age since the transmitter had been in service for greater than its recommended service life. In addition to entering this issue into their Corrective Action Program (CAP) as CRPLP201504972, the licensee replaced the failed components and planned to develop a replacement schedule for non-critical, safety-related electrolytic capacitors. The performance deficiency was determined to be 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 (i.e., core damage). The finding screened as having very low safety significance based on answering No to all of the screening questions in the Mitigating Structures, Systems, and Components (SSCs) and Functionality section of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Mitigating Systems Screening Questions. The finding had a cross-cutting aspect of Operating Experience in the Problem Identification and Resolution cross-cutting area because the licensee did not effectively and thoroughly evaluate and implement relevant industry operating experience and guidance for age-related electrolytic capacitor degradation.
05000254/FIN-2015003-012015Q3Quad CitiesFailure to Evaluate Degraded or Non-Conforming Conditions for OperabilityA finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to document degraded or non-conforming conditions in the corrective action program (CAP) and route or discuss the issue with Operations shift management so that operability of the affected components could be evaluated. Immediate corrective actions included entering the issues into the CAP and evaluating the issues for operability. The licensee captured the issue in the CAP as Issue Reports (IRs) 2537968 and 2537936. The finding was determined to be more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the failure to identify degraded, non-conforming, or unanalyzed conditions in the CAP and bring those conditions to the attention of Operations shift management so that the operability of safety-related systems, structures, and components (SSCs) may be evaluated could lead to those SSCs being in an inoperable condition without the appropriate Technical Specification (TS) actions taken. The inspectors concluded this finding was associated with the Mitigating Systems Cornerstone. The finding was determined to be of very low safety significance because the control room emergency ventilation (CREV) and high pressure coolant injection (HPCI) systems remained operable. This finding had a cross-cutting aspect of identification in the area of problem identification and resolution because the licensee did not identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, when degraded and non-conforming conditions were identified, licensee personnel failed to promptly capture the issues in the CAP (P.1).
05000254/FIN-2015003-032015Q3Quad CitiesFailure to Adequately Inspect Relay Contacts for Oxidation Results in Relay FailureA finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to establish a preventive maintenance procedure for HFA relays that was appropriate to the circumstances. Immediate corrective actions included burnishing of the associated relay contacts and testing the associated relays. In addition, the licensee revised their relay inspection procedure and planned future relay replacements during the next refueling outage. The licensee entered the issue into their CAP as IR 2485051. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the failure to perform adequate preventive maintenance on the automatic depressurization system (ADS) logic HFA relay in 2013 resulted in the build-up of oxidation on the relay contacts. This build-up caused the relay to fail its next scheduled test in 2015. A senior reactor analyst performed a detailed risk evaluation and determined the finding was of very low safety significance. This finding had a cross-cutting aspect of operating experience in the area of problem identification and resolution, because the licensee did not systematically collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, the licensee identified several internal and external operating experience events related to relay contact oxidation and failed to implement changes to their relay inspection procedures to ensure that effective corrective actions were implemented (P.5).
05000315/FIN-2015003-012015Q3CookFailure to Evaluate Fire Brigade Fire Fighting TechniquesThe inspectors identified a finding and associated NCV of Facility Operating Licenses DPR-58 condition 2.C(4) and DPR 74 Condition 2. C(3)(o), Fire Protection Program. Specifically, the licensee failed to identify and subsequently critique the failure of the Fire Brigade and Operations to de-energize a battery charger during a fire drill. On August 20, the inspectors observed an unannounced fire drill. In the scenario, the licensee simulated a fire in a nonsafety-related battery charger in the turbine building. The licensee fire brigade and on shift operations personnel responded. During the drill, the licensee failed to simulate securing direct current (DC) power to the battery charger and subsequently failed to critique this issue. The inspectors discussed the DC power issue with the licensee and the licensee agreed that the drill should have evaluated the DC power supply and the fire brigade should have simulated removing the DC power source. The licensee has briefed site personnel on de-energizing equipment with multiple power sources and entered the condition into the corrective action program. The licensees failure to demonstrate effective firefighting techniques and subsequent failure to critique the error was a performance deficiency of Green significance. The performance deficiency was more than minor because it was associated with the protection against external factors 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 undesirabl consequences. The finding screened as green using IMC 0609 Appendix M with insight from Appendix F. The finding included a cross-cutting aspect of training, H.9, in the human performance area.
05000315/FIN-2015003-062015Q3CookLicensee-Identified ViolationA finding of very low safety significance (Green) with an associated NCV of TS 5.4, Procedures, was identified by the licensee for the failure of th 1AB EDG during testing following a maintenance period. Shortly after the EDG was started, it automatically shut down on high bearing temperature Investigation revealed that the #4 main bearing had failed. The licensee performed a root cause analysis which determined that electric arcing ha occurred through the bearing which led to the failure. One of the contributors to the arcing was that air had been left in the lube oil system following maintenance. TS 5.4, Procedures, requires, in part, that the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 be established, implemented, and maintained. Section 9 of Regulatory Guide 1.33 states, in part, that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to this requirement, the procedures for performing maintenance on the lube oil system allowed air to remain in the lube oil system, which helped facilitate electric arcing in the 1AB EDG bearings. The issue was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems cornerstone, with the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened a Green based on answering no to the Mitigating Systems screening questions in IMC 0609 Appendix A, The Significance Determination Process for Findings at Power, effective July 1, 2012. The issue was entered into the CAP as AR-2015-6917. The inspectors concluded the issue was licensee-identified based on the guidance in IMC 0612, Power Reactor Inspection Reports, issue date January 24, 2013. Under the definition of licensee-identified findings, IMC 0612 states that most, but not all, licensee-identified findings or violations are discovered through a licensee program or process. One of the processes listed is post-maintenance testing, which was how the bearing failure was discovered.
05000315/FIN-2015003-072015Q3CookLicensee-Identified ViolationA finding of very low safety significance (Green) with an associated NCV of TS 3.5.2, Emergency Core Cooling System-Operating, was identified by th licensee for the failure to properly address an oil leak on the Unit 1 East RHR Pump. A leak was identified in March of 2015 and assessed to have n operability impact. Following entry into Mode 5, Cold Shutdown, for the 1AB EDG repairs in June 2015, the leak was again identified and written-up in the CAP. The operability determination performed identified that given the rate of leakage, the pump would not have been able to operate for its thirty-day mission time. A past operability assessment was performed which determined the pump would not have fulfilled the mission time when the pump was required to be operable per TS in Modes 14. This should have been identified when the leak was discovered in March. TS 3.5.2 requires two RHR trains to be operable in Modes 14. Contrary to this requirement, the East RHR train was inoperable between March and June of 2015. The issue was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems cornerstone, with the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as Green because the pump would have been able to perform for at least its 24 hour PRA mission time. The inspectors conclusions regarding significance were confirmed after consultation with a regional Senior Risk Analyst. The issue was entered into the CAP as AR20158659 and AR20157898.
05000263/FIN-2015002-032015Q2MonticelloFailure to Maintain Secondary Containment and Standby Gas Treatment System Operable During OPDRV ActivitiesThe inspectors identified a finding of very low safety significance and an associated NCV of TS 3.6.4.1, Secondary Containment and TS 3.6.4.3, Standby Gas Treatment System (SBGT) because the licensee did not maintain secondary containment and the SBGT system operable as required during activities considered OPDRVs. Specifically, on April 14, 2015, and again on May 13, 2015, the licensee failed to classify activities associated with draining reactor inventory as OPDRVs while relying on an automatic isolation function for the drain path, and as a result failed to maintain required equipment operable during these activities. Once questioned by the inspectors, the licensee took action to control other outage related draining activities as OPDRVs and placed this issue into its CAP (CAP 1479284). The inspectors determined that the failure to maintain secondary containment and SBGT operable while an OPDRV was in progress was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, RCS, and containment) protect the public from radionuclide releases caused by accidents or events because the secondary containment boundary and the SBGT were not maintained operable during an OPDRV activity. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G, the Shutdown Operations SDP since the reactor was shut down. The finding was assessed in accordance with IMC 0609 Appendix G, Attachment 1, Exhibit 4 and Appendix H for containment integrity findings. Using Appendix H, the inspectors concluded the finding had very low safety significance (Green) because decay heat was low and containment was deinerted. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Documentation aspect because of the failure of the licensee to create and maintain complete, accurate and up-to-date documentation (H.7).
05000263/FIN-2015002-072015Q2MonticelloOperations with a Potential to Drain the Reactor Vessel (OPDRV) Without Secondary Containment OperableA violation involving a failure to have secondary containment operable during Operations with the Potential to Drain the Reactor Vessel (OPDRV) was identified. Specifically, from April 23, 2015 through May 8, 2015, Monticello Nuclear Generating Plant performed a total of three activities within two work windows without setting secondary containment, which is a violation of Technical Specification (TS) 3.6.4.1. The NRC issued Enforcement Guide Memorandum (EGM) 11-003, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel, Revision 2, on December 13, 2013, allowing for the exercise of enforcement discretion for such OPDRV-related TS violations, when certain criteria are met. The NRC concluded that Monticello Nuclear Generating Plant met these criteria during the activities for which the EGM was invoked. Therefore, I have been authorized, after consultation with the Director, Office of Enforcement, and the Regional Administrator, to exercise enforcement discretion and refrain from issuing enforcement for the violation. Between April 23, 2015 and May 1, 2015 and again between May 2, 2015 and May 8, 2015, the Monticello Nuclear Generating Plant (MNGP) performed OPDRV activities while in Mode 5 without an operable secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the RPV water level below the top of fuel, without crediting the use of mitigating measures to terminate the uncovering of fuel. Secondary containment is required by TS 3.6.4.1 to be operable during OPDRV activities. The required action for this specification is to suspend OPDRV operations. Therefore, entering the OPDRV without establishing secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). The NRC issued EGM 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliance with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to (1) adhere to the NRC plain language meaning of OPDRV activities, (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times, (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a LOI event, (b) maintaining level monitoring to ensure secondary containment can be closed before inventory is drained to the RPV flange, (c) maintaining the capability to isolate the potential leakage paths, (d) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (e) prohibiting movement of recently irradiated fuel with the spent fuel storage pool gates removed in Mode 5, and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities. The inspectors reviewed this licensee event report (LER) for potential performance deficiencies and/or violations of regulatory requirements. The inspectors also reviewed the stations implementation of the EGM during the OPDRVs for which the EGM was invoked. Based on review of the following items, the inspectors determined that the licensee met the EGM requirements for discretion: 1. The inspectors observed that the OPDRV activities were logged in the control room narrative logs and that the log entry appropriately recorded that the standby source of makeup designated for the evolutions. 2. The inspectors noted that the reactor vessel water level was maintained at least 21 feet and 11 inches over the top of the RPV flange as required by TS 3.9.6. The inspectors also verified that at least one safety-related pump was available as the standby source of makeup designated in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the RPV flange was greater than 24 hours. 3. The inspectors reviewed Engineering Change documents which calculated the time to drain down during these activities and the feasibility of pre-planned actions the station would take to isolate potential leakage paths during these periods of time. 4. The inspectors verified that the OPDRVs were not conducted in Mode 4 and that the licensee did not move recently irradiated fuel during the OPDRVs. The inspectors noted that MNGP had in place a contingency plan for isolating the potential leakage path and verified that two independent means of measuring RPV water level were available for identifying the onset of LOI events. TS 3.6.4.1 required, in part, that secondary containment shall be operable during OPDRV. TS 3.6.4.1, Condition C, required the licensee to initiate action to suspend OPDRV immediately when secondary containment is inoperable. Contrary to the above, between April 23, 2015 and May 1, 2015 and again between May 2, 2015 and May 8, 2015, MNGP performed OPDRV activities while in Mode 5 without an operable secondary containment. Specifically, the station performed the following OPDRV activities without an operable secondary containment: 12 Recirculation System pump upper seal replacement; 12 Recirculation System modifications to add and replace valves; and 11 Recirculation System modifications to add and replace valves. Because the violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation (EA-15-130). In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request within 12 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The inspectors observed that Monticello is tracking the need to submit a license amendment request in its CAP (CAP 1476012). LER 05000263/2015-001-00 is now closed. This event follow-up review constituted one sample as defined in IP 71153-05.
05000263/FIN-2015002-042015Q2MonticelloFailure to Fill the Reactor Cavity in Accordance with Refueling Preparation ProcedureThe inspectors identified a finding of very low safety significance and an associated NCV of TS 5.4.1, Procedures, on April 15, 2015, when the licensee failed to implement procedure 9001, Reactor Well & Dryer-Separator Storage Pool Filling Procedure, for refueling preparation activities. Specifically, when faced with indications that the condensate storage tanks (CSTs) did not contain enough water inventory to complete outage critical path reactor pressure vessel (RPV) flooding activities, the licensee failed to implement 9001 procedure steps for using prescribed equipment and methods to fill the reactor cavity. With the proceduralized methods unavailable, operators used the site decision-making process to utilize demineralizer water hoses to fill the cavity rather than processing required 9001 procedure changes. This issue was entered into the licensees CAP (CAP 1474891). Immediate corrective actions included action to initiate the procedure change process for 9001 and department communication to Operations regarding the incident, emphasizing that the decision making process is not a substitute for the procedure change process. The inspectors determined that the failure to fill the reactor cavity in accordance with the 9001 reactor well filling procedure was a performance deficiency requiring evaluation. The inspectors evaluated IMC 0612, Appendix E, and did not find any similar examples of minor issues. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the operations crews use of the decision-making process to support outage critical path by bypassing proceduralized steps and performing activities using methods contrary to the procedure could lead to a more significant safety concern. In addition, if performed incorrectly (i.e. without flushing the hoses prior to use), the use of demineralizer hoses could introduce foreign material into the core and challenge the integrity of the fuel cladding barrier. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G, the Shutdown Operations SDP since the reactor was in Mode 5 (refueling). The finding was assessed in accordance with IMC 0609 Appendix G, Attachment 1, Exhibit 4 for Barrier Integrity and determined to have very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Conservative Bias aspect because of the failure of the individuals to use decision-making practices that emphasize prudent choices over those that are simply allowable, and the failure to ensure that proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop (H.14).
05000263/FIN-2015002-022015Q2MonticelloFailure to Measure Interpass TemperatureThe inspectors identified a Green NCV of Title 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, for a failure to measure the interpass temperature while performing welding on diesel generator fuel oil modification supports. Consequently, welding was performed without the Code and Procedure required interpass temperature being Monitored on a number of welds, a parameter which can affect the mechanical properties of the material being welded. To restore compliance, the welder proceeded to measure the interpass temperatures on the balance of the welds and verified that the interpass temperature did not exceed that allowed by procedure. The licensee entered this issue into its CAP (CAP 1475767). The inspectors determined that this issue was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because the inspectors answered yes to the more than minor question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically, absent NRC intervention, the welder would have completed all of the welds without having measured the interpass temperature, a welding parameter which can affect the mechanical properties (e.g., impact properties) of some materials being welded, and if left uncorrected could lead to a potential failure of the weld in service. In accordance with Table 2, Cornerstones Affected by Degraded Condition or Programmatic Weakness, of IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, the inspectors checked the box under the Mitigating Systems Cornerstone because leakage on the Emergency Diesel Generator (EDG) fuel oil system could cause core decay heat removal to be degraded. The inspectors determined this finding was of very-low safety significance (Green) based on answering yes to the question in Part A of Exhibit 2, Mitigating Systems Sc reening Questions, in IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued on June 19, 2012. Specifically, the inspectors answered yes to the screening question If the finding is a deficiency affecting the design or qualification of a mitigating Structure, System, or Component (SSC), does the SSC maintain its operability or functionality? The welder proceeded to measure the interpass temperatures on the balance of the welds and verified that the interpass temperature did not exceed that allowed by procedure, and the issue did not result in the actual loss of the operability or functionality of a safety system. The inspectors determined that the primary cause of the failure to monitor the interpass temperature procedure was related to the cross-cutting component of Problem Identification and Resolution, Operating Experience (P.5). Specifically, the organization failed to effectively implement external operating experience in a timely manner.
05000263/FIN-2015002-012015Q2MonticelloFailure to Maintain Portable Fire Extinguishers in Accordance with Fire StrategyThe inspectors identified a finding of very low safety significance and an NCV of TS 5.4.1.d when the licensee failed to implement procedures associated with Fire Protection Program Implementation to ensure that portable fire extinguishers were maintained in accordance with the fire strategy. Specifically, on May 1, 2015, the licensee failed to implement fire protection p an procedures when they failed to control three portable fire extinguishers in the condenser room, a room housing safe shutdown cabling, in accordance with Fire Strategy A.3-12-C. In this case, inspectors found that of the four dry chemical extinguishers required to be stationed in the condenser room, two indicated that they were partially depleted and needed to be recharged, and a third extinguisher was missing entirely. Immediate corrective actions included recharging the partially depleted extinguishers and procuring a portable extinguisher to replace the missing one. This issue was entered into the licensees CAP (CAP 1477246). The inspectors determined that the failure to implement the fire strategy procedure to ensure that condenser room portable fire extinguishers were maintained was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612 Appendix B because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and 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). Because the plant was shut down, the inspectors assessed the significance of this finding in accordance with IMC 0609, Appendix G, the Shutdown Operations SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Identification aspect because of the failure to implement a CAP with a low threshold for identifying issues, and failure to ensure that individuals identify issues completely, accur tely, and in a timely manner in accordance with the program (P.1)
05000263/FIN-2015002-062015Q2MonticelloLoss of Electrical Buses and Shutdown Cooling (SDC) Due to Inadequate Procedure AdherenceA self-revealed finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified due to the failure to properly implement Procedure 0304-01, Safeguard Bus Loss of Voltage Protection Relay Unit Calibration Safeguards Bus No. 15. Specifically, electrical maintenance workers failed to comply with Step 20 which directed the installation of a jumper between terminals ZX10 and ZX11 in an electrical panel, when they incorrectly installed the electrical jumper between terminals ZX11 and ZX12. This resulted in the loss of the Division I safety related 4160 Volts Alternating Current (Vac), 480 Vac, and 125 Volts Direct Current (Vdc) electrical buses, which subsequently led to the loss of shutdown cooling (SDC) for approximately 3 hours and 15 minutes. Initial corrective actions for this issue included immediately invoking strict plant status controls to focus efforts on recovery, restoring the electrical buses and SDC to operation, and reinforcing risk recognition and human performance tools. This issue was entered into the licensees CAP (CAP 1477351) and a root cause evaluation (RCE) was in progress at the time this inspection period concluded. The inspectors determined that the issue was more than minor because it adversely impacted the Initiating Events Cornerstone attribute of Human Performance and Configuration Control, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors utilized IMC 0609, Appendix G for shutdown operations and determined that the issue was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency aspect because of the failure of licensee individuals to implement error reduction tools and the failure of the organization to plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12).
05000263/FIN-2015002-052015Q2MonticelloInadequate Clearance Order Results in Unplanned OPDRVA self-revealed finding of very low safety significance and an associated NCV of technical specification (TS) 5.4.1, Procedures, was identified on May 16, 2015, when the licensee failed to implement procedure FP-OP-TAG-01, Fleet Tagging, for equipment control activities associated with the Scram Discharge Volume (SDV). Specifically, the licensee failed to ensure that clearance order checklist 58972-03 restored valve I-CRD-R-26, an SDV instrument vent valve, to its normal position prior to returning the SDV system to service. As a result, during subsequent reactor coolant system (RCS) pressure boundary testing, RCS water leaked out onto the reactor building floor through the open vent line, creating an unplanned operation with a potential for draining the reactor vessel (OPDRV). This issue was entered into the licensees CAP (CAP 1479307). Immediate corrective actions included termination of the leakage by closing and capping the SDV vent line and resetting the scram. The site initiated an apparent cause evaluation (ACE), which was in progress at the end of the inspection period. The inspectors determined that the failure to adequately restore the SDV system to service in accordance with fleet tagging requirements was a performance deficiency requiring evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, because it adversely impacted the Initiating Events Cornerstone attributes of Configuration Control and Procedure Quality, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G, the Shutdown Operations significance determination process (SDP) since the reactor was in Mode 4 (cold shutdown). The finding was assessed in accordance with IMC 0609 Appendix G, Attachment 1, Exhibit 2 for Initiating Events. Using IMC 0609 Appendix G, Attachment 3, for a Phase 2 analysis, the inspectors determined it to have very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Challenge the Unknown aspect because of the failure of individuals to stop when faced with uncertain conditions and the failure to ensure that risks are evaluated and managed before proceeding (H.11).
05000255/FIN-2015001-052015Q1PalisadesFailure to Evaluate the Adverse Effects of the Use of Non-Seismic Temporary JumpersA Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very low safety significance was identified by the inspectors when licensee personnel failed to maintain a written safety evaluation that provided a basis that the use of temporary alligator clip jumpers to maintain emergency diesel generator (EDG) operability during certain maintenance activities did not require a license amendment. Specifically, the licensee did not address the adverse effects of the use of alligator jumpers on the design and qualification of the diesel generator (DG) circuit breaker used per Engineering Change 50310 and changes to procedure SPSE1, 2400 Volt and 4160 Volt Allis Chalmers and Siemens Vacuum Circuit Breaker Auxiliary Switch Adjustments, Revision 34. This issue was entered into the licensees CAP as CRPLP201404859, NRC Identified 50.59 Issue, dated October 7, 2014. The performance deficiency was determined to be more than minor because it was associated with the Design Control 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. Specifically, the change that was implemented adversely affected the seismic qualification of the electrical circuit that was relied upon to ensure safety bus 1C would be loaded by the 11 DG upon a loss of offsite power. The inspectors evaluated the underlying technical issue and determined the finding was of very low safety significance. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the finding associated with this violation was determined to be of very low safety significance. This finding had a cross-cutting aspect in the Conservative Bias component of the Human Performance cross-cutting area. Specifically, the licensee did not use all available information and relevant guidance, such as Nuclear Energy Institute 9607, to demonstrate that the proposed activity was safe and did not require a license amendment prior to implementation.
05000255/FIN-2015001-042015Q1PalisadesFailure to Verify the Adequacy of Credited High Energy Line Break BarriersA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified by the inspectors when the licensee credited fire doors for High Energy Line Break (HELB) protection without a supporting test or evaluation. Specifically, Procedure 4.02 credited fire doors with protection of safety-related equipment against a HELB when the primary HELB barrier was disabled without a test or evaluation to demonstrate the doors could withstand the HELB environment. This issue was entered into the licensees CAP as CRPLP201500371, NRC Concerns with Calculation EAPSACCWHELB0217, dated January 22, 2015. The performance deficiency was determined to be more than minor because it was associated with the Design Control 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. Specifically, the licensee did not have an evaluation to demonstrate that barriers relied upon to protect mitigating systems from a HELB initiating event could perform the credited protection function. The inspectors answered No to the questions in Exhibit 2.A, Mitigating Systems Screening Questions, and as a result determined the issue was of very low safety significance. This finding was not associated with a cross-cutting aspect since the calculation in question was created in 2003 and therefore did not represent current performance.
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