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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)
05000255/FIN-2016004-022016Q4PalisadesFailure to Correct an Adverse Condition Associated with Diesel Generator Load Sequencer ModuleGreen. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to correct an adverse condition associated with the emergency diesel generator (DG) load sequencer and power supply module as revealed when the electrolytic capacitor failed two days after installation. The 12 DG was declared inoperable, the licensee replaced the failed module, and an equipment apparent cause evaluation was completed for the equipment failure. An internal operating experience review revealed that a similar issue occurred in 2005 and corrective actions to address that failure, which included establishing shelf life and age requirements for electrolytic capacitors that were part of power supply modules, were not applied to this module. The licensee entered this issue into their CAP as CRPLP201603260. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, because the performance deficiency 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). Specifically, the licensee failed to correct a condition adverse to quality, which rendered the 12 DG inoperable. This condition would have prevented the DG from automatically starting and loading on the prescribed signal. The finding was screened in accordance with IMC 0609, Appendix A, and was determined to have very low safety significance (Green) based on answering No to all the screening questions under the Mitigating Structure, System and Components, and Functionality section. The inspectors concluded that the corrective actions for the adverse condition of the aging electrolytic capacitors should have been implemented greater than three years ago, so the finding was not reflective of current licensee performance. Therefore, no cross-cutting aspect was identified.
05000255/FIN-2016004-032016Q4PalisadesFailure to Translate Design Analysis Stack-up Configuration into Specifications, Drawings, Procedures, and InstructionsGreen. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to establish measures to assure that the applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to provide instructions in procedures to construct the spent fuel dry cask loading stack-up, in the safety-related auxiliary building, in the configuration that had been analyzed for in the stack-up seismic design basis calculation. In addition, the licensee failed to provide instructions in revised procedures to construct the stack-up without certain gaps as 4 specified in the stack-up seismic design basis document. The licensee documented these issues in their CAP as CRPLP201600646, CRPLP201601308, CRPLP201601558, CRPLP201604497, and CRPLP201604826; revised the stack-up seismic analysis to address the identified issues; and translated the analyzed stack-up design configuration into stack-up installation procedures prior to performing stack-up operations with spent nuclear fuel in the multi-purpose canister. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, 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 design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a stack-up configuration that did not ensure stack-up dynamic stability or Auxiliary Building structural integrity to maintain radiological barrier functionality during a design basis seismic event. The finding screened as having very low safety significance (Green) because it did not result in the loss of operability or functionality of the Auxiliary Building. The finding had a cross-cutting aspect of Field Presence in the Human Performance cross-cutting area, because licensee senior managers failed to ensure effective supervisory and management oversight of contractor activities related to the seismic analysis and installation of the stack-up configuration (H.2).
05000255/FIN-2016004-012016Q4PalisadesFailure to Have Appropriate Controls in Place for Combustible MaterialsGreen. A finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 48(c) was identified by the inspectors for the licensees failure to appropriately implement the requirements of procedure ENDC161, Control of Combustibles. Specifically, between January 1, 2016 and October 22, 2016, the inspectors identified several examples of the licensees failure to have appropriate controls in place for the storage of combustible materials in excess of the limits required for those respective areas without a completed transient combustible evaluation (TCE). Also, on several occasions from October 19, 2016 to October 22, 2016, the required compensatory actions for a TCE related to the dry fuel storage cask transporter vehicle were not appropriately implemented as required by procedure ENDC161. The licensee entered these issues in their corrective action program (CAP) as condition reports (CRs) CRPLP201603633, CRPLP201605148, and CRPLP20160564. Corrective actions for these issues included completing the required TCEs, ensuring the combustible materials in the areas were addressed by the combustible loading calculations, and ensuring appropriate compensatory measures were implemented. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Protection Against External Factors attribute, in the area of Fire, 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). Specifically, transient combustible materials without required TCEs were stored in the charging pump cubicles and in the refueling and spent fuel pool areas. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, since none of the stored materials were self-igniting, low flashpoint liquids, or heat sources and was therefore assigned a Low degradation rating. The finding had a cross-cutting aspect of Training in the Human Performance cross-cutting area due to the common element of a lack of knowledge of the individuals with the control of combustibles process and understanding their roles in that process (H.9).
05000282/FIN-2016002-012016Q2Prairie IslandLicensee-Identified ViolationPrairie Island TS 3.6.3, Containment Isolation Valves, Required Action A.1 required, in part, isolation of the affected penetration flow path within 4 hours if one or more penetration flow paths with one containment isolation valve inoperable. Contrary to the above, since August 4, 2012 on 21 occasions for Unit 1 and 23 occasions for Unit 2 (three year reporting window), the licensee failed to isolate containment spray header penetration flow paths within 4 hours during the performance of quarterly containment spray pump surveillance procedures SP 1090A & 1090B and SP 2090A & 2090B. Specifically, the SPs inappropriately credited Note 1 of TS 3.6.3 and created open flow paths from the Unit 1 and 2 containments under administrative control while vent and/or drain valves connected to the containment spray header were opened. The opening of these valves was to facilitate draining of the header and to verify no leakage past manual isolation valves during containment spray pump operation in recirculation mode. On August 4, 2015, the licensee generated CAP 01488454 which questioned whether use of TS 3.6.3 Note 1 to open the containment spray header vent and drain valves under administrative control was permissible. The licensee performed an apparent cause evaluation and determined that because the vent and drain valves were not considered part of a containment penetration flow path, Note 1 could not be applied. A past operability review was performed and it was determined that on multiple occasions (at 1-10 hour durations) over the prior three years, the vent/drain opening resulted in a 3/8 opening in the containment pressure boundary. Because the resultant leakage at peak containment pressure during a design basis accident (approximately 4 percent of the containment volume per day) would have exceeded the maximum allowable leakage rate, conditions that could have prevented the fulfillment of the safety function of the Units 1 and 2 containments and, conditions that were prohibited by TS, had occurred. Because the inspectors answered Yes to question B.1 under Exhibit 3, Barrier Integrity Screening Questions of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors transitioned to IMC 0609, Appendix H, Containment Integrity Significance Determination Process. Because the leak rate through the vent/drain openings would not have exceeded greater than 100 percent of the containment volume per day at calculated peak containment internal pressure, the finding screened as very low safety significance (Green). The issues were entered into the licensees CAP as CAP 01488454. Corrective actions included immediate quarantine of the affected SPs and subsequent revisions to the SPs and TS Bases.
05000255/FIN-2016001-012016Q1PalisadesDesign Review of Modification to Track Alley Wall for Dry Fuel Storage ActivitiesThe inspectors identified a unresolved item (URI) associated with the design review of a modification to the Track Alley wall for dry fuel storage (DFS) campaign activities. Specifically, the licensee is currently revising the process applicability determination (50.59 and 72.48 screenings), and reviewing any necessary actions, associated with altering the newly modified wall in support of upcoming DFS campaign activities. The wall, a protective barrier with safety functions per the UFSAR, in its newly modified condition, will be altered when the steel plate covering the opening cut into it will be raised to accommodate the DFS transporter. The DFS campaign is currently on hold pending resolution of other issues. In January 2016, the licensee began work on an engineering change to permanently modify the west wall of Track Alley in order to accommodate the new transporter used for moving the casks associated with the dry fuel storage campaign. This modification removed a section of the reinforced concrete wall by cutting out an opening approximately 9 feet wide by 4 feet high by 18 inches deep into the existing wall. A three inch thick steel plate was mounted onto vertical rails which can slide down to cover the window cut into the wall and raised to open the window for when the transporter is brought into Track Alley. The west wall of Track Alley is also the east wall of the Technical Support Center (TSC). This wall is designed to withstand seismic, high wind, and tornado missile loads. It also serves as a radiation protection barrier for personnel in the TSC during emergency situations. The permanent modification of cutting the opening in the wall and installing the steel plate, to provide equivalent protection of the 18 inches of concrete that were cut out, was evaluated in Engineering Change 59170 and calculation EAEC5917001. The inspectors reviewed these documents, the supporting process applicability determination (50.59 screening), and risk assessment of implementing the design change. During this review, the inspectors identified that the licensee did not assess the alteration of the wall, a protective barrier with safety functions per the UFSAR, when the steel plate covering the window would need to be raised to accommodate the DFS transporter. The inspectors questioned this condition and the licensee subsequently completed a process applicability determination (PAD) form (72.48 and 50.59 screening). When reviewing the PAD, the inspectors questioned the licensees underlying assumption that moving the steel plate to uncover the window was considered to be in support of a maintenance activity and, hence, screened out of the 50.59 process, including not requiring certain compensatory actions for the walls safety functions during the period of time in which the opening was exposed. At the end of the inspection period the licensee was reviewing their assessment. Once their review is completed, including any changes that may be made, the inspectors will re-assess their evaluation and determine what actions, if any, will need to be accomplished in support of the DFS campaign. Since the campaign is on hold, a URI is being opened to track resolution of this issue.
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.
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.
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.
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.
05000237/FIN-2015002-042015Q2DresdenFailure to Ensure Continued Operability of Unit 2 ERV 2-02033C (2C) Following Implementation of Extended Power Uprate Plant ConditionsA finding preliminarily determined to be of lowto-moderate safety significance, and an associated Apparent Violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control; TS 3.4.3, Safety and Relief Valves; and TS 3.5.1, ECCS Operating, was self-revealed on February 7, 2015, following the discovery that one of the Unit 2 electromatic relief valves (ERVs) would not have performed its intended safety function. Vibration induced wear experienced while operating at extended power uprate (EPU) power levels resulted in the degradation of multiple ERV actuator subcomponents, which rendered the valve inoperable. This finding does not represent an immediate safety concern in that the licensee has replaced all Unit 2 and 3 ERV actuators with a hardened design successfully utilized at the Quad Cities Nuclear Power Station, which has also experienced significant steam line vibrations post EPU. The inspectors determined that the licensees apparent failure to ensure measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of SSCs, in particular ERV 2-0203-3C (2C), was a performance deficiency warranting a significance evaluation. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone attributes of design control and equipment performance, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance. The inspectors determined that this finding has a cross-cutting aspect of Resolution in the area of Problem Identification and Resolution, since it involves the failure to implement effective corrective actions to address issues in a timely manner commensurate with their safety significance. This cross-cutting issue is conditional depending on the outcome of the preliminary White finding. (P.3)
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).
05000237/FIN-2015002-032015Q2DresdenReactor Scram Due to Feedwater Level Control System Failure with a Reactor Recirculation Pump RunbackA finding of very-low safety significance (Green) was self-revealed on January 13, 2015, and again on February 6, 2015, when a loss of power to the Unit 2 feedwater level control (FWLC) system resulted in a reactor scram. The loss in power to the Unit 2 FWLC system was determined to be the result of a human performance error during the original installation of the system under Work Order (WO) 97102835, in that two spade-lug connections associated with the systems +5 Vdc power supply were not properly landed resulting in the intermittent losses in power, and reset of the FWLC system. In addition, a dual in-line package switch on a FWLC Input/Output card was improperly positioned which led to an improper anti-cavitation reactor recirculation pump runback during both events. The inspectors determined that the failure to properly land the leads associated with the Unit 2 FWLC system +5 Vdc power supply in accordance with the work instructions in WO 97102835 was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the configuration control attribute of the Initiating Events cornerstone, and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very-low safety significance (Green), because the inspectors answered "No" to the screening question, Did the finding cause a reactor trip AND the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss off condenser, loss of feedwater)? This finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee did not thoroughly evaluate repetitive alarms and a failure of the FWLC system to ensure that resolutions addressed causes and extent of condition prior to restart following the January 13, 2015, FWLC failure and reactor scram. Specifically, licensee analysis of alarms received prior to the January 13, 2015, scram and troubleshooting of the FLWC system failure on January 13, 2015, was overly focused on multi-functional processor cards which happened to be approaching their end of expected life. Activities to investigate loose wiring connections following the January 13, 2015, scram failed to identify the incorrectly landed spade-lug connections for the +5 Vdc power supply. (P.2)
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).
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-2015002-032015Q2MonticelloFailure 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-2015002-022015Q2MonticelloFailure 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)
05000237/FIN-2015002-052015Q2DresdenLicensee-Identified ViolationThe TS 5.5.1 required that the Offsite Dose Calculation Manual (ODCM), and its Radiological Environmental Monitoring Program (REMP) be established, implemented, and maintained. ODCM Radiological Environmental Control No. 12.6.1 defined the surveillance requirements for the REMP. Step E of this section provided requirements for Milk Station D-25 (Control) be sampled within 10 km to 30 km semimonthly as indicated in Table 12.6-1.4.a. Contrary to the requirements, the licensee did not sample the control Milk Station D-25. The missed samples were not identified by the licensee until March 2015. This issue was entered into the licensees CAP as AR-02469852.
05000237/FIN-2015002-022015Q2DresdenInadvertent Manipulation of a Test Switch at ESF Bus 23-1 During Surveillance Testing Results in the Inoperability of the 2/3 EDG to Unit 2A finding of very low safety significance (Green), and an associated NCV of TS 5.4.1, Procedures, was self-revealed on May 19, 2015, when the 2/3 EDG was made inoperable to Unit 2 due to the incorrect manipulation of a test switch by operations personnel during a TS required surveillance test. Specifically, while the licensee performed procedure DIS 1500-05, Division I and II Low-Pressure Coolant Injection ECCS Initiation Circuitry Logic System Functional Test, Step 106 ofChecklist B, operations personnel incorrectly opened test switch TS-159SD2/3 at motor control center 23-1 removing the under-voltage trip associated with the feed breaker for the Division I safety-related 4.16 kV engineered safeguards bus, causing the 2/3 EDG to be inoperable to Unit 2. The licensees failure to properly implement steps in the procedure was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the Mitigating Systems Cornerstone Attribute of Configuration Control, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very- low safety significance (Green), because each of the questions provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, were answered No. The finding has a cross-cutting aspect in the area of Human Performance, Field Presence, for failing to ensure senior managers applied the appropriate oversight of infrequently performed and first time work activities. Specifically, the licensee field supervisor or another senior operations manager was not present for the switching activities, which led to the configuration control error. In this instance, the surveillance test is infrequently performed (every 24 months), and the activity, which included using a maintenance procedure vice an operating procedure, was a first time evolution for both equipment perators involved. (H.2)
05000237/FIN-2015002-012015Q2DresdenFailure to Meet Technical Specification Surveillance Requirements Due to Foreign Material Left in the Unit 2 EDG Starting CircuitA finding of very-low safety significance (Green) was self-revealed on April 21, 2015 while performing TS Surveillance DOS 6600-12, Diesel Generator Tests: Endurance and Margin/Full Load Rejection/ECCS (Emergency Core Cooling System)/Hot Restart, in support of Surveillance Requirement 3.8.1.16 which requires the EDG to achieve rated frequency and voltage conditions within 13 seconds when started less than or equal to five minutes from a previously loaded run, the Unit 2 Emergency Diesel Generator (EDG) failed to complete a hot restart. Licensee troubleshooting identified a degraded pressure switch associated with main bearing lube oil pressure in the start circuit which was taking several minutes to return to a low-pressure condition upon shutting down the EDG. This resulted in a failure of the start circuit relay to be energized upon initiating a start of the EDG, until the pressure switch returned to its appropriate low-pressure state. An internal investigation of the pressure switch identified strips of Teflon tape in the bellows of the pressure switch, which resulted in the pressure switchs sluggish response to lowering lube oil pressure, and a failure to meet the TS hot restart criteria. The inspectors determined that the failure to implement Procedure MAAA716-008, Foreign Material Exclusion Program, and therefore the inability to perform TS Surveillance Requirement 3.8.1.16 was a performance deficiency, and was considered more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone, and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors utilized Attachment 0609.04, Initial Characterization of Findings, and determined that this issue was of very-low safety significance because each question provided in Inspection Manual Chapter (IMC) 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, was answered No. The inspectors concluded that this finding was cross-cutting in the Human Performance, Documentation area, because licensee procedure MA-AA716-008, Foreign Material Exclusion Program, work instructions associated with Work Order 01410972-01, and previous calibrations of pressure switch 2-6641-526 did not include specific instructions and warnings regarding the proper use of Teflon tape with regards to preventing it from becoming foreign material. Other Dresden maintenance procedures, specifically MA-DR-0300-001, Preventive Maintenance of Hydraulic Control Unit, and DEP 0300-16, Rebuilding the Unit 2 (3) ASCO Scram Solenoid Pilot Valves, have specific warnings regarding the proper use and potential for Teflon tape to become foreign material. (H.7)
05000263/FIN-2015001-062015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.1, Offsite Dose Calculation Manual, (ODCM) which requires in part, that licensee initiated changes to the ODCM shall be effective after approval of the plant manager. Contrary to the above, ODCM01.01 Revision 6 and ODCM02.01 Revision 10, were not approved by the plant manager prior to implementation. This was identified by the licensee as part of the self-assessment process. The licensee documented this issue in the corrective action program (CAPs 1455999 and 1462092). This finding was determined to be of very-low safety significance (Green) because it was not a failure to implement an effluent program and public dose did not exceed Appendix I of 10 CFR 20.1301(e) criteria.
05000263/FIN-2015001-012015Q1MonticelloFailure to Identify High Pressure Coolant Injection (HPCI) Seismic Support NonconformanceThe inspectors identified a finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify conditions adverse to quality, such as deficiencies, deviations, and nonconformances. Specifically, on February 11, 2015, the inspectors identified a safety related seismic support for high pressure coolant injection (HPCI) turbine trip instrumentation that was not rigidly attached, supported, and restrained in accordance with plant construction code and installation specifications, a nonconformance which the licensee had failed to identify since initial plant construction. Corrective actions for this issue included repairs to the seismic support to rigidly connect the instrument line restraint and installation of a standalone support for the instrument tray. This issue was entered into the licensees corrective action program (CAP 1465906). The inspectors determined that the failure to promptly identify an HPCI instrument line support nonconformance was a performance deficiency requiring evaluation. The inspectors determined that the issue was more than minor because it adversely impacted the Mitigating Systems Cornerstone attribute of Protection Against External Factors, 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 in accordance with IMC 0609 and determined that it 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 Problem Identification and Resolution, and the aspect of Identification because the licensee failed to implement a CAP with a low threshold for identifying issues (P.1).
05000263/FIN-2015001-042015Q1MonticelloInadequate Evaluation of Operating Crew During Simulator AssessmentThe inspectors identified an URI on March 16, 2015, due to the licensees potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. In accordance with IMC 0612, Power Reactor Inspection Reports, the inspectors determined that this issue represented an URI because more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. On March 16, 2015, the NRC inspectors observed a potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. Specifically, during an NRC observation of a Licensed Operator Training self-assessment and emergency preparedness objective demonstration, the inspector observed that the evaluators may not have adequately critiqued a knowledge deficiency in the Interpreting and Diagnosing Events competency area when evaluating a Shift Managers (SM) performance. The Shift Managers performance could have adversely impacted EAL classification during a graded self-assessment. This assessment included an evaluated Drill/Exercise Performance (DEP) opportunity for the EAL classification in question. During the inspectors observation, they noted that the critique session did not appear to adequately probe why the classification-related performance weaknesses occurred, and did not appear to determine a course of specific actions for the crew to take to improve individual performance relative to the SMs role in the EAL classification. Specifically, the inspectors noted that at the end of the critique, this item was not discussed as an item needing resolution, nor was it discussed that the SM had a challenge to his qualifications and needed potential remediation, which appeared to be contrary to the sites MTCP0349 procedure. These discussions and follow-up actions did not take place until after the critique had concluded and the NRC inspectors raised questions about the SMs misinterpretation of Safety Parameters Display System (SPDS) and his overall performance. This item represents an issue of concern about which more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. The NRC inspectors will work to obtain additional guidance and clarification/interpretation of the existing guidance in order to resolve this issue. Corrective actions for this issue included disqualifying the individual, developing a remediation plan, and initiating procedure changes to improve the critique process. This issue was entered into the corrective action program as CAP 1470975. (URI 05000263/201500104, Inadequate Evaluation of Operating Crew During Simulator Assessment)
05000263/FIN-2015001-032015Q1MonticelloFailure to Maintain a Standard Emergency Action Level Scheme for FloodingThe inspectors identified a finding of very low safety significance and an NCV of Title 10 CFR 50.54(q)(2) and 10 CFR 50.47(b)(4) for the licensees failure to maintain the effectiveness of the emergency plan. Specifically, from May 28, 2014, until February 26, 2015, the HA1.6 Emergency Action Level (EAL) threshold was in conflict with the EAL basis for the alert classification. Additionally, both the revised EAL threshold and original NRC-approved safety evaluation report EAL threshold were later found to be greater than the actual river level that could lead to damage of safe shutdown equipment. The licensees corrective actions documented that the current river level was 906 and if flooding were to occur the licensee would have relied on Procedure A.6, "Acts of Nature," and that an event response team would have been formed to monitor river level during the duration of a flood event. The licensee concluded that the shift manager, Event Response team, and plant management would have monitored for indication of degraded performance of equipment or structures necessary for safe shutdown for event classification escalation to the Alert level. The licensee entered this issue into the Corrective Action Program (CAP 1454593). The inspectors determined that establishing a flooding EAL threshold that was in conflict with approved EAL basis as required by 10 CFR 50.47(b)(4), and subsequent failure to determine the actual level that could lead to damage of safe shutdown equipment for the alert classification High River Level EAL HA1.6 was a performance deficiency. The inspectors determined that the issue was more than minor because it is associated with the Procedure Quality attribute of the Emergency Preparedness (EP) cornerstone and adversely affected the cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined that it 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 Problem Identification and Resolution, Evaluation aspect because the licensee did not thoroughly evaluate the identified engineering error issue to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000263/FIN-2015001-022015Q1MonticelloFailure to Maintain Fire Protection Program Procedures for Control of Portable Heater/Extension Cord Fire HazardsA finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1.d was self-revealed when the licensee failed to maintain procedures for Fire Protection Program Implementation to ensure that ignition sources (space heaters) were properly controlled to prevent plant fires. Specifically, on January 26, 2015, the licensee failed to maintain Fire Protection Program implementation procedures to include controls to ensure space heaters used in the plant stayed within allowable load ratings and were plugged directly into outlets without the use of extension cords. This resulted in a fire in the plant recombiner building which was extinguished within 13 minutes, nearing the 15 minute time limit at which a Notification of Unusual Event (NOUE) would have needed to be declared. It also resulted in a space heater causing an overloaded outlet at a location in the reactor building, near A residual heat removal (RHR) equipment. Upon discovery of the recombiner area fire, the licensee dispatched the fire brigade to ensure the fire was extinguished, performed extent of condition walkdowns in the plant, and took action to improve controls on extension cord and portable heater use in the power block. This issue was entered into the licensees corrective action program (CAP 1463506). The inspectors determined that the failure to maintain fire program procedures to ensure ignition sources (space heaters) were appropriately controlled was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor because, if left uncorrected, the failure to adequately control portable heater related fire hazards in the plant could lead to more significant safety concerns. In addition, the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of Protection Against External Factorsincluding fire, 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 assessed the significance of this finding in accordance with IMC 0609 and determined that it 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 Problem Identification and Resolution, Evaluation aspect because of the failure to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000282/FIN-2015001-032015Q1Prairie IslandUntimely Resolution of Environmental Qualification IssuesA self-revealing finding of very low safety-significance and a non-cited violation of 10 CFR 50.49 was identified on March 5, 2015, for the licensees failure to keep environmental qualification (EQ) files current and the failure to replace or refurbish EQ electrical equipment at the end of its designated life. Specifically, the licensee initiated CAP 1431268 in May 2014 to document numerous EQ file errors identified during an in-depth review of the EQ program. These file errors resulted in the EQ designated life for multiple safety-related solenoid valves being non-conservative such that some solenoids were installed beyond their designated life. Corrective actions included taking action to revise the incorrect EQ files and replacing the safety-related solenoids installed beyond their designated life. The inspectors determined that this issue was more than minor because if left uncorrected the failure to maintain the EQ files and to replace or refurbish EQ equipment could result in a more significant safety concern. Specifically, the inaccurate files could result in EQ equipment not being refurbished or replaced as required. In addition, the failure to replace or refurbish EQ equipment installed beyond its designated life could result in equipment failure during normal operation or post-accident conditions. The inspectors utilized IMC 0609, Attachment 0609.04, Initial Characterization of Findings, and determined this issue was of very low safety significance because each of the questions provided in IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, was answered No. The inspectors concluded that this issue was cross-cutting in the Problem Identification and Resolution, Evaluation area because the licensee had not thoroughly evaluated CAP 1431268 to ensure that the resolution addressed the causes and extent of condition commensurate with the safety significance (P.2).
05000282/FIN-2015001-022015Q1Prairie IslandFailure to Follow Foreign Material Exclusion Procedure during Reactor Coolant Pump Seal ReplacementA self-revealing finding of very low safety significance and associated NCV of TS 5.4.1 was identified on December 19, 2014, due to the licensees failure to follow Procedure FPMAFME01, Foreign Material Exclusion and Control. Specifically, workers failed to implement and adhere to the foreign material exclusion (FME) control requirements for a Level 1 foreign material exclusion area when replacing the Unit 1 reactor coolant pump (RCP) seals and associated piping during Refueling Outage 1R29. The failure to implement and adhere to the FME control requirements resulted in introducing foreign material into the reactor coolant system and the subsequent degradation of the #12 RCP seal in December 2014 and January 2015. The seal degradation led to two Unit 1 reactor shutdowns. Corrective actions for this issue included replacing the RCP seal, flushing the seal piping and establishing a process to review work document quality to ensure that appropriate programmatic requirements were included. The inspectors determined that the failure to follow Procedure FPMAFME01 was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors utilized Attachment 0609.04, Initial Characterization of Findings, and determined that this issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, was answered No. The inspectors concluded that this finding was cross-cutting in the Human Performance, Work Management area, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. In addition, the work process failed to include the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities (H.5).
05000306/FIN-2015001-012015Q1Prairie IslandFailure to Perform Immediate Operability Determination for 14 CFCU as Required by ProcedureAn inspector identified finding of very low safety significance and a NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," occurred on January 27, 2015, due to operations personnel failing to follow Procedure FPOPOL01, Operability/Functionality Determination, while assessing the operability of the 14 containment fan coil unit (CFCU) and the Unit 1 containment. Specifically, personnel failed to perform an immediate operability determination for the 14 CFCU and the Unit 1 containment after the inspectors identified that the 14 CFCU was potentially leaking. Corrective actions for this issue included documenting the immediate operability determination after the inspectors brought this issue to the attention of the operations department and sharing the details of this event with other operations personnel. The inspectors determined that the failure to perform an immediate operability determination on the 14 CFCU and the Unit 1 containment as required by Step 5.3.1 of Procedure FPOPOL01 was more than minor because if left uncorrected, the failure to perform operability determinations, as required by procedure could result in incorrect/untimely operability conclusions and the failure to take action to correct degraded or deficient conditions, as required by the technical specifications (TS). In addition, this is the second example of an untimely CFCU operability determination identified by the inspectors in the last ten months. The inspectors utilized IMC 0609, Attachment 0609.04, Initial Characterization of Findings, and determined that this issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, Part B, was answered No. The inspectors concluded that this finding was cross-cutting in the Human Performance, Teamwork area because individuals and work groups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained (H.4).
05000237/FIN-2015001-012015Q1Dresden10 CFR 20.1701; Failure to Implement Effective Radiological Engineering ControlsA finding of very-low safety significance, and an associated NCV of 10 CFR 20.1701 was self-revealed during work activities associated with the failure to effectively implement planned radiological engineering controls during reactor head reassembly that resulted in personal contaminations and unintended radiological intakes to workers. On November 14, 2014, during the cleaning of the reactor head studs, several workers on the refuel floor were contaminated, and received unplanned and unintended intakes of radioactive material. Corrective actions included revising applicable procedures to improve the engineering and contamination controls during reactor head reassembly. The inspectors determined that that the finding was more than minor in accordance with IMC 0612, in that the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to implement effective radiological engineering and contamination controls during the cleaning of the contaminated reactor head studs resulted in personal contaminations and intakes to several workers. The inspectors concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an as low as reasonably-achievable planning issue, there was neither overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting component in the human performance in that the licensees management did not ensures that effective radiological engineering controls was either managed or coordinated commensurate to the work activities.
05000263/FIN-2015001-052015Q1MonticelloTwo Emergency Diesels Inoperable Due to Human ErrorA self-revealing finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified on December 28, 2014, due to the failure to properly implement Procedure 0187-02B, 12 Emergency Diesel Generator /12 ESW (Emergency Service Water) Monthly Pump and Valve Tests. Specifically, operations personnel failed to comply with Step 42 which directed the 12 EDG local governor control switch to be lowered to idle setting. The failure to implement the actions directed by Step 42 resulted in the 11 EDG being inoperable. Corrective actions for this issue included procedure revisions to require: protection/flagging of redundant equipment when technical specification equipment is declared inoperable for any reason, including planned maintenance and surveillance; peer checking or concurrent verification for manipulation of operable technical specification related equipment; and all equipment manipulations require a hard match (between procedure and equipment labeling). This issue was entered into the licensees corrective action program (CAP 1460675). The issue was more than minor because if left uncorrected, the failure to properly implement procedures associated with safety-related equipment would have the potential to lead to a more significant safety concern. Specifically, the failure to follow procedure resulted in the 11 EDG being made inoperable coincident with the 12 EDG being inoperable. The inspectors utilized IMC 0609 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 a failure of individuals to implement error reduction tools (H.12).
05000263/FIN-2015001-072015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.11 which requires in part, that the Primary Containment Leakage Rate Testing (LRT) Program shall be in accordance with the guidelines contained in RG 1.163, Performance-based Containment Leak-Test Program, dated September, 1995. RG 1.163 directs use of ANSI/ANS56.81994, Containment System Leakage Testing Requirements as an acceptable testing standard. ANSI/ANS56.81994 states, in part that for pressure decay testing, temperature shall be recorded at the start and end of each test, and the leakage rate shall be calculated using a specific formula which incorporates this temperature data to temperature-compensate the volume lost. Contrary to these requirements, the licensees Containment Leakage Rate Testing Program failed to include direction to take temperature data and perform temperature compensation, which resulted in a failure to perform testing in accordance with the ANSI standard and RG 1.163. Specifically, during this time, the licensee failed to correctly perform pressure decay testing for approximately 44 containment penetrations, including the Personnel Airlock. Upon discovery, engineers performed a bounding engineering analysis which verified the containment barrier remained operable but nonconforming and entered the issue into the corrective action program (CAPs 1463917 and 1465869). The performance deficiency was more than minor because the issue is associated with the barrier performance reliability attribute of the Barrier Integrity cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that the physical containment barrier protects the public from radionuclide releases. Specifically, the repeated failure to ensure containment leakage testing met technical specification and regulatory requirements was programmatic, affected multiple components, adversely affected LRT test accuracy, and consequently impacted the licensees ability to verify the containment barrier remained operable. The finding was of very low safety significance because the finding did not represent an actual open pathway in the physical integrity of the containment barrier and did not result in a loss of containment barrier operability. (Green)
05000263/FIN-2015001-082015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Criterion V which requires in part, that activities affecting quality be prescribed by procedures appropriate to the circumstances. Contrary to this requirement, between May 22, 2011 and February 5, 2014, MNGP startup instructions and procedures, C.1 Startup Procedure, 2167 Plant Startup, and 0118 Reactor Vessel Temperature Monitoring, were not appropriate to the circumstances. Specifically, during this time these procedures allowed reactor coolant system pressure to be decreased below 0 psig seven times during reactor startup activities, which was outside of the pressure parameter inputs to the analysis that is the basis for the pressure/temperature limit curves of TS 3.4.9. The licensees analysis showed that there was no impact on RPV integrity due to the existence of the partial vacuum conditions. This issue was identified by the licensee as a result of an operating experience review. The licensee entered this issue into the corrective action program (CAPs 1425020 and 1427529) and initiated action to revise the PTLR limits and submit them for NRC review. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Procedure QualityRoutine Operations Performance, and had the potential to adversely affect the associated cornerstone objective of providing reasonable assurance that a physical design barrier, the reactor coolant system, protects the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance because analysis determined that there was no change in risk to the RCS boundary due to the performance deficiency. (Green)
05000263/FIN-2015001-092015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.47(b)(14) and 10 CFR Part 50, Appendix E, Section IV.F.1. In part, Title 10 CFR 50.47(b)(14) states, Periodic exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. Additionally, Title 10 CFR Part 50, Appendix E, Section IV.F.1 states, The program to provide for: (a) The training of employees and exercising, by periodic drills, of emergency plans to ensure that employees of the licensee are familiar with their specific emergency response duties, and (b) The participation in the training and drills by other persons whose assistance may be needed in the event of a radiological emergency shall be described. The Monticello Emergency Plan, Section 8.1.2.4, describes the required demonstration periodicity for drill and exercises. Contrary to the above, on January 1, 2015, the licensee failed to perform four emergency preparedness drill objectives at the required frequency listed in the Monticello Emergency Plan, Section 8.1.2.4. Specifically, Objectives 11.01, 11.03, and 11.04 were required to be performed annually and were not performed in 2014. Additionally, Objective 11.04 was required to be performed semi-annually and was only performed once in 2014. All missed objectives were associated with radiological exposure controls. The NRC determined that the failure to comply with the established drill and exercise program was a degradation of a planning standard function in accordance with 10 CFR 50.47(b)(14) and was a very low safety significance issue (Green) as indicated in IMC 0609, Emergency Preparedness SDP, Appendix B, Attachment 2, Failure to Comply Significance Logic. The licensee entered this issue in the corrective action program (CAP 1463920). As such, the NRC determined this to be an NCV in accordance with Section 2.3.2 of the Enforcement Policy.
05000263/FIN-2014009-062014Q4MonticelloFailure to Perform a Written Safety Evaluation10 CFR 50.59(d)(1) required, in part, that the licensee maintain records of changes to the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to Paragraph (c)(2) of this section. 10 CFR 50.59(c)(2)(ii) required that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR (as updated). Contrary to the above, on or about January 30, 2012, the licensee failed to perform and maintain a written evaluation as required by 10 CFR 50.59(d)(1) to demonstrate a change to its facility did not require a license amendment. Specifically, the licensee incorrectly concluded no 10 CFR 50.59 evaluation was required prior to implementing EC 19415, Berm Construction for A.6, Acts of Nature, and Procedure A.6, Acts of Nature, Revision 41 based on screening SCR120027. Therefore, an evaluation was not performed and without an evaluation the licensee could not have ascertained if a license amendment would have been required for this change as required by Section (c)(2)(ii). Specifically, the NRC determined the external flooding mitigation plan implemented by the licensee was unable to mitigate the consequences of the design basis PMF. In accordance with Section 6.1.c.6 of the NRC Enforcement Policy, this violation was classified as a Severity Level III Violation. The licensee entered this issue into its corrective action program as CAP 01399840 and implemented appropriate corrective actions. NRC Enforcement Policy Section 3.3, Violations Identified Because of Previous Enforcement Action, states in part, The NRC may refrain from issuing an NOV or a proposed civil penalty for a Severity Level II, III, or IV Violation that is identified after the NRC has taken enforcement action, if the violation is identified by the licensee as part of the corrective action for the previous enforcement action and the violation has the same or similar root cause as the violation for which enforcement action was previously taken. In this case, the issue was identified by the licensee after the NRC had already taken enforcement action for another related violation evaluated as having substantial safety significance by the SDP. The previous violation was documented in NRC Inspection Report 05000263/2013009 on August 29, 2013. As part of the corrective action for the previous enforcement action, the licensee documented the incorrect 10 CFR 50.59 screening in its CAP and implemented appropriate corrective actions. Therefore, in accordance with the NRC Enforcement Policy, and after consultation with the Director of the Office of Enforcement and the Region III Regional Administrator, the NRC has decided to exercise enforcement discretion in accordance with Section 3.3 of the NRC Enforcement Policy and to refrain from issuing enforcement action for the violation. In accordance with the NRCs Reactor Oversight Process, this condition will not be considered in the assessment process or the NRCs Action Matrix.
05000263/FIN-2014009-052014Q4MonticelloFailure to Update the USAR for a Change to the Mitigation Strategy for the Design Basis External Flooding Event10 CFR 50.71(e) required in part, that licensees shall periodically update the final safety analysis report (FSAR), originally submitted as part of the application for the operating license, to assure that the information included in the report contains the latest information developed. This submittal shall include the effects of all the changes necessary to reflect information and analysis submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submittal of the original FSAR, or as appropriate, the last update to the FSAR under this section. Contrary to the above, from January 10, 2002 to January 31, 2014, the licensee did not update the FSAR to assure the information included in the report contained the latest information. Specifically, the licensee failed to update the FSAR with the description and basis for maintaining an external flooding mitigation plan to protect the site against external flooding events as changes to the facility and procedures were implemented, in that it failed to fully reflect and incorporate the use of a temporary levee to mitigate the consequences of a design basis PMF. In accordance with Section 6.1.c.7 of the NRC Enforcement Policy, this violation was classified as a Severity Level III Violation. The licensee entered this issue into its corrective action program as CAP 01355853. NRC Enforcement Policy Section 3.3, Violations Identified Because of Previous Enforcement Action, states in part, The NRC may refrain from issuing an NOV or a proposed civil penalty for a Severity Level II, III, or IV violation that is identified after the NRC has taken enforcement action, if the violation is identified by the licensee as part of the corrective action for the previous enforcement action and the violation has the same or similar root cause as the violation for which enforcement action was previously taken. In this case, the issue was identified by the licensee after the NRC had already taken enforcement action for another related violation evaluated as having substantial safety significance by the SDP. The previous violation was documented in NRC Inspection Report 05000263/2013009 on August 29, 2013. As part of the corrective action for the previous enforcement action, the violation was corrected by revision to the USAR on January 31, 2014. Therefore, in accordance with the NRC Enforcement Policy, and after consultation with the Director of the Office of Enforcement and the Region III Regional Administrator, the NRC has decided to exercise enforcement discretion in accordance with Section 3.3 of the NRC Enforcement Policy and to refrain from issuing enforcement action for the violation. In accordance with the NRCs Reactor Oversight Process, this condition will not be considered in the assessment process or the NRCs Action Matrix.
05000373/FIN-2014005-022014Q4LaSalleInappropriate Instructions Led to Failure of MSIVA finding of very low safety significance and 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 instructions for an activity affecting quality that were appropriate to the circumstances. Specifically, when the Unit 2 C inboard main steam isolation valve (MSIV) failed shut due to a stem-to-disc separation on August 5, 2014, inspectors reviewed the circumstances leading to the failure and determined that engineering change (EC) 340595 was deficient. This EC was created in response to 2003 industry operating experience (OE) for the same failure mechanism (loss of pretension on the shaft-to-pilot-disc) at another facility, with the purpose of establishing inspection acceptance criteria to determine if the OE applied to LaSalle. The inspectors concluded that the acceptance criteria were inappropriate to the circumstances because they contained no guidance for identifying or dispositioning the actual failure mechanism reported in the OE. Even though two of the five MSIVs inspected at the time by the licensee displayed evidence of the OE-reported failure mechanism (loss of pretension), the acceptance criteria as written were satisfied, so the MSIVs passed their inspections and future rebuild activities were deferred based primarily on these false-negative inspection results. It was due to these deferrals that the August 5th failure occurred. All MSIV internals have since been rebuilt with a more robust design that is not susceptible to a loss of pretension failure, and a root cause evaluation was performed. The performance deficiency was determined to be more than minor because it was associated with the Initiating Events Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Since the valve failure caused a reactor scram and loss of condenser as the normal heat sink due to the Group I MSIV isolation, a detailed risk evaluation was required. The RIII Senior Reactor Analysts (SRAs) performed a detailed risk evaluation using the NRCs Standardized Plant Analysis Risk model for LaSalle, version 8.24, and calculated a conditional core damage probability estimate of 8.4E-7, which represents a finding of very low safety significance, or Green. Because this performance deficiency occurred in 2003, no cross-cutting aspect was assigned because it was not considered current performance.
05000373/FIN-2014005-012014Q4LaSalleScaffold Installed Without Engineering ReviewThe inspectors identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure on September 11, 2014, to develop and supply specific minimum clearance requirements to maintenance staff prior to erecting scaffold in close proximity to safety-related equipment. The licensee has entered this item into its corrective action program (CAP). The performance deficiency was determined to be more than minor because, if left uncorrected, the performance deficiency has the potential to become a more significant safety concern. Specifically, the method used to determine the minimum clearances did not account for the potential motion of in-place systems/components. The inspectors determined the finding could be evaluated in accordance with IMC 0609, Significance Determination Process, Exhibit 2, Mitigating System Screening Questions, dated June 2, 2011. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Training, because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values.
05000263/FIN-2014009-042014Q4MonticelloSafety/Security Interface Assessment FailureThe inspectors identified a finding of very low security significance for the licensees failure to adequately assess and manage the potential for adverse effects on safety and security associated with the development and planned implementation of its external flooding mitigation plan. Specifically, 10 CFR 73.58(b)(3)(i) requires the licensee to have the capabilities to detect, assess, interdict and neutralize threats up to and including the design basis threat of radiological sabotage at all times. The failure to adequately review and evaluate the security measures and changes that would be implemented in response to a flooding event would have resulted in the requirements of 10 CFR 73.58(b)(3)(i) not being adequately maintained. This finding is not a violation of the regulatory requirements since the licensee had not actually implemented the changes that could have adversely impacted the sites security equipment, systems, and protective measures. The licensee entered the issue into its CAP to perform and document the assessments required to manage the planned changes, and to evaluate and develop potential corrective actions. The finding was of more than minor significance because it adversely affected the Security Cornerstone objective to provide high assurance that the licensee's security system uses a defense-in-depth approach and can protect against the design basis threat of radiological sabotage from external and internal threats. Specifically, the licensee failed to assess and manage changes to security equipment, systems, and protective measures that would be required in the event of the implementation of its external flooding mitigation plan to determine whether these changes could adversely impact its ability to implement the sites protective plan, which could potentially lead to a loss of defense-in-depth. The finding was of very low security significance because the total point value of this performance issue was determined to be one (1) when it was screened using the guidance provided in IMC 0609, Significance Determination Process, Appendix E, Part 1, Baseline Security Significance Determination Process (SDP) for Power Reactors, dated January 15, 2014. The inspectors determined this finding affected the cross-cutting area of human performance with a cross-cutting aspect of change management due to the licensees failure to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the licensee did not provide validation of the security plan by conducting integrated tabletops and reviews and perform additional assessment based on feedback from its external reviewers to determine whether these changes could adversely impact its ability to implement the sites protective plan.
05000263/FIN-2014009-032014Q4MonticelloFailure to Maintain Procedures to Ensure Design Requirements Would Be Met During Construction of the External Flooding Protection LeveeThe inspectors identified a finding of very low safety significance with an associated NCV of Technical Specification 5.4.1.a for the licensees failure to maintain adequate procedures to protect the plant from external flooding events. Specifically, the licensee failed to maintain Procedure 830002, External Flooding Protection Implementation to Support A.6 Acts of Nature, in that it lacked sufficient instructions to ensure testing of materials necessary to its external flooding mitigation plan were adequately controlled. The licensee entered this violation into its corrective action program (CAP) to evaluate changes to its procedures to correct the problem. The finding was of more than minor significance because it was associated with the Protection Against External Factors and Procedure Quality attributes and adversely affected the Mitigating Systems 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 instructions for constructing the flood control levee lacked specific details on how the licensee would ensure it was constructed, compacted, and tested to at least 90 percent compaction. The finding was a licensee performance deficiency of very low safety significance because it did not involve a loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic snubbers, flooding barriers, tornado doors). This determination was based on reasonable assurance the licensee could construct and compact the levee to at least 90 percent compaction. The inspectors determined this finding affected the cross-cutting area of human performance and the work management aspect due to the licensees failure to implement a process of planning, controlling, and executing work activities such that safety is the overriding priority. Specifically, the licensees process for developing and validating the work instructions for construction of the levee did not ensure appropriate quality control steps were incorporated for critical design attributes.
05000263/FIN-2014005-022014Q4MonticelloIncorrect Emergency Action Level ThresholdAn Unresolved Item (URI) was identified because additional information is needed to determine whether a performance deficiency exists and if a violation of 10 CFR 50.54(q)(2) occurred. The inspectors identified an issue of concern associated with the licensees changing of the High River Level EAL threshold from 921 to 920 for the alert classification EAL HA1.6. Description. During the first quarter of 2014, the licensee made a change to EAL HA1.6 for High River Level. Specifically, the licensee changed the threshold for the Alert classification from 921 to 920. On November 4, 2014, the NRC questioned the reason for the EAL threshold change, noting that the change may be in conflict with the EAL basis for HA1.6. These questions prompted licensee discovery that the EAL threshold basis was associated with flooding impacts on plant equipment, rather than river level historical data, as the licensee originally believed. The inspectors observed that the basis for EAL HA1.6 was linked to the river level where flood waters would reach the top of the retention basin. The inspectors also noted that although the licensee had changed the EAL threshold, the actual level of the basin was not altered. The licensee then questioned if the known level of the retention basin was a legacy error and what the correct level was for this EAL threshold. To address these questions, the licensee requested input from engineering and documented these issues in Action Request (AR) 01454593 on that same date. As an interim action, AR 01454593 documented that the current river level was 906, and if flooding were to occur, the licensee would rely on Procedure A.6, Acts of Nature, and an event response team would be formed in accordance with the procedure to monitor river level during the duration of a flood event. The licensee noted that at a river level of 918, a Notification of Unusual Event would be declared. In addition, the licensee concluded that the shift manager, event response team, and plant management would monitor for indication of degraded performance of equipment or structures necessary for safe shutdown for event classification escalation to the Alert level. The inspectors evaluated these interim compensatory measures and found them adequate as no additional reasonable risk existed as a result of this issue. On December 3, 2014, NRC questions regarding the progress of the previous AR led to the licensees statement that the 920 level also may not be correct. Because the licensee had not yet determined the appropriate High River Level EAL threshold for the alert classification EAL HA1.6, the inspectors could not readily determine whether the error was a legacy issue with the old threshold value, a current performance issue with the new threshold value and EAL change process, or both. The interim compensatory measures identified in the previous AR remained in effect at the conclusion of this inspection and the December 3, 2014 discussions and URI determination resulted in the generation of AR 01458209 by the licensee on that same date Therefore, a URI was identified because additional information on the correct High River Level EAL threshold is needed for the inspectors to determine whether a performance deficiency existed and if a violation of 10 CFR 50.54(q)(2) occurred. (URI 05000263/201400501; Incorrect Emergency Action Level Threshold)
05000263/FIN-2014009-022014Q4MonticelloFailure to Satisfy 10 CFR 50.72 and 10 CFR 50.73 Reporting Requirements for an Unanalyzed ConditionThe inspectors identified a Severity Level IV NCV of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1), Licensee Event Report System. The licensee failed to make a required 8-hour non-emergency notification call to the NRC Operations Center and also failed to submit a required LER within 60 days after discovery in November 2012 of a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety, a condition that could have prevented the fulfillment of the safety function of structures or systems needed to mitigate the consequences of an accident, and a condition prohibited by the plants TSs. The licensee subsequently made an 8-hour notification call to the NRC Operations Center via the Emergency Notification System to report the event on August 29, 2013 (Event Notice 49314) and subsequently submitted LER 05000263/201300700, Unanalyzed Condition Due to Inadequate Flooding Procedures, on October 28, 2013. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined the performance deficiency was not of more than minor significance based on No answers to the more-than-minor screening questions. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.72(a)(1)(ii) and 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding.
05000263/FIN-2014009-012014Q4MonticelloFailure to Satisfy 10 CFR 50.73 Reporting Requirements for an Unanalyzed ConditionThe inspectors identified a NCV of the NRCs reporting requirements in 10 CFR 50.73, Licensee Event Report System. The licensee failed to submit a required Licensee Event Report (LER) within 60 days after the discovery of a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety, a condition that could have prevented the fulfillment of the safety function of structures or systems needed to mitigate the consequences of an accident, and a condition prohibited by the plants Technical Specifications (TSs). Specifically, the licensee failed to either make a separate LER or further revise an existing LER with additional information to fully describe a known unanalyzed condition affecting its ability to mitigate a design basis external flooding event based on additional problems it had discovered, the corrective actions taken to correct the condition, the safety significance, and the date when full compliance was restored. The licensee initiated a corrective action to supplement an existing LER to describe the additional issues it identified that affected its external flooding mitigation plan and to specify the noncompliance window as February 29, 2012 through January 31, 2014. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined the performance deficiency was not of more than minor significance based on No answers to the more-than-minor screening questions. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding.
05000254/FIN-2014004-022014Q3Quad CitiesInadequate Evacuation Time Estimate SubmittalsThe inspectors identified a finding of very low safety significance (Green) with an associated non-cited violation of 10 CFR 50.54(q)(2) as required by 10 CFR 50.47(b)(10) and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Quad Cities Nuclear Power Station Emergency Plan, as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations by the required date. Exelon submitted the Quad Cities Nuclear Power Station ETE to the NRC on December 12, 2012, prior to the required due date of December 22, 2012. The NRC completeness review found the ETEs to be incomplete due to Exelon fleet common and site-specific deficiencies, thereby preventing Exelon from providing the ETEs to responsible offsite response organizations and from updating site-specific protective action strategies as necessary. The NRC discussed its concerns regarding the completeness of the ETE, in a teleconference with Exelon on June 10, 2013, and on September 5, 2013, Exelon resubmitted the ETEs for its sites. The NRC again found the ETEs to be incomplete. The issue is a performance deficiency because it involves a failure to comply with a regulation that was under Exelons control to identify and prevent. The finding is more than minor because it is associated with the Emergency Preparedness Cornerstone attribute of procedure quality and because it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding is of very low safety significance (Green) because it was a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The licensee had entered this issue into their corrective action program (CAP) and re-submitted a new revision of the Quad Cities Nuclear Power Station ETE to the NRC on April 30, 2014. The cause of the finding is related to crosscutting element of Human Performance, Documentation (H.7).
05000254/FIN-2014004-012014Q3Quad CitiesAngle Iron Support Installed with Minimal Clearance to Unit 2 Torus ShellA finding of very low safety significance (Green) and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to evaluate the impact of a conduit support installed in close proximity of the Unit 2 torus shell. Specifically, during installation of the conduit support, the licensee failed to provide instructions to ensure that sufficient clearance from the torus shell was provided to accommodate the torus wall movements predicted in the Updated Final Safety Analysis Report (UFSAR) torus design basis load cases. Immediate corrective actions included performing an operability evaluation under Issue Report (IR) 1672301 that determined the torus remained operable under all design basis events. The licensee has also corrected the condition by cutting the conduit support to ensure sufficient clearance to the torus wall is maintained. The performance deficiency was determined to be more than minor because the finding was associated with the design control attribute of both the Mitigating Systems and Barrier Integrity Cornerstones. The finding adversely affected the Mitigating Systems cornerstone attribute of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding also adversely affected the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. The inspectors determined the finding screened as very low safety significance (Green) because the licensees operability evaluation determined the torus remained operable under all design basis conditions. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance because it was associated with a modification that occurred in the 1980s.
05000254/FIN-2014004-032014Q3Quad CitiesLicensee-Identified ViolationThe licensee identified a violation of TS 3.3.1.1, RPS Instrumentation, and TS LCO 3.0.4. Technical Specification 3.3.1.1 specifies that four channels of turbine condenser vacuum-low scram function are required to be operable in MODE 1. Technical Specification 3.3.1.1, Condition A, stated that if one channel is not operable, the channel of the associated trip system is to be placed in trip within 12 hours. Technical Specification 3.0.4 specifies the requirements that must be satisfied prior to making a MODE change if a limiting condition for operation (LCO) is not met. Limiting Condition for Operations 3.0.4 stated, in part, that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition for an unlimited period of time. Contrary to the above, from May 6, 2014 to May 16, 2014, the licensee failed to meet the provisions of TS 3.3.1.1 and LCO 3.0.4. Specifically, on May 16, 2014, RPS pressure switch 2-0503-B was declared inoperable. The licensee determined that the cause was the inadvertent closure of the C condenser pressure indicator root valve on May 6, 2014, at approximately 6:20 p.m. while Unit 2 was in MODE 2. Unit 2 entered MODE 1 at 10:52 p.m. on May 6, 2014. Therefore, the licensee transitioned to MODE 1 without the required number of channels and did not take the required action to place the channel or associated trip system in the trip condition within 12 hours. Section 4OA2.5 above provides additional background description for this licensee-identified violation. The licensee documented the conditions prohibited by TSs for pressure switch 2-0503-B in IR 1660714. Because the inspectors answered No to all questions in Section C of IMC 0609 Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2 Mitigating Systems Screening Questions, the finding screened as very low safety significance (Green).
05000254/FIN-2014003-022014Q2Quad CitiesPost Maintenance Test Fails to Ensure Battery Charger can Perform FunctionA finding of very low safety significance and associated non-citied violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to meet the requirements of MA-AA-716-012, Post Maintenance Testing, which states, in part that post maintenance testing ensures that a component is able to perform its intended function and that the original deficiency is corrected. Specifically, licensee procedure QCEMS 0210-01 failed to include quantitative and qualitative acceptance criteria for determining that the Unit 1 250 VDC Battery Charger could perform its intended function. This issue was placed into the licensees CAP as IR 1631541. Immediate corrective actions included replacing the float potentiometer in the battery charger circuitry, replacing a thyristor in the voltage regulation circuitry, and correcting a loose solder connection identified in the battery charger circuitry. Planned corrective actions include revising procedure QCEMS 0210-01 to include acceptance criteria that ensure the battery chargers can satisfactorily perform their intended function. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors answered, No, to all of the Exhibit 2, Mitigating Systems Screening Questions, in Section A and determined the finding was of very low safety significance. This finding had a cross-cutting aspect of design margins in the area of Human Performance because the licensee did not operate and maintain the battery charger within design margins. Specifically, the licensees post maintenance testing acceptance criteria did not give them enough margin to prevent the battery from becoming inoperable (H.6).
05000265/FIN-2014003-012014Q2Quad CitiesSeismic Scaffold in Contact with Safety-Related EquipmentA finding of very low safety significance and associated non-citied violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to meet the requirements of procedure MA-AA-796-024, Scaffold Installation, Inspection, and Removal, when scaffold Q0178 was built with one of its supports in rigid contact with the operable Unit 2 torus. Immediate corrective actions included modifying the scaffold such that it was no longer in contact with the Unit 2 torus. This issue was captured in the licensees CAP as IR 1639356. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a scaffold built in contact with safety related equipment could damage the equipment and affect its availability and reliability. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors answered, No, to all of the Exhibit 2, Mitigating Systems Screening Questions, in section A and determined the finding was of very low safety significance. This finding has a cross-cutting aspect of documentation in the area of human performance because the licensee did not create and maintain complete, accurate and, up-to-date documentation. Specifically, the licensee did not completely and accurately evaluate the acceptability of a scaffold that was in contact with safety related equipment (H.7).
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