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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000261/FIN-2010009-13 | 2010Q2 | Robinson | Dedicated Shutdown Diesel Generator Failed to Start Due to Low Starting Air Pressure | The team reviewed the circumstances which resulted in a failure of the DSDG to start. The team reviewed completed procedures, log entries, system drawings and performed a system walkdown. At 18:52 on March 28, the DS bus was automatically de-energized, as designed, due to undervoltage on 4 kV Bus 3. As a result, the DSDG support equipment, such as the starting air system compressor and battery charger, lost power. Based in part on adequate starting air pressure, the licensee considered the DSDG available for the purpose of assessing on-line risk. The log reading normal minimum value for starting air pressure is 165 psig and operators were monitoring this parameter twice per day. At 14:41 on March 31 the licensee attempted to start the DSDG and re-energize the DS bus to maintain adequate DSDG support parameters such as starting air pressure and battery voltage. Starting air pressure had decreased to 100 psig and the DSDG did not start. The licensee successfully started the DSDG on April 1 at 13:40 by pressurizing the DSDG starting air receiver tank using high pressure air bottles. Both the E-1 and E-2 safety buses were energized during this time with Bus E-1 powered from off-site power and Bus E-2 supplied from EDG B. The licensee entered Condition Reports 390954 and 390958 into their corrective action program. Additional review by the NRC will be required to determine if the DSDG was available when credited in the licensees risk assessment during the plant cooldown to Mode 4. This review will also determine whether this issue represents a performance deficiency. An Unresolved Item will be opened pending completion of this review. The issue will be identified as URI 05000261/2010009-13, Dedicated Shutdown Diesel Generator Failed to Start Due to Low Starting Air Pressure. |
| 05000261/FIN-2010009-14 | 2010Q2 | Robinson | Unexpected Loss of Instrument Bus 3 for Two Minutes | The team reviewed the circumstances which resulted in an inadvertent de-energization of Instrument Bus 3. The team reviewed completed procedures, log entries, and system drawings. The team also interviewed personnel and performed a system walkdown. At 18:52 on March 28 the B battery charger de-energized due to loss of power to Bus E-2. Per Path-1, control room operators subsequently dispatched an Auxiliary Operator (AO) to restore the B battery charger. As the AO entered the battery room he made inadvertent contact with the handle for the B Inverter Supply Breaker 72/MCC-B (1K). The contact resulted in breaking the handle off of the breaker. Based on the timeframe when the AO entered the battery room and the time when Instrument Bus 3 was unexpectedly loss, the licensees ERT concluded the contact with the breaker caused the loss of Instrument Bus 3. The Auxiliary Operator recognized the damage to the breaker handle and continued to complete the restoration the B battery charger. The B battery charger was restored at 19:31. Upon exiting the battery room the AO verified the B inverter was operating correctly and reported the damage to the breaker handle. A review of plant data indicated Instrument Bus 3 was de-energized at 19:25 and reenergized at 19:27. The loss of Instrument Bus 3 power deenergized the High Steam Flow bistables in the Engineered Safety Features system. This condition, coincident with an RCS Low Tavg signal due to the RCS cooldown, generated a Main Steam Line Isolation signal, automatically closing all MSIVs and terminating the RCS cooldown. Based on interviews with the AO, no actions were performed to reset or reclose the B Inverter Supply Breaker. The licensee generated Work Order 01735191 to repair the broken breaker handle. The licensee performed troubleshooting activities to determine the cause of the two-minute interruption in instrument bus power, but was unable to detect any problems. The licensee was continuing to perform troubleshooting at the time this report was written. The licensee entered Condition Report 390070 into their corrective action program. Additional review by the NRC will be needed to assess the adequacy of the licensee troubleshooting efforts and evaluate any problems that may be identified. This review will also determine whether any performance deficiencies exist. An Unresolved Item will be opened pending completion of this review. The issue will be identified as URI 05000261/2010009-14, Unexpected Loss of Instrument Bus 3 for Two Minutes. |
| 05000263/FIN-2011004-01 | 2011Q3 | Monticello | NOED for Emergency Diesel Generator Load Rejection Surveillance Requirement 3.8.1.7 | On September 27, 2011, during an engineering self-assessment, the licensee identified a potential issue associated with the testing methodology used to demonstrate each EDGs capability to withstand the rejection of an electrical load that is equivalent to the single largest post-accident electrical load. On September 29, 2011, the licensee verified that their existing surveillance test OSP-ECC-0566, Low Pressure ECCS (emergency core cooling system ) Automatic Initiation and Loss of Auxiliary power Test, Revision 8, did not ensure that the load rejection test was performed with sufficient load to satisfy the requirements of SR 3.8.1.7 (Verify each EDG rejects a load greater than or equal to its associated single largest post-accident load and, following load rejection, the frequency is less than or equal to 67.5 Hz.). On September 29, 2011, at approximately 1700, the licensee declared both 11 and 12 EDGs inoperable and entered the Action for TS 3.8.1.E, Two EDGs Inoperable. At approximately 2200, the licensee requested enforcement discretion to extend the Action Completion Time for TS 3.8.1.F, from twelve hours to five days, to allow time to perform the required EDG load rejection testing. At approximately 23:58, the Agency granted NOED 11-3-001. The inspectors evaluation of the issue included a review of the technical documents associated with the issue and several meetings with the licensee management and technical staff. The initial information gained by the inspectors and their assessment of the issue was communicated to senior agency managers well in advance of the licensees NOED request, significantly contributing to the Agencys understanding and appropriate disposition of the issue. Additional information associated with the inadequate surveillance procedure and EDG operability is documented in Section 1R15 of this report. |
| 05000263/FIN-2011004-03 | 2011Q3 | Monticello | Shipping and Transportation of a Radioactively Contaminated Condensate Demineralizer Vessel | On July 14, 2011, it was reported to the licensee by the driver of the vehicle that there was a puncture in the side of a container package on radioactive material shipment number 11-127. The package was a Sealand box inside an enclosed conveyance. The Sealand box contained a radioactively contaminated condensate demineralizer vessel and the puncture was a nominal 4 by 6 inch hole. There was no spread of contamination as a result of the compromised package. The inspectors initial review determined that a performance deficiency exists, in that, the shipping container contents was inappropriately braced and blocked for transport. Regulations require that licensees ensure that loads not shift under conditions normally incident to transportation. The inspectors will review the additional information provided by the licensee and determine the significance of the performance deficiency. |
| 05000263/FIN-2011005-01 | 2011Q4 | Monticello | E Condensate Demineralizer Alarm Response Procedure Limits Exceeded | The inspectors identified a finding of very low safety significance and non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures, when the operators did not take conservative action to address a high differential pressure condition on an inservice condensate demineralizer vessel. Specifically, operators allowed the E condensate demineralizer to exceed differential pressure operating limits prescribed in Alarm Response Procedure 80-DPAH-2215, Vessel T-7E D/P High, and remain above those prescribed limits for approximately a shift before taking action to correct the abnormal condition. Specific corrective actions taken by the licensee to address this issue included updating the applicable alarm response procedures and operating procedures to reflect current system limitations; engineering management reinforcing the expectation that informal processes are not acceptable when communicating technical guidance to operations staff; and site management reinforcing the expectation that, once a degrading trend is recognized, actions must be taken in sufficient time to prevent crossing established operating limits. The inspectors determined that the licensees failure to maintain the E condensate demineralizer differential pressure within prescribed operational limits was a performance deficiency because it was the result of the failure to meet a requirement or a standard; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because it impacted the Human Performance attribute of the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 1 of the Table 4a worksheet to screen the finding. For transient initiators, the inspectors answered no to the question, Does the finding contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment of functions will not be available, and determined the finding to be 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, having Work Control components, and involving aspects associated with the licensee planning and coordinating work activities, consistent with nuclear safety, specifically the need for planned contingencies, compensatory actions, and abort criteria |
| 05000263/FIN-2011005-02 | 2011Q4 | Monticello | Inadequate Completion of CAPRs Associated with 2RS to 2R Feeder Cable Testing | A finding of very low safety significance and NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was self-revealed following a reactor scram, which was the direct result of an electric plant realignment caused by a faulted feeder cable and lockout of the stations 2R transformer. Specifically, annual testing to monitor the performance of the 2R feeder cables, which was put in place as a corrective action to prevent recurrence to address issues identified subsequent to a similar event in 2008, had not been performed since the cables were placed back in service following that event. To address the identified material deficiencies, the licensee replaced and tested the electrical cables between 2RS and 2R in their entirety, employing a new route designed to avoid cable submergence. Additional corrective actions were put in place to strengthen the licensees planned maintenance deferral process and their cable condition monitoring program. The inspectors determined that the licensees failure to perform annual testing of the 2R transformer feeder cables, as required by the stations planned maintenance program, was a performance deficiency because it was the result of the failure to meet a requirement or a standard, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. The inspectors determined that the issue was more than minor because it impacted the Configuration Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors utilized Column 1 of the Table 4a worksheet to screen the finding. Because the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions would not be available, the Region III Senior Reactor Analyst (SRA) performed a Phase 3 analysis, and screened the finding to be 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, having decision-making components, and involving aspects associated with the licensees making safety-significant or risk-significant decisions using a systematic process to ensure safety is maintained |
| 05000263/FIN-2011005-03 | 2011Q4 | Monticello | Rod Worth Minimizer Inoperable During Reactor Plant Startup | A finding of very low safety significance and NCV of TS 3.3.2.1, Control Rod Block Instrumentation, was self-revealed to the operating crew, when normal startup testing could not be accomplished due to improperly configured equipment. Specifically, the operating crew transitioned from Mode 4 to Mode 2, with the rod worth minimizer (RWM) mode switch in the BYPASS position. With the RWM mode switch in the BYPASS position and the required actions of 3.3.2.1(c) not met, the requirements of TS 3.3.2.1, that the RWM be operable in Mode 1 and Mode 2 when thermal power is less than or equal to 10 percent rated thermal power, could not be met. Actions taken by the licensee in response to this event included declaring the event a reactivity management event; making an NRC notification under 50.72(b)(3)(v)(D); resetting their site event clock; providing additional training for the applicable operating crew; and revising procedures associated with this event to clarify the sequencing of key activities associated with the transition between Mode 4 and Mode 2. The inspectors determined that the licensees failure to properly control the configuration of the RWM prior to entering an operating mode that required its operability was a performance deficiency, because it was the result of the failure to meet a requirement or a standard; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors determined that the issue was more than minor because it impacted the Configuration Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors applied IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, to this finding. The inspectors answered No to the questions associated with transient initiators and screened the finding to be 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, having work practices components, and involving aspects associated with personnel work practices that support human performance, specifically in the areas of pre-job briefing, self and peer checking, and proper documentation of activities |
| 05000263/FIN-2011005-04 | 2011Q4 | Monticello | Failure to Properly Block and Brace a Radioactive Package for Transport | The inspectors reviewed a self-revealed finding of very low safety significance and an associated NCV of 10 CFR 71.5. Specifically, the licensee failed to appropriately block and brace a radioactively contaminated condensate demineralizer vessel within a transport package, such that, the package contents would not compromise and penetrate the transport package. The issue has been entered into the licensees corrective action program as CR (condition report) 01294652. Corrective actions were implemented to address supervisions responsibilities during shipment preparation regarding appropriate blocking and bracing of package contents. The finding was more than minor because the performance deficiency could be reasonably viewed as a precursor to a significant event, in that, the penetration of the transportation package by its contents could lead to the inadvertent spread of radioactive contamination in the public domain. Using IMC 0609, Attachment D, for the Public Radiation Safety SDP, the inspectors determined the finding to be of very low safety significance. The inspectors also determined that this finding had a cross-cutting aspect in the area of problem identification and resolution (operating experience) |
| 05000263/FIN-2011005-05 | 2011Q4 | Monticello | Failure to Make a Required 60 Day Event Report Per 10 CFR 50.73(a)(2)(vii)(A-D) | The inspectors identified a Severity Level IV NCV and associated finding of very low safety significance of 10 CFR 50.73(a)(2)(vii)(A-D), Licensee Event Report System, for the failure to report an event to the NRC within 60 days, where a single cause or condition caused two independent trains to become inoperable in a single system designed to help maintain safe reactor shut down, remove residual heat, control radioactive releases, or mitigate accidents. Specifically, on September 29, 2011, the licensee identified that the surveillance test procedures being used to demonstrate load reject capabilities of both EDGs had never contained the correct load rejection testing requirements from the applicable design documents. As a result, the surveillances were considered never met, and both EDGs were declared inoperable. During their evaluation and subsequent reporting of the issue, the licensee failed to recognize that the inoperability of both diesel generators caused by a single common cause was reportable to the NRC within 60 days under the 50.73 common cause criterion. The licensee entered this issue into their corrective action program (CAP 1318116). Corrective actions for this issue included plans to revise their existing licensee event report (LER) and to perform an apparent cause evaluation to further evaluate the issue. The inspectors determined that the failure to report required plant events or conditions to the NRC in accordance with reporting requirements was a performance deficiency because it was the result of the failure to meet a requirement or a standard, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. In addition, it had the potential to impede or impact the regulatory process. As a result, the NRC dispositions violations of 10 CFR 50.73 using the traditional enforcement process instead of the SDP. However, if possible, the underlying technical issue is evaluated using the SDP. In this case, the inspectors determined that the licensee failed to develop and implement adequate Emergency Diesel Generator (EDG) testing procedures during their transition to the Improved Technical Specifications in 2006, which resulted in both EDGs being declared TS inoperable, but available for use. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attributes of Human Performance and Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding had very low safety significance because they answered No to all five questions contained in Column 2 of the Table 4a worksheet. As a result, the inspectors determined that the finding had very low safety significance (Green). In accordance with Section 6.9.d.9 and 6.9.d.10 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was an example where the licensee failed to make a report required by 10 CFR 50.73; it represented a failure to identify all applicable reporting codes on an LER that may impact the completeness or accuracy of other information submitted to the NRC; and the underlying technical issue was evaluated by the SDP and determined to be of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency affected the cross-cutting area of Problem Identification and Resolution, having corrective action program components, and involving aspects associated with properly classifying and evaluating for reportability conditions adverse to quality (P.1(c)). |
| 05000263/FIN-2011005-06 | 2011Q4 | Monticello | Failure to Make a Required 60 Day Event Report Per 10 CFR 50.73(a)(2)(vii)(A-D) | The inspectors identified a Severity Level IV NCV and associated finding of very low safety significance of 10 CFR 50.73(a)(2)(vii)(A-D), Licensee Event Report System, for the failure to report an event to the NRC within 60 days, where a single cause or condition caused two independent trains to become inoperable in a single system designed to help maintain safe reactor shut down, remove residual heat, control radioactive releases, or mitigate accidents. Specifically, on September 29, 2011, the licensee identified that the surveillance test procedures being used to demonstrate load reject capabilities of both EDGs had never contained the correct load rejection testing requirements from the applicable design documents. As a result, the surveillances were considered never met, and both EDGs were declared inoperable. During their evaluation and subsequent reporting of the issue, the licensee failed to recognize that the inoperability of both diesel generators caused by a single common cause was reportable to the NRC within 60 days under the 50.73 common cause criterion. The licensee entered this issue into their corrective action program (CAP 1318116). Corrective actions for this issue included plans to revise their existing licensee event report (LER) and to perform an apparent cause evaluation to further evaluate the issue. The inspectors determined that the failure to report required plant events or conditions to the NRC in accordance with reporting requirements was a performance deficiency because it was the result of the failure to meet a requirement or a standard, the cause was reasonably within the licensees ability to foresee and correct, and should have been prevented. In addition, it had the potential to impede or impact the regulatory process. As a result, the NRC dispositions violations of 10 CFR 50.73 using the traditional enforcement process instead of the SDP. However, if possible, the underlying technical issue is evaluated using the SDP. In this case, the inspectors determined that the licensee failed to develop and implement adequate Emergency Diesel Generator (EDG) testing procedures during their transition to the Improved Technical Specifications in 2006, which resulted in both EDGs being declared TS inoperable, but available for use. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems Cornerstone attributes of Human Performance and Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding had very low safety significance because they answered No to all five questions contained in Column 2 of the Table 4a worksheet. As a result, the inspectors determined that the finding had very low safety significance (Green). In accordance with Section 6.9.d.9 and 6.9.d.10 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was an example where the licensee failed to make a report required by 10 CFR 50.73; it represented a failure to identify all applicable reporting codes on an LER that may impact the completeness or accuracy of other information submitted to the NRC; and the underlying technical issue was evaluated by the SDP and determined to be of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency affected the cross-cutting area of Problem Identification and Resolution, having corrective action program components, and involving aspects associated with properly classifying and evaluating for reportability conditions adverse to quality (P.1(c)). |
| 05000263/FIN-2011005-07 | 2011Q4 | Monticello | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Criterion XI requires, in part, that A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to this requirement, on September 29, 2011, the licensee identified that they had failed to utilize a test program which incorporated all requirements from the applicable design documents to demonstrate that both EDGs would perform satisfactorily in service. Specifically, the test procedures being used by the licensee to demonstrate operability of the EDGs did not contain the correct load rejection testing requirements from the applicable design documents. As a result, the licensee determined that they had never demonstrated that they met load rejection surveillance requirement 3.8.1.7, and that both EDGs were inoperable. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding had very low safety significance because they answered No to all five questions contained in Column 2 of the Table 4a worksheet. The licensee developed new test procedures which included the appropriate acceptance criteria and test methodologies, satisfactorily tested both EDGs, and entered this issue into their CAP as AR 01305683, CDBI FSA-Question on Definition of Post Accident Load in TS and AR 1306107, largest post-accident load greater than in test OSP-ECC-0566. |
| 05000263/FIN-2011005-08 | 2011Q4 | Monticello | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated Severity Level IV NCV of 10 CFR 50.72(b)(3)(v)(D). Title 10 CFR 50.72(b)(3)(v)(D) requires, in part, that operating reactor licensees shall notify the NRC within eight hours of the occurrence of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. Contrary to this requirement, on November 1, 2011, the licensee identified that they had failed to make a required non-emergency notification within eight hours for a safety system functional failure of the CREF and CRV systems. Specifically, on October 21, 2011, following a reactor scram, the No. 11 EDG ESW pump was declared inoperable due to low cooling water pump flow. The loss of this pump resulted in the No. 11 EDG, \'A\' CREF, and \'A\' CRV being declared inoperable when the redundant B Division CREF and CRV systems were already out-of-service due to preplanned maintenance. As a result, the licensee entered TS 3.0.3 due to both CRV and CREF systems being inoperable. The licensee failed to recognize that this represented a potential loss of safety function at the time of the event. The inspectors determined that the failure to report required plant events or conditions to the NRC was a performance deficiency, and it had the potential to impede or impact the regulatory process. The NRC dispositions violations of 10 CFR 50.72 using the traditional enforcement process, and if possible, the underlying technical issue is evaluated using the SDP. The underlying technical issue was associated with both trains of CREF/CRV being inoperable and unavailable during a scram, resulting from a lockout of the 2R transformer. This issue was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, because both trains of the CREF/CRV system were inoperable and unavailable, the regional SRAs performed a Phase 3 risk evaluation to determine the risk significance of the issue. As a result of the SRAs evaluation, the inspectors determined that the finding had very low safety significance. Because the failure to make the required 50.72 report had the potential to impede or impact the regulatory process, the inspectors used the Traditional Enforcement process to disposition the issue. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV. The licensee entered the issue into their CAP as AR 1310956, Missed 8 Hour Report, and made the required 50.72 report. |
| 05000263/FIN-2014002-01 | 2014Q1 | Monticello | RCS Pressure Boundary Leakage Operation Prohibited by Technical Specifications | A finding of very low safety significance and a non-cited violation of Technical Specification (TS) 3.4.4, RCS Operational Leakage, was self revealed when the licensee failed to comply with TS 3.4.4, Condition C, which required the plant to be in MODE 3 within 12 hours if pressure boundary leakage exists. Specifically, the licensee operated with reactor coolant system (RCS) pressure boundary leakage as a result of corrosion in the 12 recirculation pump upper seal cooler between August 9, 2013, and January 17, 2014, which is a condition prohibited by TS. The site initiated a troubleshooting team, and following confirmation of the location of the leakage, the plant was shut down in accordance with TSs. The site performed an apparent cause evaluation; implemented a modification to remove the affected seal cooler from service; and developed a periodic replacement plan for heat exchangers in a similar configuration. The inspectors determined that the licensees operation with RCS pressure boundary leakage, a condition prohibited by TSs, due to recirculation pump seal cooler leakage, 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 attribute of equipment performancebarrier integrity, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined this finding was of very low safety significance. The inspectors concluded that this finding was cross-cutting in the Problem Identification and Resolution, Evaluation area, because of the failure to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. |
| 05000263/FIN-2014002-02 | 2014Q1 | Monticello | Failure to Follow Procedure for RCS Operability Determination | The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to ensure that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. Specifically, the licensee failed to accomplish activities affecting quality in accordance with Fleet Procedure FPOPOL01, in that, on August 9, 2013, and January 3, 4, 7, and 17, 2014, the site failed to ensure that the operability determination for leakage into reactor building closed-cooling water (RBCCW) was sufficient to address the capability of a structure, system, and component (SSC) to perform its specified safety function and, as a result, the site failed to properly classify leakage from the recirculation system as reactor coolant system (RCS) pressure boundary leakage. Following NRC questions and actions by the site to confirm the location of the leakage, the site revised the operability determination and classified the leakage as reactor coolant pressure boundary (RCPB) leakage. This issue was entered into their corrective action program; a root cause evaluation was performed; and additional corrective actions were in development at the time of this report. The inspectors determined that the failure to properly classify RCS pressure boundary leakage in accordance with the fleet operability determination process was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor because, if left uncorrected, the failure to perform a thorough operability evaluation for conditions where potential RCPB leakage exists could lead to a more significant safety concern. The inspectors assessed the significance of this finding in accordance with IMC 0609 under the Initiating Events Cornerstone, and determined that it was of very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Conservative Bias area, because of the licensees failure to use decision-making practices that emphasize prudent choices over those that are simply allowable, and a failure to ensure that proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. |
| 05000263/FIN-2014002-03 | 2014Q1 | Monticello | Drywell-Torus Vacuum Breaker Inadequate Post-maintenance and Return-to-service Test | The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure that all testing required to demonstrate that SSCs will perform satisfactorily in service are identified and performed in accordance with written test procedures, which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, on May 22, 2013, the licensee failed to ensure that post-maintenance and return-to-service testing was performed on all eight safety-related drywell-torus vacuum breakers after refueling outage maintenance, to ensure that surveillance requirements for the valves opening setpoints were met prior to the valve being returned to service and prior to entry into MODE 2. The licensee entered this issue into their CAP, and additional corrective actions were in development at the time of this report. The inspectors determined that the licensees failure to perform required PMTs for vacuum breakers prior to their return-to-service and making a mode change 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 Barrier Integrity Cornerstone attribute of SSC and Barrier Performance, and affected the cornerstone objective to provide reasonable assurance that physical design barriers, including containment, protect the public from radionuclide releases caused by accidents or events. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined this finding was of very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Work Management area, because of the failure to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, and to ensure that the work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. |
| 05000263/FIN-2014002-04 | 2014Q1 | Monticello | Failure to Follow Procedure for Drywell-Torus Vacuum Breaker Operability Determination | The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, on February 14, 2014, for the licensees failure to ensure that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances. Specifically, the site changed Procedure 0143, Drywell-Torus Monthly Vacuum Breaker Check, to include allowances for multiple cyclings on the safety-related drywell-torus vacuum breaker valves to ensure they met their surveillance requirements to close, which constituted unacceptable preconditioning. The licensee entered this issue into their CAP, and corrective actions were still in development at this time of this report. The inspectors determined that the licensees failure to ensure the vacuum breaker monthly testing surveillance procedure was appropriate to the circumstances was a performance deficiency requiring evaluation. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because it adversely impacted the Barrier Integrity Cornerstone attribute of Procedure Quality, and affected the cornerstone objective to provide reasonable assurance that physical design barriers, including containment, protect the public from radionuclide releases caused by accidents or events. In addition, if left uncorrected, the proceduralized unacceptable preconditioning has the potential to lead to a more significant safety concern. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined this finding was of very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Conservative Bias area, because of the licensees failure to use decision-making practices that emphasize prudent choices over those that are simply allowable, and a failure to ensure that proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. |
| 05000263/FIN-2014002-05 | 2014Q1 | Monticello | Inadequate Drywell-Torus Monthly Vacuum Breaker Test Procedure due to Proceduralized Unacceptable Preconditioning | The inspectors identified a finding of very low safety significance and a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, on February 14, 2014, for the licensees failure to ensure that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances. Specifically, the site changed Procedure 0143, Drywell-Torus Monthly Vacuum Breaker Check, to include allowances for multiple cyclings on the safety-related drywell-torus vacuum breaker valves to ensure they met their surveillance requirements to close, which constituted unacceptable preconditioning. The licensee entered this issue into their CAP, and corrective actions were still in development at this time of this report. The inspectors determined that the licensees failure to ensure the vacuum breaker monthly testing surveillance procedure was appropriate to the circumstances was a performance deficiency requiring evaluation. The inspectors screened the performance deficiency per IMC 0612, Power Reactor Inspection Reports, Appendix B, and determined that the issue was more than minor because it adversely impacted the Barrier Integrity Cornerstone attribute of Procedure Quality, and affected the cornerstone objective to provide reasonable assurance that physical design barriers, including containment, protect the public from radionuclide releases caused by accidents or events. In addition, if left uncorrected, the proceduralized unacceptable preconditioning has the potential to lead to a more significant safety concern. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined this finding was of very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Conservative Bias area, because of the licensees failure to use decision-making practices that emphasize prudent choices over those that are simply allowable, and a failure to ensure that proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. |
| 05000263/FIN-2014002-06 | 2014Q1 | Monticello | Uncontrolled High Radiation Area Following Shut-down Cooling Re-Alignment | A finding of very low safety significance and an associated non-cited violation of Technical Specification (TS) 5.7.1 was self-revealed following a workers unexpected electronic dosimeter alarm, which resulted in the identification of an unbarricaded and unposted high radiation area. The inspectors determined a performance deficiency occurred when the licensee failed to perform radiological surveys following the implementation of noble metals chemistry which changed plant radiological conditions, and prior to authorizing entry into the 924 torus area. Specifically, on January 19, 2014, a fire watch entered this area when posted as a radiation area and received a dose rate alarm. Follow-up radiological surveys identified a high radiation area of 120 mrem/hr at 30 cm from the residual heat removal piping. This issue was entered into the licensees corrective action program as CAP 01415285. The licensee immediately barricaded and posted the area as a high radiation area. Additionally, the licensee is performing a review of radiation protection fundamentals as the result of this event. The finding was more than minor because it 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, in that, the workers entry into an unsurveyed high radiation area placed the worker at increased risk for unnecessary radiation exposure. Additionally, the inspectors reviewed the guidance in IMC 0612, Appendix E, Examples of Minor Issues, and identified Example 6(h) as similar to the performance deficiency. The finding was assessed using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and was determined to be of very low safety significance because the problem was not an as-low-as-reasonably-achievable planning issue; there were no overexposures nor substantial potential for overexposures given the highest dose rate present in the room and the scope of work; and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of this event involved a cross-cutting component in the Problem Identification and Resolution, Operating Experience area, because the licensee failed to implement known industry concerns regarding changing radiological conditions as the result of implementation of noble metals chemistry. |
| 05000263/FIN-2014002-07 | 2014Q1 | Monticello | Both Secondary Containment Access Doors Briefly Opened Simultaneously | On September 18, 2013, while performing the secondary containment airlock door interlock surveillance test, the interlock to the main plenum room did not prevent the opening of both doors to the plenum room airlock (DOOR85 and DOOR86). With the outer door to the main plenum room open, the inner door was able to be opened. The plenum airlock doors were then closed. The operator attempted a second time to verify interlock functionality. This time the inner door was opened, and again the interlock did not prevent the opening of the outer door. The plenum airlock doors were immediately closed. The total time both doors were opened was estimated to be less than 10 seconds. With both doors open, TS SR 3.6.4.1.3 was not met and secondary containment was declared inoperable. Secondary containment was declared operable, after independently verifying that at least one secondary containment access door was closed. Inspectors reviewed the LER and decided that additional information was needed to determine whether a performance deficiency exists for the event. In order to close this Unresolved Item (URI), the inspectors intend to review the sites recently performed evaluation aimed at removing this issue from being counted in the Safety System Functional Failure PI. In addition, the inspectors will factor in any insights from NRRs more generic resolution to industry wide secondary containment issues. |
| 05000263/FIN-2014002-08 | 2014Q1 | Monticello | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions. Criterion XVI requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to this requirement, on February 11, 2014, the licensee identified that they had failed to identify and correct a condition adverse to quality where a single failure could result in the EDGs picking up load on the essential busses in a time frame longer than what is required by TS SR 3.8.1.12. Surveillance Requirement 3.8.1.12 states that when required, the EDGs auto-start and energize permanently connected loads in approximately 10 seconds. Specifically, after the NRC identified an NCV on May 8, 2012, where required time delay limits would be exceeded in the sites EDG/1AR degraded voltage transfer logic, the licensees extent of condition failed to identify and correct a deficiency where relays in the EDG/1AR loss of voltage transfer logic could result in the EDGs energizing the connected loads in a slightly longer time period than allowed (< 11 seconds). The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attributes of Design Control and Equipment Performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, the inspectors determined that the finding represented a loss of system and/or function as defined for the EDGs in the TS bases; however, a detailed risk evaluation determined that there was no change in core damage frequency because exceeding the time delay would not impact the Probabilistic Risk Assessment function to respond to a loss of offsite power event. As a result, the inspectors concluded that the finding had very low safety significance. The licensee entered this issue into their CAP, and declared both EDGs inoperable until action was taken to remove 1AR from service, and relays with acceptable time delays could be installed. |
| 05000263/FIN-2014002-09 | 2014Q1 | Monticello | Licensee-Identified Violation | Technical Specification 5.7.1, High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, requires, in part, that such areas shall be barricaded and conspicuously posted as a high radiation area. Contrary to the above, on October 11, 2013, a drain hose was moved from inside the RWCU pump room to outside the room. This hose was the source of radiation which resulted in an unbarricaded and unposted high radiation area outside the pump room. This was identified by radiation protection technicians performing radiological surveys in the area. The licensee documented this issue in CAP 01401180. The finding was determined to be of very low safety significance (Green) because it was not an ALARA planning issue, there was no overexposure nor potential for overexposure, and the licensees ability to assess dose was not compromised. |
| 05000263/FIN-2014003-02 | 2014Q2 | Monticello | Operation Outside of Reactor Coolant System Pressure and Temperature Limits | An URI associated with TS 3.4.9, Reactor Coolant System Pressure and Temperature Limits, was identified. Technical Specification 3.4.9 requires, in part, that RCS pressure and RCS temperature shall be maintained within the limits specified in the PTLR, which requires that RCS pressure remain at or above 0 psig. Between May 22, 2011, and February 5, 2014, Monticello RCS pressure was decreased below 0 psig several times during reactor startup activities. During an operating experience review in April 2014, the licensee noted that a vacuum of approximately3 psig had been drawn on the RPV six times, and in one case a vacuum of17.5 psig was drawn, which was outside of the limits specified in the PTLR. The licensees analysis showed that there was no impact on RPV integrity due to the existence of the partial vacuum conditions. The licensee entered this issue into the CAP and initiated action to revise the PTLR limits and submit them for NRC review. Inspectors reviewed the results of the licensees operating review and decided that additional information was needed, including insights from NRRs more generic resolution to industrywide issues regarding TS PTLR limits, to determine whether a performance deficiency exists. |
| 05000263/FIN-2014004-01 | 2014Q3 | Monticello | Failure to Follow Reactivity Management Procedure | A finding of very low safety significance and a NCV of Technical Specification (TS) 5.4.1, Procedures, was self-revealed when the licensee failed to implement requirements specified in FPOPRM01, Reactivity Management Program. Specifically, the licensee failed to ensure that the licensed operators were aware of the consequences of the reactivity changes they were making, as required by FPOPRM01. As a result, the licensed operators were unaware that their actions to increase recirculation flow would result in the plant exceeding the minimum critical power ratio (MCPR) operating limit. This issue was entered into the licensees corrective action program (CAP) 1446848. Immediate corrective actions included restoration of the plant to within the MCPR operating limit, halting of power changes, disqualification of individuals directly involved, increased management oversight, a detailed review of the reactivity plan and procedures planned for use during the reactivity plan, and site-wide communication of the event. The site initiated a root cause evaluation (RCE), which was in progress at the end of the inspection period. The inspectors determined that the failure to perform reactivity manipulations in accordance with reactivity management 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 Barrier Integrity Cornerstone attributes of Configuration Control and Procedure Quality, and affected the cornerstone objective to provide reasonable assurance that physical design barriers, including fuel cladding, protect the public from radionuclide releases caused by accidents or events. The inspectors assessed the significance of this finding in accordance with IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria and determined this finding was of very low safety significance. The inspectors concluded that this finding was cross-cutting in the Human Performance, Documentation aspect because of the failure to ensure that the procedures being used to make the reactivity manipulations were complete, accurate, and up-to-date. |
| 05000263/FIN-2014004-02 | 2014Q3 | Monticello | Licensee-Identified Violation | Title 10 of the Code of Federal Regulations 50.54(q)(2) requires, in part, that a holder of a license under this part shall follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E, and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50, Appendix E, Section IV.A.9 states, By December 24, 2012, for nuclear power reactor licensees, a detailed analysis demonstrating that on-shift personnel assigned emergency plan implementation functions are not assigned responsibilities that would prevent the timely performance of their assigned functions as specified in the emergency plan, shall be included. Contrary to the above, on December 24, 2012, the licensees detailed analysis of on-shift staffing was deficient in that all assigned functions for on-site personnel were not evaluated. Specifically, the augmentation tasks identified in the licensees emergency plan assigned to on-shift personnel were not considered when performing A.2002, Monticello On-Shift Staffing Analysis, for the Core/Thermal Hydraulics and Radiation Waste Operator positions. The NRC determined that with no identified loss or degradation of a planning standard function, the failure to complete the detailed analysis in accordance with 10 CFR Part 50, Appendix E, Section IV.A.9 was a very low safety significance issue (Green) as indicated in IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Revision February 24, 2012. This issue was identified in a self-assessment process on May 13, 2014, and documented in corrective action entries as action requests 01430607 and 0101437840. Immediate corrective actions included interim augmentation for both on-shift positions fully analyzing and updating the on-shift staffing analysis. As such, the NRC determined this to be an NCV in accordance with Section 2.3.2 of the Enforcement Policy. |
| 05000263/FIN-2014005-01 | 2014Q4 | Monticello | Failure to Comply with ASME Code and Maintain Configuration Approved by IST Relief Request | The inspectors identified a finding of very low safety significance and NCV of 10 CFR 50.55a(f)(4) for the licensees failure to test main steam line drain containment isolation valves MO2373 and MO2374 in accordance with the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) code requirements or maintain the valves in the alternative configuration specified in an NRC-approved Relief Request (VRR05). Specifically, on October 17, 2014, the NRC identified that the licensee had failed to maintain the approved alternative configuration which had been accepted by the NRC in lieu of the required quarterly stroke testing of MO2373 and MO2374. Corrective actions for this event included immediate restoration of the NRC-approved configuration specified in the relief request, cancellation of the noncompliant procedure temporary revisions, and cancellation of the associated 10 CFR50.59 screening. The licensee also initiated an apparent cause evaluation which was in progress at the end of this inspection period. The inspectors determined that the failure to test MO2373 and MO2374 in accordance with the ASME OM code or maintain the relief request approved plant configuration was a performance deficiency. The inspectors evaluated the issue and determined that the finding was more than minor in accordance with IMC 0612, Appendix B, because it adversely impacted the Barrier Integrity Cornerstone attributes of Design Control and Configuration Control, and affected the cornerstone objective to provide reasonable assurance that physical design barriers, including containment, protect the public from radionuclide releases caused by accidents or events. The inspectors assessed the significance of this finding in accordance with IMC 0609, and determined that this finding was of very low safety significance because it did not represent an actual open pathway in the physical integrity of reactor containment, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors concluded that this finding was cross-cutting in the Human Performance Decision making aspect because of the failure to use a consistent, systematic approach to make decisions and a failure to ensure that risk insights are incorporated as appropriate. (H.13) |
| 05000263/FIN-2014005-02 | 2014Q4 | Monticello | Incorrect Emergency Action Level Threshold | An 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-2015001-01 | 2015Q1 | Monticello | Failure to Identify High Pressure Coolant Injection (HPCI) Seismic Support Nonconformance | The 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-2015001-02 | 2015Q1 | Monticello | Failure to Maintain Fire Protection Program Procedures for Control of Portable Heater/Extension Cord Fire Hazards | A 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). |
| 05000263/FIN-2015001-03 | 2015Q1 | Monticello | Failure to Maintain a Standard Emergency Action Level Scheme for Flooding | The 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-2015001-04 | 2015Q1 | Monticello | Inadequate Evaluation of Operating Crew During Simulator Assessment | The 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-2015001-05 | 2015Q1 | Monticello | Two Emergency Diesels Inoperable Due to Human Error | A 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-2015001-06 | 2015Q1 | Monticello | Licensee-Identified Violation | The 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-2015001-07 | 2015Q1 | Monticello | Licensee-Identified Violation | The 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-2015001-08 | 2015Q1 | Monticello | Licensee-Identified Violation | The 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-2015001-09 | 2015Q1 | Monticello | Licensee-Identified Violation | The 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. |
| 05000266/FIN-2011003-02 | 2011Q2 | Point Beach | Seismic Qualification of the Condensate Storage Tank and Related Flooding Barriers | The inspectors identified an unresolved item (URI) associated with the wall and flooding barrier between the CST and the non-vital switchgear room because the seismic qualifications were not clearly defined. During the performance of TI 2515/183, the inspectors identified a potential deficiency associated with the seismic qualification of the CST, the flooding barriers between the CST and the vital switchgear room, and the ability to protect the related vital switchgear. The inspectors, in consultation with a Region III civil engineer, performed an evaluation of the licensees determination of operability for the issue. The inspectors found that the licensee utilized various elements of seismic qualification utility group (SQUG) methodology and complex calculations to justify operability of the equipment. The inspectors concluded that this issue is an URI pending a review of the related calculations, use of SQUG methodology, and a review of the Point Beach licensing basis for seismic qualifications (URI 05000266/2011003-02; 05000301/2011003-02, Seismic Qualification of the Condensate Storage Tank and Related Flooding Barriers). |
| 05000266/FIN-2011003-05 | 2011Q2 | Point Beach | Diesel-Driven Fire Pump Loss of Suction During Surveillance Testing | The inspectors identified a URI when the diesel-driven fire pump failed a routine surveillance test. On June 21, 2011, the licensee unsatisfactorily performed surveillance test, O-PT-FP-014, Z-935 Portable Diesel-Driven Fire Water Pump Quarterly Functional Test, Revision 4, when the pump was unable to take suction from the lake using the portable strainer. Specifically, on the first of two attempts, the pump strainer clogged with grass; and on the second attempt, when the strainer was moved further into the lake, the strainer turned upright into the air space and the pump lost suction. The inspectors were unable, at the completion of this inspection period, to ascertain the impact of the failures relative to the regulatory requirements which established the need for the pump. This issue is unresolved pending a review of the failure and procedural adequacy relative to the current licensing basis (URI 05000266/2011003-05; 05000301/2011003-05, Diesel-Driven Fire Pump Loss of Suction During Surveillance Testing). |
| 05000266/FIN-2011004-01 | 2011Q3 | Point Beach | Potential Failure to Correctly Implement a Systems Approach to Training for the Licensed Operator Requalification Program | The inspectors reviewed the licensees procedures, training materials, and operator evaluation documentation pertaining to the LORT program to determine if the licensee was meeting the requirements of 10 CFR 55.59, Requalification. The Point Beach Nuclear Plant Licensed Operator Continued Training Program Description (TPD) required the licensee to develop, maintain, and implement the LOCT program using a SAT process as described in the TPD, to meet the requirements of 10 CFR 55.59(c). The TPD requires that the LOCT program content be composed of fixed and flexible components. Fixed components are defined by the TPD as tasks and/or topics that primarily address analyzed training needs and have been determined using the SAT process. The SAT process also determines on what frequency fixed tasks/topics should be taught, and the delivery method of the training. Flexible components are those that are responsive to other identified training needs and are used to correct actual or potential weaknesses in the performance of licensed personnel. Plant design changes are one example of a flexible topic in the facilitys TPD. During the review of the licensees Biennial and Long Range Training Plans, the inspectors identified that numerous fixed training tasks had been exempted to allow for scheduling constraints due to the licensees extended power uprate project outages. The only documentation available suggests the tasks were exempted primarily due to time constraints with little regard to the effect of the exemptions on the remaining adequacy of the SAT-based program. Site management contended that it was never their intent to remove the majority the exempted training tasks from the biannual training program and, in fact, had planned to return most if not all the material to the program. In support of this contention, the site developed a white paper explaining their decision making and included a list of 35 exempted topics/tasks that had already been trained on. Additionally, it was stated that all of the other exempted topics/tasks had now been rescheduled for completion. This issue is a URI pending further NRC review of the licensees contention and to complete an adequate assessment of the statements in the white paper related to the exempted training, URI 05000266/2011004-01; 05000301/2011004-01, Potential Failure to Correctly Implement a Systems Approach to Training for the Licensed Operator Requalification Program. |
| 05000266/FIN-2011004-02 | 2011Q3 | Point Beach | Failure to Perform an Operability Evaluation for Rod Drive Control System Failures | The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an operability evaluation as required by procedure when degraded/non-conforming conditions were identified during a surveillance of the rod drive control system. Specifically, on December 10, 2010, the licensee documented rod trouble alarms in condition report 01401564, but did not identify the degraded/non-conforming condition or evaluate the condition relative to support functions for technical specifications (TSs) 3.1.4 and 3.1.6. The licensee entered this issue into its corrective action program for evaluation and development of corrective actions. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to identify the degraded/non-conforming condition and assess the impact on operations and TS requirements resulted in latent conditions that had the potential to be of greater safety significance, and in this case resulted in the failure to evaluate the degraded/non-conforming condition relative to TSs 3.1.4 and 3.1.6. This finding has a cross-cutting aspect in the area of human performance, decision-making, because the licensee did not use conservative assumptions during related decision-making that adopted a requirement to demonstrate that the proposed action was safe in order to proceed. |
| 05000266/FIN-2011004-03 | 2011Q3 | Point Beach | Failure to Ensure Tornado Missile Protection for EDGs G01 and G02 Exhaust Stacks | The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to ensure tornado missile protection for two of the emergency diesel generator (EDG) exhaust stacks, which were considered Class I components. The licensee entered this issue into the Corrective Action Program as AR 01678709. The licensees failure to ensure tornado missile protection for EDGs G01 and G02 exhaust stacks was a performance deficiency. The performance deficiency was determined to be more than minor because there was reasonable doubt the EDG exhaust stacks would remain functional to support EDG operation in the event tornadoinduced missiles damaged the exhaust stacks The finding screened as very low safety significance because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding was determined not to have a cross-cutting aspect. |
| 05000266/FIN-2011004-04 | 2011Q3 | Point Beach | Licensee-Identified Violation | 10 CFR, Part 50, Appendix B, Criterion II requires, in part, that measures shall be established to assure that the design basis, for those SSCs that mitigate the consequences of postulated accidents ar correctly translated into procedures. Contrary to the above, PBAPS did not ensure that the CS system required flow of 6,874 gallons per minute (gpm) was correctly translated into Emergency Operating Procedure T-111, Level Restoration, to ensure long term core cooling following a loss of coolant accident. The 6,874 gpm power rate was determined by engineering analysis to account for the 624 gpm leakage through the CS sparger headers and into the reactor vessel annulus region, thereby bypassing long-term cooling of the fuel in the core shroud region. The inspectors determined that this finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Attachment 4, Phase 1 -Initial Screening and Characterization of Findings, Mitigating Systems cornerstone. because the finding did not result in the actual loss of safety function. PBAPS engineering review of quarterly surveillance tests for the last three years determined that the CS pumps have mor than sufficient margin to account for the leakage. The inspectors verified the determination through an independent inspection sampling of surveillance test data. This finding has been documented in the CAP under IR 1245207. |
| 05000266/FIN-2011005-01 | 2011Q4 | Point Beach | Failure to Disposition a Pipe Support in Accordance with ASME Code | The inspectors identified a finding of very low safety significance and associated non-cited violation of 10 CFR 50.55a(g)(4) for the licensee\'s failure earlier in 2011 to accept for continued service, by correction, or evaluation or test, a safety injection (SI) system support (SI-1501R-2-H1) whose examination detected a condition unacceptable (improper hot and/or cold setting) for continued service in accordance with American Society of Mechanical Engineers (ASME) Section XI Code. The licensee, having instead incorrectly dispositioned the condition with a system operability screening, subsequently completed an analysis to confirm that the support was operable with this configuration and entered this issue into its corrective action program. This finding was of more than minor significance because the licensee routinely failed to perform evaluations on similar issues. The failure to confirm the ability of this support to carry design loads as required by ASME Section XI Code prior to returning it to service, increased the likelihood of a component failure and adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was of very low safety significance based on answering No to the Phase I screening question identified in the Mitigating Systems column of Table 4a in Inspection Manual Chapter, Attachment 0609.04 Phase I - Initial Screening and Characterization of Findings. The finding has a cross-cutting aspect in the area of human performance, resources, because the licensees training was not adequate and failed to direct personnel to disposition an unacceptable condition in accordance with the requirements of the ASME Section XI Code |
| 05000266/FIN-2011005-02 | 2011Q4 | Point Beach | Determining an Individuals Dose of Record with Discrepant TLD/ED Data Inputs | The inspectors reviewed the licensees process and procedures for resolving discrepant information associated with thermoluminescent dosimeter (TLD) and electronic dosimeter (ED) records involving the same radiologically controlled area (RCA) entry. Specifically, the 2011 TLD blind spiking test results had dose under reports that were unexplained in the tests evaluation. In one instance with no explanation, a TLD test result indicated 142 mRem (millirem) recorded dose versus 219.5 mRem exposed dose. Similarly, when the inspectors reviewed radiation worker exposure evaluations, some individuals were assigned their ED dose as the dose of the record. In other instances, individuals were assigned their TLD dose as the dose of record. The radiation worker exposure evaluations reviewed by the inspectors were incomplete, in that, there were no bases explaining why the ED data or TLD data was used for a given RCA entry. The inspectors concluded that more information was needed from the licensee to fully understand how the licensee determined several individuals dose of record |
| 05000266/FIN-2011005-03 | 2011Q4 | Point Beach | Condition Reports and URIs Potentially Affecting Safety System Functional Failure Performance Indicator | While performing the PI validation for safety system functional failures, the inspectors found no errors in the pertinent LERs. However, the inspectors identified several CRs that require further review to determine whether the PI was affected. The issues were identified in the CAP as AR01663181 and AR01645462 and will be reviewed by the resident inspectors. Additionally, AR01678709 is being reviewed by Division of Reactor Safety. This issue related to the qualification of the EDGs has the potential to affect the PI. The Office of Nuclear Reactor Regulation is reviewing two URIs (05000266/2011003-02, 05000301/2011003-02, Seismic Qualification of the Qualification of the Condensate Storage Tank and Related Flooding, and 05000266/2011003-03, 05000301/2011003-03, RHR Pump Operability With Tanks In Auxiliary Building Not Seismically Qualified) relating to seismic qualification of SSCs important to safety, which also have the potential to impact the PI. At the end of the inspection period, the inspectors were waiting for additional information, or the completed assessments, to determine the impact on the reported data for the PI (URI 05000266/2011005-03; 05000301/2011005-03, Condition Reports and URIs Potentially Affecting Safety System Functional Failure Performance Indicator). |
| 05000266/FIN-2011005-04 | 2011Q4 | Point Beach | Licensee-Identified Violation | A licensee-identified violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the SI pump oiler modifications performed in 1995, was identified. The details of this issue are discussed in section 4OA3.2 On February 27, 2011, during testing of the Unit 2, SI Train A, an operator took initiative to inspect SI Train B. The operator found that the oiler for SI Pump B had rotated, resulting in a loss of lubricating oil to the pump inboard bearing. As a result of this condition, the licensee declared Train B inoperable. With Train A inoperable for testing, and Train B inoperable for the oiler deficiency, the licensee entered TS 3.0.3. The licensee took immediate corrective actions to restore Train B to service by refilling and reinstalling the oiler. Subsequently, the licensee performed a root cause evaluation for the issue and determined that the oiler was modified in 1995 and that the modification introduced a latent design/configuration flaw that rendered the oilers susceptible to inadvertent bumping events. The licensee instituted corrective actions to modify the design to make to oiler less susceptible to becoming inadvertently dislodged. The inspectors considered this issue as a licensee-identified violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the issue was identified as the result of an operator performing a deliberate and focused observation of the SI system, and because the issue was not discovered as a result of the condition being self-revealing. Documents reviewed are listed in the Attachment to this report. |
| 05000266/FIN-2012503-01 | 2012Q2 | Point Beach | Protective Action Recommendation Weakness | An NRC-identified finding with a preliminary low to moderate safety significance and one associated apparent violation of 10 CFR 50.47(b)(10) for failure to develop and put into place guidelines for the choice of protective actions during an emergency that were consistent with Federal guidance. Federal guidance for the choice of protective actions during an emergency is described in EPA-400-R-92-001, and states, in part, that withdrawal of protective actions from areas where they have already been implemented is usually not advisable during the early phase because of the potential for confusion and possibly impede implementation of protective actions which could place the public at additional risk. Additionally, Federal guidance described in NUREG-0654/FEMA-REP-1, Supplement 3, states, in part, licensees should not relax protective actions until the source of the threat is under control. In the case of a known impediment to evacuation, the licensees emergency implementing procedure, EPIP 1.3, Dose Assessment and Protective Action Recommendations, incorrectly directed key decision makers to withdraw protective actions to evacuate the public and replace it with a recommendation to shelter the public. After the NRC identified the finding, the licensee immediately revised its emergency implementing procedure to be consistent with Federal guidance. This finding is more than minor because it affected the Emergency Preparedness Cornerstone objective of implementing adequate measures to protect the health and safety of the public during a radiological emergency, and is associated with the cornerstone attributes of emergency response organization performance and procedure quality. Specifically, the withdrawal of implemented protective actions could cause confusion of offsite authorities and the public. The inspectors evaluated the finding using the SDP and determined this finding screened as preliminarily White. The finding has a cross-cutting aspect in the area of Human Performance, Resources, because the licensee failed to maintain complete, accurate, and up-to-date procedures as early as 2003 when the licensee returned sheltering to its range of protective action recommendation emergency plans and procedures. |
| 05000266/FIN-2013003-01 | 2013Q2 | Point Beach | Failure to Control Materials Classified as High Winds/Tornado Hazards | The inspectors identified a finding of very low safety significance for the licensees failure to maintain control over the proper storage and placement of materials that were classified as high winds/tornado hazards, in accordance with procedure NP 1.9.6, Plant Cleanliness and Storage. Specifically, the inspectors identified that the licensee failed to perform weekly high wind missile hazards inspections since April 17, 2013. As a result, unsecured wooden pallets, wooden planks, metal rods and a metallic desk were discovered by the inspectors near Units 1 and 2 transformer areas. The issue was entered into the licensees corrective action program (CAP) for resolution as action request AR01882921. The licensee took immediate corrective action to remove and/or properly store the material after the tornado warning on June 17, 2013. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected, the unsecured items would have the potential to lead to a more significant safety concern during high wind and tornado events. The inspectors determined the finding to be of very low safety significance because the inspectors answered No to each question listed in IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions . The inspectors determined that the finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not provide supervisory or management oversight of work activities such that nuclear safety was supported. Specifically, the licensee failed to provide appropriate oversight of work activities such that, when the program owner of the weekly high wind inspection changed, the requirement to perform weekly high winds tornado hazard walkdowns was not understood (H.4(c)). |
| 05000266/FIN-2013003-02 | 2013Q2 | Point Beach | Failure to Follow Operability Evaluation Process Following Water Leakage into the Control Room | The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V for the licensees failure to follow procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, following water leakage into the control room, the licensees immediate operability determination failed to evaluate the effect the leakage had on the control room envelope operability. Additionally, the licensee did not address the functionality of the degraded flood barrier and its impact on operability. This issue was entered into the CAP as AR01877185. Corrective actions for this issue included performing a test of the control room envelope to demonstrate that appropriate positive pressure could be maintained with the known degraded barrier, and repair of the degraded flood barrier following performance of a functionality assessment. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Protection Against External Factors attribute of the Initiating Event Cornerstone, and adversely affected the Cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors determined the finding to be of very low safety significance in accordance with IMC 0609, Appendix A, Exhibit 1, because they answered No to the questions under Transient Initiators and External Event Initiators. The inspectors concluded that this finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate this problem such that the resolution addressed the cause and evaluated the condition for operability (P.1(c)). |
| 05000266/FIN-2013003-03 | 2013Q2 | Point Beach | Lack of Acceptance Criteria for Containment Visual Examinations | The inspectors identified a non-cited violation of 10 CFR 50.55a(g)(4), for failure to define acceptance criteria for containment visual examinations. Consequently, active containment liner degradation (pitting) was identified and the liner returned to service without defined criteria for accepting this condition. The licensee entered this issue into the CAP as action requests AR01858862 and AR01861158, and developed visual examination acceptance criteria to restore compliance with this NRC regulation. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening dated September 7, 2012, because it adversely affected the Barrier Integrity Cornerstone attribute of maintaining the functional integrity of containment. The inspectors also answered Yes to the more-than-minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically, the lack of acceptance criteria in site procedures for containment visual examinations would become a more significant safety concern in that active liner degradation may not be properly evaluated and/or promptly corrected, resulting in a containment liner breach. 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 Barrier Integrity Cornerstone because the corrosion induced pitting degraded the containment barrier. The inspectors determined this finding was of very low safety significance based on answering No to the Exhibit 3, Barrier Integrity Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, issued on June 19, 2012. Specifically, the inspectors answered No to the screening question associated with an actual open pathway (e.g., breach) in the containment and No to the question associated with reduction in function of hydrogen igniters in containment. The inspectors determined that the primary cause of the failure to define containment visual examination acceptance criteria was related to the cross-cutting component of human performance, decision making, because licensee staff did not apply a systematic process, when faced with unexpected plant conditions, to ensure safety was maintained. Specifically, a systematic process for developing acceptance criteria was not applied for the containment visual examinations (H.1(a)). |
| 05000266/FIN-2013003-04 | 2013Q2 | Point Beach | Incorrect Equipment Selected for Ultrasonic Examination | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, for failure to select an appropriately contoured ultrasonic examination search unit wedge in accordance with procedure NDE-173, PDI Generic Procedure for the Ultrasonic Examination of Austenitic Piping Welds. Consequently, three elbow-to-pipe socket welds on the chemical and volume control system (CVCS) line were examined with the incorrectly contoured search unit and this examination would not provide a demonstrated level of accuracy necessary to reliably detect and size thermal fatigue cracks. The licensee entered this condition into the CAP as AR 01860155. To restore compliance with NRC regulations, the licensee considered the option of repeating these weld examinations using a qualified ultrasonic examination technique or the option to seek NRC approval to deviate from the American Society of Mechanical Engineers (ASME) Code Section XI requirements for ultrasonic examination. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, because the inspectors answered Yes to the more-than-minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically , the examination of three chemical and volume control system welds was presumed adequate and absent NRC intervention, would have been returned to service for an indefinite period of service, which would have placed the piping at increased risk for undetected thermal fatigue cracking, leakage, or component failure. 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 Initiating Events Cornerstone because leakage at this chemical and volume control system letdown line could result in a primary system loss of coolant accident. The inspectors determined this finding was of very low safety significance based on answering No to the questions in Part A of Exhibit 1, Initiating Events Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, issued on June 19, 2012. The inspectors answered these questions No because of the small diameter (2-inch) of the line and because the affected pipe welds were subjected to a VT-2 visual and penetrant testing (PT) examination that did not identify rejectable defects. The primary cause of the failure to select ultrasonic equipment (search unit contour) in accordance with procedure NDE-173 was related to the cross-cutting component of human performance, work practices, because the licensees management staff did not adequately set up clear expectations for procedure control and adherence for this activity. Specifically, insufficient direction was provided to vendor staff for simultaneous use of two procedures, NDE-178 and NDE-173, with different equipment requirements and restrictions (H.4(b)). |