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05000341/FIN-2018003-02Fermi2018Q3Failure to Ensure Electrolytic Capacitors Installed in the Plant Did Not Have Expired Shelf LivesA finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components was self-revealed when the reactor water cleanup system inlet flow square root converter failed, resulting in a failure of the reactor water cleanup (RWCU) differential flow instrument and loss of automatic isolation function of the RWCU isolation valves. Specifically, electrolytic capacitors were installed in the RWCU system logic that had expired shelf lives, resulting in failures of the automatic isolation function of the RWCU system.
05000387/FIN-2018002-04Susquehanna2018Q2EGM on Dispositioning BWR Licensee Noncompliance With TS Containment Requirements During Operations With A Potential For Draining The Reactor Vessel (EGM-11-03)From April 2 through April 24, 2018, Susquehanna performed OPDRVs without establishing secondary containment integrity. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measures to terminate the uncovering of fuel. TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable, and is applicable during OPDRVs. The required action for this specification if secondary containment is inoperable in this condition of applicability is to initiate actions to suspend OPDRVs immediately. As reported in LER 05000387/2018-001, Susquehanna conducted the following OPDRVs during the period of secondary containment inoperability: Recirculation system maintenance and pump replacement; Reactor water cleanup system flushes and maintenance; RHR system maintenance; Hydraulic control unit and control rod drive system maintenance; Local power range monitor replacements, including Intermediate Range Monitor 1E Dry Tube replacement; Control rod drive mechanism replacements; and Core spray instrument line flush. NRC EGM 11-03, EGM on Dispositioning BWR Licensee Noncompliance With TS Containment Requirements During Operations With A Potential For Draining The Reactor Vessel, Revision 3, provides, in part, for the exercise of enforcement discretion only if the licensee demonstrates that it has met specific criteria during an OPDRV activity. The inspectors assessed that Susquehanna adequately implemented these criteria. In accordance with EGM 11-003, in order to continue to receive enforcement discretion, a license amendment request (LAR) must be submitted and accepted for review within 12 months of the NRC staffs publication of the generic change, which occurred on December 20, 2016. The inspectors verified that Susquehanna submitted the required LAR on September 20, 2017 (ADAMS Accession No. ML17265A434), and that it was subsequently accepted by the NRC for review by a letter dated October 16, 2017 (ADAMS Accession No. ML17290A024).Corrective Action: Susquehanna submitted an LAR to adopt TS Task Force Traveler 542, Reactor Pressure Vessel Water Inventory Control, on September 20, 2017.Corrective Action Reference: AR-2015-01733 Enforcement: Violation: TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable, and is applicable during OPDRVs. The required action for this specification if secondary containment is inoperable in this condition of applicability is to initiate actions to suspend OPDRVs immediately. Therefore, failing to maintain secondary containment operability during OPDRVs without initiating actions to suspend the operation was considered a condition prohibited by TSs as defined by 10 CFR 50.73(a)(2)(i)(B). Contrary to the above,from April 2 through April 24, 2018, Susquehanna performed OPDRVs without establishing secondary containment integrity. Basis for Discretion: The NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy because all criteria described in EGM 11-003 were met, enforcement discretion was previously authorized by EA-2017-089, and the licensee submitted an LAR on September 20, 2017 which was subsequently accepted by the NRC for review on October 16, 2017, and, therefore, will not issue enforcement action for this violation. The disposition of this violation closes LER 05000387/2018-001-00.
05000220/FIN-2018001-01Nine Mile Point2018Q1Potential Failure to Submit an 8-Hour Event Notification for a Valid Actuation of HPCOn March 18, 2018,at 1:18 a.m., during the Unit 1maintenance outage while the unit was in cold shutdown, operators received multiple low level alarms on the GEMAC 11 and 12 level indications. Operators responded by adjusting reactor water cleanup reject flow and the feedwater minimum flow control valve to raise reactor water level. Upon the operators making the adjustment to reactor water level, the feedwater low flow control valve was slow to respond, but eventually opened more rapidly, and the increased flow from feedwater resulted in a rapid rise in reactor water level. At 1:28 a.m., indicated reactor water level rose to the 95-inch trip setpoint for the 11 and 12 Yarway level indication instruments, resulting in a turbine trip and HPCI initiation signal. The HPCI pumps were tagged out and thus did not inject, and the turbine was offline for the shutdown. The 11 and 12 Yarway level indication instruments provide reactor protection system logic inputs for reactor vessel water level; however, the Yarway level indication instruments are not density compensated. Therefore, under cold shutdown conditions, actual reactor vessel water level was lower than indicated water level on the 11 and 12 Yarways. During cold shutdown conditions, the GEMAC level instruments, which are calibrated to cold shutdown conditions, provide an accurate indication of actual reactor vessel water level. The GEMAC instruments both indicated well below the trip setpoint of 95 inches (indicated ~72 inches) when the turbine trip and HPCI initiation signal were received. Exelon determined that this event was not reportable under 10 CFR 50.72.Title 10 CFR 50.72(b)(3)(iv)(A) states, Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. (B) The systems to which the requirements of paragraph (b)(3)(iv)(A) of this section apply are: 10 (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system. Planned Closure Action(s): The inspectors requested the 10 CFR 50.72 subject matter experts at the Office of Nuclear Reactor Regulation (NRR) and Office of General Council (OGC) to review whether this was a valid actuation and thus reportable. The inspectors are opening an unresolved item (URI) to determine if a performance deficiency exists.Licensee Action(s): Licensee entered the concern into their corrective action program, and communicated with NRC Region I and NRR Staff. Exelons position is that the event was not reportable. Corrective Action Reference:IR 04116336 NRC Tracking Number: 05000220/2018001-01
05000263/FIN-2017004-01Monticello2017Q4Failure to Maintain Radiation Exposure ALARAA finding of very low safety significance (Green) was self-revealed due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees radiological work planning and work control program. Specifically, the licensee failed to properly incorporate ALARA strategies, insights while planning, and executing work activities during the 1R28 refueling outage. The Reactor Water Cleanup (RWCU) Inlet Outboard Isolation Valve MO2398 was scheduled for replacement during the outage. The initial dose estimate for this activity was 4.5 person-rem. However, 13.776 actual person-rem of dose was received. This issue was caused by poor radiological planning and work execution of this task. The licensee entered this issue into their Corrective Action Program (CAP) item 1558234. The finding was more than minor because it was associated with the program and process attribute of the Occupation Radiation Safety Cornerstone. Additionally, this issue affected the cornerstone objective of ensuring the adequate protection of the workers health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, the finding is very similar to IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, Example 6.i. This example provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance (Green) because Monticello Nuclear Generating Plants current 3year rolling average collective is 64.637 person-rem (20142016). This is less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. This finding had a cross-cutting aspect in the area of Human Performance, related to the cross-cutting aspect of Work Management, in that the outage plan did not adequately plan, control and execute work activities to ensure the RWCU Inlet Outboard Isolation Valve MO2398 replacement remained ALARA. (H.5)
05000461/FIN-2017003-05Clinton2017Q3Failure to Establish Secondary Containment Prior to Entering MODE 2The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS LCO 3.0.4, for the failure to follow station procedure CCAA201, Plant Barrier Control Program, Revision 11. Specifically, the licensee entered MODE 2 from MODE 4 without meeting the requirements of LCO 3.0.4 for entering a mode when an applicable LCO is not met. The licensee had not met LCO 3.6.4.1 because the doors to the B reactor water cleanup room were both opened instead of being closed to make secondary containment operable as required in MODE 2. The licensee entered this issue into their CAP as AR 04017613. As corrective actions, the licensee planned to conduct training for site personnel.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to follow the station procedure by not identifying that the open doors required a plant barrier impairment (PBI) permit that would have identified the doors as a constraint to entering MODE 2 resulted in the unit transitioning to MODE 2 with the secondary containment inoperable. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone and determined 5 to be of very low safety significance because the finding only represented a degradation of a radiological barrier function provided for auxiliary building. The inspectors determined that this finding affected the cross-cutting are of human performance in the aspect of training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent work force and instill nuclear safety values. Specifically, station personnel did not know the process for routing a PBI permit and did not know when a PBI permit was required. (H.9)
05000461/FIN-2017003-03Clinton2017Q3Failure to Perform Engineering Evaluation to Determine the Cause of Failure of SnubbersThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to demonstrate compliance with the requirement as prescribed in procedure ERCL330, CPS Snubber Program, Revisions 1 and 2. Specifically, the licensee failed to perform engineering evaluations to determine the cause of failure of snubbers that did not satisfy their functional testing acceptance criteria. The licensee entered this issue into their CAP as ARs 04015242 and 04041302. As corrective actions, the licensee evaluated the components affected by the failed snubber and determined that no operability issues existed. The performance deficiency was determined to be more-than-minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of Protection against External Factors and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability for mitigating systems to respond to initiating events. Specifically, compliance with ERCL330 would ensure the failed snubber wasevaluated for the cause of failure, to ensure the licensee identified other snubbers that may have been vulnerable to the same type of deficiency. This would ensure that any potential undesired loading on the piping system could be avoided and the affected safety-related residual heat removal and reactor water cleanup piping systems could continue to perform their design function of maintaining the pressure boundary and structural integrity following a postulated design basis seismic event. The inspectors determined the finding could be evaluated using the Significance Determination Processin accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, for the Mitigating Systems cornerstone and then Exhibit 4, External Events Screening Question. The finding screened as having very low safety significance because in each instance, the inspectors answered No to Questions 1 and 2 ofExhibit 4. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. Specifically, the licensee failed to establish a systematic approach to evaluating snubbers that did not meet the acceptance criteria to ensure all required aspects were addressed. (H.13)
05000440/FIN-2017002-02Perry2017Q2Implementation of Enforcement Guidance Memorandum 11003, Revision 3From March 17, 2017, to March 24, 2017, Perry Nuclear Power Plant (PNPP) performed Operations with the Potential to Drain the Reactor Vessel (OPDRV) while in Mode 5 without an operable primary and secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measures to terminate the uncovering of fuel. Secondary containment was required by TS 3.6.4.1 to be operable during OPDRVs. Primary containment was required by TS 3.6.1.10 to be operable during OPDRVS. The required action for these specifications was to suspend OPDRV operations. Therefore, entering the OPDRV without establishing primary and secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B).The NRC issued Enforcement Guidance Memorandum (EGM) 11003, Revision 3, on January 15, 2016, to provide guidance on how to disposition boiling water reactor licensee noncompliance with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to (1) adhere to the NRC plain language meaning of OPDRV activities; (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times; (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5; and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities.The inspectors reviewed licensee event report (LER) 201700100 for potential performance deficiencies and/or violations of regulatory requirements. The inspectors also reviewed the stations implementation of the EGM during OPDRVs:The inspectors observed that the OPDRV activities were logged in the control room narrative logs, the log entry appropriately recorded the standby source of makeup water designated for the evolutions, and that defense in-depth criteria were in place.The inspectors noted that the reactor vessel water level was maintained at least 22 feet and 9 inches over the top of the reactor pressure vessel flange as required by TS 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designated in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours.The inspectors reviewed Engineering Change documents which calculated the time to drain down during these activities and the feasibility of pre-planned actions the station would take to isolate potential leakage paths during these periods of time. The inspectors verified that the OPDRVs were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the OPDRVs. The inspectors noted that PNPP had in place a contingency plan for isolating the potential leakage path and verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events.The inspectors verified that all other TS requirements were met during the March 17, 2017, to March 24 2017, OPDRVs with primary and secondary containment inoperable.Technical Specification 3.6.4.1 required, in part, that secondary containment shall be operable during OPDRV. Technical Specification 3.6.4.1, Condition C, required the licensee to initiate action to suspend OPDRV immediately when secondary containment is inoperable. Technical specification 3.6.1.10 required, in part, that primary containment shall be operable during OPDRV. Technical specification 3.6.1.10, Condition A, required the licensee initiate action to suspend OPDRV immediately when primary containment is inoperable. From March 17, 2017, to March 24, 2017, PNPP performed OPDRV activities while in Mode 5 without an operable primary or secondary containment. Specifically, the station performed the following OPDRV activities without an operable primary or secondary containment:draining of reactor recirculation loop B; replacement of 18 control rod drive mechanisms (unbolt and install);replacement of six instrument dry tubes;replacement of reactor recirculation pump B seal;replacement of reactor recirculation loop B flow control valve actuator;plugging of drain line appendages on reactor recirculation pump B; andlocal leak rate testing of the reactor water cleanup suction line containment isolation valves.The failure to perform OPDRV activities with operable primary and secondary containments is a violation of TS 3.6.1.10 and TS 3.6.4.1. Because the violation occurred during the discretion period described in EGM 11003, Revision 3, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation.In accordance with EGM 11003, Revision 3, each licensee that receives discretion must submit a license amendment request within 12 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The inspectors observed thatPNPP is tracking the need to submit a license amendment request as commitment PYL1712101.This LER is closed. This inspection constituted one event follow-up sample as defined in IP 7115305.
05000354/FIN-2016004-01Hope Creek2016Q4Trip of Protected RWCU Pump during Maintenance ActivityGreen. A self-revealing very low safety significance (Green), non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4) was identified for inadequately assessing and managing risks associated with maintenance activities to prevent plant transients that upset plant stability. Specifically, because PSEG did not identify a conflict with the reactor water cleanup (RWCU) pump trip logic prior to conducting a planned breaker swap, the A RWCU pump tripped while it was credited to as a defense-in-depth system for decay heat removal (DHR). PSEG assigned a corrective action to perform a work group evaluation and address lessons learned from this event. The issue was more than minor because it was associated with the Equipment Performance (availability) attribute of the Initiating Event cornerstones and adversely affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown. Additionally, this issue was similar to IMC 0612, Appendix E, examples 7.e and 7.f, in that the resulting increased risk put the plant into a higher risk category. In this case, the plant risk would have been reclassified from Yellow to Orange when RWCU pump was unavailable during residual heat removal (RHR) shutdown cooling outage window. The inspectors evaluated the finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was Green because no quantitative phase 2 analysis was required, and RWCU system was not identified as a major system on Table G1 for Decay Heat Removal safety function. This finding had a cross-cutting aspect in the area of Human Performance, Work Management, because PSEG did not identify and appropriately manage risk associated with the breaker swap activity. Specifically, PSEGs work order to swap the breaker was not planned or scheduled during a RWCU system outage window where the plant shutdown safety risk would have been properly managed (H.5).
05000461/FIN-2016002-06Clinton2016Q2Failure to Perform Surface Examination Prior to Weld Repair of a Reactor Water Cleanup System PipeThe inspectors identified a finding of very-low safety significance and associated NCV of 10 CFR 50.55a(g)(4). Specifically, the licensee failed to perform a surface examination to detect cracking on reactor water cleanup small-bore piping prior to performing a weld repair. The license documented the issue in the CAP as AR 02671726 and AR 02685332 and performed an operability review. The licensee has prepared a work order to perform a surface exam of the existing weld and surrounding area. The inspectors determined that the failure to perform the surface examination prior to weld repair of RWCU pipe 1G33C001B as required by 10 CFR 50.55a(g)(4) was a performance deficiency. The inspectors determined that this issue was more-than-minor in accordance with IMC 0612, Appendix B, because it adversely affect the Initiating Events Cornerstone attribute of barrier integrity and because the answer to the question of If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? was yes. Specifically, the lack of a surface exam may result in the entire defect not being removed during the repair and the potential existed for a cracked pipe to remain in service. This could lead to a repeat leak of reactor coolant. The inspectors determined this finding was of very-low safety significance (Green) based on answering no to Question A.1 and A.2 of the Exhibit 1, Initiating Events Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power. Specifically, the inspectors answered no to the screening question associated with a reactor coolant system leak exceeding the leak rate for a small loss of coolant accident (LOCA) and no to the screening question associated with systems used to mitigate a LOCA. A subsequent visual examination of the weld repair revealed an absence of cracking and the licensee also planned to perform a follow-up surface examination of the repaired area to look for cracking. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of resources, where leaders ensure that personnel, equipment, procedures and other station resources are available and adequate to support nuclear safety. Specifically, the work order that performed the work did not specify a surface examination of the base metal prior to welding. (H.1)
05000461/FIN-2016002-07Clinton2016Q2Failure to Obtain License Amendment prior to Operating Reactor Water Cleanup Bypass SwitchesThe inspectors identified a Severity Level IV NCV of 10 CFR 50.59(a)(1), Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform an adequate written safety evaluation to provide the basis that changes to the Updated Safety Analysis Report (USAR) and station procedures did not involve an unreviewed safety question. Specifically, the licensee changed the USAR and station procedures to allow operators to a defeat the safety function of the reactor water cleanup (RWCU) isolation valves to prevent unwarranted isolation signals during normal operation without obtaining prior Commission approval. The licensee entered this issue into their CAP as AR 02685337 and will be changing station procedures to prevent placing the RWCU leak detection divisional bypass switches in bypass except for instrument channel maintenance, testing or calibration. The inspectors determined that the licensees failure to perform an adequate written safety evaluation to provide the basis that changes to the USAR and station procedures did not involve an unreviewed safety question was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small LOCA and did not affect other systems used to mitigate a LOCA resulting in a total loss of their function (e.g. Interfacing System LOCA). No cross cutting aspect was assigned because the inspectors determined the performance deficiency was not indicative of current plant performance. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
05000387/FIN-2016002-01Susquehanna2016Q2Failure to Promptly Identify a Condition Adverse to Quality Associated with Primary Containment Isolation ValvesA self-revealing Green finding and associated violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, and TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), was identified when Susquehanna did not promptly identify a condition adverse to quality. Despite observing abnormal behavior during local leak rate testing following replacement in May 2014, Susquehanna did not take any action to ensure that certain Reactor Water Cleanup (RWCU) system PCIVs passed their subsequent testing. Consequently, these valves failed their in-service and local leak rate test in March 2016 when they failed to close upon securing system flow. The failure was caused by an internal interference between the check valve hinge and body. Following the failures in March 2016, Susquehanna repaired the valves and successfully performed local leak rate testing, restoring operability of the PCIVs. The repeat failure was entered into the CAP as CRs 2016-06960 and 2016-09940. The finding was determined to be more than minor because it was associated with the Structure, System, and Component (SSC) and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to identify a condition adverse to quality during post-maintenance testing resulted in two PCIVs being rendered inoperable for longer than the TS allowed outage time. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not involve the hydrogen recombiners and did not result in an actual open pathway in the physical integrity of reactor containment. Specifically, the redundant valve for each penetration remained operable during the period in which these two valves were inoperable. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because Susquehanna did not use decision making practices that emphasized prudent choices over those that are simply allowable. Specifically, Susquehanna decided to accept elevated seat leakage for two new PCIVs, assuming that they could be declassified as PCIVs.
05000220/FIN-2016001-03Nine Mile Point2016Q1Inadequate Tagout Resulting in Reactor Building Closed-Loop Cooling Drain Down EventA self-revealing Green non-cited violation (NCV) of Technical Specification (TS) 6.4.1, Procedures, was identified when a Unit 1 Exelon operator did not maintain proper configuration control of a plant system during a system tagout for planned maintenance. Specifically, on January 25, 2016, a Unit 1 non-licensed operator manipulated a reactor building closed-loop cooling (RBCLC) system drain valve out of sequence while performing a tagout for the #13 shutdown cooling (SDC) HX for planned maintenance. This resulted in unintentional draining of the operating RBCLC system, annunciation of multiple alarms in the main control room, and operators entering abnormal operating procedures to recover the RBCLC system. As part of corrective actions, proper configuration was promptly restored and the operator involved in the event was given a remediation plan for requalification and placed on an operations excellence plan. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences; and if left uncorrected, the event had potential to lead to a more significant safety concern. Specifically, the failure to quickly isolate the drain down of the RBCLC system would have required a manual reactor scram, a manual trip of all five reactor recirculation pumps (RRPs), a manual isolation of the reactor water cleanup system, a loss of cooling to the spent fuel pool (SFP) cooling system, instrument air compressors, and the control room emergency ventilation system. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not result in the loss of a support system, RBCLC, or affect mitigation equipment. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because the non-licensed operator failed to follow Exelons procedures and the instructions he received at the pre job brief stop when manipulating the drain valve. Specifically, the non-licensed operator rationalized, without being the designated performer of the tagout, that it was acceptable to perform a valve manipulation out of sequence with the tagout plan.
05000259/FIN-2016001-04Browns Ferry2016Q1Unposted High Radiation AreasA self-revealing, NCV of 10 CFR 20.1902(b), with two examples, was identified for the failure to post multiple HRAs. Specifically, areas within the Unit 2 (U2) Control Rod Drive Rebuild Room and U2 Reactor Water Cleanup Holding Pump Room contained dose rates exceeding 100 mrem/hr at 30 cm and remained unposted for several months during 2015. These issues were entered into the licensees corrective action program as CR 1017294, CR 1023385, and CR 1119944, and the licensee took immediate corrective actions to correctly post the areas, performed surveys to evaluate the extent of condition, and performed an Apparent Cause Evaluation. The performance deficiency was greater than minor because it was associated with the Occupational Radiation Safety cornerstone attribute of Program and Process (Monitoring and RP Controls) and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors determined the finding to be of very low safety significance (Green) because it was not related to As Low As Reasonably Achievable (ALARA) planning, nor did it involve an overexposure or substantial potential for overexposure, and the ability to assess dose was not compromised. This finding involved the cross-cutting aspect of Human Performance, Documentation (H.7) because the unposted high radiation areas were a direct result of the failure to identify documented radiological conditions that required additional posting and control.
05000461/FIN-2016001-03Clinton2016Q1Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety FunctionThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.72(b)(3)(v) for failing to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within eight hours. Specifically, control room operators placed both divisions of reactor water cleanup differential flow instruments in bypass, which rendered the instruments inoperable and resulted in a loss of the isolation function. The licensee entered this issue into the CAP as AR 02645140 and created an action to submit a licensee event report under 10 CFR 50.73(a)(2)(v). The inspectors determined that the failure to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within 8 hours as required by 10 CFR 50.72(b)(3)(v) was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.72 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.72, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
05000263/FIN-2016008-01Monticello2016Q1Failure to provide acceptable Alternate Methods of Decay Heat RemovalThe inspectors identified an Unresolved Item associated with Technical Specification (TS) 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System Cold Shutdown. Specifically, the licensee failed to verify that the capability of the alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown Cooling, were adequate to combat a loss of shutdown cooling resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay heat load. The Limiting Condition for Operation (LCO) 3.4.8 of TS Residual Heat Removal Shutdown Cooling System Cold Shutdown, required in Mode 4, two RHR shutdown cooling subsystems shall be operable, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation. The TS Bases Section 3.4.8, indicated that an operable RHR shutdown cooling subsystem consisted of one operable RHR pump, one heat exchanger, the associated piping and valves, and the necessary portions of the RHR Service Water System System capable of providing cooling water to the heat exchanger. The TS Bases Section 3.4.8 further indicated that the two subsystems have a common suction source and were allowed to have a common heat exchanger and common discharge piping. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be operable. Since the piping and heat exchangers were passive components that were assumed not to fail, they were allowed to be common to both subsystems. When TS 3.4.8, LCO could not be met, Condition A, for one or two RHR shutdown cooling subsystems inoperable, the Required Action was to, verify an alternate method of decay heat removal was available for each inoperable RHR shutdown cooling subsystem. The completion time for the required action was 1 hour, and once per 24 hours thereafter. The TS Bases 3.4.8 for Condition A indicated that with one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem was capable of providing the required decay heat removal. However, the overall reliability was reduced, therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This was to ensure the re-establishment of backup decay heat removal capabilities, similar to the requirements of the LCO. The bases further stated that the required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Alternate methods that can be used included (but not limited to) the Reactor Water Cleanup System by itself or using feed and bleed in combination with Control Rod Drive System or Condensate/Feed Systems. Abnormal Procedure, Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown Cooling, provided instructions for establishing alternate methods for decay heat removal. The inspectors noticed that except for the alternate method as described below in the G-EK-1-45, the licensee was not able to show by calculation or demonstration that the systems and methods credited in this procedure would be capable of providing sufficient heat removal capability or appropriate levels of redundancy as required by TS 3.4.8. The G-EK-1-45 was a General Electric Letter to Northern States Power, Subject: Cold Shutdown Capability Report, dated April 22, 1981. This letter provided a report which described the capability of the Monticello Nuclear Generating Plant to achieve cold shutdown using only safety class systems and assuming the worst single failure. The alternate shutdown decay heat removal method used in the report credited combinations of the RHR pumps and heat exchangers in the suppression pool cooling mode of RHR to ensure suppression pool water temperatures were below the design limit. This method utilized the core spray system and safety relief valves to circulate reactor inventory to remove decay heat from the reactor. The inspectors noted that calculations supporting the above alternate strategy utilized an RHR subsystem that could be inoperable and/or unavailable and therefore may not be credited to comply with TS 3.4.8. Specifically, the inspectors were concerned that while the plant was in mode 4, with a credited one subsystem inoperable, the licensees credited alternate decay heat removal method that relied on an RHR subsystem, to perform the required suppression pool cooling function. The inspectors were concerned that relying on the only operable RHR subsystem for the alternate method did not meet the intent of the TS requirement as described in the TS Bases. Furthermore, the inspectors noticed for Mode 4 with two RHR subsystems inoperable, the licensee failed to verify by calculation or demonstrations that two additional redundant alternate decay heat removal methods existed with sufficient capacity to maintain the average reactor coolant temperature below 212 degrees Fahrenheit. During the inspection, the licensee indicated that the Boiling Reactor Owners Group was in the process of developing a draft TS Task Force Traveler to address the requirement of TS 3.4.8 and its Bases. Based on the information above, the inspectors were concerned that the plant Operations Manual was inadequate and failed to include alternate decay heat removal methods that would enable the licensee to comply with the requirement of TS 3.4.8. The Operations Manual was required per TS 5.4.1, Procedures, which required that written procedures shall be established, implemented, and maintained covering the emergency operating procedures. The inspectors determined that this issue was unresolved pending the actions by the licensee and the Boiling Reactor Owners Group and the NRC review of these actions. The licensee entered the inspectors concerns into their Corrective Action Program as AR 01516098.
05000353/FIN-2015003-01Limerick2015Q3Inadequate Procedure for RWCU Backwashing OperationsA self-revealing Green NCV of Technical Specification (TS) 6.8.1.a, Procedures and Programs, occurred because Exelon failed to establish, implement, and maintain an adequate procedure for the control of radioactivity and limiting personnel exposure during operation of a solid radioactive waste system. Specifically, the procedure for the conduct of reactor water cleanup (RWCU) filter media backwashing and collection was inadequate to ensure a sufficient receiving tank volume prior to transferring waste media. On June 28, 2015, this resulted in the overflow of a Unit 2 RWCU collection tank and back up of the reactor building floor drain system, causing high levels of radioactive contamination in accessible portions of the Unit 2 reactor building, and resulting in radioactive contamination of personnel. Exelon controlled access, decontaminated affected areas and personnel, conducted bounding dose assessments, performed extent of condition reviews, and revised affected procedures to address the issue. Exelon placed this issue into the corrective action program as issue report (IR) 2520732. This issue is more-than-minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to effectively control and manage radioactive material could result in significant unplanned, unintended occupational radiation exposure of workers. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not involve an as low as is reasonable achievable (ALARA) issue, was not an overexposure, did not result in a substantial potential for an overexposure, and did not compromise the ability to assess dose. The inspectors determined this finding has a cross-cutting aspect in the area of Human Performance, Avoiding Complacency, because Exelon did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, and therefore did not implement appropriate error reduction tools. Specifically, Exelon operated the backwash receiving tank (BWRT) to routinely accept high level alarms with associated potential for system overflow. Consequently, although this mode of operation of the system was longstanding, the issue reflects present performance.
05000219/FIN-2015002-03Oyster Creek2015Q2Reactor Water Cleanup Procedure Not Followed Resulting in a Level TransientA self-revealing NCV of Technical Specification 6.8.1(a), Procedures and Programs, was identified because Exelon did not follow procedure 303, Reactor Cleanup Demineralizer System, during the system restoration on March 26, 2015. Specifically, during startup from a forced outage (1F36), Exelon did not follow procedure 303, which required correct valve lineups for system restoration of reactor water cleanup (RWCU) after system isolation. This resulted in decreasing reactor water level, which was automatically terminated by a second RWCU isolation. Exelon entered this issue into the corrective action program. Planned corrective actions include enhancing operator training in system knowledge and procedure compliance and revising startup procedures. This finding is determined to be more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone 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. Specifically, Exelon did not properly lineup the RWCU system after isolation, which resulted in a water level transient and challenging the critical safety function of inventory control. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, the operators did not stop and fully communicate plant condition after the initial RWCU isolation. Consequently, operators opened the RWCU system inlet valve due to the increasing water level without following procedure guidance.
05000366/FIN-2015001-02Hatch2015Q1Failure to perform complete analysis of air samplesAn NRC-Identified non-cited violation (NCV) of TS 5.4.1 was identified for the failure of the licensee to perform complete quantitative analysis of air samples using approved counting equipment as required by the licensees procedures. NMP-HP-301, Step 5.6, provides guidance for quantitative evaluation of air samples. On February 16, and 25, 2015, air samples for work activities in the Reactor Pressure Vessel head (RPV) and the Reactor Water Cleanup (RWCU) System heat exchanger were not quantitatively analyzed or evaluated for alpha activity even though the areas had been identified as having elevated alpha contamination levels. The licensee entered the issue into their corrective action program (CAP) as CR 10034556. The finding was more than minor because it was associated with the Occupational Radiation Safety Program attribute of exposure control and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from airborne radioactive material during routine civilian nuclear reactor operation. Failure to identify potentially significant contributors to internal dose could lead to unmonitored occupational exposures. The finding was determined to be of very low safety significance (Green) because it did not involve: (1) an as low as is reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose related to As Low As Reasonably Achievable (ALARA) Planning and the ability to assess dose was not compromised during this instance. The cause of this finding was directly related to the cross-cutting aspect of following processes, procedures, and work instructions in the Procedure Adherence component of the Human Performance area.
05000324/FIN-2014003-03Brunswick2014Q2Licensee-Identified ViolationTechnical Specification Section 5.4.1.a, Administrative Control (Procedures), states, in part, that written procedures shall be established, implemented, and maintained, covering applicable procedures recommended in Regulatory Guide 1.33, Appendix A, November 1972 (Safety Guide 33, November 1972). Section l.1 of Regulatory Guide 1.33, Appendix A, November 1972, (Safety Guide 33, November 1972) states, in part, that maintenance that can affect the performance of safety-related equipment should be properly planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, from March 11, 2014, to April 3, 2014, the licensee failed to follow procedure OMA-NGGC-0201, Contingency Planning and Discovery Management, to properly plan for the replacement of the reactor water cleanup (RWCU) inlet line isolation valve 1-G31-F001. Specifically, a written contingency plan was not developed. As a result, significant items were missed in the planning and preparation. This contributed to the licensee exceeding the As Low as Reasonably Achievable (ALARA) dose goal for this job by 11 rem. This finding was more than minor because it was associated with the Program and Process ALARA planning attribute of the Occupational Radiation Safety Cornerstone and adversely affected the objective to ensure the adequate protection of worker health and safety from exposure of radiation from radioactive material during routine civilian nuclear reactor operations. The inspectors determined the finding to be of very low safety significance (Green) because Brunswicks three-year rolling average (2011-2013) is 185 person-rem, which is below the SDP criteria of 240 person-rem for boiling water reactors. The licensee entered this issue into the CAP as NCR 678510.
05000354/FIN-2013005-01Hope Creek2013Q4NCV Failure to Follow Procedure for Configuration Control Activity Adversely Affected Unidentified Leakage in the DrywellA Green self-revealing NCV of TS 6.8.1, Procedures and Programs, was identified regarding PSEGs conduct of maintenance and component configuration control during system restoration from an operation with a potential for draining the reactor vessel (OPDRV) activity. Specifically, PSEG did not close a reactor water cleanup (RWCU) valve in accordance with the maintenance procedure during the refueling outage. This resulted in increased RCS UIL in the reactor drywell area following startup. PSEG restored the mispositioned valves, conducted an extent of condition on other valves in the drywell, completed a prompt investigation concerning the valve mispositioning, and is in the process of conducting an Apparent Cause Evaluation (ACE) on the configuration control event under Order 70161461. PSEG has also placed this issue into CAP as notification 20632003. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, which required an analysis using Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. The finding was determined to be of very low safety significance (Green) because the finding could not result in exceeding the RCS leak rate for a small loss of coolant accident (LOCA) or have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function. This finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because PSEGs communication of human error prevention techniques did not support human performance and proper personnel work practices. Specifically, PSEG did not use adequate human performance tools and valve position verification techniques when controlling valve position for components associated with an OPDRV activity
05000237/FIN-2013004-01Dresden2013Q3Failure to Update the UFSAR for Reactor Water Cleanup Design ChangesA Severity Level IV NCV of 10 CFR 50.71(e), Periodic Update of the Final Safety Analysis Report (USFAR) and an accompanying Green finding were identified by the inspectors for the licensees failure to update the Updated Final Safety Analysis Report (UFSAR) for a design modification performed on the Unit 3 reactor water cleanup (RWCU) system. Specifically, the licensee did not update Dresden UFSAR Section 5.4.8, Reactor Water Cleanup System, to reflect changes made during a design modification installed on Unit 3 in 1997. The design changes included reducing the pipe dimension of RWCU piping outside of the primary containment and eliminating a string of regenerative and non-regenerative heat exchangers. The licensee also identified several high energy line break (HELB) calculations which did not include the design modification when determining the impact on environmentally qualified components affected by a failure of the RWCU system piping outside of the primary containment structure. Corrective actions included submitting a UFSAR change request to include the appropriate operating characteristics and specifications under the present design. In addition, the licensee reviewed all affected calculations to ensure no nonconservative outcomes resulted based on the design modifications installed. This finding was determined to be more than minor using IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012 because, if left uncorrected, the performance deficiency could have led to a more significant safety concern. Specifically, failure to update the UFSAR with the actual RWCU system configuration prevented the inspectors from readily concluding that the design change would not require additional calculational analyses for HELB. The inspectors completed a Phase 1 significance determination of this issue using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated July 1, 2012 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors answered NO to all questions in Exhibit 2, Section A, Mitigating SSCs and Functionality, therefore the finding screened as Green (very low safety significance). In accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation is categorized as Severity Level IV because the information was not used to make an unacceptable change to the facility or procedures since the design changes did not result in a reduction of the previous margin to the 10 CFR 100 guidelines nor did they challenge the environmental quality rating of safety related components in the vicinity of the RWCU system during a HELB event outside of containment. The inspectors determined that this finding did not reflect present performance because it is a legacy issue with changes made to the facility more than 16 years previously; therefore, there was no cross cutting aspect associated with this finding.
05000263/FIN-2013004-03Monticello2013Q3Loss of Accurate Level Indication During Partial RCS Drain DownA self-revealed finding of very low safety significance and non-cited violation of Technical Specification (TS) 5.4.1.a, Procedures, occurred on June 3, 2013, due to the licensees failure to implement procedures regarding maintenance or operations activities for draining and refilling the reactor vessel. Specifically, the licensee failed to follow Step 10 of Operations Manual B.02.02-05, Reactor Water Cleanup System Operation, Section G.1, Reactor Vessel Draining during Cold Shutdown Conditions, to adequately monitor water levels in the reactor during the reactor pressure vessel (RPV) partial draining process. While relying on a temporary installed level instrument, operators performed an RPV drain down which introduced pressure related inaccuracies into the temporary instrument and prevented operators from adequately monitoring vessel level. This resulted in a loss of positive configuration control of reactor coolant system (RCS) level during an infrequently conducted risk-significant evolution, and for four days thereafter. Corrective actions included transferring from the temporary level instrument to the flood up level instrument and enhancing RPV reassembly and temporary vessel installation procedures. This issue is more than minor because it is associated with the configuration control shutdown equipment lineup attribute of the Initiating Events Cornerstone and impacted the cornerstone objective to limit the likelihood of those events that challenge critical safety functions during shutdown operations. In addition, if left uncorrected, the reliance on inaccurate RPV level instrumentation could lead to a more significant safety issue because it constitutes a loss of positive control of reactor vessel level during a risk significant RCS drain down. Using IMC 0609, Appendix G, for shutdown operations, the inspectors determined that the finding had very low safety significance because it did not represent an inadvertent loss of two feet of RCS inventory or inadvertent RCS pressurization, and it did not adversely affect core heat removal, inventory control, power availability, containment control, or reactivity guidelines. The inspectors determined that this finding was cross-cutting in the Human Performance, decision making area, and involved aspects associated with using conservative assumptions in decision making and adopting a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe.
05000440/FIN-2013009-04Perry2013Q2Failure to Follow Procedural Requirements for RWCU System Fill and VentA finding of very low safety significance and associated non-cited violation of Technical Specification 5.4, Procedures, was self-revealed when the licensee failed to adhere to procedural requirements during the filling and venting of the reactor water cleanup (RWCU) system. Specifically, on April 26, 2013, valves 1G33-F008A and F556A were left in the open position, contrary to the requirements of step 7.16.9 of procedure SOI-G33, revision 36, and resulted in the RWCU system being aligned to the condensate transfer and storage system. This valve misposition event also resulted in the TS 3.6.1.3 inoperability of the containment isolation valve 1P11F0545. Upon discovery of the condition, the licensee promptly corrected the error and the entered the condition into its corrective action program as condition report 2013-07483, and performed an apparent cause evaluation. The inspectors reviewed Inspection Manual Chapter (MC) 0612, Appendix B, Issue Screening, and determined that the issue was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, Significance Determination Process. This finding has a cross-cutting aspect in the area of human performance, work practices, for the licensees failure to successfully incorporate human error prevention techniques, such as self and peer checks.
05000397/FIN-2013003-01Columbia2013Q2Failure to Obtain NRC Approval for Changes to Reactor Water Cleanup System PipingThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, because the licensee failed to obtain a license amendment, pursuant to 10 CFR 50.90, prior to implementing a change to piping classification of the reactor water cleanup system. Specifically, through a 1995 revision to the Final Safety Analysis Report, the licensee changed the classification of reactor water cleanup system piping from ASME Section III, Class 3, to ANSI B31.1 without first obtaining NRC approval. The licensee initiated Action Request AR 282022 to address the incorrect downgrading of piping in the reactor water cleanup system. The violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy because the violation could impact the ability of the NRC to perform its regulatory oversight functions. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding was of very low safety significance because the finding did not result in exceeding the reactor coolant system leak rate for a small break loss of coolant accident and because the finding did not affect other systems used to mitigate a loss of coolant accident resulting in a total loss of function. Therefore, in accordance with Section 6.1.d of the NRC Enforcement Policy, the significance was determined to be Severity Level IV. This issue was entered into the licensees corrective action program as Action Request AR 282022. This violation did not have a cross-cutting aspect because it was strictly associated with a traditional enforcement violation.
05000458/FIN-2013002-04River Bend2013Q1Failure to Properly Perform a Maintenance ActivityThe inspectors reviewed a self-revealing finding associated with the licensees failure to provide adequate instructions for installing a new seal cartridge in the reactor water cleanup A pump. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2011-09015. In that condition report, the licensee developed a corrective action to revise all reactor water cleanup procedures and model work orders to verify proper installation of the pump seal. The failure to provide adequate instructions for properly installing reactor water cleanup pump seal cartridges was a performance deficiency. The performance deficiency was more-than-minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and affected the cornerstone objective in that it caused increased collective radiation dose for occupational workers. Additionally, the finding was similar to example 6(i) in Appendix E to Manual Chapter 0612, Power Reactor Inspection Reports Examples of Minor Issues. Using Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding had very low safety significance because, although the finding involved ALARA planning and work controls, the licensees latest three-year rolling average collective dose was less than 240 person-rem. This finding had a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee failed to use complete, accurate and up-to-date procedures and work orders to perform the seal installation, which resulted in unnecessary dose
05000352/FIN-2012005-05Limerick2012Q4Licensee-Identified Violation10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components to which the appendix applies are correctly translated into specifications. Contrary to Criterion III, Exelon failed to correctly translate the design basis for 15 PCIVs on Unit 1 and 15 PCIVs on Unit 2 into specification for the motor operators for the valves. This resulted in the valves not being able to perform their intended safety function under certain conditions following a loss of coolant accident with offsite power remaining available. Exelon entered this issue into the CAP as IR 1402693 and 1416070. The inspectors determined that the finding was of very low safety significance (Green) in accordance with Section B of Exhibit 3 of NRC IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, because it did not represent an actual open pathway in the physical integrity of reactor containment. In addition, all of the effected penetrations, with the exception of two, had another primary containment isolation valve that was not impacted by the design issue. Conservatively processing the two penetrations that did not contain a non-effected valve through NRC IMC 0609, Appendix H, Containment Integrity Significance Determination Process, determined that they were not risk significant from a large early release frequency standpoint. One penetration (suppression pool clean-up suction line) did not connect with the drywell atmosphere or reactor coolant system. The other penetration (reactor water cleanup suction line) is a closed system and the design error would not have affected the ability of the systems primary containment isolation valves to isolate following their design basis event (intersystem loss of coolant accident outside of containment).
05000397/FIN-2012005-06Columbia2012Q4Licensee-Identified ViolationTechnical Specification 3.8.7 Distribution Systems - Operating , requires, in part, that the division 1 and division 2 AC electrical power distribution subsystems be operable in Modes 1, 2 and 3. Technical Specification 3.8.7, Condition A, requires that when the division 1 or 2 AC electrical power distribution subsystem is inoperable, action is taken to restore the division 1 and 2 AC electrical power distribution subsystems to operable status within 8 hours. Failure to meet Technical Specification 3.8.7, Condition A, requires entry into Technical Specification 3.8.7, Condition C. Required Action C.1 requires the unit be placed in Mode 3 within an additional 12 hours. Contrary to the above, on November 24, 2009, August 4, 2010, and September 1, 2010, the division 1 125 V AC power distribution room cooler RRA-FN-11 was inoperable for greater than 20 hours rendering the supported division 1 125 V AC power distribution subsystem inoperable. Additionally, on April 26, 2010, the division 2 125 V AC distribution room cooler RRA-FN-10 was inoperable for 20.0 hours rendering the supported division 2 125 V AC power distribution subsystem inoperable. In each instance action was not taken to place Columbia Generating Station in Mode 3 as required by Technical Specification 3.8.7, Required Action C.1. Additionally, Technical Specification 3.8.7 Distribution Systems - Operating , requires, in part, that the division 1 DC electrical power distribution subsystems be operable in Modes 1, 2 and 3. Technical Specification 3.8.7 Condition D requires that when the division 1 250 V DC electrical power distribution subsystem is inoperable, the licensee to declare the supported features inoperable immediately. On September 1, 2010, division 1 250 V DC power distribution system room cooler RRA-FN-12 was inoperable which requires the licensee to declare the reactor water cleanup system outboard isolation valve RWCU-V-4 inoperable. Valve RWCU-V-4 is a primary containment isolation valve. When this valve is inoperable, Technical Specification 3.6.1.3 Condition A, requires the licensee to isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured within 4 hours. Failure to meet Technical Specification 3.6.1.3, Condition A, requires entry into Technical Specification 3.6.1.3, Condition E. Required Action E.1 requires the unit be placed in Mode 3 within an additional 12 hours. Contrary to the above, on September 1, 2010, RRA-FN-12 was inoperable for 18.2 hours, rendering the supported division 1 250 V DC power distribution subsystem and RWCU-V-4 inoperable and action was not taken to place Columbia Generating Station in Mode 3 as required by Technical Specification 3.6.1.3 Required Action E.1. This finding was identified by the licensee and entered into the corrective action program as Action Request 268099. This finding was evaluated by a senior reactor analyst and determined to be of very low safety significance.
05000461/FIN-2011005-07Clinton2011Q4Licensee-Identified ViolationTS 5.4.1.a states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Station procedure RP-AA-1008, Unescorted Access to and Conduct in Radiologically Controlled Areas, Revision 1 implements the requirements of Regulatory Guide 1.33 Section 7.e.1. Specifically, Step 4.1.3 states that workers are responsible to read, understand, and acknowledge the appropriate copy of the RWP for any work requiring an RWP. RWP 100110101, Revision 0 states that Each worker signing on to this RWP must review the ALARA Plan for instructions pertaining to his/her job or specific evolution. In Section 9, Hold Points of the ALARA Plan states the RPT acknowledgement is required prior to starting any work activities. Contrary to the above, on January 17, 2010, a contract supervisor performed work activities without acknowledgment from radiation protection. Specifically, the contract supervisor reached into reactor water cleanup system piping to install/repair purge dams wearing only a single set of anti-contamination clothing. The inspectors determined that this finding was of more than minor significance because the work on this highly contaminated system without appropriate protective clothing resulted in the individual becoming contaminated on his skin, face, nose, and ears, as well as an internal deposition that resulted in ~3 mrem Committed Effective Dose Equivalent. The inspectors determined that this violation was associated with a licensee-identified finding of very low safety significance (i.e., Green) using the guidance in IMC 0609, Appendix C, Occupational Radiation Safety, since this issue was not related to ALARA, did not result in an overexposure, a substantial potential for overexposure, nor was the ability to assess dose compromised. This violation is being treated as a non-cited violation consistent with Section 2.3.2 of the NRC Enforcement Policy. The licensee entered this violation into its corrective action program as AR 01017724.
05000220/FIN-2011005-01Nine Mile Point2011Q4Troubleshooting Approach Not Consistent With Technical Specification BasesThe inspectors identified a Green finding for the failure of NMPNS to follow the technical specifications (TS) bases associated with limiting condition for operation (LCO) 3.0.2. Specifically on October 24, 2011, on three separate occasions, NMPNS entered TS 3.3.6.1 Condition B for operational convenience to conduct troubleshooting of a reactor water cleanup (RWCU) system differential flow high channel. The inspectors determined this action was contrary to the bases of TS LCO 3.0.2 which states, in part, intentional entry into actions should not be made for operational convenience and must not compromise safety. NMPNS immediate corrective actions included coaching the control room personnel involved in the troubleshooting process and entered the issue into the corrective action program as CR 2011-009767. This finding is more than minor because it impacted the configuration control aspect of the Barrier Integrity Cornerstone and adversely affected the Cornerstone objective to maintain functionality of containment. Specifically, as part of a planned troubleshooting activity, a protective isolation feature was removed from service on multiple occasions that collectively exceeded the allowed LCO time for the system. The inspectors determined that the finding was of very low safety significance (Green) since the finding did not represent an actual open pathway in the physical integrity of the reactor containment. This finding has a crosscutting aspect in the area of human performance in that NMPNS did not use conservative assumptions in decision making when performing multiple entries into TS
05000277/FIN-2011005-01Peach Bottom2011Q4Untimely Corrective Action to Correct MOV Degraded Stem LubricationThe inspectors determined that Exelon\'s failure to promptly correct a condition adverse to quality associated with a safety-related motor-operated valve (MOV) constituted a Green, self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion XVl, Corrective Action. Specifically, corrective actions to prevent recurrence of MOV program testing failures due to degraded stem lubrication in 2009 were not performed in a timely manner to prevent the inoperability of a safety-related MOV due to degraded lubrication, as identified on September 22, 2011. PBAPS entered this issue into the CAP via issue reports (lRs) 1266600 and 1266604. This finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity (Bl) cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the Unit 3 reactor water cleanup (RWCU) outboard isolation valve MO-3-12-018 did not develop sufficient thrust at the torque switch trip setpoint during diagnostic testing on September 22, 2011. The RWCU MOV would not have been able to perform its safety function to close during the most limiting design condition. Using the Phase \'1 worksheet in Appendix 4 of IMC 0609, SDP, the finding affected the Bl cornerstone and was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of containment. This finding had a cross-cutting aspect in the area of Problem ldentification & Resolution (PI&R), CAP, because Exelon did not take appropriate corrective actions to address the adverse trend of degraded stem lubrication on a safety-related MOV in a timely manner
05000271/FIN-2011005-03Vermont Yankee2011Q4Incomplete Inventory for Spent Resin ShipmentA self-revealing NCV of very low safety significance of 10 CFR 20.1501 and 10 CFR 20.2006(b) was identified because Entergy personnel failed to indicate an accurate total of radionuclide activity on the manifest for a radioactive waste shipment on September 19, 2011. Radiation surveys by the receiving personnel at the radioactive waste processing facility identified radiation levels exceeding those indicated on the shipping manifest. Subsequently, Entergy personnel determined that the total radionuclide activity for the shipment was 17 curies instead of 13.4 curies as originally documented. Entergy staff initiated CR-VTY-2011-03902, revised the NRC Form 541, and sent the revision to the radioactive waste processor to correct this error. The inspectors determined that the failure to indicate an accurate total of radionuclide activity on the manifest for a radioactive waste shipment was a performance deficiency that was reasonably within Entergy\\\'s ability to foresee and correct. This finding is more than minor because it affects the Public Radiation Safety cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, the failure to accurately account for all of the radioactive wastes in shipment No. 2011-85 had the potential for misclassifying wastes non-conservatively in subsequent radioactive waste processing and final shipment activities to a low level burial ground facility. The inspectors evaluated the finding using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process. The inspectors determined the finding to be of very low safety significance (Green) because the error was corrected at the waste processor rather than after shipment to a waste disposal facility, and did not affect low level burial ground nonconformance as evaluated under 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Wastes. Additionally, there were no radiological consequences (dose) to the public as a result of the shipping manifest error. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Control component, because Entergy did not appropriately coordinate work activities by incorporating actions to address the need for interdepartmental coordination and communication. Specifically, the impact of flushing a reactor water cleanup resin transfer line was not sufficiently communicated or coordinated by all groups to ensure all solid radioactive wastes discharged from the plant into the waste container were accounted for in a subsequent radioactive waste shipment
05000324/FIN-2011004-03Brunswick2011Q3Inadequate Maintenance Results in Containment Isolation Valve FailureA self-revealing Green finding was identified for inadequate maintenance on the overload relay of the unit 2 reactor water cleanup (RWCU) system inlet isolation valve 2-G31-F001. As a result of the inadequate maintenance, the overload relay actuated during operation of the valve under normal conditions, and the valve failed to shut. This was revealed while operators were attempting to isolate the RWCU system on August 2, 2011. After the valve failed to fully shut on August 2, 2011, the licensee shut the valve in series with 2-G31-F001 (2-G31-F004), repaired the overload relay for the 2-G31-F001 valve by installing the correct fasteners, returned the 2-G31-F001 valve to service, and entered the issue into their corrective action program (AR #480063). The inadequate maintenance on the 2-G31-F001 valve overload relay was a performance deficiency. The finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of structure, system, and component (SSC) and Barrier Performance, and it affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the finding prevented a primary containment isolation valve from shutting. This finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for Containment Barriers. The finding was determined to be of very low safety significance (Green) because the finding: 1) did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or the standby gas treatment system, 2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, and 3) did not represent an actual open pathway in the physical integrity of reactor containment. The cause of this finding has no cross-cutting aspect because the maintenance took place in 1992 and is not indicative of current licensee performance.
05000254/FIN-2011003-01Quad Cities2011Q2RWCU Pumps Tripped on Low FlowA self-revealed finding of very low safety significance and associated NCV of 10 CFR 50.65(a)(4) was identified for failure to adequately access and manage risks associated with maintenance activities to prevent plant transients that upset plant stability. On May 31, 2011, after a feedwater flush activity was delayed and rescheduled, operators implementing a clearance order supporting the activity failed to identify a conflict with the reactor water cleanup pumps operating in the decay heat removal mode. When the operators closed the feedwater injection valve and shut off the injection flow path, the reactor water cleanup pumps tripped on low flow. Immediate corrective actions included stopping the feedwater work, opening the feedwater injection valve, and restoring reactor water cleanup flow. The issue has been entered into the licensees corrective action program as Issue Report (IR) 1223075. The inspectors concluded the inadequate assessment and management of risk for the maintenance activity discussed above was a performance deficiency. Failure to identify operational impact and adequately evaluate the risk associated with moving the feedwater clearance activity resulted in tripping the reactor water cleanup pumps and challenging the key shutdown safety function of decay heat removal. This performance deficiency was different from the examples in IMC 0612, Appendix E, Examples of Minor Issues, in that additional reliance on manual actions by operators was required to prevent a more significant challenge to key safety functions. The performance deficiency was more than minor because it could be reasonably viewed as a precursor to a significant event using the minor screening questions of IMC 0612, Appendix B. Inspectors performed the phase 1 assessment, using both Appendix G and Appendix K of IMC 0609, and determined the finding was Green because sufficient equipment was available to meet the core heat removal guidelines, the licensees ability to recover decay heat removal was not significantly degraded, both subsystems of shutdown cooling were inoperable but available, the licensees procedure contained appropriate direction for depressurizing and placing shutdown cooling subsystems in service, and the operators had the appropriate training and briefings to accomplish the required actions in the time required. The inspectors identified that this finding had a cross-cutting aspect in Human Performance Work Control, in that, the licensee failed to appropriately coordinate work activities by incorporating actions to address the impact of changes in the schedule and conflicts between different work activities.
05000416/FIN-2011002-06Grand Gulf2011Q1Inadequate Design Control for the Mitigation Monitoring System ModificationThe inspectors reviewed a self-revealing, Green finding of EN-DC-115, Engineering Change Process, involving the failure to maintain adequate design control measures associated with the installation of the mitigation monitoring system. On November 8, 2010, a reactor coolant pressure boundary failure occurred at the skid mounted Online Noble Chemical - Mitigation Monitoring System pump inside primary containment. The positive displacement sample pump ejected the pump piston from the housing, resulting in an approximate 7 gpm leak of reactor coolant. The steam leak resulted in a reactor recirculation system flow control valve lockup (due to hydraulic power unit motor failure) and approximately 15,000 square feet of contaminated area in the primary containment structure. The licensee failed to ensure proper validation testing for the pump prior to installation. Specifically, the licensee did not ensure that the pump could withstand the operating pressures and temperatures of the system in which it was installed. The licensee removed the mitigation monitoring system from service and isolated the skid from the reactor water cleanup system. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07852. The finding is more than minor because it affects the design control attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Therefore, using inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for LOCA initiators, the inspectors concluded that the finding was of very low safety significance (Green) because the failure of the mitigation monitoring system would not have exceeded technical specifications limits for identified leakage in the reactor coolant system. This finding has a crosscutting aspect in the work practices component of the human performance area; because the licensee failed to adequately oversee the design of the mitigation monitoring system such that nuclear safety is supported. (H.4(c))
05000263/FIN-2010007-01Monticello2010Q4HELB Analysis Potentially Non-ConservativeAs part of the review of the ACE for an adverse trend in double disc gate valve (DDGV) local leak rate testing (LLRT) performance documented in CAP 1202466, the inspectors noted that the ACE had determined that the valves performance degradation did not prevent the valves from performing their safety function. The ACE only addressed the valves safety function of providing containment isolation. The inspectors questioned if the safety function of the high pressure coolant injection (HPCI), reactor core isolation cooling (RCIC) and reactor water cleanup (RWCU) steam supply valves to close after detection of a HELB should have been considered. The licensee responded that the ACE did not need to consider the effect of the valves increased leakage on the HELB analyses because any leakage would not impact the alternate shutdown path. The inspectors reviewed the assumptions and acceptance criteria of the HELB calculations for HPCI, RCIC and RWCU line breaks and identified potential inconsistencies between the calculations assumptions with Technical Specifications and UFSAR allowed values for valve closure times, incorporation of delay actuations, and isolation initiation signals. The licensee entered the NRC concerns with these potential inconsistencies into the CAP by initiating CAP 01252363 on October 1, 2010. The licensee stated that the calculations were appropriate and provided the inspectors with some original licensing documents for the HELB analyses; however, additional questions remained. This issue will be tracked as URI 05000263/201007-01 pending further NRC review of the licensee responses and the HELB analyses and determination of the original and current licensing bases.
05000461/FIN-2010005-02Clinton2010Q4Failure to Perform Preventative Maintenance of Division 1 Self Test System (STS) Power Supply Results in Spurious Repositioning of Safety-Related ValvesA finding of very low safety significance was self-revealed on August 24, 2010, when the Reactor Water Cleanup (RT) System return line outboard primary containment isolation valve went closed. Many other unintended valve repositioning events occurred from August 25 through August 26, 2010. The licensee failed to perform preventive maintenance on the Division 1 Self Test System (STS) safety-related 5 Volt (V) power supply. As a result, a degraded voltage condition existed in the test circuit, which was identified as the cause for the above valve repositioning events. As a corrective action, the licensee has since installed a temporary plant modification of dual 5 V power supplies for all four divisions of the STS. No violation of regulatory requirements was identified. The finding was of more than minor significance because the failure to perform preventative maintenance on critical components, if left uncorrected, would potentially lead to a more significant safety concern. This finding was of very low safety significance based on answering no to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, in the area of resources, the licensee did not adequately maintain long-term plant safety by the maintenance of design margins, minimizing preventive maintenance deferrals, and ensuring maintenance and engineering backlogs are low enough to support safety.
05000254/FIN-2010003-02Quad Cities2010Q2PCIS Relay Common Neutral BrokenA self-revealed finding of very low safety significance and a NCV of Technical Specification (TS) 5.4.1 was identified on April 8, 2010, when a Unit 2 Group III containment isolation signal was received during replacement of a primary containment isolation system (PCIS) relay as a result of a disconnected common neutral wire. Immediate corrective actions for this event included restoration of the reactor water cleanup system and rewiring for the PCIS relay to the proper configuration. The inspectors determined that the licensees failure to identify and provide instructions to mitigate the common neutral during the work planning process was a performance deficiency. The inspectors determined that this finding was cross-cutting in the area of Human Performance, Work Control, because the licensee failed to assess the impact of changes to the work scope during the maintenance activity when plant operating conditions had changed (H.3(b)) The inspectors determined the finding was more than minor because the performance deficiency impacted the Mitigating Systems Cornerstone attribute of Configuration Control for Operating Equipment Lineup to ensure the availability, reliability and capability of safety systems to respond to initiating events to prevent undesirable consequences. The inspectors performed a Phase 1 SDP evaluation. Using IMC 0609, Attachment 4, Table 4a, Mitigating Systems Cornerstone, all questions were answered No, and this finding screened as Green, or having a very low safety significance.
05000237/FIN-2009005-08Dresden2009Q4Procedural Deficiency Causing a Pressure Pulse Resulting in a Reactor Water Level Low-Low Group 1 Isolation Signal and Unit 3 Reactor ScramA self-revealed finding involving a non-cited violation (NCV) of Technical Specification 5.4.1 was identified on October 3, 2009, due to the licensees failure to include essential information in DOP 1200-03, RWCU System Operation with the Reactor at Pressure, Revision 51, regarding startup of the reactor water cleanup system with the reactor at pressure. This procedural deficiency caused a pressure pulse that resulted in a reactor water level Low-Low Group 1 Isolation Signal and Unit 3 reactor scram. This event was entered into the licensees corrective action program (CAP) as Issue Report (IR) 974426. Corrective actions by the licensee included revising procedure DOP 1200-03. This finding was considered more than minor because it affected 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 at power operations. The finding was determined to be of very low safety significance because it did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigating equipment or functions will not be available. This finding has a cross-cutting aspect in the area of Human Performance (Resources) because the licensee did not provide complete, accurate and up-to-date procedures to plant personnel. H.2(c
05000219/FIN-2009003-08Oyster Creek2009Q2Licensee-Identified ViolationTechnical Specification 3.5.A.3, Primary Containment Integrity requires that with one or more of the automatic containment isolation valves inoperable: maintain at least one isolation valve operable in each affected penetration that is open and within 4 hours: restore the inoperable valve(s) to operable status, isolate the affected penetration by use of at least one deactivated automatic valve secured in the isolation position, or isolate each affected penetration by use of at least one closed manual valve or blind flange. Contrary to this, on February 2, Exelon personnel did not identify that automatic containment isolation valve V-16-14(reactor water cleanup (RWCU) heat exchanger inlet isolation valve) was inoperable and take the required actions as stated above. This violation was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of the reactor containment because the redundant PCIV V-16-1 was operable during the time period when V-16-14 was inoperable. This issue is described in corrective action program condition report IR 875329. Exelons corrective actions included briefing operations personnel of this event and reinforcing proper technical rigor is applied when issues are identified
05000373/FIN-2009002-01LaSalle2009Q1Licensee-Identified ViolationTechnical Requirements Manual (TRM) 3.4.a, Structural Integrity, requires the structural integrity of ASME Code class 1, 2, and 3 components shall be maintained in accordance with the ISI and Testing Programs. Technical Specification 3.4.a.1 requires that the structural integrity of ASME l Code class 1, 2, and 3 components be verified in accordance with the ISI Program. Contrary to this, during the L2R12 RFO, the licensee discovered two reactor water cleanup flow element dissimilar metal welds were misclassified and, therefore not volumetrically examined. The flow elements received a system pressure test during L1R12 and L2R11 with no indication of leakage. The flow element was replaced during the U2 current RFO with the U1 element to be replaced during the next outage. Based upon this, the violation was of very low safety significance. The licensee entered this issue into the CAP as IR 865730
05000416/FIN-2008005-04Grand Gulf2008Q4Failure to Correct Leaking Reactor Water Cleanup System Primary Containment Isolation ValvesThe finding was more than minor because it was associated with systems, structures, and components and the reactor coolant system barrier performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has across cutting aspect in the area of human performance associated with resources in that the licensee failed to take actions to correct a long-standing equipment issue associated with excessive leakage from primary containment isolation valves H.2(a). (Section 1R20
05000440/FIN-2008002-02Perry2008Q1Failure to Implement Adequate Configuration Control Affecting \\\'A\\\' Reactor Water Cleanup System(Section 1R13)A finding of very low safety significance and a non-cited violation of Technical Specification (TS) 5.4, Procedures, was self-revealed on January 4, 2008, when reactor steam was observed coming from the from the \\\'A\\\' reactor water cleanup (RWCU) system as operators opened the pump suction shutoff valve. A system isolation valve that was danger-tagged as shut to provide double-boundary protection from the reactor coolant system was found in the open position. At the time of the event, licensee personnel were in the process of restoring the \\\'A\\\' RWCU pump to service following maintenance and the reactor was at rated power and pressure. As part of their immediate corrective actions, licensee personnel isolated the leak, performed a system alignment, and entered this issue into their corrective action program. The finding was considered more than minor because it was associated with the Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power functions. Specifically, the finding resulted in a reactor coolant leak to the safety-related auxiliary building. The finding was determined to be of very low safety significance because the reactor water leak was readily isolable. The primary cause of this finding was related to the cross-cutting area of Human Performance as defined by IMC 0305 H.1(b) because licensee personnel failed to use conservative assumptions in decision making associated with the valve tagging procedure. Section (1R13
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