Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000263/FIN-2018003-012018Q3MonticelloLicensee-Identified ViolationThis violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants; which requires, in part, that equipment qualified by test must be preconditioned by natural or artificial aging to its end of life or a shorter designated life considering all significant types of degradation which can have an effect on equipment function. Contrary to the above, on June 2, 2018, the licensee determined that EQ evaluation 608000000032, of MO2034, MO2035, MO2075, and MO2076 (HPCI and RCIC Steam Line Isolation Valves) internal actuator cables, failed to consider the temperature rise due to the high temperature process fluid in the vicinity of the affected components when aging (preconditioning) them and the unaccounted temperature rise shortened the life of some components to the point that they were no longer EQ qualified to the end of planned life. The unaccounted for process fluid temperature increases were verified by the licensee when thermography of the associated valves was performed. The licensee performed a prompt operability determination, entered the issue into the corrective action program (CAP) as CAP 501000012766 and performed a thermal life analysis engineering evaluation. Long-term corrective actions include replacement of the internal actuator cables during the next refueling outage. 10 Significance/Severity Level: This finding was more than minor because the performance deficiency was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, HPCI and RCIC Steam Line Isolation Valves are designed to provide reactor coolant pressure boundary, required for a safe reactor shutdown following a Design Basis Accident or transient. The finding was of very low safety significance (Green) because it was a design or qualification deficiency, did not involve an actual loss of safety system, did not represent actual loss of a safety function of a single train for greater than its Technical Specification (TS) allowed outage time, and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hrs. Corrective Action Reference: 501000012766
05000263/FIN-2018002-012018Q2MonticelloLicensee-Identified Violation

This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section2.3.2 of the Enforcement Policy.Enforcement: Violation: Title 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements of 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(8) requires, in part, that a licensee must provide and maintain adequate emergency facilities and equipment to support the emergency response plan.Contrary to the above requirements, on March 23, 2018, the licensee identified the site failed to maintain the effectiveness of the emergency plan by not providing and/or maintaining equipment capable of measuring the Immediately Dangerous to Life and Health (IDLH) concentrations for several toxic chemicals as required to properly classify an Alert Emergency Action Level (EAL). Specifically, while performing an emergency equipment inventory, the licensee identified that detector tubes (Draeger tubes) available to measure chlorine gas concentrations were not capable of measuring the IDLH concentration of 10 ppm required to identify the threshold level for classifying an Alert EAL (HA 3.1) since the measurement range of the available sample tubes was 50500 ppm.The inability to properly classify the Alert EAL represented a Loss of Emergency Assessment Capability and resulted in the licensees submission of Event Notification Report # 53298 in accordance with the requirements of 10 CFR 50.72(b)(3)(xiii). An immediate extent of condition review performed by the licensee identified additional deficiencies in adequate sampling methods for determining IDLH concentrations for Butadiene, Ethylene Dichloride, and Gasoline. Additionally, the licensee identified that in April 2015 there was missed opportunity to correct this deficiency when an Emergency Preparedness (EP) Coordinator, performing a Control Room Emergency Equipment Inventory, identified the need to order and replace the existing chlorine detector tubes. The EP Coordinator added the incorrect detector tubes to the existing inventory form without validating the tubes detection range and accuracy to ensure it was capable of detecting the IDLH threshold concentration level of 10 ppm.Upon identification of the issue, the licensee implemented compensatory measures for determining the EAL classification and entered the issue into the corrective action program (CR 501000009876). On May 08, 2018, the licensee implemented the sites new EAL classification procedure that was developed using NEI 9901, Revision 6, which does not require atmospheric sampling (use of detection tubes) for classification of EAL HA 3.1.Significance/Severity Level: Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.81, the inspectors determined this finding was 

10 of very low safety significance (Green) because a significant amount of equipment necessary to implement the E-plan was not available or functional to the extent that any key ERO member could not perform his/her assigned functions, in the absence of compensatory measures (Degraded Planning Standard), specifically the ability to accurately classify the Alert EAL. Determining the finding significance using IMC 0609, Appendix B, Table 5.41, results in the same finding significance (very low significance) since the performance deficiency would have rendered an EAL initiating condition ineffective such that the Alert would have been declared in a degraded manner.Corrective Action Reference: 501000009876, CR Toxic Gas Detector Tube.
05000263/FIN-2018001-022018Q1MonticelloLicensee-Identified ViolationViolation: Title 10 CFR 50.59(d)(1) requires, in part, that the licensee maintain records of changes to the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c).These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to Paragraph (c)(2) of this section.Title 10 CFR 50.59(c)(2)(ii) requires that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the Final Safety Analysis Report (FSAR) (as updated).Technical Specification (TS) 3.3.1.1, Reactor Protection System (RPS) Instrumentation, states the RPS instrumentation for each function in Table 3.3.1.11 shall be operable. As specified in Table 3.3.1.11, Function 5, Main Steam Isolation Valve (MSIV) - Closure (8 channels) and Function 8, Turbine Stop Valve (TSV) Closure (4 channels) are required to be operable in Mode 1. TS 3.3.1.1, Condition C.1 states with one or more functions with RPS trip capability not maintained, to restore RPS trip capability in 1 hour and was applicable to both the MSIV and TSV RPS logic functional testing.Contrary to the above, on March 7, 2009 and July 11, 2009, the licensee failed to perform and maintain a written evaluation as required by 10 CFR 50.59(d)(1) to demonstrate a change to its facility did not require a license amendment. Specifically, the licensee incorrectly concluded in its 10 CFR 50.59 evaluation SCR080319, dated September 29, 2008, that no license amendment was required prior to implementing two surveillance test procedures; 0009 Turbine Stop Valve Closure Scram Test Procedure, Revision 16 on March 7, 2009 and; 0008 Main Steam Line Isolation Valve Closure Scram Test Procedure, Revision 20 on July 11, 2009. The test fixture was applied during quarterly surveillance testing through September 16, 2017.Implementation of procedures 0008 and 0009, respectively, resulted in the loss of RPS trip Function 5 (MSIV) and Function 8 (TSV) by bypassing more than the TS minimum allowed inputs per channel to maintain functionality, thereby violating the requirements of TS 3.3.1.1. Loss of these functions resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR (as updated) as specified by 10 CFR 50.59(c)(2)(ii).On November 14, 2017, the licensee generated CAP 501000005391 after conducting an operating experience evaluation of a similar event at another station concluding the event was applicable to the Monticello Plant. The surveillance procedures were immediately quarantined and subsequently revised on December 8, 2017 and December 11, 2017, to remove the use of the RPS test fixture.Significance/Severity Level:Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined this finding was of very low safety significance (Green) because it did not affect a single RPS trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown.The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV The disposition of this violation closes LER 05000263/201700600.Corrective Action Reference: 501000005391
05000263/FIN-2018001-012018Q1MonticelloFailure to Follow Procedure for Storage of Equipment Near Safety-Related EquipmentThe inspectors identified a finding of very low safety significance (Green) with an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B Criterion V for the failure to accomplish activities affecting quality as prescribed by documented procedures. Specifically, the licensee failed to follow procedure 4 AWI04.02.01, Housekeeping for storage of items or equipment near safety-related equipment. On two separate occasions, the inspectors identified items being stored near safety-related equipment that did not comply with procedure requirements.
05000440/FIN-2018001-012018Q1PerryFailure to Notify the NRC within 60 Days of a Condition Prohibited by Technical SpecificationsThe inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to report a condition that was prohibited by the plants Technical Specifications to the U.S. Nuclear Regulatory Commission (NRC) within 60 days. Specifically, the licensee did not report a condition that, as determined by the NRC, rendered the Division 2 Diesel Generator (DG) inoperable for a period longer than the Technical Specification allowed completion times of its associated required actions.
05000440/FIN-2017003-012017Q3PerryLicensee-Identified ViolationTechnical Specification 5.5.1, states in part, that the ODCM shall contain the conduct of the Radiological Environmental Monitoring Program (REMP). The ODCM, Revision 20, includes Table 5.1 1 ODCM REMP Locations and Section 3.12.1.c, which states in part, With milk or broadleaf vegetation samples unavailable from one or more of the sample locations...identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. Contrary to the above, as of August 11, 2017, substantive changes to the REMP identified by the licensee in 2015 were not incorporated into the ODCM. Specifically, the licensee identified that a milk sampling location was no longer available and that the expansion of broadleaf vegetation sampling was required. Additionally, the licensee relocated collection sites for water and sediment samples that were not reflected in the ODCM. The licensee documented this issue in CR 2017 08353. The inspectors determined that this REMP issue was of very low safety significance (Green) after reviewing IMC 0609, Appendix D, Public Radiation Safety SDP, dated 23 February 12, 2008. The inspectors determined that this finding was associated with the Environmental Monitoring Program, therefore, the finding screened as Green (verylow safety significance).
05000263/FIN-2017002-022017Q2MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of TS 3.7.1, Residual Heat Removal Service Water (RHRSW) System; which requires, in part, that two RHRSW subsystems shall be operable in Modes 1, 2, and 3 or per Condition A, One RHRSW subsystem inoperable; the RHRSW subsystem must be restored to OPERABLE status within 7 days or the applicable conditions and required actions of Limiting Condition forOperations 3.4.7, Residual Heat Removal Shutdown Cooling System Hot Shutdown, for RHR shutdown cooling made inoperable by RHRSW System must be entered. Contrary to the above, on March 27, 2017, the licensee exited the requirements in TS 3.7.1, with a Tag Section still hanging, rendering B RHRSW subsystem inoperable, while in Mode 1. This was identified by the licensee when the maintenance organization notified operations that work was complete, and the Tag Section was released. The licensee reentered TS 3.7.1, Condition A, entered the issue as CAP 1554105 and assigned a Human Performance Event Investigation. A crew clock reset was also taken as well as communicating lessons learned to the entire plant organization.This finding was more-than minor because the performance deficiency wasassociated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, RHRSW System is designed to provide cooling water for the RHR System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident or transient. Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post-accident heat loads, assuming the worst case single active failure occurs coincident with the loos of offsite power. The finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not involve an actual loss of safety system, did not represent actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not represent an actual loss of function of one or more non-Tech Spec Trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hours.
05000263/FIN-2017002-012017Q2MonticelloLow Reactor Water Level During Shutdown of 11 Reactor Feedwater PumpA self-revealed finding of very-low safety significance and a Non-Cited Violationof Technical Specification 5.4.1.a occurred on April 15, 2017, due the licensees failure to establish, implement and maintain procedures regarding shutdown operations. Specifically, Operations Manual B.06.05-05 did not account for the state of the opposite train of feedwater when shutting down the 11 Reactor Feedwater Pump. Licensee use of the inadequate procedure placed equipment in a configuration where no condensate flow path to the reactor existed causing reactor water level to lower to a point where trip/isolation set-points were reached. This caused an unplanned Reactor Protection System (RPS) trip and Partial Group II Isolation. The licensee initiated Corrective Action Program (CAP) 1555785 to document the reactor water level transient, RPS trip and Partial Group II Isolation. Immediate corrective actions includedopening the 11 Reactor Feedwater Pump discharge valve to restore reactor water level allowing reset of the Group II isolation and RPS trip. Subsequent licensee actions included development of expectations via an Operations Memo and revision to Operations Manual B.06.0505 as well as Procedure 2204 and Procedure 2167 to ensure abnormal equipment lineups are addressed such that unexpected procedure interactions are avoided.The inspectors determined the failure to establish, implement and maintain procedures regarding shutdown operations as required by Technical Specification 5.4.1.a was a performance deficiency that required an evaluation. The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, and IMC 0609, Appendix A, Exhibit 1, Section B, and determined a detailed risk evaluation was required because the finding caused a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of feedwater). A Senior Reactor Analyst performed a detailed risk evaluation using bounding assumptions and the change in Core Damage Frequency was calculated to be 9E7/year (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Change Management aspect, because licensee leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000263/FIN-2016007-012016Q4MonticelloInadequate Procedure for Identification of Significant Conditions Adverse to QualityThe inspectors identified a finding of very low safety significance and non-cited violation of Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to prescribe a procedure appropriate to the circumstances with respect to the identification of a significant condition adverse to quality (SCAQ). Specifically, FPPAARP01, CAP Action Request Process, provided an overly restrictive definition of what constituted a SCAQ. Consequently, the failure to provide an adequate definition of a SCAQ could result in a failure to identify a SCAQ and therefore, failure to implement corrective actions that preclude repetitive failures of safety-related equipment. The licensee entered this issue into the CAP as action request (AR) 1536735. The inspectors determined that the licensees failure to prescribe a procedure appropriate to the circumstances under FPPAARP01 was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because, if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Although, this issue could potentially affect each of the Reactor Safety Cornerstones, the inspectors elected to evaluate this issue under the Mitigating Systems Cornerstone because inspectors concluded it impacted the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage) more than the attributes of the other Cornerstones. The inspectors utilized IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding screened as very low safety significance (Green) since the inspectors answered No to each of the questions in Exhibit 2, Section A, Mitigating Systems Screening Questions. The inspectors determined that the performance characteristic of the finding that was the most significant causal factor of the performance deficiency was associated with the cross-cutting aspect of Problem Identification and Resolution, Self-Assessment, and involving the organization routinely conducting self-critical and objective assessments of its programs and practices. Specifically, the failure to identify the overly restrictive definition of SCAQ during previous audits of the CAP was caused by an insufficiently self-critical audit focus.
05000266/FIN-2016004-012016Q4Point BeachScaffolds Constructed Without Required Engineering ApprovalGreen: A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the licensees failure to follow step 4.1.3 (2) of procedure MAAA1001002, Scaffold Installation, Modification, and Removal Requests. Specifically, the licensee failed to obtain and document engineering approval for multiple scaffolds constructed in the cable spreading room that did not meet the separation criteria of Attachment 1 of MAAA1001002. The licensees short-term corrective actions included obtaining the appropriate engineering evaluations for the affected scaffolding and conducting a stand-down and information sharing with the scaffold builders to ensure they were aware of the importance of obtaining engineering approvals. The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, if the licensee continued to construct scaffolding without obtaining required engineering approvals, scaffolding could be constructed that was not seismically qualified and adversely affect the operability of surrounding structures, systems, and components (SSCs). The inspectors concluded this finding was associated with the Mitigating Systems cornerstone. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on October 7, 2016. Specifically, the inspectors used IMC 0609, Appendix A, SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered "No" to the screening questions. This finding has a cross-cutting aspect of Teamwork (H.4), in the area of Human Performance, for the failure of individuals and work groups to communicate and coordinate their activities across organizational boundaries to ensure nuclear safety is maintained. Specifically, the scaffold building team failed to communicate with the engineering organization to ensure the engineering evaluations were complete.
05000266/FIN-2016004-022016Q4Point BeachLicensee-Identified ViolationThe following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. The licensee identified a finding of very low safety significance (Green) and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, due to the failure to properly implement instructions in Work Order (WO) 40461957 for the replacement of the power range nuclear instrument (NI) 1N-43 gain potentiometer vernier dial. Specifically, step 5 of the WO stated, Replace the gain pot vernier with the preset spare. Prevent movement of the potentiometer shaft as much as possible. Contrary to the WO instructions, the technician performing the work believed it was necessary to dial the gain potentiometer to zero before replacing the dial and in doing so caused the 1N-43 NI high flux trip function to become inoperable. This was identified when the control room operators observed the indicated NI power reading for 1N-43 decrease to 82 percent and questioned the technician performing the work about the observed power change. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings required, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, on December 2, 2016, the licensee did not accomplish activities affecting quality in accordance with the documented instructions. Specifically, the licensee did not follow step 5 of the work instructions in WO 40461957, causing the NI high flux trip function for 1N43 to become inoperable. The licensee entered this issue into the CAP as AR 02172378. The inspectors determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016 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 C, Reactivity Control Systems. This resulted in the finding screening as Green.
05000263/FIN-2016004-012016Q4MonticelloLicensee-Identified ViolationWelding Blanket Partially Covered Reactor Building Ventilation Intake (CAP 1539781) The following violation of very-low significance (Green) was identified by the licensee and was a violation of NRC requirements and met the criteria of the NRC Enforcement Policy for being dispositioned as a Non-Cited Violation (NCV). The licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, on October 28, 2016, when the licensee failed to follow procedures while performing activities affecting quality. Specifically, the licensee failed to identify and control modifications of safety-related SSCs in accordance with FPEMOD03; Temporary Modifications, in that operators installed a welding blanket which partially blocked the suction of VAC10A and 10B (intakes of the Reactor Building main supply fans) to prevent welding sparks from being sucked into the intakes and failed to follow steps in that procedure. Procedure FPEMOD03, Revision 13 states, in part, that This procedure shall be applied to Safety-Related SSCs, should be applied to augmented quality or reliability related SSCs, and may be applied to commercial facility changes. Contrary to these requirements, the licensee failed to use FPEMOD03 to evaluate the physical change of installing welding blankets over Safety-Related Reactor Building Ventilation main supply fan intakes for potential plant impact prior to installation. Specifically, this resulted in an increase in negative pressure of the reactor building and an increase of steam chase temperatures which had the potential to upset plant stability by initiating a Group 1 Isolation. This was identified by the licensee during a deliberate observation process by the Shift Manager. Immediate corrective actions included stopping the welding, removing the welding blanket, reducing the steam chase temperature. The licensee documented this issue in the corrective action program (CAPs 1539781, 1541340, and 1541514). The performance deficiency was determined to be more than minor because it adversely affected the Configuration Control attribute of the Initiating Events Cornerstone, with the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as Green based on answering no to the Initiating Events screening questions in inspection manual chapter (IMC) 0609 Appendix A, The Significance Determination Process for Findings at Power, effective July 1, 2012. The issue was entered into the corrective action program as CAPs 1539781, 1541340, and 1541514. The inspectors concluded the issue was licensee-identified based on the guidance in IMC 0612, Power Reactor Inspection Reports, issue date May 06, 2016.
05000263/FIN-2016004-022016Q4MonticelloLicensee-Identified ViolationPast Reactor Building Wide Range Gas Monitor Settings Prevented Transition to Mid/High Range (CAP 1537833) The following violation of very-low significance (Green) was identified by the licensee and is a violation of U.S. Nuclear Regulatory Commission (NRC) requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a Non-Cited Violation. Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) requires, in part, that a holder of a license under this part shall follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E, and the planning standards of Title 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4) requires standard emergency classification and action level scheme, the bases of which includes facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, from June 30, 1994, through September 1, 2016, the licensee failed to maintain the effectiveness of the sites emergency plan and the emergency classification and action level scheme. Specifically, the licensee changed the Engineering Unit Conversion Factor (EUCF) for the Reactor Building (RB) Vent Wide Range Gas Monitor (WRGM), resulting in non-conservative monitor indications that were 13 times lower than the actual effluent levels. The EUCF error impacted the licensees Emergency Plans effectiveness (emergency classification and action level scheme) by reducing the licensees ability to rely on the monitor to identify radiological conditions that exceed the Emergency Action Level (EAL) initiating condition threshold for the declaration of Emergency Classification Levels ranging from an Unusual Event (UE) up to and including a General Emergency (GE). The use of the inaccurate RB Vent WRGM readings would delay the classification of an UE (RA1.2) due to actual effluent levels exceeding the threshold initiating conditions for the respective EAL, while the WRGM was erroneously indicating a much lower value. While this condition would prevent using the RB Vent WRGM for the declaration of an Alert (RA1.2), Site Area Emergency (SAE) (RS1.1), or GE (RG1.1), the licensee remained capable of performing the timely and accurate declaration of an Alert, SAE or GE by monitoring radiological conditions (releases) using the Off-gas Stack WRGM, in accordance with the bases identified in the respective EALs. Consequently, the Alert, SAE, and GE declaration (based on radiological conditions) would not be delayed or missed due to the RB WRGM issue. The NRC determined that since the change in the EUCF would have only prevented the timely and accurate classification of a potential UE (as required by 10 CFR 50.47(b)(4)) the issue was determined to be of a very-low safety significance (Green) as indicated in Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 22, 2015. On September 1, 2016, this issue was identified through the licensees self-assessment process and documented in the CAP as Action Request 01533526, Reactor Building Vent Wide Range Gas Monitor Effluent Channel Reading Non-Conservative. The licensee implemented corrective actions to correct the EUCF for the RB Vent WRGM and restore compliance. As such, the NRC determined this to be a Non-Cited Violation in accordance with Section 2.3.2 of the Enforcement Policy.
05000263/FIN-2016010-012016Q3MonticelloFailure to Plan and Perform maintenance to Correct HPCI Oil LeakA self-revealing finding preliminarily determined to be of low to moderate safety significance (White), and an associated apparent violation of Technical Specification 5.4.1.a, were identified for the licensees failure to plan and perform maintenance affecting the safety-related high pressure coolant injection (HPCI) system in accordance with written documents appropriate to the circumstance as required by Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, improperly planned and performed pre-April 2005 maintenance initiated a crack in a safety-related HPCI oil pipe and, for numerous years, the licensee failed to perform maintenance to resolve repeated identification of HPCI oil leakage. These failures resulted in a sudden increase in oil leakage on March 22, 2016, extending the unavailability of HPCI during a maintenance window and causing a loss of safety function. The licensee documented the issue in the corrective action program (CAP) as CAP 1516361 prior to repairing the oil leak and restoring the HPCI safety function. The inspectors determined that the licensees failure to pre-plan and perform maintenance on safety-related equipment was a performance deficiency; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors determined the issue was more than minor because it adversely impacted the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, improperly planned and performed 2005 maintenance initiated a crack in a safety-related HPCI oil pipe and, for numerous years, the licensee failed to perform maintenance to resolve repeated identification of HPCI oil leakage. These failures resulted in a sudden increase in oil leakage on March 22, 2016, extending the unavailability of HPCI during a maintenance window and causing a loss of safety function. The inspectors applied IMC 0609, Attachment 4, and IMC 0609, Appendix A, Exhibit 2, Section A, for Mitigating Systems to screen this finding and determined a detailed risk evaluation was required because the finding represented a loss of system and/or function. Therefore, a coordinated effort between inspection staff and regional Senior Reactor Analysts (SRAs) was required to arrive at an appropriate risk evaluation for the degraded condition that resulted from the finding. The SRA used the Monticello Standardized Plant Analysis Risk (SPAR) model, version 8.24 for the detailed risk evaluation. This evaluation concluded that the HPCI system was degraded for over 10 years and significantly degraded for approximately 4 months. The system is risk-important and is used to mitigate many internal and external initiating events. The total delta CDF for the 121 day portion of the exposure period is 3.8E6/yr., which is a finding of low to moderate safety significance (White). HPCI is an important high pressure injection system that is used to mitigate internal events, internal flooding, and internal fire events at Monticello. The inspectors determined the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Conservative Bias because licensee individuals failed to use decision-making practices that emphasize prudent choices over those that are simply allowable (H.14). Specifically, licensee Operations and Engineering management did not ensure entry into formal evaluation processes to address a potentially degraded condition for the HPCI oil leaks.
05000263/FIN-2016011-012016Q3MonticelloFailure to Plan and Perform maintenance to Correct HPCI Oil LeakTitle 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, requires in part, that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are corrected. Licensee procedure FP-PA-ARP-01, CAP Action Request Process, Revision 11, Section 1.2 and Revision 42, Section 2.1, required, in part, that conditions adverse to quality are promptly identified and corrected. Contrary to the above, between March 14, 2006 and March 21, 2016, the licensee failed to correct oil leakage from the safety-related HPCI system, a condition adverse to quality. Specifically, as documented in Condition Reports nos. 1018528; 1508130; and 1515945, the licensee initiated a number of work orders and subsequently closed them without any further work performed to correct these conditions adverse to quality, which resulted in gradual degradation and loss of HPCI system safety function. This violation is associated with a White Significance Determination Process finding.
05000263/FIN-2016003-012016Q3MonticelloFailure to Follow Procedures While Performing Activities Affecting QualityInspectors identified a self-revealed finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of Technical Specification 5.4.1.a, on June 24, 2016, when the licensee failed to follow procedures while performing activities affecting quality. Specifically, the licensee failed to accomplish activities affecting quality in accordance with FPGDOC03; Procedure and Work Instruction Use and Adherence, in that operators performed the Standby Gas Treatment (SBGT) A Train, Quarterly Test (025301) and failed to follow steps in that procedure. This resulted in an unanticipated trip of the turbine building ventilation and reactor building exhaust plenum fans causing an increase of steam chase temperatures which had the potential to upset plant stability by initiating a Group 1 Isolation. Immediate corrective actions included restoring ventilation to reduce the steam chase temperature, and entering the issue into the licensees Corrective Action Program (CAP 1526310). The inspectors determined that the licensees failure to follow procedures while performing activities affecting quality was a performance deficiency requiring evaluation. The finding was determined to be more than minor because it adversely impacted the Initiating Events Cornerstone attribute of Human Performance in the area of human error, 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, the failure to follow procedures resulted in conditions that had the likelihood to upset plant stability and challenge critical safety functions, in this case, the potential to initiate a Group 1 Isolation due to high steam chase temperatures. The inspectors evaluated the finding in accordance with IMC 0609 and determined it to be of very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency; Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools (H.12).
05000263/FIN-2016001-012016Q1MonticelloFailure to Use Procedures While Performing Activities Affecting QualityAn NRC identified finding of very low safety significance (Green) and associated of 10 CFR 50, Appendix B, Criterion V; Instructions, Procedures, and Drawings, was identified on February 5, 2016, as a result of the licensees failure to use procedures while performing activities affecting quality. Specifically, the licensee failed to accomplish activities affecting quality in accordance with FP-G-DOC-03; Procedure and Work Instruction Use and Adherence, in that documented procedures were not used to install a conduit support on safety related Emergency Filtration Train (EFT) Division II conduits. Immediate corrective actions included removal of the support and entering the issue into the licensees Corrective Action Program (CAP) 1511349. The finding was determined to be more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the inspectors based this determination on the fact that performing activities affecting quality without using procedures has the potential to adversely affect the design/qualification of a Structure, System, and Component (SSC) or impact the operability or functionality of a system or component. The inspectors determined the finding to have very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, teamwork because of the licensees work group failures to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained.
05000373/FIN-2015002-012015Q2LaSalleBoth Ammonia Detectors Non-Functional on Unit 0 Train A of Control Room Ventilation(Opened) Unresolved Item 05000374/2015002-01: Both Ammonia Detectors Non-Functional on Unit 0 Train A of Control Room Ventilation. The inspectors identified an unresolved item (URI) concerning the failure of both Unit 0 Train A (0A) control room ventilation ammonia detectors. These detectors were previously removed from the TSs by amendment. On June 26, licensee staff identified that both ammonia detectors in the 0A Train of control room ventilation were non-functional. The inspectors identified that Amendments 61 and 42 contained language referencing these detectors. The amendments indicated that a commitment was made with respect to maintenance of these detectors. The licensee maintains a listing of current commitments, but this item was not on that list. The actual commitment predated the implementation of the licensees formal commitment tracking program. The particulars of the commitment were contained in referenced correspondence. A review of the correspondence will allow the inspectors to determine if additional regulatory action is warranted. This issue has been entered into the CAP under AR 2520223, NRC Question on Ammonia Detector Commitment. This issue is a URI pending NRC evaluation of the details found in the correspondence referenced above (URI 05000373/2015002-01, Both Ammonia Detectors Non-Functional on Unit 0 Train A of Control Room Ventilation).
05000374/FIN-2015002-042015Q2LaSalleFailure to Include Limiting Conditions for Operation in the Technical SpecificationsThe inspectors identified a Severity Level IV NCV of 10 CFR 50.36, Technical Specifications, having very low safety significance (Green), for the licensees failure to ensure that limiting conditions for operation (LCOs) were contained in the stations Technical Specifications (TSs). Specifically, as of March 15, 2015, through the Unit 2 Core Operating Limits Report (COLR), Cycle 16, Revisions 1 and 2, the licensee introduced new Operating Limits for Lost Jet Pump Plug Seals Mitigation Strategy, that created new LCOs as defined by 50.36(c)(2) but did not incorporate these LCOs into the TSs. The licensee incorrectly believed that because the COLR was revised via the 50.59 process and the special content that accounted for the existence of the plugs was developed using NRC-approved methodologies, the change was acceptable and no change to the TSs was obtained from the NRC. This finding was considered more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the Unit 2 COLR was revised in a manner that created new LCOs, and further, could have resulted in the operation of Unit 2 outside of its approved TSs and license. Operating the unit in accordance with its NRC-approved TSs could have resulted in the plant operating in an unanalyzed condition that could have resulted in fuel failure. The finding involves the potential for a failed safety/relief valve (SRV) or turbine bypass valve concurrent with complete flow blockage to a peripheral fuel assembly, with a simultaneous breakdown of control room operator knowledge of the special steps required by the COLR revision. Given standard operating practices and the significant amount of extra attention and sensitivity placed on the jet pump plugs and their potential effect, an error that results in licensed operators failing to comply with the restrictive limits of the COLR would be very unlikely. Additionally, a read-and-sign was required of all Unit 2 control room operators and supervisors delineating the special compensatory measures to be taken in the event that a COLR base case component, such as an SRV, were to fail. Further, the inspectors considered the relatively short duration of time (March 1523, 2015) where the plug material parameters were sufficient to cause plugging of an orifice coincident with plant power levels that could challenge the fuel integrity limits. Given these factors, the inspectors determined that the likelihood of a failed COLR base case component, combined with the operation of the unit in an unanalyzed condition in accordance with the NRC-approved TSs, combined with a blocked orifice that could result in fuel clad damage was very low. Given the very low likelihood of the event scenario to occur and the low consequences if it were to occur, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Change Management, because the licensee leaders did not ensure the use of a systematic process for evaluating and implementing change so that nuclear safety remained the overriding priority (H.3).
05000374/FIN-2015002-032015Q2LaSalleInadequate 10 CFR 50.59 Evaluation for Jet Pump Plugs Affecting Fuel Bundle CoolingThe inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.59, Changes, Tests, and Experiments, having very low safety significance (Green), for the licensees failure to provide a written safety evaluation supporting the determination that a license amendment was not required for operation with jet pump seal plugs (lost in the reactor vessel in February 2015 during a refueling outage) that could negatively impact fuel bundle cooling during an anticipated operational occurrence (misplaced fuel bundle). The licensee entered this issue into the corrective action program (CAP) as action report (AR) 02486215 and considered the core operable because additional testing demonstrated that with sufficient time (approximately 11 days) at operating temperature, the rubber plugs would degrade and pass through the affected flow orifices. The finding was determined to be more than minor because the inspectors could not reasonably determine that the activity, to operate with jet pump plugs blocking peripheral fuel bundle flow, would not have required prior NRC approval. Specifically, if the licensee operated a peripheral blocked fuel bundle coincident with a misplaced fuel bundle, the minimum critical power ratio limits/margins may not have been assured. Additionally, this finding was more than minor because the underlying technical issue adversely affected the Barrier Integrity Cornerstone objective of design control and cladding performance. The finding involves the potential for a misplaced fuel bundle concurrent with complete flow blockage to a fuel assembly. Given standard refueling practices, an error that results in plant operation with a misplaced fuel bundle is very unlikely due to strict procedural controls and multiple verifications of fuel assembly placement. In addition, the misplaced fuel assembly would have to be located at a peripheral core location to be susceptible to a jet pump plug that could possibly block bundle cooling and this was very unlikely. Further, the inspectors considered the relatively short duration of time where the plug material parameters were sufficient to cause plugging of an orifice coincident with plant power levels that could challenge the fuel integrity limits. Given these factors, the inspectors determined that the likelihood of a misplaced fuel assembly combined with a blocked orifice that could result in fuel clad damage was very low. Given the very low likelihood of the event scenario to occur and the low consequences if it were to occur, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors identified a cross-cutting aspect associated with this finding in the area of Human Performance, Conservative Bias, because the licensee staff did not use a decision-making practice that emphasized prudent choices over those that are simply allowable (H.14).
05000374/FIN-2015002-022015Q2LaSalleInadvertent Operation of Circuit Breaker Affecting Unit 2 Train A Residual Heat Removal Suppression Chamber Spray Isolation Valve (235Y-2 C3)A finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.4.1, Procedures, was self-revealed when the licensee failed to properly preplan and perform maintenance in accordance with written procedures and instructions appropriate to the circumstances. Specifically, on May 14, 2015, the Work Order (WO 1643222) for testing of the motor for the Unit 2 reactor core isolation cooling (RCIC) water leg pump and involving operation of the motors breaker did not include precautions or restrictions to prevent the inadvertent operation, by bumping, of the adjacent breaker for the safety-related Unit 2 A residual heat removal (RHR) suppression chamber spray isolation valve. Workers inadvertently bumped and opened the breaker for the RHR valve and rendered the system inoperable. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to provide a work order appropriate to the circumstances of the juxtaposed breakers. The subsequent, inadvertent opening of the 2A RHR suppression chamber spray isolation valve breaker, unexpectedly rendered the valve inoperable. This negatively impacted the RHR suppression chamber spray systems ability to reduce suppression chamber pressure by removing one of the required two spray paths. The inspectors determined the finding to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because configuration control and error prevention techniques (robust barriers) in an existing licensee procedure were not appropriately implemented due to the failure of individuals to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12). Specifically, licensee staff failed to implement the guidance found in procedure HU-AA-101, Human Performance Tools and Verification Practices.
05000374/FIN-2015001-032015Q1LaSalleCOLR Revision Potentially Created Non-Conservative Technical SpecificatAs part of the overall review of the Unit 2 Jet Pump Plug issue (described in greater detail in Section 4OA2.4 of this report), the inspectors reviewed the changes made to the Unit 2 COLR, Cycle 16, Revisions 1 and 2. The inspectors assessed the changes with respect to their potential impact on the current licensing basis, i.e., TSs and regulations such as 10 CFR 50.36. In Revision 1 of LaSalles Unit 2 Cycle 16 COLR, the licensee introduced a new section in the form of an Appendix, entitled Operating Limits for Lost Jet Pump Plug Seals Mitigation Strategy. This appendix states The following limits apply while the jet pump plug peripheral bundle blocked orifice condition exists. Specifically, item 4 entitled Other Requirements, states in part that All equipment must be in-service. This includes the EOOS (equipment out-of-service) assumed in the Base Case mentioned in Footnote 1 of COLR Section 10 EXCEPT LPRMs (local power range monitors) and TIPOOS (traversing in-core probe out-of-service) (...) In the event of an EOOS, take action in accordance with TS 3.2.2 ACTION statements. Those TS actions were to Reduce THERMAL POWER to < 25% RTP (rated thermal power) within a 4-hour completion time. The equipment referenced in the COLR Section 10 Base Case that have associated TS LCOs are safety relief valves (SRVs) (LCOs 3.4.4 and 3.5.1) and turbine bypass valves (TBVs) (LCO 3.7.7). LCO 3.4.4 states The safety function of 12 SRVs shall be OPERABLE. Unit 2 has a total of 13 SRVs, so this LCO essentially allows one SRV to be OOS indefinitely with no further action required; however, since the COLR created a new operational restriction to prohibit any SRVs from being OOS in order to maintain the unit in an analyzed condition, the inspectors questioned the apparent non-conservatism that the COLR created for LCO 3.4.4. Specifically, under an identical condition of 1 SRV OOS, the COLR would have required the unit to downpower to less than 25 percent power, while the TSs would have allowed continuous operation at full power. LCO 3.5.1 states (...) the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE. Unit 2 has a total of 7 ADS SRVs, so this LCO essentially allows one ADS SRV to be OOS indefinitely with no further action required; however, since the COLR created a new operational restriction to prohibit any SRVs from being OOS in order to maintain the unit in an analyzed condition, the inspectors questioned the apparent non-conservatism that the COLR created for LCO 3.5.1. Specifically, under an identical condition of 1 ADS SRV OOS, the COLR would have required the unit to downpower to less than 25 percent power, while the TSs would have allowed continuous operation at full power. LCO 3.7.7 states The Main Turbine Bypass System shall be OPERABLE. OR LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR), limits for an inoperable Main Turbine Bypass System, as specified in the COLR, are made applicable. The Cycle 16 COLR Base Case was analyzed to allow 2 TBVs to be OOS without taking any further action or incurring any operational penalty; however, since the COLR created a new operational restriction to prohibit any TBVs from being OOS in order to maintain the unit in an analyzed condition, the inspectors questioned the apparent non-conservatism that the COLR created for LCO 3.7.7. Specifically, under an identical condition of 2 TBVs OOS, the COLR would have required the unit to downpower to less than 25 percent power, while the TSs would have allowed continuous operation at full power. This issue is considered a URI pending additional internal discussion with the NRC Office of Nuclear Reactor Regulation to seek guidance on whether the above examples classify as LCOs and further, how NRC Administrative Letter 9810 may apply.
05000373/FIN-2015001-042015Q1LaSalleBreakers Installed Beyond DesignEarly in plant life, Westinghouse HFB breakers were installed in various systems onsite, including safety-related applications. The design life of these breakers was 20 years, according to Westinghouse. A portion of these breakers remained in-use at LaSalle and were installed more than 30 years ago. The licensee monitored breaker performance and tested these breakers routinely. Records show that failures of this type of breaker were researched and all of the previous failures at LaSalle have been attributed to manufacturing defects. While the lack of degradation-related failures for these breakers supported the disposition under the 50.65 Maintenance Rule, at this time, it is not clear that a lack of such failures provided adequate basis for an extension to the previously established design life. This issue is unresolved pending the inspectors review of additional information regarding the details of the components qualifications as noted in the original purchase order, in order for the inspectors to evaluate the acceptability of the licensees current method of extending the life of the component.
05000373/FIN-2015001-062015Q1LaSalleLicensee-Identified ViolationTitle 10 CFR 50.72(b)(3)(xiii) states, in part, a licensee shall report (notify the NRC as soon as practical, and in all cases within 8 hours of the occurrence) any event that results in a major loss of emergency assessment capability. Contrary to this requirement, on March 24, 2015, the licensee identified a failure to submit a report for the loss of emergency assessment capability when the site declared seismic monitoring instrumentation inoperable. Specifically, on January 28, 2015, the Instrument Maintenance Department discovered the seismic monitoring program on the seismic laptop computer in the auxiliary electrical equipment room was not running; thereby, preventing the seismic monitoring instrumentation from providing indications required for emergency assessment of a potential seismic event. The system degradation would have adversely impacted the sites ability to declare an ALERT Emergency Action Level in accordance with EPAA1005, Radiological Emergency Plan Annex for LaSalle Station, in the event of an earthquake of sufficient magnitude. The licensee entered the issue into the CAP as AR 02473472, Need to Assess Seismic Monitor Reportability, and conducted an extent of condition review for the prior 3-year period. The licensee identified a total of six times in which the seismic monitoring system experienced this degradation, and the licensee failed to submit an event report at the time, as required by 10 CFR 50.72(b)(3)xiii). Upon completion of the extent of condition review, the licensee initiated AR 02474658, Emergency Notification System Notification Required for Past Seismic Monitor Inoperative, and submitted the required notification to the NRC on March 26, 2015, to restore compliance (Event Number 50926, Seismic Monitor Not Available for Emergency Plan Assessment ). The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the generic communications input that 10 CFR 50.72 reports serve. Since the issue impacted the regulatory process, it was dispositioned through the Traditional Enforcement Process. The inspectors determined that this issue was a Severity Level IV violation based upon Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, Example d.9 in the NRC Enforcement Policy. Example d.9 specifically states, The licensee fails to make a report requirement by 10 CFR 50.72, or 10 CFR 50.73. Because the issue was entered into the licensees CAP (as AR 02473472 and AR 02474658), the violation is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000373/FIN-2015001-012015Q1LaSalleLiquid Penetrant Testing Procedure Was Not Qualified for Its Full Applicability RangeThe inspectors identified a Green NCV of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure, as of February 13, 2015, to properly qualify a non-destructive testing procedure in accordance with applicable codes. Specifically, a liquid penetrant test (PT) procedure was not qualified for its full applicability temperature range in accordance with American Society of Mechanical Engineers (ASME) Code, Section V, Non-Destructive Examination. The licensee entered this issue into its corrective action program (CAP) as Action Request (AR) 02451872. The failure to qualify a liquid PT procedure in accordance with ASME Section V was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, since the liquid PT procedure was not qualified for its full applicability temperature range, liquid penetrant examinations would not be assured to detect flaws in the unqualified temperature range. As a consequence, the potential would exist for a rejectable flaw to go undetected affecting the operability of the affected system. This finding affected the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality; thus, the inspectors answered No to all of the screening questions. Specifically, the licensee review completed liquid penetrant examination records and did not find an example where the procedure was implemented at the unqualified temperature ranges. The inspectors determined that the primary cause of the failure to properly qualify the PT procedure was related to the Problem Identification and Resolution cross-cutting area, Operating Experience aspect (P.5). Specifically, the organization failed to effectively implement external operating experience in a timely manner.
05000373/FIN-2015001-022015Q1LaSalleFailure to Measure Interpass TemperatureThe inspectors identified a Green NCV of Title 10, CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, for a failure of the licensee on February 12, 2015, to measure the interpass temperature while performing welding on the 2 diesel generator cooling water (DGCW) piping system. Consequently, welding was performed without the Code-and procedure-required interpass temperature being monitored on a number of welds, a parameter, which could have affected the mechanical properties of the material being welded. To restore compliance, the welders proceeded to measure the interpass temperatures on the balance of the welds, and verified that the interpass temperature did not exceed that allowed by procedure. The licensee entered this issue into its CAP as AR 02451583. The inspectors determined that this issue was a performance deficiency that was more than minor because it had the potential to lead to a more significant safety concern. Specifically, absent NRC inspector intervention, the welders would have completed all of the welds without having measured the interpass temperature, a welding parameter which can affect the mechanical properties (e.g., impact properties) of some materials being welded, and, if left uncorrected could lead to a potential failure of the weld in service. The inspectors determined this finding was of very low safety significance (Green) because the DGCW system maintained its operability or functionality. The welders proceeded to measure the interpass temperatures on the balance of the welds, and verified that the interpass temperature did not exceed that allowed by procedure, and the issue did not result in the actual loss of the operability or functionality of a safety system. The inspectors determined that the primary cause of the failure to measure the interpass temperature while performing a manual welding process was related to the cross-cutting area of Human Performance, Procedure Adherence aspect (H.8). Specifically, the welders failed to follow procedures.
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Don-20Krause-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"