Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000282/FIN-2018002-012018Q2Prairie IslandResults of ISFSICask Array Dose Calculation Not Incorporated into FSARPrairie Island ISFSI FSAR, as updated, Revision 18, Section A7A.7 evaluates off-site dose rates for an array of ISFSI casks. In this dose rate calculation, explicit modeling credit is given to the earthen berm that surrounds the Prairie Island ISFSI as discussed in Section A7A.7.1. The earthen berm provides radiation shielding for the ISFSI. This calculation allows the licensee to demonstrate, in part, compliance with Title 10 of the Code of Federal Regulations (CFR) 72.104(a) which requires, in part, that, During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ. Calculation TN40HT0502, TN40HT Far Field Shielding Calculations, Revision 0, was performed by the licensee in support of a License Amendment Request (LAR) to modify the Prairie Island ISFSI TN40 cask design (designated as TN40HT casks). The TN40HT LAR was submitted to the NRC by the licensee on March 28, 2008. This dose rate calculation does not credit the earthen berm and, in part, also allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a). The licensee also provided this calculation directly to the NRC in a February 29, 2012, letter in response to a Request for Supplemental Information (RSI) from the NRC associated with the license renewal application for the Prairie Island ISFSI. Although the results from calculation TN40HT0502 for a single cask was incorporated into the Prairie Island ISFSI FSAR, Revision 18, in Tables A7A.22 and A7A.61, the results from TN40HT0502 for an array of casks which, in part, allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a), has not been incorporated into the ISFSI FSAR, Revision 18.Title 10 CFR 72.70, Safety analysis report updating requires, in part, that (a) Each specific licensee for an ISFSI shall update periodically, as provided in paragraphs (b) and (c) of this section, the FSAR to assure that the information included in the report contains the latest information developed (b) Each update shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last update to the FSAR under this section. The update shall include the effects of: (2) All safety analyses and evaluations performed by the licensee in support of approved license amendments.This Unresolved Item is being opened to determine whether or not the licensee is required to update the ISFSI FSAR with the results of calculation TN40HT0502 for an array of casks in accordance with 10 CFR 72.70.Planned Closure Action: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Material Safety and Safeguards to determine whether or not calculation TN40HT0502 is subject to the FSAR updating requirements of 10 CFR 72.70 for the Prairie Island ISFSI.
05000529/FIN-2017001-012017Q1Palo VerdeFailure to establish station procedure instructions for denial work authorizationsThe inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the failure to establish procedure instructions for work authorization denials or deferrals. Specifically, this led to a 60 day extended unavailability of the diverse auxiliary feedwater actuation system when corrective maintenance was inappropriately deferred by the operations department. Failure to provide adequate procedural guidance in the event of a denied work authorization, a circumstance anticipated to occur, is a performance deficiency. The performance deficiency is more than minor, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability and reliability of equipment that responds to an initiating event. Specifically, because the corrective maintenance was not performed in a timely manner, both trains of the diverse auxiliary feedwater actuation system remained in bypass for an additional 60 days whereby the system was not capable of performing its required safety function. The inspectors evaluated the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, Section A, Question 2, which required a detailed risk evaluation because the finding involved a loss of system safety function. A Region IV senior reactor analyst performed a detailed risk assessment of the finding and determined that the finding was of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect in the human performance area of Work Management. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the Unit Operations Managers decision to deny the work authorization was based on conservative but faulty assumptions, and if other work groups with greater specific technical knowledge had been involved, the corrective maintenance likely would have proceeded (H.5)
05000382/FIN-2016004-012016Q4WaterfordFailure to Ensure Appropriate Post-Maintenance Testing on Essential Chiller BGreen. A self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, occurred because the licensee did not assure that the procedures for post-maintenance testing of activities affecting quality included appropriate quantitative or qualitative acceptance criteria for determining that maintenance activities were satisfactorily accomplished. Specifically, the licensee did not assure that post-maintenance testing of essential chiller B would identify inappropriately assembled guide vanes, following maintenance on April 11, 2016, resulting in the unexpected inoperability of essential chiller B on August 12, 2016. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2016-05155. The corrective action taken to restore compliance was to issue work instructions for post-maintenance testing of the essential chillers that ensures the guide vanes respond to load changes. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform maintenance with procedures appropriate to the circumstances resulted in the inoperability of essential chiller B. The inspectors determined the significance of the finding using NRC Inspection Manual Chapter 0609, Significance Determination Process. Using Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green) because all the screening questions in Exhibit 2, Mitigating Systems Screening Questions, were answered No. The finding had a cross-cutting aspect in the area of human performance, teamwork, because the licensee did not ensure that individual and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, electrical and mechanical maintenance personnel did not communicate and coordinate work to ensure that the guide vane arm and actuator linkage were assembled appropriately (H.4).
05000263/FIN-2015001-012015Q1MonticelloFailure to Identify High Pressure Coolant Injection (HPCI) Seismic Support NonconformanceThe inspectors identified a finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify conditions adverse to quality, such as deficiencies, deviations, and nonconformances. Specifically, on February 11, 2015, the inspectors identified a safety related seismic support for high pressure coolant injection (HPCI) turbine trip instrumentation that was not rigidly attached, supported, and restrained in accordance with plant construction code and installation specifications, a nonconformance which the licensee had failed to identify since initial plant construction. Corrective actions for this issue included repairs to the seismic support to rigidly connect the instrument line restraint and installation of a standalone support for the instrument tray. This issue was entered into the licensees corrective action program (CAP 1465906). The inspectors determined that the failure to promptly identify an HPCI instrument line support nonconformance was a performance deficiency requiring evaluation. The inspectors determined that the issue was more than minor because it adversely impacted the Mitigating Systems Cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined that it was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, and the aspect of Identification because the licensee failed to implement a CAP with a low threshold for identifying issues (P.1).
05000263/FIN-2015001-052015Q1MonticelloTwo Emergency Diesels Inoperable Due to Human ErrorA self-revealing finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified on December 28, 2014, due to the failure to properly implement Procedure 0187-02B, 12 Emergency Diesel Generator /12 ESW (Emergency Service Water) Monthly Pump and Valve Tests. Specifically, operations personnel failed to comply with Step 42 which directed the 12 EDG local governor control switch to be lowered to idle setting. The failure to implement the actions directed by Step 42 resulted in the 11 EDG being inoperable. Corrective actions for this issue included procedure revisions to require: protection/flagging of redundant equipment when technical specification equipment is declared inoperable for any reason, including planned maintenance and surveillance; peer checking or concurrent verification for manipulation of operable technical specification related equipment; and all equipment manipulations require a hard match (between procedure and equipment labeling). This issue was entered into the licensees corrective action program (CAP 1460675). The issue was more than minor because if left uncorrected, the failure to properly implement procedures associated with safety-related equipment would have the potential to lead to a more significant safety concern. Specifically, the failure to follow procedure resulted in the 11 EDG being made inoperable coincident with the 12 EDG being inoperable. The inspectors utilized IMC 0609 and determined that the issue was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency aspect because of a failure of individuals to implement error reduction tools (H.12).
05000263/FIN-2015001-062015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.1, Offsite Dose Calculation Manual, (ODCM) which requires in part, that licensee initiated changes to the ODCM shall be effective after approval of the plant manager. Contrary to the above, ODCM01.01 Revision 6 and ODCM02.01 Revision 10, were not approved by the plant manager prior to implementation. This was identified by the licensee as part of the self-assessment process. The licensee documented this issue in the corrective action program (CAPs 1455999 and 1462092). This finding was determined to be of very-low safety significance (Green) because it was not a failure to implement an effluent program and public dose did not exceed Appendix I of 10 CFR 20.1301(e) criteria.
05000263/FIN-2015001-072015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.11 which requires in part, that the Primary Containment Leakage Rate Testing (LRT) Program shall be in accordance with the guidelines contained in RG 1.163, Performance-based Containment Leak-Test Program, dated September, 1995. RG 1.163 directs use of ANSI/ANS56.81994, Containment System Leakage Testing Requirements as an acceptable testing standard. ANSI/ANS56.81994 states, in part that for pressure decay testing, temperature shall be recorded at the start and end of each test, and the leakage rate shall be calculated using a specific formula which incorporates this temperature data to temperature-compensate the volume lost. Contrary to these requirements, the licensees Containment Leakage Rate Testing Program failed to include direction to take temperature data and perform temperature compensation, which resulted in a failure to perform testing in accordance with the ANSI standard and RG 1.163. Specifically, during this time, the licensee failed to correctly perform pressure decay testing for approximately 44 containment penetrations, including the Personnel Airlock. Upon discovery, engineers performed a bounding engineering analysis which verified the containment barrier remained operable but nonconforming and entered the issue into the corrective action program (CAPs 1463917 and 1465869). The performance deficiency was more than minor because the issue is associated with the barrier performance reliability attribute of the Barrier Integrity cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that the physical containment barrier protects the public from radionuclide releases. Specifically, the repeated failure to ensure containment leakage testing met technical specification and regulatory requirements was programmatic, affected multiple components, adversely affected LRT test accuracy, and consequently impacted the licensees ability to verify the containment barrier remained operable. The finding was of very low safety significance because the finding did not represent an actual open pathway in the physical integrity of the containment barrier and did not result in a loss of containment barrier operability. (Green)
05000263/FIN-2015001-082015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Criterion V which requires in part, that activities affecting quality be prescribed by procedures appropriate to the circumstances. Contrary to this requirement, between May 22, 2011 and February 5, 2014, MNGP startup instructions and procedures, C.1 Startup Procedure, 2167 Plant Startup, and 0118 Reactor Vessel Temperature Monitoring, were not appropriate to the circumstances. Specifically, during this time these procedures allowed reactor coolant system pressure to be decreased below 0 psig seven times during reactor startup activities, which was outside of the pressure parameter inputs to the analysis that is the basis for the pressure/temperature limit curves of TS 3.4.9. The licensees analysis showed that there was no impact on RPV integrity due to the existence of the partial vacuum conditions. This issue was identified by the licensee as a result of an operating experience review. The licensee entered this issue into the corrective action program (CAPs 1425020 and 1427529) and initiated action to revise the PTLR limits and submit them for NRC review. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Procedure QualityRoutine Operations Performance, and had the potential to adversely affect the associated cornerstone objective of providing reasonable assurance that a physical design barrier, the reactor coolant system, protects the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance because analysis determined that there was no change in risk to the RCS boundary due to the performance deficiency. (Green)
05000263/FIN-2015001-092015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.47(b)(14) and 10 CFR Part 50, Appendix E, Section IV.F.1. In part, Title 10 CFR 50.47(b)(14) states, Periodic exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. Additionally, Title 10 CFR Part 50, Appendix E, Section IV.F.1 states, The program to provide for: (a) The training of employees and exercising, by periodic drills, of emergency plans to ensure that employees of the licensee are familiar with their specific emergency response duties, and (b) The participation in the training and drills by other persons whose assistance may be needed in the event of a radiological emergency shall be described. The Monticello Emergency Plan, Section 8.1.2.4, describes the required demonstration periodicity for drill and exercises. Contrary to the above, on January 1, 2015, the licensee failed to perform four emergency preparedness drill objectives at the required frequency listed in the Monticello Emergency Plan, Section 8.1.2.4. Specifically, Objectives 11.01, 11.03, and 11.04 were required to be performed annually and were not performed in 2014. Additionally, Objective 11.04 was required to be performed semi-annually and was only performed once in 2014. All missed objectives were associated with radiological exposure controls. The NRC determined that the failure to comply with the established drill and exercise program was a degradation of a planning standard function in accordance with 10 CFR 50.47(b)(14) and was a very low safety significance issue (Green) as indicated in IMC 0609, Emergency Preparedness SDP, Appendix B, Attachment 2, Failure to Comply Significance Logic. The licensee entered this issue in the corrective action program (CAP 1463920). As such, the NRC determined this to be an NCV in accordance with Section 2.3.2 of the Enforcement Policy.
05000263/FIN-2015001-022015Q1MonticelloFailure to Maintain Fire Protection Program Procedures for Control of Portable Heater/Extension Cord Fire HazardsA finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1.d was self-revealed when the licensee failed to maintain procedures for Fire Protection Program Implementation to ensure that ignition sources (space heaters) were properly controlled to prevent plant fires. Specifically, on January 26, 2015, the licensee failed to maintain Fire Protection Program implementation procedures to include controls to ensure space heaters used in the plant stayed within allowable load ratings and were plugged directly into outlets without the use of extension cords. This resulted in a fire in the plant recombiner building which was extinguished within 13 minutes, nearing the 15 minute time limit at which a Notification of Unusual Event (NOUE) would have needed to be declared. It also resulted in a space heater causing an overloaded outlet at a location in the reactor building, near A residual heat removal (RHR) equipment. Upon discovery of the recombiner area fire, the licensee dispatched the fire brigade to ensure the fire was extinguished, performed extent of condition walkdowns in the plant, and took action to improve controls on extension cord and portable heater use in the power block. This issue was entered into the licensees corrective action program (CAP 1463506). The inspectors determined that the failure to maintain fire program procedures to ensure ignition sources (space heaters) were appropriately controlled was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor because, if left uncorrected, the failure to adequately control portable heater related fire hazards in the plant could lead to more significant safety concerns. In addition, the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined that it was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Evaluation aspect because of the failure to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000263/FIN-2015001-032015Q1MonticelloFailure to Maintain a Standard Emergency Action Level Scheme for FloodingThe inspectors identified a finding of very low safety significance and an NCV of Title 10 CFR 50.54(q)(2) and 10 CFR 50.47(b)(4) for the licensees failure to maintain the effectiveness of the emergency plan. Specifically, from May 28, 2014, until February 26, 2015, the HA1.6 Emergency Action Level (EAL) threshold was in conflict with the EAL basis for the alert classification. Additionally, both the revised EAL threshold and original NRC-approved safety evaluation report EAL threshold were later found to be greater than the actual river level that could lead to damage of safe shutdown equipment. The licensees corrective actions documented that the current river level was 906 and if flooding were to occur the licensee would have relied on Procedure A.6, "Acts of Nature," and that an event response team would have been formed to monitor river level during the duration of a flood event. The licensee concluded that the shift manager, Event Response team, and plant management would have monitored for indication of degraded performance of equipment or structures necessary for safe shutdown for event classification escalation to the Alert level. The licensee entered this issue into the Corrective Action Program (CAP 1454593). The inspectors determined that establishing a flooding EAL threshold that was in conflict with approved EAL basis as required by 10 CFR 50.47(b)(4), and subsequent failure to determine the actual level that could lead to damage of safe shutdown equipment for the alert classification High River Level EAL HA1.6 was a performance deficiency. The inspectors determined that the issue was more than minor because it is associated with the Procedure Quality attribute of the Emergency Preparedness (EP) cornerstone and adversely affected the cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined that it was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Evaluation aspect because the licensee did not thoroughly evaluate the identified engineering error issue to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000263/FIN-2015001-042015Q1MonticelloInadequate Evaluation of Operating Crew During Simulator AssessmentThe inspectors identified an URI on March 16, 2015, due to the licensees potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. In accordance with IMC 0612, Power Reactor Inspection Reports, the inspectors determined that this issue represented an URI because more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. On March 16, 2015, the NRC inspectors observed a potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. Specifically, during an NRC observation of a Licensed Operator Training self-assessment and emergency preparedness objective demonstration, the inspector observed that the evaluators may not have adequately critiqued a knowledge deficiency in the Interpreting and Diagnosing Events competency area when evaluating a Shift Managers (SM) performance. The Shift Managers performance could have adversely impacted EAL classification during a graded self-assessment. This assessment included an evaluated Drill/Exercise Performance (DEP) opportunity for the EAL classification in question. During the inspectors observation, they noted that the critique session did not appear to adequately probe why the classification-related performance weaknesses occurred, and did not appear to determine a course of specific actions for the crew to take to improve individual performance relative to the SMs role in the EAL classification. Specifically, the inspectors noted that at the end of the critique, this item was not discussed as an item needing resolution, nor was it discussed that the SM had a challenge to his qualifications and needed potential remediation, which appeared to be contrary to the sites MTCP0349 procedure. These discussions and follow-up actions did not take place until after the critique had concluded and the NRC inspectors raised questions about the SMs misinterpretation of Safety Parameters Display System (SPDS) and his overall performance. This item represents an issue of concern about which more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. The NRC inspectors will work to obtain additional guidance and clarification/interpretation of the existing guidance in order to resolve this issue. Corrective actions for this issue included disqualifying the individual, developing a remediation plan, and initiating procedure changes to improve the critique process. This issue was entered into the corrective action program as CAP 1470975. (URI 05000263/201500104, Inadequate Evaluation of Operating Crew During Simulator Assessment)
05000263/FIN-2014005-022014Q4MonticelloIncorrect Emergency Action Level ThresholdAn Unresolved Item (URI) was identified because additional information is needed to determine whether a performance deficiency exists and if a violation of 10 CFR 50.54(q)(2) occurred. The inspectors identified an issue of concern associated with the licensees changing of the High River Level EAL threshold from 921 to 920 for the alert classification EAL HA1.6. Description. During the first quarter of 2014, the licensee made a change to EAL HA1.6 for High River Level. Specifically, the licensee changed the threshold for the Alert classification from 921 to 920. On November 4, 2014, the NRC questioned the reason for the EAL threshold change, noting that the change may be in conflict with the EAL basis for HA1.6. These questions prompted licensee discovery that the EAL threshold basis was associated with flooding impacts on plant equipment, rather than river level historical data, as the licensee originally believed. The inspectors observed that the basis for EAL HA1.6 was linked to the river level where flood waters would reach the top of the retention basin. The inspectors also noted that although the licensee had changed the EAL threshold, the actual level of the basin was not altered. The licensee then questioned if the known level of the retention basin was a legacy error and what the correct level was for this EAL threshold. To address these questions, the licensee requested input from engineering and documented these issues in Action Request (AR) 01454593 on that same date. As an interim action, AR 01454593 documented that the current river level was 906, and if flooding were to occur, the licensee would rely on Procedure A.6, Acts of Nature, and an event response team would be formed in accordance with the procedure to monitor river level during the duration of a flood event. The licensee noted that at a river level of 918, a Notification of Unusual Event would be declared. In addition, the licensee concluded that the shift manager, event response team, and plant management would monitor for indication of degraded performance of equipment or structures necessary for safe shutdown for event classification escalation to the Alert level. The inspectors evaluated these interim compensatory measures and found them adequate as no additional reasonable risk existed as a result of this issue. On December 3, 2014, NRC questions regarding the progress of the previous AR led to the licensees statement that the 920 level also may not be correct. Because the licensee had not yet determined the appropriate High River Level EAL threshold for the alert classification EAL HA1.6, the inspectors could not readily determine whether the error was a legacy issue with the old threshold value, a current performance issue with the new threshold value and EAL change process, or both. The interim compensatory measures identified in the previous AR remained in effect at the conclusion of this inspection and the December 3, 2014 discussions and URI determination resulted in the generation of AR 01458209 by the licensee on that same date Therefore, a URI was identified because additional information on the correct High River Level EAL threshold is needed for the inspectors to determine whether a performance deficiency existed and if a violation of 10 CFR 50.54(q)(2) occurred. (URI 05000263/201400501; Incorrect Emergency Action Level Threshold)
05000382/FIN-2014003-012014Q2WaterfordFailure to Control Entry into a High Radiation AreaThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638. The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12).
05000382/FIN-2014003-022014Q2WaterfordFailure to Maintain Adequate Public Address System to Implement Onsite Protective ActionsThe inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10). The licensee implemented compensatory measures while the system was being restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860. The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings; and the corresponding Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes (P.2).
05000263/FIN-2014003-022014Q2MonticelloOperation Outside of Reactor Coolant System Pressure and Temperature LimitsAn URI associated with TS 3.4.9, Reactor Coolant System Pressure and Temperature Limits, was identified. Technical Specification 3.4.9 requires, in part, that RCS pressure and RCS temperature shall be maintained within the limits specified in the PTLR, which requires that RCS pressure remain at or above 0 psig. Between May 22, 2011, and February 5, 2014, Monticello RCS pressure was decreased below 0 psig several times during reactor startup activities. During an operating experience review in April 2014, the licensee noted that a vacuum of approximately3 psig had been drawn on the RPV six times, and in one case a vacuum of17.5 psig was drawn, which was outside of the limits specified in the PTLR. The licensees analysis showed that there was no impact on RPV integrity due to the existence of the partial vacuum conditions. The licensee entered this issue into the CAP and initiated action to revise the PTLR limits and submit them for NRC review. Inspectors reviewed the results of the licensees operating review and decided that additional information was needed, including insights from NRRs more generic resolution to industrywide issues regarding TS PTLR limits, to determine whether a performance deficiency exists.
05000382/FIN-2014002-012014Q1WaterfordFailure to Establish Procedures for Using the Alternate Emergency Fuel Oil Storage Tank Fill LineThe inspectors identified a non-cited violation of Technical Specification 6.8.1.a because the licensee did not establish written procedures to fill the diesel fuel oil storage tanks for their emergency onsite power sources. Specifically, the licensee did not establish procedures to fill the fuel oil storage tanks for the emergency diesel generators using the credited safety-related, seismic category 1 emergency fill line. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2014-00636. The immediate corrective action taken to restore compliance was to develop procedures to fill the emergency diesel generator fuel oil storage tanks using the safety-related, seismic category 1 emergency fill line and evaluate other alternative methods. The inspectors concluded that the failure to establish procedures to fill the fuel oil storage tanks for the emergency onsite power sources was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it reduced the licensees reliability and capability to fill the fuel oil storage tanks for the onsite power sources following a loss of offsite power or extreme weather event (e.g., a seismic or flooding event) that may last longer than seven days. The inspectors performed the initial significance determination for this issue. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate the issue. The initial screening directed the inspectors to use Inspector Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Section A, to determine the significance of the issue. The finding required a detailed risk evaluation because the performance deficiency could have resulted in a loss of safety function (onsite ac power) because the system may not have remained operable for its 30- day design basis accident mission time. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation for this issue. The analyst determined that the finding was of very low safety significance (Green) because the diesel generators would have remained functional for the 24-hour probabilistic risk assessment mission time. This detailed risk evaluation used the shorter mission time because after 24 hours, the NRC assumed that the licensee had substantially more resources available to help mitigate the accident. The dominant core damage sequences included longer-term loss of offsite power events and the common cause failure of the diesel generators due to potential problems to refill the diesel fuel oil storage tanks after seven days. The relatively long period prior to ultimate diesel generator failure helped to minimize the risk. Additionally, the finding was not a significant contributor to the large early release frequency. The inspectors concluded that the finding reflected current licensee performance and involved an avoiding complacency cross-cutting aspect of the human performance area in that the licensee did not recognize and plan for the possibility of mistakes, latent issues and inherent risk, even while expecting successful outcomes.
05000382/FIN-2014002-022014Q1WaterfordFailure to Replace an Essential Chiller Oil Pump prior to the Vendor Service LifeA self-revealing, non-cited violation of Technical Specification 6.8.1.a. occurred because the licensee did not develop a preventative maintenance schedule to inspect or replace an item that has a specific lifetime. Specifically, the licensee did not develop a preventative maintenance schedule to inspect or replace the essential chiller oil pump motors prior to exceeding their duty life. As a result, the essential chiller oil pump B motor failed in-service. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2014-0095. The immediate corrective action taken to restore compliance was to issue an action request to establish the periodic replacement of the essential chiller oil pumps prior to the end of their vendor recommended service life. The inspectors concluded that the failure to develop a preventative maintenance schedule to inspect or replace the essential chiller oil pump motors prior to the end of the vendor provided duty life was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it affected the availability and reliability of the essential chillers to provide a heat sink for the removal of process and operating heat from selected safety-related equipment during design basis accidents. The initial screening directed the inspectors to use Inspector Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Section A, to determine the significance of the finding. The inspectors categorized the finding as having very low safety significance (Green) because the finding did not affect the design or qualification of the system; it did not represent a loss of the system or function; and, the loss of the essential chiller was less than the technical specification allowed outage time. The inspectors also concluded that the finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency occurred more than 3 years ago, and did not reflect current licensee performance.
05000382/FIN-2014002-032014Q1WaterfordFailure Implement a Fire Protection Program Procedure to Perform an Evaluation for Transient CombustiblesThe inspectors identified a non-cited violation of License Condition 2.C.9 because the licensee did not implement Procedure EN-DC-161, Control of Combustibles, which requires, in part, that a transient combustible evaluation shall be processed or compensatory actions shall be established if a flammable liquid exceeds one pint in an approved container. Specifically, the licensee did not implement Section 5.6 of Procedure EN-DC-161 after a fuel oil leak from the standby fuel oil pump for the train B emergency diesel generator exceeded one pint in an approved container which eventually failed to hold the fuel oil while in service. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2013-6020 and CR-WF3-2013-06123. The immediate corrective actions taken to restore compliance was to remove the leaking fuel oil from around the emergency diesel generator, implement an hourly fire watch, and repair the standby fuel oil pump leak and returned the emergency diesel generator to an operable status on January 3, 2014. The inspectors concluded that the failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform a transient combustible evaluation when a flammable liquid above one pint in an approved container was present in the B emergency diesel generator room prevented the licensee from implementing required compensatory measures in response to the presence of transient combustibles surrounding and on the B emergency diesel generator. In addition, similar to NRC Inspection Manual Chapter 0612, Appendix E, Section 4, Example k, of a more than minor violation, the failure of the leak collection device resulting in fuel oil around the B emergency diesel generator represented a credible fire scenario involving transient combustibles that could affect equipment important to safety. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. Since this finding was related to controls for transient combustible materials, the initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, to determine the significance of the finding. The inspectors categorized the finding under Task 1.4.1, Fire Prevention and Administrative Controls, and qualitatively screened it as very low safety significance (Green) because the impact of the fire finding was limited to no more than one train of equipment important to safety. The inspectors concluded that the finding reflected current licensee performance and involved a conservative bias cross-cutting aspect in the human performance area in that the licensee did not use decision making practices that emphasized prudent choices over those that are simply allowable.
05000382/FIN-2014002-042014Q1WaterfordFailure to Establish Adequate Design Control Measures for the Selection and Review for the Suitability of Application of Molded Case Circuit BreakersA self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, occurred because the licensee did not establish design control measures for the selection and review for the suitability of application of a molded case circuit breaker that was essential to the safety-related function of a shutdown cooling heat exchanger fan cooler. Specifically, the licensee did not select and review for the suitability of the correct safety-related circuit breaker for the application to provide circuit fault protection to the train B shutdown cooling heat exchanger air handling unit fan motor. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2013-02316 and CR-WF3-2013-04644. The immediate corrective action taken to restore compliance included the replacement of the breaker with a breaker more suitable for the application to protect the air handling unit fan motor. The planned corrective actions included an extent of condition review for other installed breakers and the revision of work order instructions to eliminate the practice of substituting and using the factory acceptance testing for pre-installation and post-maintenance tests, respectively. The inspectors concluded that the failure to establish design control measures for the selection and review for suitability of application for the correct safety-related circuit breaker was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect breaker affected the availability, reliability, and capability of the shutdown cooling heat exchanger fan coolers to remove heat from the shutdown cooling heat exchanger areas following a design basis accident. The inspectors performed the initial significance determination. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Section A, to determine the significance of the finding. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety-related equipment for longer than the technical specification allowed outage time. The total exposure period was 23 days. The allowed outage time was 7 days. A Region IV senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was 5E-13/year (Green). The dominant core damage sequences included loss of offsite power events, failure of both trains of containment spray, and the failure of a pressurizer safety relief valve to remain closed. The equipment that helped mitigate the risk included the emergency diesel generators and the essential feedwater systems. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect of avoiding complacency in the human performance area because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk on relying on 21 year old vendor information and installing a breaker without pre-installation and adequate post-maintenance testing.
05000263/FIN-2014002-072014Q1MonticelloBoth Secondary Containment Access Doors Briefly Opened SimultaneouslyOn September 18, 2013, while performing the secondary containment airlock door interlock surveillance test, the interlock to the main plenum room did not prevent the opening of both doors to the plenum room airlock (DOOR85 and DOOR86). With the outer door to the main plenum room open, the inner door was able to be opened. The plenum airlock doors were then closed. The operator attempted a second time to verify interlock functionality. This time the inner door was opened, and again the interlock did not prevent the opening of the outer door. The plenum airlock doors were immediately closed. The total time both doors were opened was estimated to be less than 10 seconds. With both doors open, TS SR 3.6.4.1.3 was not met and secondary containment was declared inoperable. Secondary containment was declared operable, after independently verifying that at least one secondary containment access door was closed. Inspectors reviewed the LER and decided that additional information was needed to determine whether a performance deficiency exists for the event. In order to close this Unresolved Item (URI), the inspectors intend to review the sites recently performed evaluation aimed at removing this issue from being counted in the Safety System Functional Failure PI. In addition, the inspectors will factor in any insights from NRRs more generic resolution to industry wide secondary containment issues.
05000382/FIN-2013008-012013Q4WaterfordFailure to Establish an Adequate Test Program to Demonstrate that the Train B Emergency Diesel Generator Exhaust Fan Would Perform Satisfactorily In ServiceA self-revealing apparent violation of 10 CFR Part 50, Appendix B, Criteria XI, Test Control, occurred because the licensee failed to establish an adequate test program to demonstrate that a safety-related component associated with the train B emergency diesel generator would perform satisfactorily in service. Specifically, the licensee failed to identify and perform adequate testing on the train B emergency diesel generator exhaust fan to demonstrate that the exhaust fan would perform satisfactorily in service. The test did not incorporate requirements and acceptance limits contained in applicable design documents such as the Final Safety Analysis Report, as updated. As a result, the licensee failed to ensure that for all operational tests that the safety-related exhaust fan would perform satisfactorily such that it would provide sufficient flow and remove heat during accident conditions. The licensee entered this condition into its corrective action program as Condition Report CR-WF3-2013-02530. The immediate corrective actions taken to restore compliance included the replacement of the train B emergency diesel generator exhaust fan assembly. The planned corrective actions included the review of the emergency diesel generator ventilation system monitoring plan. The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed the initial significance determination for the diesel generator room ventilation fan failure. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings. The finding required a detailed risk evaluation because it involved a potential loss of one train of safety-related equipment for longer than the technical specification allowed outage time. The emergency diesel generator needed the ventilation exhaust fan to remain operable. The unit was not recoverable. The total exposure period was 25 days. The allowed outage time was 72 hours. The analyst determined the best estimated change to the core damage frequency was 4.4E-6/year. This finding was preliminarily characterized as low to moderate safety significance (White). The dominant core damage sequences included loss of offsite power events, leading to station blackout, coincident with the failure of the turbine driven auxiliary feedwater pump. Equipment that helped to mitigate the risk included recovery of an emergency diesel generator or manually starting a temporary emergency diesel generator set. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the resource component of the human performance area in that the licensee did not have complete, accurate and up-to-date operational surveillance test procedure.
05000382/FIN-2013005-012013Q4WaterfordFailure to Critique Weaknesses During an Evaluated ExerciseThe inspectors identified a non-cited violation of 10 CFR Part 50.47(b)(14) for the failure to identify deficiencies resulting from the licensees 2013 biennial evaluated exercise. Specifically, the licensee did not identify as part of the critique process two examples of failure to provide a range of protective actions for emergency workers. First, actions were not taken to minimize radiological dose for one in-plant repair team; second, the licensee did not perform habitability evaluations to determine the suitability for continued use of emergency response facilities during the simulated radiological emergency. This violation was entered into the licensees corrective action program as Condition Reports CR-WF3-2013-5895 and CRWF3-2013-5905. The failure to identify weaknesses occurring in an exercise is a performance deficiency. The performance deficiency is more than minor because it is associated with the emergency response organization performance attribute of the emergency preparedness cornerstone and it adversely impacted the objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In addition, if left uncorrected, continuing these behaviors could result in unnecessary radiological dose to emergency workers and the public in an actual event. Using NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green). The finding had a cross-cutting aspect in the corrective action program component of the problem identification and resolution crosscutting area because the licensee failed to thoroughly evaluate two issues during the exercise critique process (P.1(c))
05000382/FIN-2013005-032013Q4WaterfordLicensee-Identified ViolationTitle 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to the above, prior to September 5, 2013, the licensee did not establish test program requirements that incorporated acceptable limits to ensure that safety-related auxiliary component cooling water temperature control valves ACC-126A and ACC-126B would perform satisfactorily in service when manually closed to conserve wet cooling tower inventory following a loss of coolant accident. The licensee identified this issue during a review of a degraded trend in the auxiliary component cooling water high point header pressure for the automatic closure function. As part of troubleshooting effort, the licensee identified that the operational surveillance procedure used to conduct the manual closure function of the valve did not incorporate requirements and acceptance limits contained in applicable design documents. Specifically, Procedure OP-903-118, Primary Auxiliaries Quarterly IST Valve Test, Sections 7.12 and 7.13 did not contain requirements or acceptable limits to ensure that the manual closure function for the valves would maintain the required high point header pressure contained in applicable design documents. The licensee entered this condition into their corrective action program as CR-WF3-2013-04290 and CR-WF3-2013-04324. A senior reactor analyst performed a bounding detailed risk evaluation and determined that the finding was of very low safety significance (Green).
05000382/FIN-2013005-022013Q4WaterfordProtective Action Recommendations Under Conditions of Changing Wind Vectors Not Consistent with Federal GuidanceThe inspectors identified an unresolved item related to the adequacy of the licensees guidelines for the choice of protective actions during an emergency in accordance with the requirements of 10 CFR 50.47(b)(10). Specifically, the licensees implementation of guidelines for extending existing protective action recommendations into additional geographical areas of the emergency planning zone under conditions of changing wind vectors may not be consistent with the guidance of EPA-400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents. The inspectors observed during the December 4, 2013, evaluated exercise that the licensee expanded an existing protective action recommendation for the public into a geographical area for which protective actions may not have been warranted. Specifically, with an existing recommendation of evacuate all sectors within two miles of the reactor and to five miles in three downwind sectors, the licensee subsequently expanded the five-mile recommendation to an adjacent (fourth) downwind sector following a wind vector change. The licensee applied deterministic, plant-condition based criterion in expanding the five-mile evacuation recommendation. The expansion of the protective action recommendation into a fourth sector may not have been warranted because the licensee had valid dose assessments showing that protective action guides were not exceeded at two miles in the newly-affected sector at the time when wind direction changed. The licensee did not apply those results in making the decision to expand protective action recommendations. The inspectors concluded the licensees recommendation was not based on EPA guidance, which states, in part, that protective action guides are the approximate levels at which protective measures are justified, and that evacuation is seldom warranted at less than 1 rem Total Effective Dose Equivalent. The inspectors identified that Procedure EP-002-052, Protective Action Guidelines, Revision 23, allows the licensee to generate evacuation protective action recommendations for members of the general public in areas of the emergency planning zone where radiological protective action guides are not exceeded. The procedure required the user determine the plant is in a stable condition before allowing the application of radiological assessment results when wind vectors change. Specifically, Step 5.4.1.1(A) required as a precondition that plant conditions are well understood and changes can be reasonably predicted, and Step (B) that radiological releases have a high degree of predictability in terms of isotopic composition...and release rate. The inspectors observed that licensee staff understood Step (A) to require that core damage had been arrested and that plant conditions precluded any future change in core state with a high degree of confidence. The inspectors observed that licensee staff used deterministic plant-based protective action recommendations instead of radiological based assessments in expanding protective action recommendations because core damage had not been arrested and future changes in core state were not precluded with a high degree of confidence. In addition, licensee staff also concluded that future release rates were not predictable. This issue was identified as an unresolved item because the NRC has not determined whether the licensee has adequately implemented Planning Standard 10 CFR 50.47(b)(10), which states, in part, that ...guidelines for the choice of protective actions, consistent with Federal Guidance, are developed and in place.... Specifically, the NRC has not determined whether the restrictions on the application of radiological assessments in Procedure EP-002-052, Revision 23, Step 5.4.1.1, adequately implement the guidance of EPA-400-R-92-001. No additional information is required from the licensee. The licensee has entered this issue into their corrective action program as Condition Report CR-WF3-2013-5900. This issue is identified as URI 05000382/2013005-02, Protective Action Recommendations Under Conditions of Changing Wind Vectors Not Consistent with Federal Guidance.
05000382/FIN-2013004-022013Q3WaterfordFailure to Implement Fire Protection Program Procedure Requirements When Securing from a Fire WatchThe inspectors identified a non-cited violation of Waterfords Facility Operating License Number NPF-38, License Condition 2.C.9, because the licensee did not implement fire protection procedure FP-001-014, Duties of a Fire Watch. Specifically, the licensees fire watch personnel did not implement Section 6.5 of FP-001-014 to remove firefighting equipment from work areas when securing from a fire watch. As a result, multiple undercharged fire extinguishers were left in a fire area. The inspectors determined that this would affect safety-related equipment because it would delay the response to fires in the fire areas. The licensee entered this condition into their corrective action program as CR-WF3-2013-03398 and CR-WF3-2013-03523 for resolution. The immediate corrective actions taken to restore compliance included the removal of all undercharged fire extinguishers from deactivated posts and returning them to their proper storage location. The failure to implement a fire protection program procedure was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to remove undercharged fire extinguishers from work areas that contained safe shutdown equipment could hinder responses to fires in the fire area. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix F, Fire Protection Significance De termination Process, to determine the significance of the finding. The inspectors determined that the finding had a low degradation rating because it reflected a fire protection program element whose performance and reliability would be minimally impacted. Specifically, in all cases identified, there were permanent fully charged portable fire extinguishers of the proper type nearby. Therefore, the finding was of very low safety significance (Green). The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work practices component of the human performance area in that the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported.
05000382/FIN-2013004-032013Q3WaterfordFailure to Accomplish Activities Affecting Quality on a Degraded Safety-Related Solenoid Valve In Accordance With Procedure RequirementsThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not accomplish activities affecting quality on a degraded safety-related train B component cooling water (CCW) bypass valve (CC-134B) in accordance with maintenance procedure EN-MA-101, Fundamentals of Maintenance. Specifically, the licensee did not control and perform testing on a leaking solenoid valve related to the operation of a safety-related bypass valve (CC-134B) after maintenance personnel removed the degraded equipment from service as required by Section 5.10 of EN-MA-101. As a result, the licensee could not characterize and determine the cause of the leakage for the safety-related valve. The inspectors determined that this would challenge the safety function of the valve to provide CCW to the ultimate heat sink following a tornado event. The licensee entered this condition into their corrective action program as CR-WF3-2012-05991, CR-WF3-2012-06288, and CR-WF3-2013-04047. The immediate corrective actions taken to restore compliance included the installation of a new valve and debriefing personnel about controlling equipment removed from service when combining preventative and corrective maintenance tasks in one work order. The failure to control failed equipment removed from the plant to determine the cause in accordance with maintenance procedure requirements was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System 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 degraded condition challenged the safety function of the valve (CC-134B) to limit the loss of CCW through damaged portions of the dry cooling tower fans following a tornado-generated missile strike. The inspectors used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it was potentially risk significant for an external event (tornado). Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The senior reactor analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 3E-7/year. The finding was not significant with respect to the large early release frequency. The dominant core damage sequences included tornado induced losses of offsite power, failure of the dry cooling tower pressure boundary, failure to isolate the damaged dry cooling tower, and failure to recover instrument air. The redundant train A component cooling water system combined with the tornado frequency helped to reduce the risk exposure. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the work control component of the human performance area in that the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of changes to work scope or activity on plant and human performance.
05000382/FIN-2013004-042013Q3WaterfordLicensee-Identified ViolationTitle 10 CFR, Part 55.9, Completeness and Accuracy of Information, requires that information provided to the Commission by an applicant for a license or by a licensee ... shall be complete and accurate in all material respects. Contrary to the above, on September 13, 2012, an NRC Form 396, Certification of Medical Examination by Facility Licensee, was submitted to the NRC for a licensed operator applicant with inaccurate information. Specifically, a restriction for corrective lenses was omitted, even though the applicants medical exam stated that the individual required corrective lenses. An operating license was granted by the NRC to the individual without a corrective lens restriction. The error was identified during the operators subsequent annual medical examination in July 2013, after which the operator reported to licensing that an additional restriction was being placed on his license though his vision had not changed. The licensee confirmed that the operator had not performed any licensed duties and a revised NRC Form 396 was submitted to Region IV on July 29, 2013. The licensee documented the deficiency in Condition Report 2013-03181. The submission of inaccurate information to the NRC is a violation. The violation was evaluated using the traditional enforcement process because it impacted the NRCs ability to perform its regulatory function. The violation was determined to be Severity Level IV because it fits the example of Enforcement Policy Section 6. 4.d.1(d), Violation Examples: Licensed Reactor Operators. This section states, SL IV violations involve, for example ... an individual operator who met ANSI/ANS 3.4, Section 5, as certified on NRC Form 396, required by Title 10, Part CFR 55.23, but failed to report a condition that would have required a license restriction to establish or maintain medical qualification based on having the undisclosed medical condition. In this case, the individual operator did report the condition to the licensee, but the licensee failed to include that information in its original license application to the NRC.
05000285/FIN-2013008-402013Q2Fort CalhounFailure to Obtain Prior NRC Approval for a Facility ChangeThe team identified a non-cited violation of 10 CFR 50.54(q)(2) for the licensees failure to maintain the effectiveness of an emergency plan. Specifically, since May 14, 2009, the licensee failed to maintain a proper value for low river level associated with the declaration of an emergency at the ALERT classification level. The licensee did not maintain a standard emergency action level scheme in accordance with the requirements of 10 CFR 50.47(b)(4), which states in part, that a standard emergency classification and action level scheme is in use by the nuclear facility licensee. The emergency action level scheme was not maintained because emergency action levels HU1 and HA1, Natural or destructive phenomena affecting the Protected Area, contained an inaccurate river level of 973 feet 9 inches. The river level was inaccurate because the basis document, Procedure EPIP-OSC-1, Emergency Classification, Revision 46, stated the emergency action level was based on the minimum elevation of the raw water pump suction. Based on available plant data, the minimum elevation of the raw water pump suction was above the Alert declaration point of 973 feet 9 inches. This issue has been entered into the corrective action program as Condition Reports CR2013-04198 and CR 2013-04169. This performance deficiency is more than minor, and therefore a finding, because it is associated with emergency response organization performance attribute of the Emergency Preparedness Cornerstone and affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, inaccurate emergency action levels degrade the licensees ability to implement adequate measures to protect public health and safety. The finding was evaluated using the Emergency Preparedness Significance Determination Process, and was determined to be of very low safety significance (Green) because the finding was not a lost or degraded risk significant planning function. The planning standard function was not degraded because the Notification of Unusual Event and Alert emergency classifications would have been declared although potentially in a delayed manner. This finding was not assigned a cross-cutting aspect because the performance deficiency is not reflective of current performance
05000285/FIN-2013008-412013Q2Fort CalhounInappropriate Modification of Turbine Driven Auxiliary Feedwater Pump Back Pressure Protection TripThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with an inappropriate modification of the auxiliary feedwater system. Specifically, from April 2011 through February 2013, measures established by the licensee did not assure that the modification to remove the turbine driven auxiliary feedwater pumps exhaust back pressure trip, properly considered and addressed the open configuration of the pumps exhaust piping to prevent blockage of the exhaust piping. This issue has been entered into the corrective action program as Condition Report CR 2013-05026, and an immediate operability determination was performed. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected, the continued practice of modifying the facility without evaluating for adverse impacts had the potential to lead to a more significant safety concern. Specifically, unevaluated modifications to the facility could introduce adverse changes that result in systems not able to perform their intended safety function which would not be recognized. This finding was associated with the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-422013Q2Fort CalhounFailure to Make Timely Event Notifications for Unanalyzed ConditionsThe team identified four examples of a non-cited violation of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, for the licensees failure to make required event notifications within 8 hours following discovery of an event requiring a report. Specifically, on April 12, 2012, February 7, 2013, February 25, 2013, and February 27, 2013, the licensee failed to notify the NRC within 8 hours of the occurrence an event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. This issue has been entered into the corrective action program as Condition Report CR 2013-05070. The violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy, because the failure to required event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. The team determined that a cross-cutting aspect was not applicable to this performance deficiency because the failure to make a required report was strictly associated with a traditional enforcement violation
05000285/FIN-2013008-432013Q2Fort CalhounRepetitive Issues Involving Untimely Submittal of Required Licensee Event ReportsThe team identified nine examples of a non-cited violation of 10 CFR 50.73, Immediate Notification Requirements for Operating Nuclear Power Reactors, for the licensees failure to make required licensee event reports within 60 days following discovery of an event requiring a report. Specifically, on nine occurrences between May 9, 2011, and August 30, 2012, the licensee failed to submit a licensee event report for an event meeting the requirements for reporting specified in 10 CFR 50.73. This issue has been entered into the corrective action program as Condition Report CR 2012-03796. The violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. The team determined that a cross-cutting aspect was not applicable to this performance deficiency because the failure to make a required report was strictly associated with a traditional enforcement violation
05000285/FIN-2013008-392013Q2Fort CalhounFailure to Properly Implement Applicable ASME OM Code RequirementsThe team identified two examples of a non-cited violation of 10 CFR 50.55a.(f)(4)(ii), Codes and Standards, associated with the licensees failure to properly implement applicable code requirements for in-service testing of safety-related pumps and check valves. Specifically, prior to March 11, 2013, the licensee failed to ensure that the testing of safety-related pumps and valves met the requirements of the American Society of Mechanical Engineers Operation and Maintenance Code. The applicable Code for the current in-service test program is the 1998 Edition through the 2000 Addenda. This issue has been entered into the corrective action program as Condition Reports CR 2013-04680, CR 2013-05018, CR 2013-05514, and CR 2013-05569. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions addressed the causes
05000285/FIN-2013008-382013Q2Fort CalhounDeficient Evaluation for Known Degraded Conditions - AFW Pumps Discharge Check Valve Leakage and Potential Overpressure of AFW Pump Suction PipingThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to properly evaluate a known degraded condition regarding the auxiliary feedwater pump discharge check valve leakage and potential over-pressurization of the pumps suction piping. Specifically, from October 10, 2012, to March 15, 2013, the licensee failed to properly evaluate concerns regarding the auxiliary feedwater pump discharge check valves which resulted in the failure to implement adequate corrective actions to verify leak tightness of the check valves and prevent potential over pressurization of the pumps suction piping. This issue has been entered into the corrective action program as Condition Reports CR 2013-04806 and CR 2013-05018. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate it is unsafe in order to disapprove the action
05000285/FIN-2013008-372013Q2Fort CalhounImproper Storage of the RAW Water to Auxiliary Feedwater Emergency Tank Fill HoseThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to properly store the raw water to emergency feedwater storage tank fill hose. Specifically, from July 1996 to February 27, 2013, the licensee failed to provide adequate instructions or procedures to ensure proper storage and temperature qualification of the auxiliary feedwater emergency fill hose. This issue has been entered into the corrective action program as Condition Report CR 2013-52276. This performance deficiency is more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-362013Q2Fort CalhounDeficient Evaluation of NRC Bulletin 88-04, Strong Pump Weak Pump Due to Failure to Consider the Effect of AFW Pumps Discharge Check Valves LeakageThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to properly evaluate NRC Bulletin 88-04, Potential Safety-Related Pump Loss, regarding the auxiliary feedwater pumps. Specifically, from November 28, 2010, through February 2013, the licensee failed to properly evaluate NRC Bulletin 88-04, for strong pump, weak pump, interaction regarding auxiliary feedwater pumps FW-6 and FW-10. The evaluation failed to consider pump-to-pump interaction that may result due to pump discharge check valve leakage. In addition, the licensee failed to re-evaluate the condition after surveillance testing performed on November 28, 2010, and September 1, 2012, identified leakage past both pump discharge check valves. This issue has been entered into the corrective action program as Condition Reports CR 2013-04680 and CR 2013-04806. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that appropriate corrective actions were promptly implemented
05000285/FIN-2013008-352013Q2Fort CalhounFailure to Correct Condition Adverse to Quality Associated with Corrective Action Program Procedures and the Operability ProcessThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to implement corrective actions to address inadequate procedures involving the degraded/nonconforming condition evaluation and operability determination process. Specifically, prior to March 1, 2013, the licensee failed to correct the procedural inadequacies associated with Procedure FCSG-24-3, Condition Report Screening, Revision 3, as identified in the root cause analysis for Condition Report CR 2012-09494. This issue has been entered into the corrective action program as Condition Report CR 2013-04380. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected, inadequate corrective action program procedures could become a more significant safety concern. This finding is associated with the Mitigating Systems Cornerstone. Since the finding was discovered while in a shutdown condition, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to implement a corrective action program with a sufficiently low threshold. Specifically, although the licensee identified significant flaws in Fort Calhoun Station procedures while performing the root cause analysis for Condition Report CR 2012-09494, the licensee failed to initiate the appropriate corrective action documents to drive the necessary procedure changes
05000285/FIN-2013008-342013Q2Fort CalhounFailure to Follow ASME Code Requirements When Establishing New Pump Reference Values as Corrective ActionsThe team identified a non-cited violation of 10 CFR 50.55a, Codes and Standards, for the failure of the licensee to follow the ASME Code when establishing new reference curves as corrective action to address the performance of component cooling water pump AC-3A within the low required action range of the in-service testing program. Specifically, on July 29, 2011, the licensee failed to follow ASME Code, Subsection ISTB 6200(c), in that, the new reference curves were established without performing an analysis which included verification of the pumps operational readiness at a pump level and a system level, without determining the cause of the change in pump performance, and without an evaluation of all trends indicated by available data. The team confirmed that while the pump was inoperable from an in-service testing perspective during this period, required surveillance testing showed that pump flows and differential pressures were still sufficient to meet the assumptions used in the Fort Calhoun Station safety analysis. This issue has been entered into the corrective action program as Condition Report CR 2013-04010. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since this finding was discovered during plant shutdown and involved plant equipment needed during shutdown conditions, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to fully evaluate the degraded performance of component cooling water pump AC-3A to ensure that resolutions correctly addressed causes of the degraded performance and the cumulative impact on system operational readiness
05000285/FIN-2013008-332013Q2Fort CalhounInadequate Operability Determination Due to Failure to Establish Component Cooling Water System Leakage CriteriaThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, involving the licensees failure to follow procedures when evaluating the impact of component cooling water system leakage on the containment air coolers. Specifically, on October 6, 2010, and December 29, 2010, the operability determinations for Condition Reports CR 2010-04955 and CR 2010-06905 were not performed in accordance with Procedure NOD-QP-31, Operability Determination Process, Revision 43-44, Step 4.1.3 J, and consequently, failed to evaluate the impact of component cooling water system leakage on containment air coolers operability. This issue has been entered into the corrective action program as Condition Report CR 2013-05630. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective action program component. Specifically, the team identified that the licensee failed provide an adequate technical discussion such that a reasonable expectation of operability was demonstrated for containment air coolers with known leakage in the component cooling water system
05000285/FIN-2013008-322013Q2Fort CalhounMultiple Examples of Inadequate RISK-BASED Operability DeterminationsThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, involving multiple examples of the licensees use of probability or probabilistic risk assessment when performing operability determinations. The use of probability or probabilistic risk assessment when determining operability is contrary to Procedure NOD-QP-31, Operability Determination Process, Revision 49-53. Specifically, on January 26, 2012 and twice on February 21, 2013, the operability determinations performed for Condition Reports CR 2012-00626, CR 2013-03839, and CR 2013-03842 used probability and/or probabilistic risk assessment to justify the operability of structures, systems, and components. This issue has been entered into the corrective action program as Condition Reports CR 2013-05590, CR 2013-05466, and CR 2013-05597. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved inadequate operability determinations that occurred while in a shutdown condition and involved plant equipment needed during shutdown conditions, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee failed to use conservative assumptions in decision making when performing operability determinations. Specifically, the licensee proposed that a degraded/nonconforming condition was safe by relying on a non-conservative assumption that an event such as a tornado generated missile or external flooding at the site were not likely to occur
05000285/FIN-2013008-312013Q2Fort CalhounMultiple Examples of Operability Determinations That Lacked Adequate Technical JustificationThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, involving multiple examples of the licensees failure to perform an adequate operability determination as required by Procedure NOD-QP-31, Operability Determination Process. In each example, the team identified that the operability determination lacked adequate technical justification for why the structure, system, or component was operable with the degraded or nonconforming condition. Specifically, on January 24, 2012, June 6, 2012, December 27, 2012, January 22, 2013, and February 5, 2013, the operability determinations for Condition Reports CR 2012-00580, CR 2012-04973, CR 2012-20806, CR 2013-00907, and CR 2013-02260 were not performed in accordance with Procedure NOD-QP-31, Revision 49-53, Step 4.1.3 J. This issue has been entered into the corrective action program as Condition Reports CR 2013-08343, CR 2013-05596, CR 2013-08590, CR 2013-04163, and CR 2013-05353. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involving inadequate operability determinations occurred while in a shutdown condition, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective action program component. Specifically, the team identified that the licensee failed provide an adequate technical discussion such that a reasonable expectation of operability was demonstrated for several degraded or nonconforming conditions
05000285/FIN-2013008-302013Q2Fort CalhounEvaluation of Change to Alternate Shutdown Cooling FlowpathThe team identified an unresolved item related to engineering change modifications that changed a procedure to include the replacement of automatic actions with manual actions. Specifically, the10 CFR 50.59 evaluation proposed a change to substitute automatic flow control of shutdown cooling flow and temperature with manual control using the low pressure safety injection loop injection valves alternate shutdown cooling flow control. During a review of Engineering Change Modification 54058, Procedure Change to Allow Closing of HCV-335 while on Alternate Shutdown Cooling, the team identified that the licensee changed a procedure to include the replacement of an automatic action with a manual action. Specifically, the engineering change proposed to close both shutdown cooling heat exchanger isolation valve HCV-335 and flow control valve FCV-326 while pinning open valve HCV-341 and manually throttling low pressure safety injection loop injection valves to maintain the desired RCS temperature and flow rate. The team questioned whether the licensee required prior NRC review and approval to make this change since flow control valve FCV-326 normally controls temperature and flow automatically as described in Section 9.3.4.3 of the USAR. The licensee entered this issue of concern into the CAP. Additional NRC review and follow up will be required to determine if this issue represents a performance deficiency associated with meeting the 10 CFR 50.59 requirement of more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component important to safety previously evaluated in the USAR. This item is unresolved pending review of the licensees evaluation. This issue is identified as URI 05000285/2013008- 30, Evaluation of Change to Alternate Shutdown Cooling Flowpath.
05000285/FIN-2013008-282013Q2Fort CalhounFailure to Perform an Evaluation for a Change to Component Cooling Water MAKE-UPThe team identified a Severity Level IV non-cited violation of 10 CFR Part 50.59, with an associated Green finding, because the licensee failed to perform an evaluation for a design change that may have required NRC review and approval. Specifically, from June 2008, the licensee did not evaluate a change that would permanently substitute manual actions for an automatic action to add water and nitrogen gas to the component cooling water surge tank, which is an updated safety analysis report described design function for the component cooling water system. The licensee entered this condition into their corrective action program and planned to perform an evaluation to determine if prior NRC review and approval is needed for this design change. This issue has been entered into the corrective action program as Condition Report CR 2013-04417. The team determined that it was reasonable for the licensee to be able to foresee and prevent the occurrence of this deficiency. The team evaluated this performance deficiency as both a traditional enforcement violation, and a reactor oversight process finding. The violation of 10 CFR Part 50.59 was more than minor because it involved a change to an updated safety analysis report design function in that there was a reasonable likelihood that the change would require NRC review and approval. This finding is associated with the Mitigating Systems Cornerstone. The team used the NRC Enforcement Manual and Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding is determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in the loss of operability or functionality. The violation of 10 CFR 50.59 impacted the ability of the NRC to perform its regulatory oversight function and was determined to be Severity Level IV because the resulting changes were evaluated by the significance determination process as having very low safety significance, in accordance with the NRC Enforcement Policy. The NRC concluded that the finding did not reflect current licensee performance
05000285/FIN-2013008-262013Q2Fort CalhounFailure to Properly Inspect, Maintain, and Test Emergency Feedwater Tank EquipmentThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to ensure proper inspection, maintenance, and testing of equipment associated with emergency feedwater tank FW-19. Specifically, from initial construction until February 27, 2013, the licensee failed to ensure proper inspection, maintenance, and testing was performed on the emergency feedwater storage tanks sight glass ball check isolation valves, to prevent draining of the tank following failure of the sight glass. The licensee performed an analysis and concluded that operators have adequate time to respond to such a loss of tank FW-19 inventory. This issue has been entered into the corrective action program as Condition Reports CRs 2012-15687, CR 2013-03974, and CR 2013-06170. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-252013Q2Fort CalhounDeficient Evaluation for Known Degraded Conditions: SAFETY-RELATED Air Operated Valve Elastomers Not Qualified for Helb/Loca TemeraturesThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensee\'s failure to properly evaluate a known degraded condition regarding safety-related air operated valve elastomers that were not qualified for high energy line break or loss of coolant accident temperatures. Specifically, from January 11 through January 18, 2013, due to a an improper application of the single failure criteria, the licensee failed to properly evaluate and correct a known degraded condition associated with safety-related air operated valve elastomers that were not qualified for high energy line break or loss of coolant accident temperatures. This issue has been entered into the corrective action program as Condition Reports CR 2013-01396 and CR 2013-02611. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected, the failure to correct the degraded condition had the potential to lead to a more significant safety concern. Specifically, the affected air operated valves would have been in a condition where they would not have been qualified to perform their intended safety function. This issue was associated with the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-242013Q2Fort CalhounFailure to Effectively Monitor the Performance of Penetration SealsThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct conditions adverse to quality. Specifically, between July 2012 and March 2013, the team identified 6 instances where the licensee failed to identify a deficiency or a condition adverse to quality and to enter them into the corrective action program. As a result, conditions adverse to quality may not be corrected in a timely manner commensurate with the safety significance. This issue has been entered into the corrective action program as Condition Report CR 2013-07959. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to identify conditions adverse to quality and enter them into the corrective action program, has the potential to lead to a failure to correct conditions adverse to quality in a timely manner commensurate with the safety significance. This finding was associated with the Mitigating Systems Cornerstone. The team determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, and conducted a Phase 1 characterization and initial screening. Using Phase 1, Table 3, SDP Appendix Router, the team answered yes to the following question: Does the finding pertain to operations, and event, or a degraded condition while the plant was shutdown? As a result, the team used IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, the finding is determined to have very low safety significance (Green) since it did not need a quantitative assessment. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues and did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance
05000285/FIN-2013008-212013Q2Fort CalhounFailure to Ensure That Design Requirements Associated with the Containment Electrical Penetration Assemblies Were Correctly Translated Into Installed Plant EquipmentThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to translate applicable regulatory requirements and the design basis into specifications, drawings, procedures, and instructions. Specifically, from initial construction to present, the licensee did not perform adequate analysis and/or post-accident condition functional testing of the teflon insulated and teflon sealed Conax electrical penetration assemblies to determine if they were suitable for expected post accident conditions. The licensee has decided to replace or cap all Teflon-insulated containment electrical penetration assemblies prior to returning to power operations. This issue has been entered into the corrective action program as Condition Report CR 2013-03571. This performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Barrier Integrity Cornerstone and affected the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to implement a corrective action program with a low threshold for identifying issues and identify such issues completely, accurately, and in a timely manner commensurate with their safety significance
05000285/FIN-2013008-202013Q2Fort CalhounFailure to Identify Conditions Adverse to QualityThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct conditions adverse to quality. Specifically, between July 2012 and March 2013, the team identified 6 instances where the licensee failed to identify a deficiency or a condition adverse to quality and to enter them into the corrective action program. As a result, conditions adverse to quality may not be corrected in a timely manner commensurate with the safety significance. This issue has been entered into the corrective action program as Condition Report CR 2013-07959. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to identify conditions adverse to quality and enter them into the corrective action program, has the potential to lead to a failure to correct conditions adverse to quality in a timely manner commensurate with the safety significance. This finding was associated with the Mitigating Systems Cornerstone. The team determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, and conducted a Phase 1 characterization and initial screening. Using Phase 1, Table 3, SDP Appendix Router, the team answered yes to the following question: Does the finding pertain to operations, and event, or a degraded condition while the plant was shutdown? As a result, the team used IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, the finding is determined to have very low safety significance (Green) since it did not need a quantitative assessment. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues and did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance
05000285/FIN-2013008-192013Q2Fort CalhounFailure to Initiate Condition Reports in Accordance with the Corrective Action Program ProceduresThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to initiate condition reports when problems or conditions adverse to quality were identified in accordance with Procedure FCSG-24-1, Condition Report Initiation, Revision 3. Specifically, between July 2012 and March 2013, the team identified 11 instances where licensee staff failed to initiate a condition report after identifying a deficiency or a condition adverse to quality. In some instances, licensee personnel had to be prompted by the team to initiate a condition report. As a result, the corrective actions taken to address the conditions could have been potentially untimely. This issue has been entered into the corrective action program as Condition Report CR 2013-06991. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, if the licensee does not enter conditions adverse to quality into the corrective action program, the conditions adverse to quality may not be evaluated and corrected in a timely manner. This finding is associated with Mitigating Systems Cornerstone. The team determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, and conducted a Phase 1 characterization and initial screening. Using Phase 1, Table 3, SDP Appendix Router, the team answered yes to the following question: Does the finding pertain to operations, and event, or a degraded condition while the plant was shutdown? As a result, the team used IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, the finding is determined to have very low safety significance (Green) since it did not need a quantitative assessment. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues
05000285/FIN-2013008-182013Q2Fort CalhounFailure to Establish Adequate Instructions for Restoring Temporary ModificationsThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to establish adequate instructions for restoring temporary modifications. Specifically, from January 17, 2013, to the present, the licensees temporary modification control procedure did not include appropriate criteria for determining that control room and operations control center references reflect current plant configuration and were updated in a timely manner. The licensee initiated Condition Report CR 2013-04286, which stated that the licensees transition to a new procedure will help ensure that control room and operations control center documents were updated in a timely manner and that the licensee is determining whether any near-term action is necessary to address the issue until then. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the procedure inadequacy could become a more significant issue because it could allow operators to continue to reference material that does not reflect current plant configuration. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, the team determined that because this finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. Therefore, the finding is determined to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the work control component because the licensee failed to appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities. Specifically, the licensee did not incorporate actions into the procedure that would address the impact of out-of-date control room references on operator performance
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Marlone-20Davis-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"