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	Quarter	Site	Title	Description
05000397/FIN-2015002-03	2015Q2	Columbia	Failure to Maintain Doses ALARA for the Alternate Fuel Pool Cooling Modification Job	The inspectors identified a finding associated with the licensees failure to maintain doses as low as reasonably achievable (ALARA) while performing the Alternate Fuel Pool Cooling Modification job. Specifically, the licensee failed to effectively apply dose reduction methods, evaluate dose rates in a timely manner, prevent loitering and minimize workers in high dose fields, and implement in-field supervision as needed. As immediate corrective action, the licensee held site ALARA committee meetings and in-progress reviews to discuss the issues and developed lessons learned to incorporate into future job activities. The issue was documented into the licensees corrective action program as Action Request 321415. The failure to maintain doses ALARA while performing the Alternate Fuel Pool Modification job was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding was of very low safety significance because: (1) it was associated with ALARA planning and (2) the licensees current three-year rolling average collective dose of 102 person-rem was less than the 240 person-rem threshold for boiling water reactors. The finding had a field presence cross-cutting aspect, in the human performance cross-cutting area, because the licensee failed to have both radiation protection and engineering leaders commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations (H.2)
05000397/FIN-2015002-04	2015Q2	Columbia	Failure to Report a Major Loss of Emergency Assessment Capability	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.72(b)(3)(xiii) for the licensees failure to make a required event notification within eight hours for a major loss of assessment capability. Specifically, the licensee failed to make a report for an unplanned loss of rod position indication on April 29, 2015 that resulted in the inability to evaluate the position of all control rods for emergency action levels involving failures of the reactor protection system and anticipated transient without scram scenarios. As corrective actions to address a late 8-hour report, the licensee submitted Event Notification EN 51027 on April 30, 2015 and initiated Action Request AR 326719. The inspectors determined that the failure to make a required event notification within the time limits specified in regulations was a violation 10 CFR 50.72. 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 inspectors determined that a cross-cutting aspect was not applicable because the issue involving untimely reports to the NRC was strictly associated with a traditional enforcement violation.
05000397/FIN-2015002-05	2015Q2	Columbia	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Columbia Generating Station Final Safety Analysis Report, Section 8.3.1.4 established the design basis for the stations electrical distribution system and required, in part, that the physical independence of electrical systems comply with the requirements of IEEE 308-1974, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations. IEEE 308-1974 required, in part, that non-Class 1E circuits may be supplied from Class 1E power systems, provided that the Class 1E systems are not degraded below an acceptable level. Contrary to the above, prior to April 29, 2015, the licensee failed to establish measures to assure that the design basis was correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to translate the design basis into specifications for Class 1E electrical power panel E-PP-7AA because Circuit 13 of this power panel supplied a non-Class 1E circuit without appropriate isolation devices such that the non-Class 1E circuit would degrade the Class 1E system below an acceptable level. The finding was of very low safety significance because the finding is a deficiency affecting the design or qualification of a mitigating system where the system maintained its functionality. This issue was entered into the licensees corrective action program as AR 326573.
05000397/FIN-2015002-06	2015Q2	Columbia	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, in part that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Licensee procedure PPM 1.3.57, Barrier Impairment, Revision 0-32 is a safety related procedure used to establish compensatory measures to maintain equipment operable following a high energy line break when certain hazard barriers are removed. Contrary to the above, prior to April 15, 2015, the licensee failed to prescribe procedures of a type appropriate to the circumstances for activities affecting quality. Specifically, licensee procedure PPM 1.3.57, Barrier Impairment, Revision 0-32 failed to establish appropriate measures to maintain electrical instrument racks operable following a postulated high energy line break because the licensee incorrectly concluded that high energy line break events are slow developing events where there would be sufficient time to implement compensatory measure to restore disabled barriers. The finding was of very low safety significance because the finding did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This issue was entered into the licensees corrective action program as AR 325768.
05000397/FIN-2015002-07	2015Q2	Columbia	Licensee-Identified Violation	Title 10 CFR 50.54(q)(2) requires a licensee to follow and maintain in effect an emergency plan that meets the requirements of 10 CFR 50.47(b). Planning standard 10 CFR 50.47(b)(4) requires a licensee have a standard emergency action level scheme. Licensee procedure 13.1.1, Classifying the Emergency, Revision 47, implements the licensees standard emergency action level scheme. Emergency action levels 5.1.U.1 and 5.1.A.1 require classification of an emergency based, in part, on a valid reading exists which exceeds the Table 3 unusual event or alert effluent monitor thresholds for plant service water radiation indicting switch TSW-RIS-5. Contrary to the above, between April 3, 2015 and April 25, 2015, the licensee did not follow an emergency plan meeting the requirements of 10 CFR 50.47(b)(4). Specifically, the licensee ability to classify emergency action levels 5.1.U.1 and 5.1.A.1 was degraded because the modify select parameter was set incorrectly for TSW-RIS-5 resulting in an unintended attenuation factor of 15. This finding was identified by the licensee and entered in the licensees corrective action program as Action Request AR 326313. This finding was determined to be of very low safety significance because it did not involve Emergency Action Levels greater than Alert per Table 5.4-1 of Inspection Manual Chapter 0609 Appendix B, Emergency Preparedness Significance Determination Process.
05000397/FIN-2017002-01	2017Q2	Columbia	Mechanism Operated Cell Switch Failure	Green . The inspectors reviewed a self -revealed finding for the licensees failure to follow plant Procedure SWP -CA P-01, Corrective Action Program, that ensures corrective actions are timely. As a corrective action for failures associated with mechanism operated cell switches for nonsafety 4160 VAC circuit breakers in 2013 and 2015, the licensee assigned modifications to the mechanism operated cell switches but failed to implement t hem in a timely manner. Consequently, on July 20, 2016, circuit breaker E -CB -S/3 mechanism operated cell switches failed to change state resulting in a loss of a main feed pump and an unplanned runback to 70 percent reactor power. As corrective action, the licensee declared the startup transformer inoperable, modified the mechanism operated cell assembly for circuit breaker E -CB -S/3 to remove one switch, and performed post -maintenance testing. The licensee also initiated Action Request 352504 to perform an apparent cause review and address long -term corrective actions. The failure to follow plant Procedure SWP -CAP -01, Corrective Action Program, that ensures corrective actions are timely was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Initiating Event Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the loss of major loads on E -SM -3 upset plant stability by causing a loss of feed and reactor runback transient. The inspector performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, the licensee remained at power and maintained diverse feed and condensate pumps. This finding had a cross -cutting aspect in the area of human performance, consistent process, in that the licensee failed to use a systematic approach to make decisions including incorporating risk insights. Specifically, circuit breaker E -CB -S/3 is utilized at least monthly 3 for emergency diesel generator surveillance testing and a failure could render the startup transformer inoperable. The mechanism operated cell assembly modification, recommended in 2013 and assigned for action in 2015, was not planned or scheduled as a work order at the time of the failure in 2016 (H.13).
05000397/FIN-2017002-02	2017Q2	Columbia	Failure to Conduct Adequate Surveys of Spent Filters Moved from the Spent Fuel Pool	Green . The inspectors reviewed a self -revealed, non- cited violation of 10 CFR 20.1501 resulting from the licensee's failure to conduct radiation surveys necessary to establish appropriate controls to support movement of spent filters from the spent fuel pool to a shipping cask. This issue was entered into the licensee's corrective action program as Action Requests 356390 and 358265. The licensees failure to perform surveys necessary to establish appropriate controls to support the movement of filters from the spent fuel pool to a shipping cask was a performance deficiency. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and adversely affected the associated cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. Specifically, the inadequate radiation surveys resulted in inadequate controls being implemented causing unplanned and unintended personnel dos e. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green), because it did not involve: (1) ALARA planning and controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. The finding had a cross- cutting aspect in the area of human performance, associated with work management, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees organization and work processes failed to include the identification and management of radiological risk commensurate with the spent fuel pool filter project and the need for strict coordination with different groups or job activities (H.5).
05000397/FIN-2017002-03	2017Q2	Columbia	Inadequate Corrective Actions Causes Failure of HPCS Room Normal Supply Fan	Green . The inspectors reviewed a self -revealed, non- cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly identify and correct a condition adverse to quality. Specifically, since 2012, the licensee failed t o implement prompt corrective actions to correct an adverse condition related to the use of a contactor coil for a motor starter in the high pressure core spray room normal supply fan. As an immediate corrective action, the licensee replaced the contactor for the high pressure core spray room normal supply fan. The licensee entered this issue into the corrective action program as Action Request 360595. The failure to correct an adverse condition related to the use of a contactor coil for a motor starter in the HPCS room normal supply fan, though the licensee had an opportunity and plan to do so, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected 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 licensees failure to correct the use of a contactor coil for a motor starter in the high pressure core spray room normal supply fan resulted in an inoperable fan, high pressure core spray bus 4160 VAC switchgear, and high pressure core spray pump during the January 25, 2017, event when smoke was observed from the motor control center. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety -significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined that this finding did not have a cross -cutting aspect as the decision to not replace the contactor occ urred in 2014 and was not reflective of current performance.
05000397/FIN-2017002-04	2017Q2	Columbia	Licensee-Identified Violation	Title 10 CFR 50.55a(g)4, Inservice Inspection Standards Requirement For Operating Plants , requires , in part, that thro ughout the service life of a boiling water -cooled nuclear power facility, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the ASME Code. The ASME Code, Section XI, Article IWA -2610, requires that all welds and components subject to a surface or volumetric examination be included in the licensees inservice inspection program. This includes identifying each system support that is subject to Section XI requirements. Contrary to the above, prior to March 9, 2017, the licensee did not apply the applicable inservice inspection requirements to all system pressure boundaries within ASME Code Class 1, 2, and 3 boundaries. Specifically, the licensee failed to include the control rod d rive housing welds, as well as portions of the residual heat removal and high pressure core spray systems in their inservice inspection program. The licensee entered this issue into their corrective action program as AR 00343761 and reasonably determined the affected components and system remained operable. The licensee restored compliance by entering the components and systems into the ASME Section XI program. The finding was of very low safety significance (Green) because the finding did not represent an actual loss of safety function of a system or train, and did not result in the loss of a single train for greater than technical specification allowed outage time.
05000397/FIN-2017002-05	2017Q2	Columbia	Licensee-Identified Violation	Title 10 CFR 50.55a(g)(5 )(i) , ISI Program Update: Applicable ISI Code Editions and Addenda, requires , in part, that the inservice inspection program for a boiling water - cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of this section. Paragraph (g)4 (ii ), Applicable ISI Code: Successive 120- Month Intervals, requires, in part, that inservice examination of components and system pressure tests conducted during successive 120- month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section, 12 months before the start of the 120- month inspection interval. Contrary to these requirements, the licensee failed to issue the inspection plan for the fourth 10- year inservice inspection interval in a timely manner . Specifically, the licensee failed to issue the inservice inspection plan until January 27, 2016, even though the third 10- year inservice inspection interval had ended on December 13, 2015. A relief request to allow emergent repairs to be completed under the third 10 -year inservice inspection plan was requested by the licensee on December 16, 2015, and was approved by the NRC ; however, no repairs needed to be completed. The finding was of very low safety significance (Green) because the finding did not represent an actual loss of safety function of a system or train and did not result in the loss of a single train for greater than technical specification 31 allowed outage time. This issue was entered into the licensees corrective action program as A R 00341506.
05000397/FIN-2017002-06	2017Q2	Columbia	Licensee-Identified Violation	On October 13, 2016, several ARMs unexpectedly alarmed when six filters were simultaneously lifted from the SFP to be placed into a radioactive waste liner . The radiation work permit (RWP) governing performance of the job, RWP 3003788, Revision 00, dated September 7, 2016, had the following , Hold Point , requirements in the event that unexpected radiological conditions occurred during the movement of spent filters: Stop work immediately and notify RP personnel if an unanticipated ARM alarms. If a reading greater than 10 rem/hour contact or 800 millirem/hour at 30 centimeters was detected, but not expected, place the filter back into the SFP . The six filters that had been raised from the SFP had radiation levels as high as 14,000 rem/hour on contact and over 300 rem/hour at almost 30 centimeters. However, the filters were placed in the liner rather than back into the SFP , as specified in the RWP and instructed by RP staff during the evolution. Technical Specification 5.4.1.a requires, in part, that procedures be written, implemented, and established for those areas recommended in Regulatory Guide 1.33, Appendix A, Revision 2, 1978. Section 7(e) of Appendix A recommends written procedures for RWP systems to control access to radioactive materials and limit personnel exposure. Radiation Work Permit 3003788 stated, in part, in the event of unexpected radiological conditions during movement of spent filters, stop work immediately if an unanticipated area radiation monitor alarms, and if a reading greater than 10 rem/hour contact was detected but not expected, place the filter back into the SFP. Contrary to the above, on October 13, 2016 , the licensee f ailed to stop work immediately when several area radiation monitors unexpectedly alarmed and failed to place the filters back into the SFP when readings greater than 10 rem/hour contact were detected but not expected . Subsequently, 16 workers received an addition al 63.5 millirem when the instructions of the RWP and RP staff were not followed. The finding was of very low safety significance (Green) because it did not involve: (1) as-low- as-reasonably achievable (ALARA) planning and controls ; (2) a radiological overexposure; (3) a substantial potential for an exposure; or (4) a compromised ability to assess the dose. This issue was entered into the licensees corrective action program as ARs 356390 and 358265.
05000416/FIN-2011003-01	2011Q2	Grand Gulf	Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found the entrance door to the diesel generator building and the entrance door to the division 2 diesel generator in a degraded condition. The inspectors identified that the door seals did not make complete contact with the door frames all the way around as required by procedure. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-02575. The finding is more than minor because it is associated with the protection against external factors attribute of 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. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multiple trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis. The result was a delta-core damage frequency of 3.3E-7/yr and a delta-large early release frequency of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The inspectors determined the apparent cause of this finding was that licensee personnel were not adequately trained to perform these inspections. Therefore this finding has a cross-cutting aspect in the area of human performance associated with resources in that the licensees training of personnel was not adequate in performing inspection of the probable maximum precipitation door seals.
05000416/FIN-2011003-02	2011Q2	Grand Gulf	Failure to Follow Scaffold Control Procedure	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement scaffolding control procedural requirements related to post-installation inspections and engineering safety evaluations for scaffolding constructed within 2 inches of safety-related or fire protection equipment. During plant walkdowns, inspectors identified multiple examples of the licensee not properly implementing Entergys corporate and site procedures for the control of scaffolding. The licensees immediate corrective actions included inspecting the scaffolding that had been installed, modifying or removing it where appropriate, and properly posting the scaffolds. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2011-03480, CR-GGN-2011-03601, CR-GGN-2011-03602, and CR-GGN-2011-03603. The inspectors determined that this finding is more than minor because it is associated with the external factors and equipment performance attributes 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 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green), because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the apparent cause of this finding was lack of supervisor oversight during scaffold construction. Therefore the finding has a cross-cutting aspect in the area of human performance associated with work practices, in that the licensee did not provide effective supervisor oversight of workers constructing scaffolding to ensure these activities were performed per procedural requirements.
05000416/FIN-2011003-03	2011Q2	Grand Gulf	Failure to Identify Conditions Adverse to Fire Protection	The inspectors identified a noncited violation of License Condition 2.C(41) for the failure to identify conditions adverse to the fire protection program. Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of bent or misaligned sprinkler head deflector plates and a painted sprinkler head. Corrective action included correcting bent or misaligned plates and replacing the painted sprinkler head. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-03132. The finding is more than minor because it is associated with the protection against external factors 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 safety concern is that the number of bent or misaligned sprinkler heat canopies and painted sprinkler heads would not provide an adequate area-wide coverage of suppression. The inspectors evaluated the significance of this finding using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the inspectors determined that the deficiency had low degradation since less than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1of Manual Chapter 0609, Appendix F. This finding has a cross-cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain specific criteria for identifying unacceptable sprinkler conditions.
05000416/FIN-2011003-04	2011Q2	Grand Gulf	Failure to Ensure that Safety Related Manholes were Properly Sealed to Prevent the Entry of Flammable Liquid	The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that manholes MH01, MH20 and MH21 were properly sealed to prevent the entry of flammable liquid. During the performance of the manhole/vault inspection, the inspectors were reviewing engineering change packages associated with solar sump pumps for MH20 and MH21. During their review, they determined that the licensee was not meeting the requirements of their license bases documents for MH20 and MH21, which contain safe shutdown cables for standby service water trains A and B. The licensees immediate corrective action included placing hazmat barricades around each manhole to prevent flammable fluids from entering the manholes. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00562. This finding was more than minor because it was associated with the protection against external factors 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. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b, Item 1 directs the inspectors to Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. This noncited violation was therefore determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to initiate a condition report when the issue was identified during the development of their engineering change package, which resulted in the failure to ensure the safety related manholes were sealed in accordance with their license based documents.
05000416/FIN-2011003-05	2011Q2	Grand Gulf	Failure to Provide Adequate Procedures for High Pressure Core Spray Minimum Flow Valve Surveillance Testing	The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the licensees failure to provide adequate testing procedures, which resulted in the high pressure core spray minimum flow valve inadvertently stroking approximately 11 times during a surveillance test. The excessive stroking of the valve resulted in the unplanned inoperability of the high pressure core spray system because the valves feeder breaker overcurrent instantaneous trip setpoint had drifted below the manufacturers tolerance for the existing setting. As immediate corrective action, the licensee replaced the degraded breaker. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01901. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone\'s objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function since the high pressure core spray system would still have been functional even with the minimum flow valve potentially failing open. Additionally, it did not represent a loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with operating experience in that licensee had not incorporated operating experience from a similar event that had occurred at another Entergy site.
05000416/FIN-2011003-06	2011Q2	Grand Gulf	Loose Fuse Clips in Division 3 Emergency Diesel Generator	The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee\'s failure to take adequate corrective actions for a significant condition adverse to quality associated with the division 3 emergency diesel generator. While performing a maintenance effectiveness review of the diesel generators, the inspectors noted on October 17, 2009, at 9:07 p.m., the FU-7 fuse for the division 3 diesel generator was determined to have a faulty fuse clip, resulting in the inoperability of the diesel generator due to loss of power to the direct current powered fuel pumps. Then on March 18, 2011, the division 3 emergency diesel generator was again rendered inoperable due to a faulty fuse clip on the FU-8 fuse holder, which is of the same design and function as the FU-7 fuse holder in the previous occurrence. Short term corrective action included replacing the fuse holder. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01868. The finding is more than minor because it is associated with equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of human performance associated with resources because the training provided to correct the initial event was not adequate to ensure proper fuse installation and verify good connection existed between the fuse and fuse holder.
05000416/FIN-2011003-07	2011Q2	Grand Gulf	Failure to Assure Configuration Control of Safety Related Systems	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to review the suitability of leaving test fittings on reactor coolant system flow transmitter equalizing block drain ports instead of the design specified manifold plugs. As corrective action, the licensee replaced the test fittings with the correct drain plugs. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-04485. This finding is more than minor because it is associated with the design control 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. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability of functionality, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, associated with work practices, because the licensee failed to ensure that human error prevention techniques, such as holding pre-job briefings, self- and peer-checking, and proper documentation of activities were utilized such that work activities were performed safely and personnel did not proceed in the face of uncertainty or unexpected circumstances Specifically, the licensee failed to review the suitability of installing test and brass fittings on pressure, differential pressure and flow transmitter block valve drain ports instead of the design specified manifold plugs.
05000416/FIN-2011003-08	2011Q2	Grand Gulf	Failure to Follow a Procedure Resulting in the Inoperability of the Reactor Core Isolation Cooling System Primary Containment Isolation Valve	The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, for failure to follow a procedure resulting in the inoperability of the reactor core isolation cooling system primary containment isolation valve. This occurred while the licensee was performing surveillance on the reactor core isolation cooling system and incorrectly attached a jumper to the wrong terminal point resulting in blowing a fuse that caused a loss of control power to the reactor core isolation cooling primary containment isolation valve 1E51-F031. As immediate corrective action, the licensee removed the jumper and replaced the control power fuse. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01932. The finding is more than minor since it is associated with the human 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. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. In addition, this finding had a human performance cross-cutting aspect associated with work practices in that the licensee did not use the proper human performance techniques of self-checking to prevent the loss of control power to a primary containment isolation valve.
05000416/FIN-2011003-09	2011Q2	Grand Gulf	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, states, in part, that activities affecting quality shall be accomplished in accordance with prescribed procedures. Specifically EN-OP-104, Operability Determination Process, Revision 5, Section 5.3(1), states in part to Confirm the existence of a Degraded or Nonconforming Condition for the Technical Specification System Structure or Component. Contrary to this requirement, on March 18, 2011, the on-shift senior reactor operator failed to perform a proper operability determination for the high pressure core spray pump after minimum flow valve 1E22-F012 cycled approximately 11 times during testing causing the supply breaker to trip open, resulting in high pressure core spray being inoperable. After resetting the breaker for 1E22-F012, ensuring the breaker was not faulted, and performing a one-time stroke test, the system was declared operable. Engineering personnel evaluated the event several hours later and questioned the operability of valve 1E22-F012 and the high pressure core spray system due to repeated cycling of the valve motor, which resulted in the breaker tripping. Based on engineering input, operations performed a second operability determination and determined that the system was operable with evaluation required. The licensee performed testing of the breaker for valve 1E22-F012 and determined its over-current trip setting had drifted to approximately 60 amps when its minimum allowed setting was 85 amps. This confirmed that the high pressure core spray system was inoperable the entire time. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-02240. The finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
05000416/FIN-2014002-01	2014Q1	Grand Gulf	Failure to Ensure Scaffold Activity Would not Interfere with Fire Brigade Response	The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, for the failure to adhere to procedural requirements to ensure that scaffold installed in the plant would not prevent or restrict the fire brigade from accessing a certain route used for response to a fire in the area. On February 4, 2014, the licensee installed a scaffold in the containment building for an inspection. The licensees procedure required a walkdown of proposed scaffold to determine if the scaffold would prevent or restrict fire brigade access. The initial reviewer identified that the ladder to access the scaffold would restrict fire brigade access, thus the ladder was not installed until it was required. On March 1, 2014, the ladder was installed for the four hour inspection. Once completed, the licensee failed to remove the scaffold ladder to restore normal access to the area. On March 4, 2014, the inspectors identified that the scaffold ladder was still installed. The inspectors brought their concern to the licensee, who determined that the scaffold would adversely affect the response of fire brigade members to that area of containment. As an immediate corrective action, the licensee removed the scaffold ladder to allow adequate access for the fire brigade members. The licensee documented this issue in Condition Report CR-GGN-2014-02363. The failure to ensure fire brigade members had adequate access passed a scaffold installed in the containment building was a performance deficiency. The performance deficiency was more than minor and therefore a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone in that the fire brigades inability to gain access to certain areas in containment could result in preventing prompt extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was shutdown for refueling outage RF19. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, dated February 28, 2005. The inspectors determined that Appendix G did not address fire brigade issues and solicited input from the senior reactor analyst. The senior reactor analyst performed a detailed risk evaluation and determined that Inspection Manual 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, June 19, 2012, Exhibit 2, Mitigating System Screening Questions, adequately bounded the performance deficiency. The inspectors determined that the finding involved the response time of the fire brigade to a fire, and the finding was of very low safety consequence (Green) because the fire brigades response time was mitigated by other defense-in-depth elements such as area combustible limits were not exceeded, installed fire detection systems were functional, and alternate means of safe shutdown were not impacted. Specifically, there were no combustibles in the area beyond limits, all fire detectors for the area were functional, and the plant was in a shutdown condition with the cavity flooded at the time. The apparent cause of this finding was the work groups involved did not communicate the significance of the impact the scaffold ladder had on fire brigade access to the area and the importance of having the ladder removed upon completion of the work. Therefore, the finding has a cross-cutting aspect in the human performance area associated with team work, in that the individuals and workgroups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained.
05000416/FIN-2014002-02	2014Q1	Grand Gulf	Failure to Control a Locked High Radiation Area Due to Unsecured Highly Radioactive Materials Stored in the Pool	The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.7.3, resulting from the licensees failure to control a high radiation area with radiation levels greater than 1000 millirem per hour. As immediate corrective actions, the licensee stopped the work activity, placed a senior radiation protection technician in control of the area, surveyed all affected areas, and properly posted and controlled the area. The licensee also checked qualifications of the involved individuals and conducted a root cause evaluation for the event. This event was documented in the licensees corrective action program as Condition Reports CR-GGN-2014-02219, CR-GGN-2014-02221, and CR-GGN-2014-02224. The failure to control a high radiation area with radiation levels greater than 1000 millirem per hour was a performance deficiency and a violation of Technical Specification 5.7.3. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because it removed a barrier intended to prevent the worker from receiving unexpected dose. 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 an 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 procedure adherence, because the licensee failed to follow process, procedures, and work instructions when they did not inventory and ensure control of the dry tube plunger end as it was stored in the horizontal fuel transfer system pool within containment.
05000416/FIN-2014002-03	2014Q1	Grand Gulf	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that design control measures be established and implemented to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, the licensee failed to implement applicable design bases for the Standby Service Water System Pump 4160 VAC cables being submerged. Specifically, on January 31, 2014, the licensee did not prevent water from submerging the cables in Manhole MH-01 due to a failed sump pump. The inspectors verified that the latest megger tests for the standby service water pump cables were acceptable for demonstrating operability. This finding has been entered into the licensees corrective action program as Condition Reports CR-GGN-2014-00616 and CR-GGN-2014-00768. Using Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that this finding had very low safety significance (Green) because it did not result in the standby service water system becoming inoperable.
05000416/FIN-2014002-04	2014Q1	Grand Gulf	Licensee-Identified Violation	Technical Specification (TS) 3.3.6.1, Primary Containment and Drywell Instrumentation, requires the primary containment and drywell isolation instrumentation be operable while in Modes 1, 2, and 3. Contrary to the above, on August 3, 2013, the licensee failed to ensure the primary containment and drywell isolation instrumentation was operable prior to changing from Mode 4 (Cold Shutdown) to Mode 2 (Startup). On August 6, 2013, during a supervisory review of procedures in progress, the licensee determined that they were not incompliance with TS 3.3.6.1 due to jumpers that were installed to disable the function of the instrumentation. The licensee immediately entered the TS 3.3.6.1 Limiting Condition for Operation and associated actions. The licensee restored compliance with the TS by removing the jumpers and restoring the primary containment and drywell instrumentation to operable status and documented this issue in the corrective action program under Condition Report CR-GGN-2013-5101. Using Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that this finding had very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment or drywell and did not involve the hydrogen igniters in the reactor containment.
05000416/FIN-2014009-01	2014Q2	Grand Gulf	Failure To Correct Degraded Viewing Ports In A Timely Manner	The inspectors identified a Green finding resulting from the licensees failure to follow Procedure EN-LI-102, Corrective Action Process, Revision 23, and Procedure EN-OP-104, Operability Determination Process, Revision 7, for an adverse condition. The licensee failed to repair degraded viewing ports on the isophase bus ducting in a timely manner. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2013-00319. The failure to implement adequate corrective actions in a timely manner after the discovery and evaluation that the viewing windows on the isophase bus duct had the potential to cause a reactor scram is a performance deficiency. The performance deficiency was more than minor because it was associated with the Initiating Events cornerstone attribute of Human Performance and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, Exhibit 1, Section B, Transient Initiators, the inspectors determined that the issue has a very low safety significance (Green) because it only caused a reactor trip and did not cause a loss of mitigating equipment relied on to transition the plant from the onset of a trip to a stable shutdown condition. This finding has a cross-cutting aspect in the problem identification and resolution area, associated with operating experience, because the licensee failed to systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner (P.5).
05000416/FIN-2014009-02	2014Q2	Grand Gulf	Failure to Correct A Significant Condition Adverse to Quality And Preclude Repetition	The inspectors reviewed a self-revealing, Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, resulting from the licensees failure to prevent the repetition of a break of the first stage turbine sensing line, which resulted in a reactor scram. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2014-02824. The failure to implement adequate corrective actions from the previous first stage turbine pressure sensing line break to preclude repetition of a significant condition adverse to quality was the performance deficiency. The performance deficiency was more than minor because it was associated with the Initiating Events cornerstone attribute of Human Performance and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, Exhibit 1, Section B, Transient Initiators, the inspectors determined that the issue required a detailed risk evaluation by the senior reactor analyst because the violation caused a reactor trip and the loss of mitigation equipment. The licensee performed an inadequate evaluation of the root cause of the 2012 steam sensing line break, resulting in inadequate corrective actions to prevent repetition. Therefore, this violation has a cross-cutting aspect in the problem identification and resolution performance area, associated with evaluation, because the licensee failed to thoroughly evaluate the issue to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000416/FIN-2015301-01	2015Q4	Grand Gulf	Inadequate Plant Operating Procedures with Eight Examples	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to this, The licensees Off-Normal Procedure ONEP 05-1-02-I-1, Reactor Scram, Revision 125, does not provide all necessary guidance on how to scram the reactor. Once the immediate action of placing the mode switch in the shutdown position is completed, all additional guidance for shutting down the reactor using alternate methods is contained in EP-2A. However, the first backup method of using the scram pushbuttons is missing from both of these procedures. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensee is missing several off-normal procedures that are required by Technical Specifications based on commitments to NRC Regulatory Guide 1.33, Revision 2. Specifically, there are no off-normal procedures for 1) a total or partial loss of DC power, 2) electrical grounds, and 3) partial or total loss of all annunciators. The licensee is committed to revision 2 of this regulatory guide in its Technical Specifications. These procedure deficiencies were entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Emergency Procedure 05-1-02-II-1, Attachment III, Shutdown from the Remote Shutdown Panel, Revision 47, does not include all of the required steps to complete the attachment. Step 3.2.5a of this procedure requires an operator to obtain one key while two keys are actually required to complete the task. One key is required to open the protective box covering the switch and a different key is required to operate the switch. This procedure discrepancy led to delays and confusion during examination administration by applicants and during examination validation by licensed operators. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Emergency Procedure 05-S-1-EP-1, Attachment 6, Defeating Reactor Feed Pumps RPV Level 9 Trips, Revision 32, contains labeling discrepancies in that the relay nomenclature in the procedure does not match the nomenclature in the main control room cabinet 1H13-P612 Bay B. This caused confusion among both the applicants and licensed operators. The confusion delayed the completion of the task administered during the examination. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees System Operating Instruction 04-1-01-P41-1, Standby Service Water System, Revision 140, Section 4.2, contains labeling discrepancies in that the control board labeling for several switches do not match the nomenclature listed in the procedure for the associated switches. Specifically, steps 4.2.2A(4)(a), 4.2.2A(4)(b), and 4.2.2A(6) each have a discrepancy. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Alarm Response Instruction 04-02-1H13-P870-2A-E1, Revision 134, for the residual heat removal (RHR) alarm RHR A PMP RM FLOODED contains non-conservative guidance to close the suction valve (valve 1E12-F004A) for RHR pump A without regard to ensuring that the pump is secured first. This creates a condition where the safety-related residual heat removal pump is tripped on interlock only in order to prevent damage. The expectation provided to the NRC by the operations staff is that the operators should first trip the residual heat removal pump and then shut the suction valve. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensee was unable to locate any written guidance for placing a safetyrelated diesel generator in maintenance mode to prevent automatic start and subsequent overheat of the machine when cooling water is unavailable. According to the Updated Final Safety Analysis Report, Section 9.5, Revision LDC 05077, the diesel generator jacket cooling water system provides sufficient heat sink to permit the standby diesel engines to start and operate for 2 minutes without cooling water available. Procedures that were reviewed included SOI 04-1-01-P75-1, SOI 04-1-01-Y47, and ONEP 05-1-02-I-4. An additional NRC concern for this sequence is that there is no time critical action associated with securing these diesel generators when cooling water (standby service water) is not available. The licensee needs to review the risk management program and ensure that this is not assumed in the risk management profile or if it is assumed, then operators are trained and can implement the shutdown in the appropriate time to prevent equipment damage. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Equipment Performance Instruction 04-1-03D21-1, Monthly Area Radiation Monitors Functional Test, Revision 37, has confusing guidance which led several applicants in not being able to complete the task administered during the NRC initial license examination. Specifically the procedure has a limit and precaution stating that not all ARM module function switches spring return to OPERATE after being taken to ALARM. Some must be manually returned to OPERATE after being taken to ALARM while the specific steps in the procedure have the operator place and hold function switch in alarm and then release. No guidance is given within the step to return the switch to operate and this creates a situation where the observation of indication returning to normal does not occur. A precaution in the front matter in the procedure stating that the equipment may not function as the procedure is written is not sufficient to meet the quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The failure of these eight procedures to have the appropriate qualitative and/or quantitative criteria to complete these activities was a performance deficiency. The finding was more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, inadequate procedures could adversely affect the operating crews ability to take appropriate actions to ensure reactor safety is being maintained. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, the team determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program for greater than 24 hours. The finding has a cross-cutting aspect in the area of human performance associated with procedure adherence because individuals did not follow the processes to change or correct procedures that contained incorrect, missing, or non-conservative guidance (H.8).
05000416/FIN-2016004-01	2016Q4	Grand Gulf	Failure to Incorporate Design Requirements for Switchgear Room Cooling	Green. The inspector identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to implement appropriate design control measures associated with a safety-related service water flow calculation. Specifically, several unverified and potentially nonconservative inputs were identified associated with Calculation MC-QIP41-97020, Revision 11, Determination of Minimum Allowable SSW Flows (LOCA Lineup) to Safety Related Heat Exchangers, used to analyze minimum service water flow to the vital switchgear room coolers. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2016-07597, initiated action to update Calculation MC-QIP41-97020, and initiated actions to analyze the ability of vital switchgear room cooling to meet its specified safety function. This performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not assure that the vital switchgear ventilation system was capable of maintaining the rooms temperature below design requirements under all conditions. The NRC performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding had very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding had a cross-cutting aspect in the documentation aspect of the human performance cross-cutting area because the licensee failed to maintain complete, accurate, and up-to-date documentation of the design temperature limits for safety-related equipment. Specifically, the licensee failed to document and evaluate a change to temperature limits related to switchgear cooling to ensure that its use as a design parameter was consistent with original design specifications of the equipment (H.7).
05000416/FIN-2016004-02	2016Q4	Grand Gulf	Failure to Use Procedures and Engineering Controls to Maintain Occupational Doses ALARA	Green. The inspectors identified a non-cited violation of 10 CFR 20.1101(b) for the licensees failure to implement radiation exposure reduction procedures and engineering controls to minimize unplanned and unintended radiation dose to workers and to maintain occupational doses as low as is reasonably achievable (ALARA). Several radiological work permits exceeded initial dose estimates with minimal or no actions taken to evaluate the basis for the dose overrides and to develop mitigating strategies. The primary contributor to the unplanned exposures was elevated dose rates from increased cobalt-60 activity associated with a failure to properly plan and execute spent fuel pool and reactor cavity cleanup operations. In addition, the licensee failed to observe radiological work permit hold points, to initiate ALARA Management Committee meetings, and to perform radiological assessments of radiological work permit dose estimates as procedurally required. As immediate corrective actions, the licensee reviewed the work activity, documented lessons learned, and generated Condition Reports CR-GGN-2016-03151 and CR-GGN-2016-08543 to address these programmatic weaknesses for future outages. The failure to implement procedures and engineering controls to minimize unplanned and unintended radiation dose and to maintain occupational doses as low as is reasonably achievable was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (ALARA planning) and adversely affected the cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, inadequate ALARA planning and radiological controls resulted in unplanned, unintended dose for a number of work activities in which the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined this finding to be of very low safety significance (Green) because the finding involved ALARA planning and controls, and because the licensees latest 3-year rolling average did not exceed 240 person-rem per unit for boiling water reactors. The finding had a cross-cutting aspect in the area of problem identification and resolution, associated with operating experience, in that, the licensees organization failed to systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, the licensee failed to implement and incorporate relevant internal operating experience from Refueling Outage 18, which was of similar radiological circumstances, to mitigate the effects of cobalt-60 activity in the reactor cavity and unplanned spent fuel pool cleanup outages (P.5).
05000445/FIN-2015003-01	2015Q3	Comanche Peak	Failure to Take Appropriate Maintenance Rule Corrective Actions for the Instrument Air System	The inspectors identified a non-cited violation of 10 CFR Part 50.65(a)(1) for the failure to take appropriate corrective actions for a system that did not meet established goals. Specifically, the Unit 1 instrument air system had been in maintenance rule (a)(1) status since 2011 due to dryer component failures. In 2014, the instrument air system experienced additional failures that resulted in water accumulating in air operated valve actuators on Unit 1. The water intrusion resulted in abnormal operation of the air operated valves in the Unit 1 main feedwater system. These failures were determined to be due to inadequate maintenance on the instrument air dryers unrelated to the 2011 failures. However, the licensee failed to revise their corrective actions to address the causes of the water intrusion. The licensee entered these issues into corrective action program as Condition Report CR-2015-009077. The licensees failure to take appropriate corrective actions for a system that did not meet established goals was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to take appropriate corrective actions adversely affected the reliability of a system scoped in the plants maintenance rule program. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings AtPower, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance (Green) because the finding affected a support system initiator but did not involve the loss of a support system that contributed to the likelihood of an initiating event and affected mitigation equipment. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation, in that, the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes. Specifically, the licensee performed an inadequate cause evaluation and failed to identify the cause of the water intrusion (P.2).
05000445/FIN-2015003-02	2015Q3	Comanche Peak	Inadequate Maintenance Procedure Results in Power Reduction	The inspectors reviewed a self-revealing finding associated with an inadequate procedure which resulted in a unit down power. Specifically, the procedure used for over speed testing of the main feedwater pumps did not provide adequate guidance for operation of the test push button which resulted in a trip of main feedwater pump 1A and subsequent unit power reduction. The licensee entered this issue into the corrective action program as Condition Report CR-2015-005195, and took actions to increase the maintenance frequency on the mechanical trip device, and to reduce power when performing mechanical over speed testing in the future. The failure to provide adequate procedures for main feedwater pump over speed testing was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the procedural quality attribute of the Initiating Events Cornerstone, and directly affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and is therefore a finding. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that this finding does not have a cross-cutting aspect because the most significant contributor of this finding would have occurred more than three years ago, in 2001, and is not reflective of current licensee performance.
05000445/FIN-2015008-01	2015Q3	Comanche Peak	Failure to Evaluate the Lack of Missile Protection on the Turbine Driven Auxiliary Feedwater Pumps Steam Exhaust Piping	The team identified a cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the lack of missile protection on the turbine driven auxiliary feedwater pumps steam exhaust piping. Specifically, since June 13, 2012, the licensee failed to verify the adequacy of design of the turbine driven auxiliary feedwater pumps steam exhaust piping to withstand impact from a tornado driven missile hazard, or to evaluate for exemption from missile protection requirements using an approved methodology. This issue does not represent an immediate safety concern because the licensee performed an operability evaluation, which established a reasonable expectation of operability. The licensee entered this issue into the corrective action program for resolution as Condition Report CR-2015-007869. The licensees failure to analyze the effects of a tornado missile strike on the turbine driven auxiliary feedwater pumps steam exhaust piping was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to evaluate a design nonconformance on the turbine driven auxiliary feedwater pumps steam exhaust piping for lack of missile protection. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the team determined that the finding is of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The finding has a human performance cross-cutting aspect associated with conservative bias because individuals failed to use decision-making practices that emphasize prudent choices over those that are simply allowable (H.14).
05000445/FIN-2015008-02	2015Q3	Comanche Peak	Failure to Properly Assess and Document the Basis for Operability associated with the Turbine Driven Auxiliary Feedwater Pumps Steam Exhaust Piping not being Evaluated for Tornado Generated Missile Impacts	The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated the licensees failure to perform adequate operability assessments when a degraded or nonconforming condition was identified associated with the turbine driven auxiliary feedwater pumps steam exhaust piping not being evaluated for tornado generated missile impacts. Specifically, operators used probabilistic assumptions and failed to adequately assess and document the basis for operability when a degraded or nonconforming condition was identified associated with the turbine driven auxiliary feedwater pumps steam exhaust piping not being evaluated for tornado generated missile impacts. This issue does not represent an immediate safety concern because the licensee performed a subsequent operability evaluation, which established a reasonable expectation of operability. The licensee entered this issue into the corrective action program for resolution as Condition Report CR-2015-007919. The licensees failure to properly assess and document the basis for operability when a degraded or nonconforming condition associated with the turbine driven auxiliary feedwater pumps steam exhaust piping not being evaluated for tornado generated missile impacts was identified, was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to evaluate a design nonconformance on the turbine driven auxiliary feedwater pumps steam exhaust piping for lack of missile protection. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the team determined that the finding is of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The finding has a human performance cross-cutting aspect associated with conservative bias because individuals failed to use decision-making practices that emphasize prudent choices over those that are simply allowable (H.14).
05000445/FIN-2015008-03	2015Q3	Comanche Peak	Inadequate Procedure for Surveillance on Safety-Related Service Water Systems	The team identified a non-cited violation of Technical Specification (TS) 5.4.1, Procedures, for an inadequate procedure for performing surveillances on the station service water (SSW) systems in units 1 and 2. Specifically, Procedures OPT-207 A and B, Service Water System, were modified in September 2010 so that failure of any SSW vacuum breaker to OPEN was considered a degraded condition and not an inoperable condition of the associated SSW System train. However, per DBD-ME-233, Station Service Water, Revision 33, Active Valves, vacuum breakers are required by ASME (Code Section) III on the inlet and outlet piping to the diesel generator jacket water coolers to mitigate the effects of water hammer due to water column separation and subsequent rejoining following a pump trip. This issue does not represent an immediate safety concern because the licensee confirmed that all of the vacuum breakers in service had passed their most recent surveillance test. The licensee entered this issue into the corrective action program for resolution as Condition Report CR-2015-010800. The finding is more than minor because it is associated with the procedure quality 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 (i.e., core damage). Specifically, the licensee did not ensure the guidance incorporated into quality related procedures was accurate and consistent with the design basis analysis for the systems and this conflict resulted in inadequate operability determinations associated with the SSW System. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the team determined that the finding is of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a human performance cross cutting aspect associated with design margins because the licensee failed to operate and maintain the SSW system equipment within design margins. Rather than ensure that margins are carefully guarded and changed only through a systematic and rigorous process, the licensee failed to re-evaluate SSW system operability with failed vacuum breaker valves even when additional test information indicated previous assumptions were incorrect (H.6).
05000445/FIN-2015008-04	2015Q3	Comanche Peak	Failure to Maintain Adequate Controls for Design Calculations	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, with two examples associated with the licensees failure to ensure that design changes were subject to design control measures commensurate with those applied to the original design and were approved by the designated responsible organization. Specifically: (1) The licensee instituted an engineering change package to modify the design and setpoints for the station service water (SSW) system vacuum breaker valves (CP1/2-SWVAVB-01/02/03/04) and did not consider the allowable tolerance for the setpoint for all design basis events and operating conditions. The licensee adequately addressed this issue by reperforming the calculation incorporating the setpoint allowable tolerance. (2) The licensee failed to account for system design leakage in design calculation DBD-CS-096, for the safe shutdown impoundment minimum level. The licensee evaluated the water loss from the impoundment due to evaporation, but failed to account for losses due to system design leakage. The licensee adequately addressed this issue by applying the design system leak rate for a 30-day mission time to the available water in the safe shutdown impoundment. The licensees failure to evaluate properly the effects of modifying the setpoint including allowable tolerances for all modes of operation and all sources of water loss from the safe shutdown impoundment was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the configuration control 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 (SDP) for Findings AtPower, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the team determined that the finding is of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined that this finding does not have a cross-cutting aspect because the most significant contributor of this finding occurred more than three years ago and does not reflect current licensee performance.
05000445/FIN-2015008-05	2015Q3	Comanche Peak	Failure to Perform Adequate Operability Assessments associated with Failures of Service Water System Vacuum Breaker during Surveillance Tests	The team identified an unresolved issue associated with the failures of the vacuum service water breakers that remained in service. During these failures, the licensee had documented the surveillance failures as degraded conditions and concluded that they did not have an impact on the operability of the service water system. The team reviewed the licensees operability assessments associated with surveillance tests where at least one of the service water system vacuum breakers failed to meet acceptance standards. During these failures, maintenance personnel mechanically agitated the vacuum breakers in order to get them to operate but did not replace the vacuum breakers until a future date. The inspectors noted that design basis calculations indicate that the larger of the two vacuum breakers (check valve) was required in order to protect the EDG jacket service water coolers and concluded that the licensee did not have appropriate justification to conclude that the service water system remained operable with a failed vacuum breaker if it was the larger breaker. During the inspection period the team was not able to determine which vacuum breakers were found in a degraded condition, therefore more information is required to determine if a non-compliance exists. Specifically, since September 2010, the licensee issued twenty six operability evaluations associated with failed surveillance test on vacuum breakers in the service water system where operators used incorrect information when assessing operability, which failed to establish a reasonable expectation of operability. This issue does not represent an immediate safety concern because at the time of discovery, there were no failed vacuum breakers in service. The licensee entered the finding into corrective action program as Condition Report CR-2015-008334. This issue will remain unresolved until the NRC is provided sufficient information regarding the particulars associated with the check valve/vacuum breaker failures in order to determine if a non-compliance exists. Specifically, the team requires information associated with the specific valve(s) that failed the length of time that the failed valve remained in service prior to replacement; whether the opposite train diesel generator was ever inoperable during the period the failed valve remained in service. (URI 05000445/2015008-05; 05000446/2015008-05, Failure to Perform Adequate Operability Assessments associated with Failures of Service Water System Vacuum Breaker during Surveillance Tests)
05000445/FIN-2016002-02	2016Q2	Comanche Peak	Failure to Determine Dose Rates Prior to Allowing Entry into a High Radiation Area	The inspectors reviewed a self-revealed non-cited violation of Technical Specification 5.7.1.e associated with the licensee allowing a worker access into the 2-077-B penetration valve room, a high radiation area, without an adequate knowledge of the radiological conditions. Specifically, the licensee briefed the worker on the conditions with outdated radiation survey information even though the 2-077-B penetration valve room was subject to changing radiological conditions. As a result, an individual entered areas with general area dose rates of 210 mrem per hour rather than the briefed dose rates of less than 50 mrem per hour. This issue was entered into the licensees corrective action program as Condition Report CR-2015-010211. Corrective actions included performing follow-up radiation surveys and implementing improvements to the high radiation area access control program. The inspectors determined that allowing a worker access into a high radiation without an adequate knowledge of the radiological conditions was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, entry into a high radiation area without adequate knowledge of the radiological conditions placed the individual at risk for unnecessary exposure. The finding was assessed using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, and was determined to be of very low safety significance (Green) because the performance deficiency was not an ALARA planning issue, there was not an overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The finding has a human performance cross-cutting aspect associated with work management, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority (H.5).
05000445/FIN-2016002-03	2016Q2	Comanche Peak	Licensee-Identified Violation	Comanche Peak Unit 1, Operating License NPF-87, Condition 2.G, Fire Protection, requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 78 and as approved in the Safety Evaluation Report and its supplements through Supplement 24. Comanche Peak Unit 2, Operating License NPF-89, Condition 2.G, Fire Protection, requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87 and as approved in the Safety Evaluation Report and its supplements through Supplement 27. The stations approved fire protection program includes Fire Protection Report, Revision 29, Section 5.3.8, Fire Area EO Control Room, includes Deviation 3c-1, Control Room Missile Door, which requires, in part, that since the control room missile door in the west wall is not a three hour rated fire door, the area of the turbine deck within 100 feet of the door is to be void of combustibles. Contrary to the above, on May 5, 2016, the licensee stored combustible materials within 100 feet of the control room missile door in the west wall. Specifically, licensee personnel identified that contractors had stored combustibles within the combustible free zone, and that no compensatory measures had been implemented for the deviation from the Fire Protection Report. The licensee implemented a periodic roving fire watch to compensate for the reduction in fire protection. The violation is more than minor because if left uncorrected, it could lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspector determined that the violation is of very low safety significance (Green) because the finding did not affect the ability of either unit to achieve safe shutdown. The violation was entered into the licensees corrective action program as CR-2016-004167.
05000446/FIN-2015003-03	2015Q3	Comanche Peak	Notice of Enforcement Discretion 15-4-02 for One Inoperable Train of Emergency Core Cooling Systems	Unit 2 Train B Safety Injection System Inoperable for Longer Than Allowed by Technical Specifications and Notice of Enforcement Discretion 15-4-02 Introduction. The inspectors opened an unresolved item associated with a potential noncompliance with Technical Specification 3.5.2 that occurred on July 10, 2015. Notice of Enforcement Discretion 15-4-02 was granted by the NRC staff agreeing not to enforce compliance with the technical specification completion time for an additional 25 hours. On July 7, 2015, a potential through wall leak from pipe segment SI-2-070 in the Unit 2, train B Safety Injection (SI) pump room was discovered during routine system walkdowns by a licensee engineer. Approximately 1-2 cups of boric acid accumulation was identified on the floor underneath valve 2SI-0055 (SIP 2-02 Suction Test Connection). The pipe insulation was removed to identify the source of the leakage, which was determined to be from a socket weld connection between the six inch suction piping for SI Pump 2-02 and the 34 inch vent piping to 2SI-0055. At 1:04 p.m. on July 7, 2015, the licensee declared unit 2, train B, Emergency Core Cooling System (ECCS) inoperable and entered Technical Specification 3.5.2, Condition B, for one or more (ECCS) trains inoperable for reasons other than one inoperable centrifugal charging pump, and at least 100 percent of the ECCS flow equivalent to a single operable ECCS train available. Required Action B.1 of Technical Specification 3.5.2 required restoration of the train(s) to an operable status within 72 hours. Further, Technical Specification 3.5.2 required that if Required Action B.1 could not be met within 72 hours, unit 2 would be required to enter Technical Specification 3.5.2 Condition C, Required Actions C.1 and C.2, and be in Mode 3 in 6 hours and Mode 4 in 12 hours. The licensees initial assessment determined the likely cause of the socket weld failure to be vibration induced fatigue failure. An attempted repair utilizing ASME Code Case N-666 was conducted on July 8, 2015. During the welding activity a small pinhole leak A-16 developed in the vent piping. The licensee then initiated alternate repair activities including a freeze seal on the affected piping, installation of a new vent line and valve (to facilitate post-repair filling and venting of the SI piping), and repair of the affected weld. The licensee requested a notice of enforcement discretion and an additional 25 hours to restore safety injection pump 2-02, such that the completion time of Required Action B.1 would expire at 2:04 p.m. on July 11, 2015. A notice of enforcement discretion was granted by the NRC staff at 9:20 a.m. on July 10, 2015. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific technical specifications in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine if there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item. (URI 05000446/2015003-03, Notice of Enforcement Discretion 15-4-02 for One Inoperable Train of Emergency Core Cooling Systems)
05000446/FIN-2016002-01	2016Q2	Comanche Peak	Failure to Correct Conditions Adverse to Quality	The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to correct a condition adverse to quality in safety-related equipment. Specifically, following an in-service testing failure of auxiliary feedwater check valve 2FW-091 in November 2012, the licensee performed an operability evaluation of the auxiliary feedwater system. However, the inspectors identified that the licensee failed to take corrective action to address the condition adverse to quality that resulted in the valve failing to seat properly. Consequently, the same valve failed a subsequent inservice test in November 2015. Following discovery of this issue, the licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The licensee entered this issue into corrective action program as CR-2015-10961. The licensees failure to correct a condition adverse to quality was a performance deficiency. 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 affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to correct auxiliary feedwater check valve 2FW-0191 failure to seat in November 2012 resulting in an additional failure in November 2015. Using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and 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 allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation, in that, the licensee failed to thoroughly evaluate issues to ensure that resolutions address extent of conditions. Specifically, the licensee failed to appropriately classify the issue of the check valve not seating and recognize this as a degraded condition (P.2).
05000446/FIN-2016002-04	2016Q2	Comanche Peak	Licensee-Identified Violation	Technical Specification 5.4.1.a states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a., identifies procedures for maintenance as required procedures. Work order 4831032 is a procedure established by the licensee for performing maintenance on diesel generator 2-02. The work order provided instructions for installation of the magnetic speed pickup sensor cable. Contrary to the above, from October 1996 through March 2, 2016, the licensee failed to install the unit 2 diesel generator 2-02 magnetic speed pickup sensor cable in accordance with the approved instructions. Specifically, the speed sensor cable conduit was not fully threaded onto the cable plug. This inadequate installation was present until 2016, when the conduit threaded connection was physically impacted at an undetermined time. The impact caused the conduit connection to break and the conduit to separate from the plug, leaving the cable leads exposed but intact. A licensee technician identified the broken connection during a system walk down on March 2, 2016. The licensee declared the diesel generator inoperable and restored the cable to its design configuration. The licensee analyzed the apparent thread engagement, and determined that, prior to the break in the conduit connection, the cable would have maintained its function in a seismic event, but after the break, the cable function could not be assured. The licensee determined that a failure of the cable would result in the diesel generator exceeding its allowed frequency, but would not result in a diesel generator failure to run. Because the time that the break occurred could not be determined, the diesel generator was assumed to be inoperable at the time of discovery. The violation is more than minor because it affected the configuration control attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspector determined that the violation is of very low safety significance (Green) because the finding did not represent a loss of system or function, and did not represent a loss of function of a single train for greater than its technical specification allowed outage time. The violation was entered into the licensees corrective action program as CR-2016-001941.
05000458/FIN-2002007-01	2002Q4	River Bend	Failure to properly lock open Valve CNM-FCV200	The inspectors identified an apparent violation of Technical Specification 5.4.1.a, which requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, Item 4.n requires instructions for operation of the condensate system. System Operating Procedure SOP-0007, Condensate System, Revision 21, required Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 to be locked open. On September 18, 2002, Valve CNM-FCV200 was found to be not properly locked open. The failure to properly lock Valve CNM-FCV200 in the open position resulted in unexpected closure of the valve and a loss of feedwater flow to the reactor vessel following a reactor scram. The final significance of this issue will be determined using the Significance Determination Process (Section 3.5).
05000458/FIN-2015009-01	2015Q2	River Bend	Vendor and Industry Recommended Testing Adequacy on Safety-related and Safety-significant Circuit Breakers	The team identified an unresolved item related to the licensees breaker maintenance and troubleshooting programs for safety-related and safety-significant circuit breakers. The charter tasked the team with inspecting the issues associated with Magne Blast breaker problems that occurred during and after the December 25, 2014, scram. The NRC team determined that breaker maintenance and troubleshooting practices extended beyond the Magne Blast breakers. The team identified that there were potential issues with safety-related Master Pact breakers and determined that maintenance procedures used to ensure that 4160 V and 13.8 kV safety-related and safety-significant breakers were being maintained and overhauled in a timely manner may not conform to industry recommended standards. The team identified that the licensees maintenance programs for Division I, II, III, and non-safety 4160 V and 13.8 kV breakers installed in the plant may not meet the standards recommended by the vendor, corporate, or Electric Power Research Institute (EPRI) guidelines. The licensees programs were based on EPRI documents TR-106857-V2 and TR-106857-V3, which were preventive maintenance program bases for low and medium voltage switchgear. However, the licensee appeared to only implement portions of the recommended maintenance program, and were not able to provide the team with engineering analyses or technical bases to justify the changes. The EPRI guidance was developed specifically for Magne Blast breakers based on industry operating experience, NRC Information Notices, and General Electric SILs/SALs. The NRC team was concerned that the licensee may not have performed the entire vendor or EPRI recommended tests, inspections, and refurbishments on the breakers since they were installed. The aggregate impact of missing these preventive maintenance tasks needs to be evaluated to determine if the reliability of the affected breakers has been degraded. Pending further evaluation of the above issue by the licensee and subsequent review by NRC inspectors, this issue will be tracked as URI 05000458/2015009-01, Vendor and Industry Recommended Testing Adequacy on Safety-related and Safety-significant Circuit Breakers.
05000458/FIN-2015009-02	2015Q2	River Bend	Failure to Establish Adequate Procedures to Perform Maintenance on Equipment that can Affect Safety-Related Equipment	The team reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to establish adequate procedures to properly preplan and perform maintenance that affected the performance of the B reactor protection system motor generator set. Specifically, due to inadequate procedures for troubleshooting on the B reactor protection system motor generator set, the licensee failed to identify a degraded capacitor that caused the B reactor protection system motor generator set output breaker to trip, which resulted in a reactor scram. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2014-06605 and replaced the degraded field flash card capacitor. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Event Screening Questions, this finding is determined to have a very low safety significance (Green) because the transient initiator did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not have been available. This finding has an evaluation cross-cutting aspect within the problem identification and resolution area because the licensee failed to thoroughly evaluate this issue to ensure that the resolution addressed the cause commensurate with its safety significance. Specifically, the licensee failed to thoroughly evaluate the condition of the field flash card to ensure that the cause of the trip had been correctly identified and corrected prior to returning the B reactor protection system motor generator set to service (P.2).
05000458/FIN-2015009-03	2015Q2	River Bend	Failure to Provide Adequate Procedures for Post-scram Recovery	The team reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to establish, implement and maintain a procedure required by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, Procedure OSP-0053, Emergency and Transient Response Support Procedure, Revision 22, which is required by Regulatory Guide 1.33, inappropriately directed operations personnel to establish feedwater flow to the reactor pressure vessel using the startup feedwater regulating valve as part of the post-scram actions. The startup feedwater regulating valve operator characteristics are non-linear and not designed to operate in the dynamic conditions immediately following a reactor scram. To correct the inadequate procedure, the licensee implemented a change to direct operations personnel to utilize one of the main feedwater regulating valves until the plant is stabilized. This issue was entered in the licensees corrective action program as Condition Report CR-RBS-2015-00657. 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 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 procedure directed operations personnel to isolate the main feedwater regulating valves and control reactor pressure vessel level using the startup feedwater regulating valve, whose operator was not designed to function in the dynamic conditions associated with a post-scram event from high power, and this challenged the capability of the system. The team performed an initial screening of the finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the team determined that the finding is of 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has an evaluation cross-cutting aspect within the problem identification and resolution area because the licensee failed to thoroughly evaluate this issue to ensure that the resolution addressed the cause commensurate with its safety significance. Specifically, the licensee failed to properly evaluate the design characteristics of the startup feedwater regulating valve operator before implementing the procedure to use the valve for post-scram recovery actions (P.2).
05000458/FIN-2015009-04	2015Q2	River Bend	Failure to Identify High Reactor Water Level as a Condition Adverse to Quality	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure a condition adverse to quality was promptly identified. Specifically, the licensee failed to identify, that reaching the reactor pressure vessel water Level 8 (high) setpoint, on December 25, 2014, was an adverse condition, and as a result, failed to enter it into the corrective action program. To restore compliance, the licensee entered this issue into their corrective action program as Condition Report CR-RBS-2015-00620 and commenced a causal analysis for Level 8 (high) trips. 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 adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to identify Level 8 (high) conditions and unplanned automatic actuations as conditions adverse to quality, would continue to result in the undesired isolation of mitigating equipment including reactor feedwater pumps, the high pressure core spray pump, and the reactor core isolation cooling pump. The team performed an initial screening of the finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the team determined that the finding is of 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has an avoid complacency cross-cutting aspect within the human performance area because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee tolerated leakage past the feedwater regulating valves, did not plan for further degradation, and the condition ultimately resulted in the Level 8 (high) trip of the running reactor feedwater pump on December 25, 2014 (H.12).
05000458/FIN-2015009-05	2015Q2	River Bend	Failure of the Plant-Referenced Simulator to Demonstrate Expected Plant Response	The team identified an apparent violation of 10 CFR 55.46(c)(1), Plant-Referenced Simulators, for the licensees failure to maintain the simulator so it would demonstrate expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to respond. As of January 30, 2015, the licensee failed to maintain the simulator consistent with actual plant response for normal and transient conditions related to feedwater flows, alarm response, and behavior of the startup feedwater regulating valve controller. Specifically, the River Bend Station simulator failed to correctly model feedwater flows and resulting reactor vessel level response following a scram, failed to provide the correct alarm response for a loss of a reactor protection system motor generator set, and failed to correctly model the behavior of the startup feedwater regulating valve controller. As a result, operations personnel were challenged in their control of the plant during a reactor scram that occurred on December 25, 2014. This issue has been entered into the corrective action program as Condition Report RBS-CR-2015-01261, which includes actions to initiate simulator discrepancy reports, investigate and resolve the potential fidelity issues, and provide training to operations personnel on simulator differences. 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 adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the incorrect simulator response adversely affected the operations personnels ability to assess plant conditions and take actions in accordance with approved procedures during the December 25, 2014, scram. The team performed an initial screening of the finding in accordance with inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Attachment 4, Initial Characterization of Findings. Using Inspection Manual Chapter 0609, Attachment 4, Table 3, SDP Appendix Router, the team answered yes to the following question: Does the finding involve the operator licensing requalification program or simulator fidelity? As a result, the team used Inspection Manual Chapter 0609, Appendix I, Licensed Operator Requalification Significance Determination Process (SDP), and preliminarily determined the finding was of low to moderate safety significance (White) because the deficient simulator performance negatively impacted operations personnel performance in the actual plant during a reportable event (reactor scram). This finding has an evaluation cross-cutting aspect within the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate this issue to ensure that the resolution addressed the extent of condition commensurate with its safety significance. Specifically, the licensees evaluation of the fidelity issue identified by the NRC in March 2014, focused on other training areas that used simulation, rather than evaluating the simulator modelling for additional fidelity discrepancies (P.2).
05000458/FIN-2015009-06	2015Q2	River Bend	Failure to Identify and Classify Operator Workarounds that Impacted Scram Recovery Actions	The team identified a finding for the licensees failure to follow written procedures for classifying deficient plant conditions as operator workarounds and providing compensatory measures or training in accordance with fleet Procedure EN-OP-117, Operations Assessment Resources, Revision 8. A misclassification of these conditions resulted in the failure of the operations department to fully assess the impact these conditions had during a plant transient. The failure to identify operator workarounds contributed to complications experienced during reactor scram recovery on December 25, 2014. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2015-00795. This performance deficiency is more than minor, and therefore a finding, because it had the potential to lead to a more significant safety concern if left uncorrected. Specifically, the performance deficiency contributed to complications experienced by the station when attempting to restore feedwater following a scram on December 25, 2014. The team performed an initial screening of the finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the team determined this finding is of 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has a consistent process cross-cutting aspect in the area of human performance because the licensee failed to use a consistent, systematic approach to making decisions and failed to incorporate risk insights as appropriate. Specifically, no systematic approach was enacted in order to properly classify deficient conditions (H.8).
05000458/FIN-2015010-01	2016Q1	River Bend	Technical Specification Allowed Outage Time During Loss of Non-Technical Specification Supported Systems	The team identified an unresolved item related to the licensees treatment of the control building chilled water system (HVK) chillers as a non-technical specification support system for other technical specification systems. The team noted that when an entire division of HVK chillers is out of service, such as chillers 1A and 1C for division I, the licensee would only enter the Technical Specification (TS) 3.7.3, Control Room Air Conditioning (AC) System, action statement for the condition of one control room AC subsystem being inoperable (condition A). The licensee did not enter TS action statements associated with inoperability of other components cooled by HVK chillers, such as the AC switchgear, DC switchgear, and vital inverters. The licensee, instead, has incorporated a safety evaluation for the Perry Plant (ML020950074), dated April 5, 2002, into the bases for TS 3.0.6 and applied that document as guidance: ...no TS limits the duration of the non-TS support subsystem outage, even though the single-failure design requirement of the supported TS systems is not met. However, by assessing and managing risk in accordance with 10 CFR 50.65(a)(4), an appropriate duration for the maintenance activity can be determined. The NRC team questioned whether the Perry Plants safety evaluation could be applied generically, if the licensee improperly incorporated the safety evaluation via the 10 CFR 50.59 process, if the guidance conflicted with section 9.2.10.3 of the Updated Safety Analysis Report (USAR) for River Bend Station, and if the safety evaluation for the Perry Plant conflicted with guidance found in Generic Letter 80-30, Clarification of the Term Operable As It Applies to Single Failure Criterion For Safety Systems Required by TS. The aggregate impact of these decisions resulted in the River Bend Station placing TS systems cooled by HVK, such as the AC switchgear, DC switchgear, and vital inverters, in a single-failure vulnerable configuration for durations exceeding the allowed outage time specified in the TS. Pending further evaluation of the above issue by NRC Headquarters staff via a Technical Specification interpretation request (ML15231A111) and subsequent review by NRC inspectors, this issue will be tracked as URI 05000458/2015010-01, Technical Specification Allowed Outage Time During Loss of Non-Technical Specification Supported Systems. Further discussion of performance deficiencies associated with the HVK chiller system is included in Section 2.6.a of this report.
05000458/FIN-2017001-01	2017Q1	River Bend	Failure to Follow Station Guidance on Control of Scaffolding	Green . The inspectors identified a non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to follow station maintenance procedures related to the control of scaffolding in the reactor building. Specifically, the licensee installed scaffolding less than two inches from safety -related containment unit cooler HVR -UC1 B without completing an engineering evaluation. The licensee entered this issue into their corrective action program as Condition Report CR- RBS -2016- 07963 . Corrective actions included removing the scaffolding. The licensees installation of scaffolding within two inches of a safety -related containment unit cooler , without completing an engineering evaluation, was a performance deficiency. 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, containment unit cooler HVR- UC1B was rendered inoperable by the incorrectly installed scaffolding and remained inoperable until the scaffolding was removed. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance ( Green) because the finding did not represent an actual loss of function of one or more trains of safety-related equipment for greater than its technical specification allowed outage time. This finding has a cross -cutting aspect in the area of human performance, avoid complacency , because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risks, even while expecting successful outcomes . Specifically, the station failed to implement appropriate error reduction tools when it did not perform and document Procedure EN -MA -133, Control of Scaffolding, Attachments 9.5 and 9.6 , which could have prevented the scaffolding construction error (H. 12).

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