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05000382/FIN-2008004-052008Q3WaterfordLicensee-Identified ViolationSection 3.33 of licensee Refueling Procedure RF-005-002, Refueling Equipment Operation, states that using the key override to move the refueling machine hoist in the outward direction with a fuel assembly in the core region would require entering the limiting condition for operation for Technical Specification 3.9.6. Technical Specification 3.9.6, requires suspending movement of fuel assemblies when the refueling mast overload cut off limit of less than or equal to 3350 pounds was unavailable. Contrary to the above, on May 18, 2008, the overload cut off limit was unavailable and operators placed the refueling machine in key override and moved a fuel bundle in the outward direction. The operators did not enter Technical Specification 3.9.6. The operators were in the process of moving fuel when the refueling machine computer had \"locked-up.\" In an effort to reboot the computer, licensee personnel placed the refueling machine in key override, which bypassed the refueling equipment interlocks. The intent was to place the mast in a position that would allow the computer to be rebooted. The personnel failed to realize that the actions were not permitted by Technical Specification and plant procedures
05000382/FIN-2008004-062008Q3WaterfordLicensee-Identified ViolationSection 3.2 of licensee Refueling Procedure RF-005-002 states, that while fuel movement is in progress, PEER check/verifications are required for grapple operation of the refueling machine including verification of proper Z-coordinate, and weight verification when raising the hoist. Section 3.31 states that operation of fuel handling equipment with an interlock bypassed raises the risk of damaging fuel assemblies and equipment. Finally, a caution statement in Section 5.1.8 states that if the refueling machine load varies more than 100 pounds during withdrawal of a fuel assembly, fuel movement should be terminated. Contrary to this, on May 18, 2008, operators proceeded to lift a fuel assembly without verifying that the load did not vary by more than 100 pounds. The load varied by about 1400 pounds. The senior reactor operator (PEER) noted the initial load of the fuel assembly of approximately 1500 lbs, and failed to check the load again during travel. Both the refueling bridge operator and the senior reactor operator failed to note that the grapple and fuel assembly had rotated approximately 25 degrees out of position with the hoist box. At approximately 193 inches, the grapple actuator came in contact with the hoist box and began lifting the hoist box. The refueling machine weight gauge indicated an increase in weight to 2900 lbs. The finding was of very low safety significance (GREEN) because it did not represent an actual event that upset plant stability or damaged fuel cladding. The licensee entered the violation in their corrective action program as Condition Report CR-WF3-2008-2423
05000382/FIN-2008005-012008Q4WaterfordEssential Chiller B Tube Rupture Due to Untimely Corrective ActionsThe inspectors reviewed a self revealing noncited violation of 10 CFR 50,Appendix B, Criterion XVI due to the failure by the licensee to take prompt corrective actions following the identification of an inadequate testing method used for determining the integrity of the Essential Chiller B heat exchanger tubing. Failure to take this timely action resulted in an inadvertent tube rupture and inoperability of Essential Chiller B. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5342.This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because, assuming worst case degradation of both the B and AB Essential Chillers failing, the redundant A Essential Chiller would still have been available for accident mitigation. This finding had a crosscutting aspect in Problem Identification and Resolution, specifically the Corrective Action Program aspect P.1(d) because the licensee failed to take appropriate corrective actions to address a degrading condition in a timely manner. Specifically, the failure to perform timely tube inspections of Essential Chiller B, following the identification of an inadequate testing methodology used for identifying Essential Chiller heat exchanger tubing degradation (Section 4OA2)
05000382/FIN-2008005-022008Q4WaterfordEssential Chiller AB Component Failure Due to Inadequate Procedural GuidanceThe inspectors reviewed a self revealing noncited violation of Technical Specification 6.8.1.a for failure to provide documented instructions appropriate to the circumstances as recommended in Appendix A of Regulatory Guide 1.33.The failure by the licensee to provide adequate guidance for the replacement of the Essential Chiller AB compressor motor temperature sensor resulted in there introduction of a failure mechanism that had previously been corrected. This subsequently led to the failure of the temperature sensor wiring and inoperability of Essential Chiller AB. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-5471.This finding was more than minor because it is associated with the Mitigating Systems attributes for Equipment Performance and would impact the availability and reliability of systems that respond to initiating events. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, and determined that it was of very low safety significance (Green) because the redundant Essential Chillers A and B would still have been available for accident mitigation. Based on the guidance provided in Manual Chapter 0612, Appendix B, Section1-5, Screen for Cross-Cutting Aspects, this finding did not have a crosscutting aspect because it was not considered to be reflective of current licensee performance. Specifically, the licensees failure to update the model work instructions in 2000 was a latent issue, whereby the licensee did not have a reasonable opportunity to identify the problem prior to August, 2008. In addition, the licensee has since instituted programs and processes such that the problem would not reasonably occur today (Section 4OA2)
05000382/FIN-2008005-032008Q4WaterfordOperability of safety Injection Valves SI-405A(B)The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III to address three examples of inadequate calculations involving shutdown cooling Valves SI-405A and SI-405B. The calculations were also used, in part, to support valve operability, which made an existing operability assessment invalid. First, a calculation performed by a contractor to estimate the bounding thrust requirements for pressure locking contained errors and used mathematical formulas out of their intended context without applying uncertainties to account for the differences. Recent operational experience with these valves was inconsistent with the calculation\'s conclusions. In addition, the licensee failed to meet their quality assurance program requirements that specified that engineers perform a design verification of the calculation prior to use. Second, the licensee\'s calculation, that demonstrated valve actuator thrust capabilities, contained errors. Specifically, it failed to account for the friction between the actuator piston disk and walls as well as the weight of components. Third, a calculation that determined that the temperature within the valve bonnet would not heat up during small break loss of coolant accidents and faulted steam generator accidents was inadequate, in that it failed to address a faulted steam generator event, it used heat transfer calculation methods on water that were intended only for solid materials, it failed to model all components, and it failed to determine the temperatures inside the valve bonnets, which was the overriding variable of interest. The licensee entered the finding into the corrective action program as Condition Report CR-WF3-2009-00127.This finding was more than minor because it was similar to non-minor finding Example 3.j in NRC Inspection Manual Chapter 0612 Appendix E, Examples of Minor Issues, in that there was a reasonable doubt concerning the operability of Valves SI-405A(B). The inspectors utilized NRC Manual Chapter 0609,Appendix G, Shutdown Operations Significance Determination Process, to characterize the significance of the issue. Using the worst case scenario of having both SI-405A(B) valves inoperable, the finding was of very low safety significance because multiple systems or components would still be available to remove decay heat and respond to a loss of inventory event. These systems included the emergency feedwater system, main feedwater system, auxiliary feed water system, atmospheric dump valves, charging pumps, safety injection tanks, and the high-pressure safety injection system. This performance deficiency would not result in any loss of instrumentation needed for safe shutdown and cool down of the plant. The finding had a crosscutting aspect in the area of problem identification and resolution (P.1(c)) because engineers failed to thoroughly evaluate the potential for valve pressure-locking. The calculations were completed in 2008 and were indicative of current performance(Section 4OA2)
05000382/FIN-2009002-012009Q1WaterfordFailure to Follow Commitment Tracking ProceduresThe inspectors identified a finding because the licensee inadvertently deleted procedural steps to recover an emergency diesel generator during a severe accident. The steps were part of a formal commitment to the NRC. The licensee had failed to follow the site commitment management program when making the procedure change and the procedure writer failed to understand the basis for the steps prior to deleting them. The licensee entered this finding in their corrective action program as Condition Reports CR-WF3-2009-0193 and CR-WF3-2009-1616. The finding was more than minor because, if left uncorrected, it could result in a more significant safety concern. Specifically, during a severe accident, operators would not have an appropriate mitigation strategy for starting an emergency diesel generator under certain severe accident conditions. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because the finding: (1) could result in a loss of functionality of an emergency diesel generator; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a crosscutting aspect in the area of Human Performance, Decision Making component (H.1(a)), because the licensee failed to use a systematic process when removing the procedural steps (Section 1R13)
05000382/FIN-2009002-022009Q1WaterfordFailure to Obtain Voltage Readings Following a Single Cell Battery ChargeThe inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion V (Instructions, Procedures and Drawings) because the licensee failed to implement instructions that were intended to help troubleshoot a defective 125 Vdc battery cell. In response to the degraded cell, the licensee had established additional measures to monitor the cell following charging to ensure proper cell operation. However, the licensee did not perform the monitoring. Once identified by the inspectors, the licensee performed more frequent cell tests. The licensee subsequently replaced the faulty cell. The licensee entered this finding into their corrective action program as Condition Reports CR-WF3- 2009-1088 and CR-WF3-2009-1099. The finding was more than minor because it could have resulted in a more significant safety concern if left uncorrected. Specifically, the normal monitoring period for the cell was weekly. The cell may not have remained operable between weekly tests. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it: (1) could have resulted in a loss of operability of the 125 Vdc battery; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of Problem Identification and Resolution, because the licensee failed to implement corrective measures intended to address a condition adverse to quality (P.1(d)) (Section 1R18). period for the cell was weekly. The cell may not have remained operable between weekly tests. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 screening worksheet, the finding was of very low risk significance because it: (1) could have resulted in a loss of operability of the 125 Vdc battery; (2) did not represent a loss of safety function; (3) did not represent an actual loss of a single train of equipment for more than its technical specification allowed outage time; (4) did not involve non-technical specification equipment; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a crosscutting aspect in the area of Problem Identification and Resolution, because the licensee failed to implement corrective measures intended to address a condition adverse to quality (P.1(d)) (Section 1R18)
05000382/FIN-2009005-012009Q4WaterfordFailure to follow radiation protection procedural requirementsThe inspectors reviewed a self-revealing noncited violation of Technical Specification 6.8.1 which resulted from a worker failing to follow radiation protection procedures. A contract radiation worker went to work near steam generator 1 rather than the area for which he/she was briefed and received multiple electronic dosimeter dose rate alarms, but did not leave the area until receiving a continuous dose alarm. In response, the licensee investigated the occurrence and restricted the individuals access. Additional actions were being evaluated. This issue was entered into the licensees corrective action program as Condition Reports CR-WF3-2009-05648 and WF3-2009-06852. This finding is greater than minor because it involved the program attribute of exposure control and affected the cornerstone objective in that the failure of the worker to follow procedural guidance resulted in the worker being unknowledgeable to the dose rates in all areas entered. The inspectors used the Occupational Radiation Safety Significance Determination Process and determined the finding had very low safety significance because it was not: (1) an as low as reasonably achievable (ALARA) finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding had a crosscutting aspect in the area of human performance, work practices component, because the worker failed to use human error prevention techniques such as self and peer checking (H.4(a)) (Section 2OS1)
05000382/FIN-2009005-022009Q4WaterfordReactor Coolant Pump Vapor Seal LeakageA self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee did not promptly correct reactor coolant pump vapor seal leakage that resulted in boric acid accumulation on the component cooling water heat exchanger and cover areas of three reactor coolant pumps. Corrective actions for this condition were implemented during Refueling Outage 15, but these corrective actions failed to correct the condition and the vapor seal leakage continued through operating Cycle 16. This resulted in some additional boric acid corrosion and degradation to reactor coolant pump covers and carbon steel component cooling water flanges. The licensee implemented a design modification to correct the condition and documented the condition in Condition Report CR-WF3-2009-5501. The licensees failure to promptly correct a condition adverse to quality is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment was still available. This finding had a crosscutting aspect in the area of human performance associated with work control in that the licensee did not effectively plan for the resources necessary to implement the postmaintenance testing associated with the corrective actions implemented during Refueling Outage 15, and therefore failed to discover that those corrective actions were inadequate to correct the condition (H.3(a)) (Section 4OA2)
05000382/FIN-2009005-032009Q4WaterfordFailure to Update Drawings after Design ChangeA self-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, was identified for the licensees failure to prescribe an activity affecting quality by documented instructions, procedures, or drawings appropriate to the circumstance. Specifically, for all reactor coolant pump heat exchanger to pump cover bolted connection gasket replacements between the refueling outage of 1986 (Refueling Outage 1) and the refueling outage of 2009 (Refueling Outage 16), the licensee prescribed the wrong gasket material, gasket size, and fastener preload because they had failed to incorporate a design change implemented during Refueling Outage 1 into their instructions, procedures, or drawings. Station Modification Package SMP-1427, an engineering change implemented during Refueling Outage 1 in response to industry operating experience, called for a thicker gasket, different gasket material, and an increased bolt preload in order to increase gasket compression and reduce the probability of leakage. As a consequence of failing to incorporate Station Modification Package SMP-1427 changes into procedures, all heat exchanger gasket replacements since Refueling Outage 1, four gasket replacements in total, have utilized thinner gaskets with less than the vendor recommended compression. The licensee documented this condition in Condition Report CR-WF3-2009-5501. The licensees failure to prescribe appropriate gasket replacement requirements is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability. The finding has very low safety significance because, although the finding contributes to the likelihood of a reactor trip, mitigation equipment is still available. This finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee did not institutionalize operating experience through changes to the station procedures (P.2(b)) (Section 4OA2)
05000382/FIN-2009005-042009Q4WaterfordLicensee-Identified ViolationTechnical Specification 6.8.1 requires written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Appendix A lists procedures for access control to radiation areas. Procedure EN-RP-100, Radworker Expectations, Revision 3, Section 5.3(9) requires the radiation work permit to be read, understood, and obeyed as a condition of radiologically controlled area access. Procedure EN-RP-100, Radworker Expectations, Revision 3, Section 5.4(3)(h) requires the worker know where to properly perform his/her task. Section 5.3(17) requires the worker be briefed and sign on the appropriate radiation work permit. Section 5.3(11) requires the worker know the radiological conditions in the work area. The licensee identified an example of a worker entering a high radiation area using an inappropriate radiation work permit and without knowing the dose rates in the area. On October 24, 2009, a security officer entered shutdown heat exchanger Room B and received an electronic dosimeter dose rate alarm. The room was posted as a high radiation area and dose rates within the area were as high as 140 millirem per hour. The officer entered the radiological controlled area using Radiation Work Permit 2009005, Tours and Inspection in All Radiological Controlled Areas, Except High Radiation Areas, Locked High Radiation Areas, Very High Radiation Areas, and the Reactor Containment Building. Because the radiation work permit did not allow entry into high radiation areas, radiation protection personnel did not anticipate the officer would enter the room and did not brief the officer on the dose rates in the area. In response, the licensee conducted a human performance error review and counseled the officer. This finding was of very low safety significance because it did not involve an actual or substantial potential of an overexposure. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2009-05648.
05000382/FIN-2010004-012010Q3WaterfordLicensee-Identified ViolationOn May 23, 2008, the licensee identified that one transmitter for Steam Generator No. 1 and one transmitter for Steam Generator No. 2 were inoperable during a previous cycle of operation with the associated technical specification of limiting condition for operation had not been entered. Specifically, the licensee operable with one less than the total number of channels required for startup or power operations without placing the channel in bypass or a trip condition within one hour. This was a violation of Technical Specification Table 3.3-1. The licensee determined that the cause was an inadequate procedure that did not provide enough detail to assure proper valve manipulation and venting to properly calibrate the transmitters. The licensee documented this issue in Condition Report CR-WF3-2008-02107. The corrective actions included enhancing training for temporary outage workers on how to properly calibrate the equipment, revising calibration procedures to provide specifics on appropriate set-up of sensing lines and to identify other reference leg level transmitter calibration procedures which are inadequate. This finding is more than minor because it is associated with the procedure quality attribute of the Mitigating System Cornerstone and affects 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). This finding was considered to have very low safety significance (Green) after using NRC Manual Chapter 0609, Phase 1 screening worksheet under the Mitigation Systems Cornerstone. This licensee-identified finding involved a violation of Technical Specification 3.3.1, Reactor Protective Instrumentation.
05000382/FIN-2011002-012011Q1WaterfordFailure to follow Operability Determination Process for a Degraded and Non-Conforming condition Related to Reactor Coolant Pump N9000 SealsThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because the licensee did not adequately implement the operability determination process requirements in accordance with EN-OP-104, Operability Determination Process. Specifically, the licensee did not frequently and regularly review a degraded and nonconforming condition associated with the reactor coolant pump N-9000 stage seals as required by EN-OP-104. As a result, the licensee did not perform a new operability determination after assumptions and compensatory measures identified in the original operability determination changed. This also led to compliance issues with technical specifications and missed maintenance rule functional failures. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1965. The immediate corrective actions included revising the operability determination to account for the current configuration. The planned corrective actions included the licensee replacing the degraded reactor coolant pump seals during the next two refueling outages. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects 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. Specifically, the licensee did not frequently and regularly review a degraded and nonconforming condition that had the potential to lead to a small loss of coolant accident. The senior resident inspector performed the initial significance determination for the failure to perform an adequate operability determination associated with increased reactor coolant pump seal leakage when transitioning the plant to a shutdown condition, using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because the finding involved reactor coolant system leakage in excess of the technical specification limit of 1 gallon per minute for unidentified leakage (leakage was up to 6 gallons per minute). A Region IV senior reactor analyst performed a Phase 2 significance determination using Inspection Manual 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. However, this particular finding was not well suited for the Phase 2 process. The senior reactor analyst subsequently performed a Phase 3 significance determination. The analyst found that the finding was of very low safety significance (Green). Potentially risk important sequences included those involving reactor coolant pump seal failures (where leakage could surpass the capacity of the charging system) and long term station blackout. The relatively small amount of leakage helped to mitigate the significance. The finding has a cross-cutting aspect in the corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and report ability conditions adverse to quality (P.1.c of IMC 0310)
05000382/FIN-2011002-022011Q1WaterfordControl Room Envelope PreconditioningThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because the licensee did not conduct required technical specification surveillance testing on equipment in an as-found condition. Specifically, the licensee performed corrective maintenance (preconditioning) on the system to achieve better results, prior to completing the surveillance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1927. The immediate corrective action included the performance of the control room envelope tracer gas test. The finding is more than minor because it is associated with the barrier performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not properly perform testing on equipment to evaluate barrier performance. The inspectors evaluated this finding using IMC 0609 Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the finding doesnt represent a degradation of the radiological barrier, or the smoke and toxic gas barrier functions provided for the control room. The finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately plan work activities by incorporating the need for planned contingencies, compensatory actions, and abort criteria (H.3.a of IMC 0310)
05000382/FIN-2011003-012011Q2WaterfordFailure to Evaluate and Adequately Monitor Activities Associated with the Internal Conditions of the Condensate and Refueling Water Storage Pool StructuresThe inspectors identified a noncited violation of 10 CFR 50.65(a)(3) because the licensee did not evaluate or adequately monitor activities associated with the condition of the condensate and refueling water storage pool structures. Specifically, the licensee did not evaluate the internal condition of the storage pools through the performance of appropriate preventive maintenance activities and did not evaluate these activities at least every refueling cycle, where practical, per industry-wide operating experience. As a result, there was no preventive maintenance developed for this activity even though industry-wide operating experience documented previous issues of concrete deterioration due to contact with boric acid over a long period of time. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-1168. The planned corrective actions include the development of appropriate preventive maintenance activities to examine the internal condition of the storage pool structures during refuel outages. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects 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, no preventive maintenance to monitor the internal condition of the storage pools, would impact the reliability of the structures. The inspectors evaluated this finding using Inspection Manual Chapter 0609 Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding had a crosscutting aspect in the operating experience component of the problem identification and resolution area because the licensee did not implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs.
05000382/FIN-2011003-022011Q2WaterfordFailure to Update the FSAR following Modifications to the Reactor Coolant Pump Vapor SealsThe inspectors identified a Severity Level IV noncited violation of 10 CFR 50.71(e) because the licensee did not revise the Final Safety Analysis Report (FSAR), as updated, with information consistent with plant conditions. Specifically, the licensee did not update Section 5.4.1.3 of the FSAR for Waterford Steam Electric Station, Unit 3, following modifications to the reactor coolant pump vapor seals in 2007 and 2009, respectively. As a result, the licensee did not promptly identify and correct FSAR noncompliance. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7421. The planned corrective actions include revising the FSAR, as updated, and replacing the degraded reactor coolant pump seals during the next two refueling outages. The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC\'s ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation was more than minor because the longstanding and incorrect information in the FSAR, as updated, had a material impact on safety and licensed activities. The material impact was that the modifications to the reactor coolant pump vapor seals, created the conditions for a reactor coolant pump seal loss of coolant accident inside containment, which could have potentially impacted licensed activities. The inspectors determined the violation was a Severity Level IV (very low safety significance) since the information that was not updated in the FSAR, was not used to make an unacceptable change to the facility nor did it impact a licensing or safety decision by the NRC. The inspectors determined there was a cross-cutting aspect in the corrective action component of the problem identification and resolution area. Specifically, the licensee did not thoroughly evaluate and take adequate actions in a timely manner to update the FSAR to be consistent with plant conditions.
05000382/FIN-2011003-032011Q2WaterfordFailure to Implement Written Procedures for Restoring a Time Delay Relay Associated with the A Output BreakerA self-revealing noncited violation of Technical Specification 6.8.1.a occurred because the licensee did not implement written procedures and work order instructions. Specifically, maintenance personnel did not follow Procedure ME-007-005, Time Delay Relay Setting Check, Adjustment, and Functional Test , during the lifting leads process for restoration of a time delay relay (EG EREL2327-C) associated with an A emergency diesel generator (EDG) maintenance activity. As a result, the A EDG output breaker did not automatically close during technical specification surveillance testing because the leads on the relay were wired incorrectly. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3190. The immediate corrective action included the re-wiring of the relay to allow the A EDG to automatically close to the safety-related bus. The finding was more than minor because it was associated with the human 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 (i.e., core damage). Specifically, the licensee did not ensure the availability, reliability and capability of the A EDG through the use of human error prevention techniques. The inspectors performed the initial significance determination for the diesel generator output breaker failure. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. A Region IV senior reactor analyst performed a Phase 2 significance determination and used the pre-solved worksheet from the Risk Informed Inspection Notebook for the Waterford-3 Nuclear Power Plant, Revision 2.01a. The senior reactor analyst considered the output breaker a part of the emergency diesel generator component boundary. Assuming a one year exposure period, the finding was potentially Yellow, which warranted further review. Therefore, the senior reactor analyst performed a bounding Phase 3 significance determination. The analyst determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was approximately 5.4E-7/year. The dominant core damage sequences included loss of offsite power events, failure of the output breaker recovery action, independent failure of the other emergency diesel generator and failure to recover offsite power in 4 hours. Equipment that helped mitigate the risk included the ability of an operator to recover the output breaker. The finding had a crosscutting aspect in the work practices component of the human performance area because the licensee did not communicate human performance error prevention techniques, such as self and peer checking, and proper documentation of activities.
05000382/FIN-2011003-042011Q2WaterfordFailure to Implement Work Order Instructions to Restore a Feedwater Heater Drain ValveA self-revealing finding occurred because maintenance technicians did not follow written procedures during the calibration of a level switch that controls feedwater heater drain valve FHD703A. Specifically, the technicians did not perform concurrent verification checks as required by documented work order instructions (WO-00180716) and procedures to ensure that personnel restore and/or manipulate components to the correct position following maintenance. As a result, the feedwater heater drain valve was left in a closed position, which caused a spurious isolation of a string of feedwater heaters. The isolation of the feedwater heaters caused operators to down power the reactor to approximately 72 percent. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2009-7420. The immediate corrective actions included restoring the feedwater heater drain valve to its proper position. The finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and 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. Specifically, the human error caused an event that upset plant stability during power operation. The inspectors evaluated this finding using Inspection Manual Chapter 0609 Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The finding had a crosscutting aspect in the work practices component of the human performance area because the licensees personnel proceeded in the face of uncertainty or unexpected circumstances.
05000382/FIN-2011003-052011Q2WaterfordLoss of Reactor Coolant Inventory during the Assembly of Incore Instrumentation FlangesOn April 30, 2011, while the plant was in Mode 5 during Refueling Outage 17 (RFO-17), operators identified reactor coolant inventory coming from the incore instrumentation flanges. Based on the restoration from other maintenance activities, operators locked closed the pressurizer spray line vent valve RC-309 on the prior shift to return the valve to its normal position. However, the plant conditions at the time of restoration required RC-309 to be open to provide a reactor coolant vent path during the assembly of incore instrumentation (ICI) flanges. The inspectors noted that the loss of reactor coolant inventory occurred twice prior to the licensee securing from assembly of the ICI flanges. The inspectors questioned whether the operators understood the true condition of the plant since the closed vent path caused inaccurate reactor coolant system level indication. The inspectors also questioned the amount of reactor coolant inventory loss because operators identified water coming out of the ICI flanges twice while charging with the charging pumps. The inspectors reviewed the initial condition report and the maintenance work activities surrounding this evolution. However, the licensee had not completed the apparent cause evaluation prior to the end of the inspection period. This item was unresolved pending further review and investigation of the licensees apparent cause evaluation such that the inspectors can determine if there are performance deficiencies associated with this loss of reactor coolant inventory event.
05000382/FIN-2011003-062011Q2WaterfordLicensee-Identified ViolationCriterion III of Appendix B to 10 CFR Part 50 requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, prior to September 13, 2009, the licensee did not verify the adequacy of design basis calculation ECM98-067, Limiting Single Failure Thermal-Hydraulic Analysis of Spent Fuel Pool, through the performance of a design review. As a result, a single failure of the spent fuel pool level switch would have caused a loss of all the spent fuel pool cooling pumps. The inspectors determined that the finding was of very low safety significance because it did not result in a loss of cooling to the spent fuel pool, whereby operators could preclude restoration of cooling prior to pool cooling, did not result from fuel handling errors that caused damage to fuel clad integrity or dropped assembly, and did not result in a loss of spent fuel pool inventory greater than ten percent of the spent fuel pool volume. The licensee entered this issue into their corrective action program as CR-WF3-2009-4908.
05000382/FIN-2011003-072011Q2WaterfordLicensee-Identified ViolationCriterion III of Appendix B to 10 CFR Part 50 requires, in part, that measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to safety-related functions of the structures, systems, and components. Contrary to the above, prior to February 8, the licensee did not select and review the proper clamp for the suitability of the application that was essential to the safety-related function of the EDG fuel oil supply line. As a result, the inadequate clamp would have rendered the EDG inoperable prior to its 30 day mission time. The inspectors determined that the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability for the probabilistic risk analysis (PRA) mission time of twenty-four hours. The licensee entered this issue into their corrective action program as CR-WF3-2010-0889.
05000382/FIN-2011004-012011Q3WaterfordFailure to Evaluate and Adequately Perform Preventive Maintenance Activities Associated with Dry Cooling Tower Process Analog Control CardsThe inspectors identified a non-cited violation of 10 CFR 50.65 (a)(3) because the licensee did not adequately evaluate and take into account, where practical, industry operating experience related to preventive maintenance activities for the dry cooling tower process analog control cards. Specifically, internal and industry-wide operating experience documented previous failures of process analog control cards due to age-related degradation after about 15 years of services. The vendor and industry performed a benchmark in 2003, and noted that the average service life is about 12 to 15 years. The licensee initially provided a preventive maintenance activity to replace the cards on a 20 year interval. However, the licensee deleted the preventive maintenance activities in March of 2009. The licensee determined that the cards were non-critical and had no justification of deleting the preventive maintenance activities. The inspectors noted that after the deletion of the preventive maintenance activities and prior to the 15 year service internal, the licensee experienced additional unplanned failures of several process analog control cards that challenged dry cooling tower reliability. The licensee entered this issue into their corrective action program for resolution as CR-WF3- 2011-1356. The immediate corrective action includes the evaluation of the preventive maintenance activity for the dry cooling tower process analog control cards. The planned corrective action includes the reinstatement of the preventive maintenance activity that aligns with industry operating experience. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The process analog control card failures challenged the system availability and reliability. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because the condition is not a design or qualification deficiency, did not represent the loss of a system safety function, did not represent an actual loss of a single train of equipment for more than its Technical Specification completion time, and did not screen as potentially risk-significant due to an external initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalizes operating experience through change to station processes, procedures, equipment, and training programs
05000382/FIN-2011004-022011Q3WaterfordFailure to Adequately Implement a Reactor Coolant System Drain Down ProcedureThe inspectors documented a self-revealing non-cited violation of Technical Specification 6.8.1.a because the licensee did not adequately implement Operating Procedure OP-001-003, Reactor Coolant System Drain Down, during the installation of the incore instrumentation flanges. Specifically, the licensee did not establish a reactor coolant system vent path while maintaining reactor coolant level below 26 feet for the assembly of the incore instrumentation flanges as required by OP-001-003. As a result, the licensee experienced a loss of reactor coolant inventory from three unassembled incore instrumentation flanges, which spilled onto the reactor vessel head insulation and filled the upper annulus cavity of the reactor vessel. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3163 and CR-WF3-2011-3636. The immediate corrective actions included opening the pressurizer spray line vent valve (RC-309) to establish a reactor coolant system vent path. The finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and affects 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. The inspectors performed the initial significance determination for the failure to adequately implement operating procedures using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on the conditions of the plant at the time of the event. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding has a cross-cutting aspect in the work control component of the human performance area because the licensee did not appropriately coordinate work activities in incorporating actions to address the impact of the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities
05000382/FIN-2011004-032011Q3WaterfordFailure to Provide Adequate Testing for a Shutdown Cooling Heat Exchanger Outlet Stop Check ValveThe inspectors documented a self-revealing non-cited violation of 10 CFR 50.55a, Codes and Standards, because the licensee did not establish and maintain an adequate testing program for a shutdown cooling heat exchanger outlet stop check valve (CS-117A) in accordance with Mandatory Appendix II, Check Valve Condition Monitoring Program, of the American Society of Mechanical Engineers Operation and Maintenance Code 2001 through 2003. Specifically, the licensee did not provide adequate inservice testing to detect degradation of seat leakage on the stop check valve CS-117A. As a result, the operating train of shutdown cooling experienced a flow diversion when the licensee opened the upstream containment spray isolation header valve to fill the containment spray riser. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-3350 and CR-WF3-2011-5841. The immediate corrective action included the closure of the upstream isolation valve and the initiation of a work order to address seat leakage on the stop check valve CS-117. The planned corrective action includes the development of an augmented test to determine appropriate seat leakage criteria for the stop check valve. The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affects 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. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, since the degraded stop check valve upsets plant stability and challenge critical safety functions during shutdown conditions. The inspectors evaluated the significance of the finding and determined that it did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the inspectors determined that the finding is of very low safety significance (Green). This finding did not have a cross-cutting aspect associated with it because the licensee established the check valve condition monitoring program prior to the past three years. Therefore it is not reflective of current plant performance.
05000382/FIN-2011004-042011Q3WaterfordFailure to Promptly Identified and Correct Work Order Instructions used for Technical Specification Surveillance RequirementsThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly identify and correct work order instructions used to perform technical specification surveillance requirements. Specifically, the licensee did not provide adequate work order instructions or acceptance criteria to perform technical specification surveillance requirements associated with safety-related dry cooling tower fans and control room air handling units. The inspectors initially identified the issue of concern with the control room air handling units in December 2010. However, the licensee did not perform an adequate extent of condition review to determine if other work order instructions used to perform technical specification surveillance requirements contained adequate instructions and acceptance criteria. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-7223 and CR-WF3-2011-6254. The immediate corrective actions include revisions to the work order instructions in order to provide appropriate quantitative and qualitative acceptance criteria. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects 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 inspectors concluded that without appropriate quantitative and qualitative acceptance criteria this would affect the availability, reliability, and capability of the dry cooling tower fans and control room air handling units. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in corrective action program component of the problem identification and resolution area because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary
05000382/FIN-2011004-052011Q3WaterfordFailure to Comply with Technical Specifications Surveillance Requirement 4.0.3 and the Limiting Conditions for Operation for Technical Specifications 3.0.3The inspectors identified a non-cited violation of Technical Specification (TS) because the licensee did not enter or comply with the technical specification action requirements. Specifically, the licensee did not enter or comply with Technical Specification Surveillance Requirement 4.0.3 upon discovery of a never performed surveillance related to a safety-related relay contact for the Essential Chilled Water system. The licensee discovered the issue on July 27, 2011. However, the licensee did not enter TS 4.0.3 until August 12, 2011. Subsequently, when the licensee entered TS 4.0.3, the licensee did not perform a risk evaluation within 24 hours, as directed by the technical specification surveillance requirement. The licensee, per Technical Specification 4.0.3, has up to 24 hours to perform a risk evaluation or enter the applicable technical specification limiting condition for operation immediately. The inspectors determined that the licensee exceeded the allowed 24 hours and then did not enter the limiting condition for operation for Technical Specification 3.0.3 once the requirements for Technical Specification 4.0.3 and other applicable technical specifications had not been met. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-5779. The immediate corrective action included the performance of a special test instruction to demonstrate operability of the safety-related relay. The finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affects 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 inspectors concluded that a failure to comply with TS 4.0.3 and 3.0.3 affects the availability and reliability of the Essential Chill Water system. The inspectors evaluated this finding using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen potentially risk significant due to external events. The finding has a cross-cutting aspect in decision-making component of the human performance area because the licensee did not make a safety-significant or risk-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained
05000382/FIN-2011004-062011Q3WaterfordUntimely Actions to Correct Repetitive Dry Cooling Tower Fan FailuresThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because the licensee did not promptly correct a condition adverse to quality related to repetitive failures of the dry cooling tower fans to start and run in fast speed. Specifically, the licensee did not perform corrective actions to resolve the failure mechanism of the fast speed breaker relay in a timely manner. As a result, additional failures occurred over a period of several years prior to the implementation of corrective action in March 2011. The licensee entered this issue into their corrective action program for resolution as CR-WF3- 2011-2546. The corrective action includes a plan to replace the affected components inside the dry cooling tower fan breakers with a new design. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems 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 inspectors concluded that the reoccurrence of the problem challenged the reliability, and capability of the dry cooling tower fans. The inspectors performed the initial significance determination for the failure to start the dry cooling tower fans in fast speed using NRC Inspection Manual Chapter 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Attachment 1 of Appendix G, Shutdown Operations Significance Determination Process, based on fact that the failures of the breaker relay to start in fast speed occurred during refueling outages. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment since adequate mitigating equipment remained available and it did not constitute a loss of control, as defined in Appendix G. This finding has a cross-cutting aspect in the resource component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals
05000382/FIN-2011004-072011Q3WaterfordFailure to Follow Apparent Cause Evaluation Process ProcedureThe inspectors identified a finding because the licensee did not implement procedure EN-LI-119, Apparent Cause Evaluation Process. Specifically, the licensee did not follow the requirements provided in procedure EN-LI-119, Section 5.3.3 (k), to complete corrective actions in a timely manner for the intersystem leakage in the gas waste management system. As a result, no corrective action implementation occurred prior to additional equipment failures for the system. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2011-0934. The immediate corrective action included the reevaluation of the causal determination and development of an implementation plan to complete the corrective actions in a timely manner. The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events cornerstone and affects 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. The intersystem leakage of the gas decay tanks increase the likelihood of a potential explosive mixture and continued to challenge technical specification oxygen concentration limits. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The Initial screening directed the inspectors to use Appendix F, Fire Protection Significance Determination Process, because the finding is a contributor to a fire initiation event. The inspectors assigned a degradation rating of low to the finding since the oxygen concentration levels in the gas decay tanks were below the limit of an explosive mixture. The inspectors determined that the finding is of very low safety significance (Green) because the finding minimally impacted the fire protection capabilities of the fire area. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not minimize long-standing equipment issues and maintenance deferrals
05000382/FIN-2011004-082011Q3WaterfordFailure to Submit a LER within 60 days after Discovery of an EventThe inspectors identified a non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit required Licensee Event Reports (LERs) within 60 days after discovery of conditions that required a report. Specifically, the inspectors identified three instances of untimely LERs submittals for conditions related to an inoperable emergency feedwater pump, a single point vulnerability of spent fuel pool pumps, and a degraded fuel oil supply line for the Train A emergency diesel generator. The licensee submitted the reports at 332,163, and 101 days after discovery of the conditions, respectively. As a result, the licensee exceeded the 60 days for each condition that required a report. The inspectors noted that this is also contrary to the licensees reportability procedure UNT-006-010, Event Notification and Reporting. The licensee entered this issue into their corrective action program for resolution as CR-WF3-2010-5923. The immediate corrective actions include the performance of a human performance error review. The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC\'s ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor because it occurred repeatedly within a two year period and the licensee missed opportunities to identify the issue. The NRC relies on the licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done, this impacts the NRCs ability to carry out its statutory mission. The finding has a cross-cutting aspect in the work practices component of the human performance area because the licensee did not define and effectively communicate expectations regarding procedural compliance
05000382/FIN-2012002-012012Q1WaterfordFailure to Develop Preventive Maintenance Tasks for Critical Limit Switches on Component Cooling Water Inlet Isolation ValvesA Green self-revealing, non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a occurred because the licensee did not establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches. Specifically, the licensee had not developed preventive maintenance tasks to lubricate or replace critical limit switches that provide a permissive for the operation of the dry cooling tower fans. As a result, on February 4, 2011, the limit switch on valve CC-135A failed to operate as designed and rendered an entire train of fans inoperable. The licensee entered this condition into their corrective action program as CR-WF3-2011-0679 for resolution. The immediate corrective action included the lubrication of the limit switch and the manual stroking of the valve to obtain free and smooth movement of the degraded equipment. The planned corrective actions included the development of a preventive maintenance task to lubricate and replace the limit switches on a scheduled frequency. The failure to establish procedures for performing preventive maintenance tasks on the dry cooling tower component cooling water inlet isolation valves CC-135A and CC-135B limit switches is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, since there was no preventive maintenance task for lubrication and replacement of the equipment, the limit switches can become stuck and render an entire train of dry cooling tower fans inoperable. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to an external initiating events. The inspectors also concluded that no cross-cutting aspect is applicable to this finding because the performance deficiency is not reflective of current performance
05000382/FIN-2012002-022012Q1WaterfordFailure to Identify and Perform Testing to Demonstrate Performance of SAFETY-RELATED ValvesThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, because the licensee did not identify and perform testing on a safety-related component to demonstrate that it would perform satisfactorily in service in accordance with requirements contained in applicable design documents. Specifically, the licensee did not identify and perform proper testing for the essential chiller hot gas bypass valves RFR-106A, B, and C. As a result, the licensee could not demonstrate that the safety-related valves would perform satisfactorily in service without performing a test and operability evaluation. The licensee entered this condition into the corrective action program as CR-WF3-2012-0632 and CR-WF3-2012-0659. The immediate corrective action included testing the hot gas bypass valves to demonstrate the proper performance of their safety function. The failure to identify and perform testing to demonstrate that a safety-related component would perform satisfactorily in service in accordance with requirements contained in applicable design documents is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the hot gas bypass valve closure is required to ensure the essential chiller can perform its safety function during all design basis accident conditions. The inspectors determined the significance of the finding using the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of a safety function of a system or a single train for greater than its technical specification completion time, and did not screen as potentially risk-significant due to any external initiating events. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date test procedures were available to demonstrate that equipment performance is adequate to assure nuclear safety
05000382/FIN-2012003-012012Q2WaterfordFailure to Establish Adequate Procedural Guidance to Control Feedwater Heater Level Control ValvesA self-revealing finding occurred because the licensee did not establish adequate procedural guidance to control feedwater heater level control valves. Specifically, the procedures used to control the settings for the valves did not contain guidance that properly adjusted the proportional gain and air pressure input to ensure the valves open quickly during a transient. As a result, multiple failures in the feedwater heater drain system resulted in a feedwater pump A trip and a subsequent reactor power cutback. The licensee entered this condition into their corrective action program as CR-WF3-2012-1729 for resolution. The corrective actions included a revision of the procedure and loop calibration settings for the feedwater heater level control valves. The failure to provide adequate guidance that properly adjusted the proportional gain to ensure the valves open as designed is a performance deficiency. The performance deficiency is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affects 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. Specifically, multiple feedwater heater control valve failures resulted in a reactor power cutback that upset plant stability. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it only contributed to the likelihood of a reactor trip and not the likelihood that mitigation equipment or functions would not be available. This finding has a cross-cutting aspect in the resources component of the human performance area in that the licensee did not ensure that complete, accurate, and up-to-date design documentation for loop calibration settings was available to assure nuclear safety
05000382/FIN-2012003-022012Q2WaterfordFailure to Provide Adequate Design Control Measures for Verifying or Checking the Adequacy of the Ultimate Heat Sink Thermal Performance AnalysisThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III because the licensee did not provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis. Specifically, the licensee did not ensure that the design calculation used to determine the required number of wet cooling tower fans needed to operate the plant under normal and design conditions utilized the correct equation. As a result, the incorrect calculation provided reasonable doubt as to the operability of the wet cooling tower fans. The licensee entered this issue into their corrective action program as CR-WF3-2012-1395. The immediate corrective actions taken to restore compliance included a preliminary analysis of the condition and actions to perform a review of the methodology, inputs, and assumptions for the ultimate heat sink thermal performance calculations. The failure to provide adequate design control measures for verifying or checking the adequacy of the ultimate heat sink thermal performance analysis is a performance deficiency. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the wet cooling tower fans are required to be operable for heat removal following all accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to determine the significance. The inspectors determined that the finding is of very low safety significance (Green) because it is a design deficiency confirmed not to result in a loss of operability or functionality of the ultimate heat sink. This finding has a cross-cutting aspect in the decision making component of the human performance area in that the licensee did not conduct effectiveness reviews of safety-significant decisions to verify the validity of the underlying assumptions, identify possible unintended consequences, and determine how to improve future decisions
05000382/FIN-2012003-032012Q2WaterfordLicensee-Identified ViolationCriterion III of Appendix B to the 10 CFR Part 50 requires, in part, that measures shall be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to safety-related functions of structures, systems, and components. Contrary to the above, prior to April 22, 2012, the licensee did not select and review for the suitability of the application of tubing support and routing that was essential to the safety-related function for essential chiller A. Specifically, the licensee did not assure that adequate tubing support and routing was used to preclude wear related problems due to equipment vibration. As a result, the use of a tie-wrap did not provide adequate support or meet clearance requirements for instrument tubing for essential chiller A. This would have rendered the essential chiller inoperable prior to its 30 day mission time. The inspectors determined that the finding was of very low safety significance because it was a design deficiency confimed not to result in a loss of operability for the probabilistic risk analysis (PRA) mission time of twenty-four hours. The licensee entered this issue into their corrective action program as CR-WF3-2012-2022.
05000382/FIN-2012004-012012Q3WaterfordFailure to Establish Procedural Controls to Ensure That Licensed Operators Could Perform Immediate and Time Critical Operator Actions Associated with Security and Fire EventsThe inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification 6.8.1.a because the licensee did not establish procedural controls to ensure that the assigned minimum staff of licensed operators could perform immediate and time critical operator actions associated with a security or fire event. Specifically, the licensee did not establish procedural guidance to restrict licensed operators from leaving the protected area. As a result, the licensee could not ensure that operators would respond in a timely manner to perform immediate and time critical operator actions required by a fire or security event. The licensee entered this issue into their corrective action program as CR-WF3-2012-3815. The immediate corrective actions taken to restore compliance included the issuing of a standing instruction to instruct the assigned minimum staff of licensed operators to remain in the protected area unless officially relieved of their duties. The failure to establish procedural controls to ensure that licensed operators could perform immediate and time critical steps associated with security and fire events was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee would have been challenged to complete immediate and time critical steps with licensed operators being outside the protected area. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it could be risk significant for external events. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.0 E-7/year. The risk important sequences included control room fires that required a control room evacuation. The short duration of the operator being outside the protected area reduced the risk significance. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement and institutionalize operating experience through changes to station processes and procedures
05000382/FIN-2012004-022012Q3WaterfordFailure to Identify and Correct Degraded Conditions Associated with the Auxiliary Component Cooling Water Heat Exchanger Outlet Temperature Control ValveThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct conditions adverse to quality related to the header A auxiliary component cooling water heat exchanger outlet temperature control valve ACC-126A. Specifically, the licensee did not promptly identify and correct degraded conditions associated with the valves shaft bushings, a pneumatic transducer that controls the valve actuator, and its soft seat. As a result, the licensee declared the valve inoperable on several occasions. The licensee entered this issue into their corrective action program as CR-WF3-2012-03280. The immediate corrective actions taken to restore compliance included the replacement of all the degraded components. The failure to promptly identify and correct multiple degraded conditions associated with the auxiliary component cooling water heat exchanger outlet temperature control valve ACC-126A was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded components challenged the closed safety function of the valve and its ability to maintain an adequate water inventory for the wet cooling tower following a loss of coolant accident. The inspector used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding was of very low safety significance (Green) because the bounding change to the core damage frequency was less than 4.2E-7 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary
05000382/FIN-2012004-032012Q3WaterfordFailure to Identify and Correct a Torn Diaphragm of a SAFETY-RELATED Air Operated Valve Associated with the Emergency Feedwater SystemThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI because the licensee did not promptly identify and correct a condition adverse to quality associated with the B emergency feedwater backup control valve EFW-223B. Specifically, the licensee did not promptly identify and correct internal leakage from tears in the EFW-223B actuator diaphragm. As a result, these internal tears in the diaphragm caused excessive leakage that affected two nitrogen accumulators used to operate EFW-223B and other safety related valves. The licensee entered this issue into their corrective action program as CR-WF3-2012-0860. The immediate corrective actions taken to restore compliance included the replacement of the diaphragm and to determine the extent of condition for other air-operated valves with the same type, make, and model diaphragm. The planned corrective action included the revision of the air operated valve program post maintenance tests to identify similar problems. The failure to promptly identify and correct tears in the internal actuator diaphragm of the B emergency feedwater backup control valve EFW-223B was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the internal leakage of EFW-223B affected two safety-related nitrogen accumulators and their ability to provide nitrogen gas to other connected safety related valves following a loss of offsite power event. The inspector used the NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding required a detailed analysis because it involved a potential loss of one train of safety related equipment for longer than the technical specification allowed outage time. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the finding is of very low safety significance (Green) because the bounding change to the core damage frequency is less than 1E-9 per year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary
05000382/FIN-2012004-042012Q3WaterfordFailure to Adequately Access and Manage Risk Before Performing Maintenance Activities Associated with NON-STANDARD LiftsThe inspectors identified a non-cited violation of 10 CFR 50.65(a)(4) because the licensee did not assess and manage the increase in online risk involved with maintenance activities that lifted heavy loads over safety related equipment. Specifically, the licensee did not assess and manage the integrated plant online risk prior to performing heavy load lifts in the train B dry cooling tower fan area when installing a temporary work platform to support the steam generator replacement project. As a result, the licensee did not implement additional risk management actions as required by their procedure EN-WM-104, OnLine Risk Assessment. The licensee entered this condition into the corrective action program as CR-WF3-2012-4195 and CR-WF3-2012-4489. The immediate corrective action taken to restore compliance was to re-evaluate and change the integrated risk classification from a normal risk to a high-risk level and implement the required risk management actions. The failure to adequately assess and manage overall plant risk prior to performing maintenance activities that lifted heavy loads over the train B dry cooling tower fan area was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to identify non-standard lifts over safety related equipment as high risk prevented the licensee from taking additional risk management actions to limit the likelihood of an event that would upset plant stability. The inspectors used NRC Inspection Manual 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The initial screening directed the inspectors to use Appendix K Maintenance Risk Assessment and Risk Management Significance Determination Process to determine the significance of the finding. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, a senior reactor analyst determined that the finding was very low safety significance (Green) because the bounding risk deficit was approximately 1E-7/year. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision making component of the human performance area because the licensee did not make safety significant or risk significant decisions using a systematic process to ensure safety was maintained
05000382/FIN-2012005-032012Q4WaterfordReactor Auxiliary Building Roof Flood Protection IssueThe inspectors identified an unresolved item (URI) related to a probable maximum precipitation (PMP) flooding event that could potentially affect safety-related equipment located on the roof of the reactor auxiliary building. During a walk-down of the reactor auxiliary building roof, the inspectors identified safety-related components, electrical conduit, and power cables that are vulnerable to a PMP flooding event. The inspectors questioned whether the licensee had performed an analysis to demonstrate that the safety-related components and electrical equipment related to conduits and power cables were adequately protected from a design basis PMP. At that time, the licensee could not identify an analysis that addressed the roof drains and scupper capacities of the area or if the safety-related equipment identified on the roof would function during and after a PMP event. The licensee entered this issue of concern into their corrective action program as CR-WF3- 2012-7520. The inspectors opened this unresolved item to determine if there is a performance deficiency associated with design control since the licensee did not have an analysis demonstrating adequate flooding protection. This item is unresolved pending a review of a flooding design analysis and other related documentation: URI 05000382/2012005-03: Reactor auxiliary building roof flood protection issue.
05000382/FIN-2013003-012013Q2WaterfordFailure to provide design control measures to withstand the effects of flooding on the reactor auxiliary building roofThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, because the licensee did not provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a probable maximum precipitation (PMP) flooding event on the reactor auxiliary building (RAB) roof areas. Specifically, the licensee did not provide an analysis to demonstrate that adequate flood protection existed from the effects of a PMP flooding event on safety-related components and electrical equipment located on the roof of the RAB in the main steam isolation valve (MSIV) wing areas. As a result, the licensee did not perform an analysis to determine if expected ponding levels from a PMP flooding event would challenge safety-related components and electrical equipment such as the emergency feedwater flow control and isolation valves and cables, main steam isolation valves and cables, atmospheric dump valves, and back-up nitrogen accumulator components. The licensee entered this issue into their corrective action program as CR-WF3-2012- 7520. The immediate corrective actions taken to restore compliance included the performance of a preliminary analysis to show that the installed scuppers and roof drains have margin to protect against a local PMP flooding event and that the ponding depth would have little or no affect on the safety-related equipment and cables located in the MSIV wing areas. The failure to provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a local PMP on the RAB roof areas was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating System cornerstone and 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-related equipment located on the RAB roof in the MSIV wing areas are required to safely shutdown and maintain the reactor in a cold shutdown condition following accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not identify potential flooding issues completely, accurately, and in a timely manner commensurate with their safety significance.
05000382/FIN-2013003-022013Q2WaterfordFailure to Update Fuel Handling Accident Analysis in the Updated Final Safety Analysis ReportThe inspectors identified a Severity Level IV non-cited violation for the licensees failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e). Specifically, from July 1981 to April 18, 2013, the licensee failed to update the methodology, the data input, and the resulting limits for the fuel bundle drop accident analysis in the Waterford Steam Electric Station, Unit 3, Updated Final Safety Analysis Report (UFSAR), Section 15.7.3.4, Design Basis Fuel Handling Accidents. This violation was entered into the licensees corrective action program as Condition Report CR-WF3-2013-0193. The failure to update the methodology, the data input to the calculation, and the resulting limits for the fuel bundle drop accident analysis in Section 15.7.3.4 of the UFSAR in accordance with 10 CFR 50.71(e) is a performance deficiency. This performance deficiency was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, the inspectors determined that the performance deficiency is a Severity Level IV non-cited violation. This noncited violation had no cross-cutting aspect because there was no finding associated with this traditional enforcement violation.
05000382/FIN-2013003-032013Q2WaterfordFailure to comply with Action 4 of TS 3.3.1 during shutdown in Modes 4 and 5The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 because the licensee did not take action to suspend operations that involved reactivity changes to accomplish startup activities with only one excore nuclear instrumentation (ENI) logarithmic (log) channel operable. Specifically, the licensee did not take action to suspend operations involving diluted water additions to the volume control tank and temperature increases with a positive moderator temperature coefficient (MTC) without the required number of operable log channels. As a result, the licensee did not comply with Action 4 of TS LCO 3.3.1 because they did not suspend all operations involving positive reactivity changes with the exception of minimum reactivity additions due to temperature fluctuations or operations, which are necessary to maintain fluid inventory. The licensee entered this issue into their corrective action program as CR-WF3-2013-2166 and CR-WF3- 2013-3182. The immediate corrective actions taken to restore compliance included the discontinued use of water additions to the volume control tank and the increase of RCS temperatures with a positive MTC until the licensees personnel returned an additional log channel to service. The failure to comply with TS LCO 3.3.1, Action 4, was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the LCO for the log power channels ensures that adequate information is available to verify core reactivity conditions while shutdown to minimize the probability of the occurrence of postulated events. The inspectors used Checklist 4 contained in Attachment 1 of the NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists, to evaluate this finding. The inspectors determined that the finding did not meet the reactivity guidelines because the licensee did not comply with TS LCO 3.3.1, Action 4. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment and was not similar to any of the examples requiring a phase two or phase three analyses. The inspectors also determined that the licensee maintain the required shutdown margin to preclude inadvertent criticality in the shutdown condition. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision-making component of the human performance area in that the licensee did not make a safety-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety was maintained. This included obtaining interdisciplinary input and reviews on safety-significant decisions.
05000382/FIN-2013003-042013Q2WaterfordFailure to submit an LER after discovery that manual handwheels on AOVs were not functionalThe inspectors identified a non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit a Licensee Event Report (LER) in a timely manner after the discovery of a reportable event. Specifically, the licensee failed to submit a required LER within 60 days after the discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corresponding back-up nitrogen accumulators. The licensee determined that the manual hand-wheel function on the essential chiller and emergency feedwater isolation and backup flow control valves did not work. The licensee was aware of the condition that existed but did not adequately evaluate the condition as a part of their reportability review. The licensee entered this issue into their corrective action program as CR-WF3-2013-2564. The immediate corrective actions taken to restore compliance included a new reportability review of the condition and the development of an LER. The failure to submit a required LER within 60 days after discovery of a condition that required a report was a violation of NRC requirements. The inspectors determined that this violation was also a performance deficiency. However, the inspectors determined that the performance deficiency was minor. The inspectors considered this issue to be within the traditional enforcement process because it had the potential to impact the NRC\'s ability to perform its regulatory oversight function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors determined that the violation was a Severity Level IV because it was similar to an example provided in Section 6.9 of the NRC Enforcement Policy. The inspectors did not assign a cross-cutting aspect to this non-cited violation because there was no finding associated with this traditional enforcement violation.
05000382/FIN-2013003-052013Q2WaterfordFailure to provide adequate post modification testing instructions for vibration monitoring on the feedwater piping systemThe inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, because the licensee did not provide post modification testing instructions for activities affecting quality that were appropriate to the circumstances and that included appropriate acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring of the feedwater piping system that included appropriate acceptance criteria following the installation of the new replacement steam generators. As a result, the plant experienced an automatic reactor trip and a subsequent down power due to an increase in vibrations on the feedwater piping system without appropriate acceptance criteria and monitoring during power ascension. The licensee entered this issue into their corrective action program as CR-WF3-2013-0445. The immediate corrective actions taken to restore compliance included the implementation of a revised vibration-monitoring plan to include appropriate acceptance criteria and the development of engineering changes to mitigate vibration effects on the feedwater piping system. The failure to provide adequate post modification testing instructions for vibration monitoring of feedwater piping system following steam generator replacement was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control 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 power operations. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because the transient initiator did not contribute to the likelihood that mitigation equipment or functions would not be available. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement operating experience through changes to station equipment to support plant safety.
05000416/FIN-2014002-012014Q1Grand GulfFailure to Ensure Scaffold Activity Would not Interfere with Fire Brigade ResponseThe 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-2014002-022014Q1Grand GulfFailure to Control a Locked High Radiation Area Due to Unsecured Highly Radioactive Materials Stored in the PoolThe 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-2014002-032014Q1Grand GulfLicensee-Identified ViolationTitle 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-2014002-042014Q1Grand GulfLicensee-Identified ViolationTechnical 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-2015001-012015Q1Grand GulfFailure to Take Timely Corrective Actions Associated with Division 1 and 2 Standby Service Water Pump House Ventilation System Due to Degraded RelaysThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take timely corrective actions to correct a condition adverse to quality associated with the division 1 and 2 standby service water pump house ventilation systems. Specifically, in June 2011, the licensee identified that relays associated with the standby service water system pump house ventilation system failed due to age/environmental degradation, which resulted in an unplanned inoperability of the standby service water system. However, the licensee did not implement timely corrective actions for replacing these relays, which resulted in the inoperability of the division 1 standby service water system in December 2014, and again in January 2015. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2015-00739. The short-term corrective actions included replacing all of the division 1 and 2 standby service water ventilation pump house relays in February and early March 2015. The inspectors determined that the failure to take timely corrective actions to replace degraded relays in the standby service water pump house ventilation system was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, the inspectors determined the issue to be of very low safety significance (Green) because all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. The inspectors determined that this performance deficiency was not indicative of current plant performance, and therefore no cross-cutting aspect was considered.
05000416/FIN-2015001-022015Q1Grand GulfFailure to Follow a Procedure Resulting in the Unplanned Inoperability of the Reactor Core Isolation Cooling SystemThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, for failure to follow a procedure which resulted in the unplanned inoperability of the reactor core isolation cooling system. This occurred when licensee technicians tested for continuity between incorrect points, while performing surveillance activities related to the residual heat removal system. This resulted in an invalid group 4 isolation signal and an isolation of the reactor core isolation cooling steam supply. The licensee entered this issue into the corrective action program as Condition Report CR-GGN- 2015-01532, and took immediate corrective actions to stop the residual heat removal system surveillance activity and restore the reactor core isolation cooling system to service. The failure to properly follow the surveillance procedure, which resulted in the unplanned inoperability of the reactor core isolation cooling system, was a performance deficiency. 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. Specifically, the licensees failure to properly follow the surveillance procedure resulted in the unplanned inoperability of the reactor core isolation cooling system, which adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and 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 inspectors determined that the finding was of very low safety significance (Green) in that the issue did not affect the design or qualification of the reactor core isolation cooling system; did not represent a loss of the reactor core isolation cooling system function (in that the isolation could have been promptly reset by procedures, had the system operation been required); and did not represent loss of function for greater than the Technical Specification allowed outage time. The inspectors determined this finding had cross-cutting aspect in the area of human performance associated with avoiding complacency, in that the I&C technicians did not implement appropriate error reduction tools to ensure the meter was connected to the correct points, which resulted in the invalid group 4 isolation signal, and inoperability of the reactor core isolation cooling system (H.12).
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