Semantic search

This page or action requires JavaScript to work. Please enable JavaScript in your browser or use a browser where it is supported, so that functionality can be provided as requested. For further assistance, please have a look at the noscript help page.

Options | Search Clear all entries |

Condition

Printout selection

Options

Parameters [+]

limit:

The maximum number of results to return

offset:

The offset of the first result

link:

Show values as links

headers:

Display the headers/property names

mainlabel:

The label to give to the main page name

intro:

The text to display before the query results, if there are any

outro:

The text to display after the query results, if there are any

searchlabel:

Text for continuing the search

default:

The text to display if there are no query results

class:

An additional CSS class to set for the table

transpose:

Display table headers vertically and results horizontally

sep:

The separator between results

prefix:

Control display of namespace in printouts

Sort options

Delete

Add sorting condition

ResultCodeDebugExtra

To embed this query inline into a wiki page use the code below.

{{#ask: [[Category:Finding]] [[Inspector::Ray Azua]]
 |?Quarter
 |?Site
 |?Title
 |?Description
 |format=table
 |limit=50
 |offset=0
 |sort=
 |order=asc
 |mainlabel=
 |searchlabel=[Findings List]
}}

Results 1 – 50 previous 50 20 50 100 250 500 next 50

JSON CSV RSS RDF

Query log

{
    "query_string": "[[Category:Finding]] [[Inspector::Ray Azua]]",
    "query_source": "SMWSQLStore",
    "query_time": "0.1901",
    "from_cache": false
}

	Quarter	Site	Title	Description
05000285/FIN-2011004-01	2011Q3	Fort Calhoun	Failure to incorporate design information into procedures	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to incorporate design information into procedures for operation of the component cooling water system for temporary off-normal system conditions during refueling. The failure to ensure that the minimum flow assumption contained in calculation FC06700 was incorporated in component cooling water operating procedures is a performance deficiency. This was reasonably within the licensee ability to foresee and correct. The performance deficiency is more than minor as it affected the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as, power operations. Since the finding affects the safety of the reactor during refueling outages, forced outages, and maintenance outages, it was evaluated using Inspection Manual Chapter 0609, Appendix G. The finding did not require quantitative assessment and therefore is of very low safety significance or green. A crosscutting aspect was not assigned as none were reflective of current plant performance
05000285/FIN-2011004-02	2011Q3	Fort Calhoun	Failure to provide adequate procedures to ensure leak before break commitment	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V for failure to have adequate instructions, procedures, or drawings including appropriate quantitative or qualitative acceptance criteria to ensure they can detect reactor coolant leakage, as required by the Updated Safety Analysis Report, using the containment dew point instrument or containment sump level instruments. Title 10 CFR Part 50, Appendix B, Criterion V states, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to this, the inspectors determined that the licensees failure to have adequate instructions, procedures, or drawings including appropriate quantitative or qualitative acceptance criteria to ensure they can detect a one gallon per minute leak in four hours was a performance deficiency. This was within the licensees ability to foresee and correct. The performance deficiency is more than minor as it affected the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Since the finding occurred during power operation and included structures, systems, and components where existing Significance Determination Process guidance is not adequate to provide reasonable estimates of the finding significance within the established Significance Determination Process timeliness goal of 90 days, the finding was evaluated using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. Using Table 4.1, Qualitative Decision Making Attributes for NRC Management Review, the finding was determined to be of very low safety significance (Green). This finding does not have a crosscutting aspect as the performance characteristic described by a potential crosscutting aspect did not occur within the last three years
05000285/FIN-2011004-03	2011Q3	Fort Calhoun	Failure to Identify and Correct the Lack of Calibration for the HE-2 Crane Load Weighing System	The inspectors identified a noncited violation of 10 CFR 50 Appendix B Criterion XVI for the failure to identify and correct a condition adverse to quality. Specifically, with regard to the calibration of the load weighing system for the HE-2 crane prior to its use in lifting the spent fuel transfer cask, loaded with spent fuel, out of the spent fuel pool. This issue was entered into the licensees corrective action program as Condition Report 2009-3186. The failure by the licensee to promptly identify and correct the condition whereby the HE-2 crane load weighing system had not been calibrated or tested for an extended period of time leading up to its use during the lift of the spent fuel transfer cask on July 7, 2009, is a performance deficiency. The performance deficiency was determined to be more than minor because it adversely impacted the spent fuel pool fuel handling attribute of the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed on more than one occasion to identify and correct a condition whereby the load cell for the HE-2 crane was neither calibrated nor tested prior to lifting the spent fuel transfer cask, loaded with spent fuel, out of the spent fuel pool. Using Attachment 4 of Inspection Manual Chapter 0609, the inspectors determined that this finding has a very low safety significance (Green) because it did not result in a fuel handling error that caused damage to fuel clad integrity or a dropped assembly. The finding was not found to be indicative of current plant performance and thus no crosscutting aspect was identified
05000285/FIN-2011004-04	2011Q3	Fort Calhoun	Licensee-Identified Violation	Technical Specification 2.1.4 states, in part, that the Reactor coolant systems operational LEAKAGE shall be limited to: No Pressure Boundary LEAKAGE. Contrary to this, pressure boundary leakage occurred during the operating cycle prior to the April 2011 refueling outage. The licensee removed the cracked line and replaced it. The finding was determined to be Green as it would not have resulted in exceeding any technical specification limit for reactor coolant system leakage. Because this violation was of very low safety significance and it was entered into the licensees corrective action program as CR 2011-3198, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000285/FIN-2014005-01	2014Q4	Fort Calhoun	Failure to establish Appropriate Preventive Maintenance and Failure to Identify Raw Water SSC Maintenance Rule Performance Criteria Exceeded and thereby establish Monitoring Requirements for the SSC	The inspectors identified an NCV of very low safety significance of 10 CFR 50.65 paragraph (a)(2) Requirements for Monitoring the Effectiveness of Maintenance of Nuclear Power Plants, because the licensee did not demonstrate that performance of a component was being effectively controlled through appropriate preventive maintenance, and did not monitor the performance of the component against licensee-established goals to provide reasonable assurance that the component was capable of fulfilling its intended function. Specifically, the licensee failed to demonstrate that the performance of raw water system valve HCV-2875A was being effectively controlled through appropriate preventive maintenance and failed to monitor the valves performance against licensee established goals when performance criteria were exceeded. Corrective actions taken for this violation included revising the Maintenance Rule performance criteria assessment for this component, classifying the component as 10 CFR 50.65 (a)(1), and specifying goals, corrective actions, and additional monitoring for the component. The licensees failure to demonstrate component performance through appropriate preventive maintenance, and the failure to identify that system performance criteria had been exceeded, and as a result, the failure to perform an evaluation of the system for 50.65 (a)(1) goals, corrective actions, and monitoring, was a performance deficiency within the licensees ability to foresee and correct. The finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the Cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify that valve HCV-2875A performance criteria had been exceeded resulted in a delayed assessment of this component and additional failures occurred in the intervening timeframe which adversely affected the overall reliability of the raw water system. The inspectors screened the finding in accordance with NRC IMC 0609, Appendix A, the Significance Determination Process (SDP) for Findings at Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its TS allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution and the Evaluation aspect because the licensee failed to appropriately evaluate the preventive maintenance for valve HCV-2875A to demonstrate component performance and failed to correctly evaluate a functional failure against system performance criteria to ensure system goals, corrective actions, and monitoring were identified.
05000285/FIN-2014005-02	2014Q4	Fort Calhoun	Failure to determine the availability of local population data for use in estimating changes in the EPZ population	The NRC identified a Green non-cited violation for the licensees failure to determine the availability of year 2013 state and local population data in estimating annual changes in the plume exposure emergency planning zone population. The failure to determine whether State and/or local population data was available for 2013 was a performance deficiency within the licensees ability to forsee and correct. Appendix E to 10 CFR Part 50, Section IV.5, states, in part, that during the years between decennial censuses, nuclear power reactor licensees shall estimate emergency planning zone permanent resident population changes once a year using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. Contrary to the above, Fort Calhoun Station failed in 2013 to estimate emergency planning zone permanent resident population changes using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. Specifically, Fort Calhoun Station failed to determine whether State and local government population data was available prior to performing the analysis. The issue was entered into the licensees corrective action system as Condition Report 2014-12474. This finding is more than minor because the issue is associated with procedure quality and offsite Emergency Preparedness cornerstone attributes and adversely affected the Emergency Preparedness cornerstone objective. The finding was evaluated using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated February 24, 2014, and was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements, was not a loss of planning standard function, and was not a degraded planning standard function. The planning standard function was not degraded because including state and local 2013 data would not have required the current emergency planning zone time estimate to be updated. There are no immediate safety or security concerns associated with this finding. This finding was assigned a cross-cutting aspect in the area of human performance associated with work management because the licensee failed to understand the scope of work performed by a contractor on their behalf, and failed to ensure the contractor fully complied with regulatory requirements.
05000285/FIN-2015009-01	2015Q4	Fort Calhoun	Failure to Take Adequate Corrective Action to Preclude Repetition of a Significant Condition Adverse to Quality Associated with Emergency Diesel Generator Room Water Intrusions	The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take corrective actions to prevent repetition of a significant condition adverse to quality. Specifically, since February 2009, the licensee failed to take corrective actions to prevent repetitive water intrusions from the Auxiliary Building HVAC room (Room 82) into the number one Emergency Diesel Generator room (Room 63). The inspectors determined that the licensees failure to implement corrective actions to preclude repetitive water intrusions into Room 63 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone. Specifically, water intrusion events from Room 82 into Room 63 could challenge the reliability of the emergency diesel generator when relied upon during a loss of offsite power. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Question, inspectors determined that the finding was of very low safety significance (Green). The finding has a problem identification and resolution cross-cutting aspect within the area of Resolution, because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3).
05000285/FIN-2015009-02	2015Q4	Fort Calhoun	Failure to Revise Procedures and Perform Additional Training	The team evaluated a self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies... are promptly identified and corrected. Specifically, prior to September 30, 2015, the licensee failed to revise procedures, and perform additional operator training, to prevent the inadvertent opening of steam bypass and steam dump valves during plant startup, and any subsequent plant impacts. In response to this issue, the licensee initiated a condition report to document these corrective actions. This finding was entered into the licensees corrective action program as Condition Report CR-FCS-2015-13718. The team determined that the failure to take timely corrective actions to revise procedures and complete additional training to correct a condition adverse to quality, was a performance deficiency. This finding was more than minor because it was associated with the initiating events cornerstone objective of configuration control to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to take recommended corrective actions to revise procedures and perform additional operator training to ensure proper alignment of the steam dump and bypass valves controller during startup. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the team determined that the finding was determined to have very low safety significance (Green) since the transient did not result in a reactor trip or loss of mitigation equipment. The finding has a problem identification and resolution cross-cutting aspect in the area of Operating Experience, because the licensee failed to systematically and effectively collect, evaluate, and implement relevant internal operating experience in a timely manner (P.5).
05000298/FIN-2014005-01	2014Q4	Cooper	Failure to Follow Procedure for Post Maintenance Testing	The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to follow Special Procedure GEH-TP-116, Procedure for the Operation and Maintenance of the REMTAKE-2/D-100 Modified REMTAKE 2, Revision 3, for postmaintenance testing following corrective maintenance. Specifically, the licensee did not follow post-maintenance testing requirements associated with the calibration of the bleeder valve for the REMTAKE-2/D-100 tool following corrective maintenance to address water intrusion. This resulted in the bleeder valve being misadjusted and nullifying the fail-safe feature of the REMTAKE-2/D-100 tool. With the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REMTAKE-2/D-100 tool and dropped onto the reactor core top guide when the supplemental employee inadvertently pressed the disengage button. No reactor fuel was damaged as indicated by normal radiation levels and air samples on the refuel floor and reactor water coolant samples. The licensees immediate corrective actions for the event was to suspended all in-vessel maintenance activities and remove REMTake-2/D-100 grapple from service and determined functionality of the tool. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2014-06809. The licensees failure to follow the post-maintenance testing requirements in Special Procedure GEH-TP-116 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the associated objective of maintaining functionality of fuel cladding. Specifically, with the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when a supplemental employee inadvertently pressed the disengage button. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 09, 2014, inspectors determined that the finding was of very low safety significance (Green) because the finding did not impact the fuel barrier because it: (1) does not increase the potential for failure of the freeze seal or if unmitigated have the potential to cause a disruption of residual heat removal/decay heat removal or a loss of inventory event; (2) does not involve two or more adjacent control rods with the potential to, or actually, add postive reactivity; and (3) does not degrade the ability to isolate a drain down or leakage path. The finding has a cross-cutting aspect in the area of human performance associated with the field presence component because the licensee failed to ensure supervisory and management oversight of work activities including contractors and supplemental personnel (H.2).
05000298/FIN-2014005-02	2014Q4	Cooper	Implementation of Enforcement Guidance Memorandum 11-003, Revision 2, Causes Conditions Prohibited by Technical Specifications	During Refueling Outage 28, Cooper Nuclear Station performed Operations with a Potential for Draining the Reactor Vessel (OPDRV) activities while in Mode 5 without an operable secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measure to terminate the uncovering of fuel. Secondary containment is required by TS 3.6.4.1 to be operable during OPDRV. The required action for this specification is to suspend OPDRV operations. Therefore, entering the OPDRV without establishing secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). The NRC issued Enforcement Guidance Memorandum (EGM) 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliances with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to: (1) adhere to the NRC plain language meaning of OPDRV activities, (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times, (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5, and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities. The inspectors reviewed this Licensee Event Report for potential performance deficiencies and/or violations of regulatory requirements. The inspectors reviewed the stations implementation of the Enforcement Guidance Memorandum 11-003, Revision 2, during operations with a potential for draining the reactor vessel. Specific observations included: 1. The inspectors observed that the operations with a potential for draining the reactor vessel activities were logged in the control room narrative logs, and that the log entry appropriately recorded the standby source of makeup designated for the evolutions. 2. The inspectors noted that the reactor vessel water level was maintained at least greater than 21 feet above the top of the reactor pressure vessel flange as required by Technical Specification 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designed in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours. 3. The inspectors verified that the operations with a potential for draining the reactor vessels were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the operations with a potential for draining the reactor vessels. The inspectors verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events. Technical Specification 3.6.4.1 requires, in part, that secondary containment shall be operable during operations with a potential for draining the reactor vessel. Technical Specification 3.6.4.1, Condition C, requires the licensee to initiate actions to suspend operations with a potential for draining the reactor vessel immediately when secondary containment is inoperable. Contrary to the above, from October 3, 2014 to October 22, 2014, Cooper Nuclear Station performed operations with a potential for draining the reactor vessel activities while in Mode 5 without an operable secondary containment. Specifically, the station conducted the following seven operations with a potential for draining the reactor vessel activities without an operable secondary containment: Draining reactor recirculation pump without the jet pump plugs fully installed Control rod drive maintenance Removal of jet pump plugs associated with reactor recirculation pump B maintenance Venting the control rod drives Defeating the scram function for two control rod drives and support IVVI inspections Reactor recirculation pump A seal maintenance Control rod drive freeze seal These conditions were reported as conditions prohibited by Technical Specifications. The licensee entered this issue into its corrective action program as Condition Reports CR-CNS-2014-06293. Since this violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy, and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request within 4 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The Licensee Event Report is closed.
05000298/FIN-2015001-01	2015Q1	Cooper	Inadequate Operations Procedure	The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the inadequate Operations Procedure 2.2.7, Condensate Storage and Transfer System, Revision 56. Specifically, the procedure did not require that the affected system, either the high pressure coolant injection system or the reactor core isolation cooling system, be declared inoperable when one or more of the high pressure coolant injection or reactor core isolation cooling test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, or RCIC-MOV-33, were moved off of their closed (passive safety function position) seats. The license entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-00274. The failure to establish and maintain a correct filling procedure required by Technical Specification 5.4.1.a. was a performance deficiency and resulted in the licensees failure to declare the high pressure coolant injection and reactor core isolation cooling systems inoperable when required to do so. The performance deficiency is more than minor, and therefore a finding, because it is associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the high pressure coolant injection and reactor core isolation cooling systems were not declared inoperable when their test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, and RCIC-MOV-33, were taken off their normally closed (passive safety function position) seats. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-ofservice for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction techniques. Specifically, licensee personnel fell into a pattern of acceptance regarding Procedure 2.2.7. This resulted in a failure to question the lack of an operability caution statement, even though there was other guidance in the inservice inspection program to that effect (H.12).
05000298/FIN-2015002-01	2015Q2	Cooper	Failure to Prevent Reactor Thermal Power from Exceeding 2419 MWt for Preplanned Activity	The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the licensees failure to appropriately implement General Operating Procedure 2.1.10, Station Power Changes, Revision 107. Specifically, the procedure required in Step 10.3 that the licensee, Ensure any pre-planned evolution (e.g., pressure change, flow change, etc.) will not result in operation greater than 2419 MWt. On May 8, 2015, the licensee failed to implement Step 10.3 of General Operating Procedure 2.1.10, when they failed to reduce power to ensure that reactor power did not exceed 2419 MWt as the reactor recirculation motor generator B scoop tube was unlocked. As a result of this failure to reduce power for this planned evolution, reactor power increased to 2422 MWt. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-04259. The performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown as well as power operations. Specifically, the licensee did not know the condition of the reactor recirculation motor generator set B potentiometer prior to unlocking it and failed to reduce power such that when the scoop tube was unlocked, the resulting power increase would not exceed 2419 MWt. The inspectors screened the finding using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Section C, Reactivity Control Systems, which resulted in a Yes answer to Question 2 since the finding involved control manipulations that unintentionally added positive reactivity. This referred the inspectors to Inspection Manual Chapter 0609, Appendix M, Significance Determination Using Qualitative Criteria. A Senior Reactor Analyst performed a bounding qualitative evaluation and determined that the finding was of very low safety significance (Green) because of the relatively small magnitude of the overpower event, the prompt operator actions to return power to below the licensed limit upon discovery, and the fact that the overpower event did not result in any failure of the fuel cladding. This finding has a cross-cutting aspect in the area of human performance associated with conservative bias. Specifically, the affected evolution was known in advance to have the possibility of a positive reactivity impact; however, operators did not take appropriate actions to reduce power sufficiently prior to unlocking the reactor recirculation motor generator set B scoop tube in order to prevent the reactor from exceeding 2419 MWt (H.14).
05000298/FIN-2018003-01	2018Q3	Cooper	Failure to Provide Complete and Accurate Information in a License Amendment Request	The inspectors identified that the licensee provided inaccurate information to the NRC in a license amendment request for an emergency action level scheme change. Specifically, the licensee provided information about the measurement ranges of a liquid effluent radiation monitor used in emergency action levels that was not accurate.
05000298/FIN-2018003-02	2018Q3	Cooper	Failure to Perform Process Applicability Determination	The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow Administrative Procedure 0.9, Tagout, Revision 88, for performing a monthly audit and Process Applicability Determination. Specifically, the inspectors noted that a clearance order on the safety-related residual heat removal service water booster pump room fan coil unit was hanging for greater than 90 days with no Process Applicability Determination performed, which resulted in the power switch for the fan coil unit being unintentionally tagged out of its normal configuration for almost 2 years
05000298/FIN-2018003-03	2018Q3	Cooper	Failure to Provide Adequate Lubrication for Drywell Fan Coil Units	The inspectors reviewed a self-revealed finding for the licensees failure to implement Work Order 5060136 during maintenance on the drywell fan coil units. Specifically, on October 26, 2016, during bearing replacement work on drywell fan coil, unit D, maintenance personnel failed to properly reinstall auto-lubricator injection connectors after removing the interferences per the work order instructions. This error resulted in the failure of drywell fan coil, unit D, due to inadequate bearing lubrication, and ultimately led to a downpower and reactor shutdown.
05000313/FIN-2009005-01	2009Q4	Arkansas Nuclear	Failure to Follow Procedure Result in an Inadequate Operability Determination	The inspectors identified a noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instruction, Procedures, and Drawing, regarding th licensees failure to follow the requirements of Procedure EN-OP-104 Operability Determination Process, Revision 4. Specifically, on October 15 2009, following removal of a seismic restraint from the Train B Containmen Spray Valve 2CV-5672-1 for preventive maintenance purposes, the inspectors identified that the shift manager approved and documented an operabilit determination using a cancelled engineering change document. The license entered this into their corrective action program as Conditio Report CR-ANO-2-2009-3794 The failure of the licensee to follow the requirements of Procedure EN-OP-104 Operability Determination Process, Revision 4, and approve an adequate basi for operability was a performance deficiency. The performance deficiency wa determined to be more than minor because the condition of not performin adequate operability determinations could become more significant if lef uncorrected and is therefore a finding. Using Manual Chapter 0609 Significance Determination Process, Phase 1 Worksheet, the finding wa determined to have very low safety significance because it did not result in th loss of safety function of any technical specification required equipment. It wa determined that the finding had a crosscutting aspect in the area of proble identification and resolution associated with the corrective actio program (P.1(c)), in that, the licensee failed to thoroughly evaluate problems suc that the resolutions addressed causes and extent of conditions, as necessary (Section 1R15.1)
05000313/FIN-2009005-02	2009Q4	Arkansas Nuclear	Failure to Correct a CAQ Associated with Removal of Rigid Seismic Restraints	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure of licensee personnel to correct a condition adverse to quality - removal of rigid seismic restraint for valve 2CV-5672-1, containment spray pump 2P-35B minimum recirculation valve, in the support of motor-operated valve actuator maintenance with an invalid engineering change to support the containment spray systems seismic operability licensing basis. This condition should have caused Unit 2 to enter Technical Specification 3.0.3 for 31 minutes on October 15, 2009. The inspectors had previously identified that the licensee was incorrectly applying ASME Code, Section III, Appendix F allowables to maintain operability for planned preventative maintenance. This issue was originally entered into the corrective action program as Condition Report CR-ANO-C-2009-1408. The licensee took action to cancel several engineering change documents, but did not review previously approved work orders to ensure that the removal of rigid seismic restraints would be prevented. This issue has been entered into the licensees corrective action program as Condition Reports CR-ANO-C-2009-2193, CR-ANO-2-2009-3356, and CR ANO-2-2009-3794. The failure to correct a condition adverse to quality associated with the removal of motor-operated valve actuator seismic restraints without a valid engineering evaluation was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the protection against external events attribute of the Mitigating Systems Cornerstone and directly affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Specifically, the engineering change used to justify seismic operability was invalid and should not have been used to support continued operability and had been cancelled for future use. Using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, Mitigating Systems Cornerstone, the finding was determined to have very low safety significance because it did not represent an actual loss of safety function and did not screen as potentially risk significant due to a seismic initiating event. The cause of this finding was determined to have a crosscutting aspect in the area of human performance associated with resources (H.2(c)) in that the licensee failed to have complete and accurate procedures to prevent engineering changes that had been cancelled from being used in work orders that had been previously planned and approved for work
05000313/FIN-2009005-03	2009Q4	Arkansas Nuclear	Failure to Follow Procedure Led to Loss of Shutdown Cooling	The inspectors documented a self-revealing, noncited violation of Technical Specification 6.4.1.a for the licensees failure to follow Operating Procedure OP-1015.008, Unit 2 SDC Control, Revision 30. Specifically, Unit 2 operators did not obtain permission from operations or plant management prior to performing maintenance on any protected train components. In this particular case, both trains of shutdown cooling, and their associated power supplies, were declared protected trains by operations. On September 20, 2009, operations personnel decided to perform an offsite power fast transfer test on the train A and train B vital buses. During the performance of the test on the train A vital bus, a fast transfer relay failed to actuate causing the slow transfer of the bus power supply. This caused the bus to de-energize and caused the inservice shutdown cooling pump to trip. The loss of shut down cooling resulted in a reactor coolant system temperature rise of 5 degrees. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-C-2009-2002. The inspectors determined that the failure of the operations staff to follow Operating Procedure OP-1015.008, Unit 2 SDC Control, Revision 30, was a performance deficiency. Specifically, the Unit 2 operations test team failed to obtain operations manager or plant manager permission prior to performing surveillance testing on the protected systems or equipment. The performance deficiency was determined to be more than minor because it was associated with the human error attribute and adversely affected the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown conditions and is therefore a finding. The failure to follow procedures resulted in the loss of the only train of shutdown cooling that was in service. This finding was evaluated for significance using NRC Manual Chapter 0609, Significance Determination Process, Appendix G, Checklist 3, for shutdown operations, and was determined to be of very low safety significance because the core heat removal guidelines associated with instrumentation, training and procedures, and equipment were met. Specifically, both trains of shutdown cooling remained operable with necessary support systems. This finding was determined to have a crosscutting aspect in the area of human performance, associated with decision making (H.1(a)) in that the licensee failed to make safety-significant or risksignificant decision using a systematic process, especially faced with uncertain or unexpected plant conditions, to ensure safety was maintained. In this case, although the licensee formally defined the authority and roles for decisions affecting nuclear safety, the shift manager and the shift operations manager oversight failed to implement their roles and authorities in deciding to conduct the offsite power transfer test on both protected trains of shutdown cooling (Section 4OA3.1)
05000313/FIN-2009005-04	2009Q4	Arkansas Nuclear	Failure to Take Timely and Effective Corrective Action for Fish Influx and Blockage of Circulating Water Intake Structure Leads to Unit 1 Reactor Down Power	The inspectors documented a self-revealing finding for the licensees failure to implement timely corrective action for industry operating experience associated with intake water blockage and for failure to implement effective corrective action stemming from a very similar event in 2006 where Unit 1 was forced to decrease reactor power due to an unexpected Shad run. The licensee entered this into their corrective action program as Condition Report CR-ANO-1-2009-1880. The licensees failure to take timely and effective corrective actions in response to industry and site specific operating experience was determined to be a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the external events attribute and directly affected the cornerstone objective to limit the likelihood of those events that upset plant stability and is therefore a finding. Specifically, the licensees failure to take timely and effective action led to the October 12, 2009, Unit 1 reactor down power due to a Shad (fish) influx into the intake structure. Using NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, Initiating Events Cornerstone, the finding was determined to have a very low safety significance 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 did not have a crosscutting aspect because the cause of the performance deficiency was not associated with any of the crosscutting aspects listed in Manual Chapter 0305, Operating Reactor Assessment Program, dated August 11, 200
05000313/FIN-2009005-05	2009Q4	Arkansas Nuclear	Failure to Report a Safety System Functional Failure	The inspectors identified a noncited violation of 10 CFR 50.73, Licensee Event Report System, associated with the licensees failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria as specified. Specifically, on September 22, 2009, the licensee completed their analysis of an issue associated with degradation of the latching mechanism of a station high energy line break door. The licensee determined that an unanalyzed condition may have existed for the period that the door was unlatched. The licensee reported the unanalyzed condition per 10 CFR 50.73. The licensee further determined that, due to this door latch issue, a main feedwater pipe critical crack high energy line break event would force the door open which would create a harsh environment in the adjoining emergency feedwater pump room, which would result in both trains of emergency feedwater being inoperable. The licensee determined that this was a safety system functional failure. Based on this, the inspectors determined that this condition was reportable per 10 CFR 50.73(a)(2)(v) since this resulted in a condition which affected both trains of a system described in the Safety Analysis Report that was needed to mitigate the consequences of an accident. Although the licensee submitted the licensee event report indicating that Unit 1 was in an unanalyzed condition, they failed to report the safety system functional failure aspect. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-C-2009-2590
05000313/FIN-2009005-06	2009Q4	Arkansas Nuclear	Failure to Notify the NRC with 8 Hours of a Nonemergency	The inspectors identified a noncited violation of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, for the licensees failure to notify the NRC Operations Center within 8 hours following discovery of an event meeting the reportability criteria as specified. Specifically, on September 22, 2009, the licensee initiated a 10 CFR 50.72 (b)(3)(xiii) 8-hour nonemergency report at 12:46 p.m. CST to the NRC Operations Center based on an event time of 5:11 a.m. Operations staff notified the resident inspectors of the 8-hour event notification to the NRC Operations Center later that afternoon. The inspectors questioned whether the timing of the NRC notification met the requirements of the applicable regulation. The inspectors determined that the initial loss of power to the emergency offsite facility occurred at approximately 10:40 p.m. on September 21, 2009, the emergency offsite facility diesel generator K8 started but failed to supply power to the facility, and this was reported to the control room at 11:45 p.m. on September 21, 2009. Normal power was restored at 4:20 a.m. Due to the time that the emergency offsite facility was degraded, this was considered a major loss of assessment, communications, and response capability, and the licensee initiated a 10 CFR 50.72 (b)(3)(xiii) 8-hour nonemergency report, but not within the 8-hour reporting period of the discovery. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-C-2008-2024. The failure to report an applicable nonemergency 8-hour event notification report within the required time frame was determined to be a performance deficiency. The finding was determined to be applicable to traditional enforcement because the NRCs ability to perform its regulatory function was potentially impacted by the licensees failure to make a required notification within the specified time frame. The finding was not suitable for evaluation using the significance determination process and was therefore evaluated in accordance with the NRCs Enforcement Policy. The finding was reviewed by NRC management and was determined to be of very low safety significance (Severity Level IV) consistent with the NRC Enforcement Policy. The cause of this finding was determined to have a crosscutting aspect in the area of human performance associated with resources (H.2(c)) in that the licensee failed to have complete and accurate procedures to properly evaluate problems when faced with unexpected condition
05000313/FIN-2009005-07	2009Q4	Arkansas Nuclear	Diesel Generator Ventilation Systems Susceptibility to the Depressurization Effects of a Tornado	The inspectors identified an unresolved item associated with the licensees lack of analysis to demonstrate the capability of the emergency diesel generator ventilation systems of either Units 1 or 2 to withstand the differential pressure effects of a tornado. During an NRC inspection in May 2005, NRC inspectors questioned whether the licensees ventilating and air conditioning system and other components in an emergency diesel generator room would be able to operate safely during and after a tornado event. Specifically, the NRC staff questioned whether wind pressures and differential pressures caused by a tornado passing directly over the emergency diesel generator building could adversely affect safety-related systems and components inside the emergency diesel generator building. The emergency diesel generator combustion air intake and exhaust system was constructed in such a way that it was exposed to ambient pressure from the outside and therefore would be exposed to the pressure differential that would be created by a tornado passing over the building. In response to the NRC questions, the licensee conducted an industry wide survey revealing approximately 25 other plants with a licensing basis similar to their own. As a result, on December 6, 2006, the NRC issued Regulatory Information Summary 2006-23, Post-Tornado Operability of Ventilating and Air-Conditioning Systems Housed in Emergency Diesel Generator Rooms. The purpose of Regulatory Information Summary 2006-23 was to notify licensees of the NRCs regulatory position regarding loading effects caused by natural phenomena to safety-related systems and components housed inside a structure partially exposed to the outside environment. In particular, ventilating and air conditioning systems housed in the emergency diesel generator room. The NRC expects licensees to consider natural hazards during the design of systems and components housed inside safety-related structures if these systems and components may be exposed to the outside environment and if their malfunction or loss may prevent or impact the operability of safety-related systems and components. Vented ventilating and air conditioning ducts, and other internal safety-related systems and components, may be subjected to the effects of rapid room depressurization and repressurization and other effects associated with a tornado event. In some cases, the loss of structural integrity of ventilating and air conditioning systems may pose a challenge to the safe operation of the facility. In such cases, licensees should take any necessary measures to ensure the operability of ventilating and air conditioning duct systems located in emergency diesel generator rooms. On December 6, 2006, Entergy initiated Condition Report LO-LAR-2006-0171 to have all sites perform a review of Regulatory Information Summary 2006-023. Specifically, each site was to determine if the sites design had adequately considered tornado wind and pressure drop effects on safety-related systems and components inside building structures open to the outside environment. On April 12, 2007, the licensee completed the review and concluded that the plants design criteria to comply with General Design Criteria GDC-2 requires that the structure remain fully functional before, during, and after a tornado event without exceeding code allowables. The original designers accomplished this by (1) designing the external structure (walls, ceilings, floors) to resist tornado winds, missiles, and depressurization; and (2) providing missile barriers near openings into the building where a missile trajectory could potentially directly strike a safety-related system/component. The temporary effects associated with a rapid external depressurization of systems and components were not considered in the original analyses. The safety-related components of Arkansas Nuclear Ones heating, ventilation, and air conditioning system are protected from tornados and other natural events by being located within the protection of reinforced concrete structures. Arkansas Nuclear Ones reinforced concrete structures that house safety-related equipment are designed to resist the effects of tornado conditions. For these structures, the ventilation system intakes and exhausts are designed to resist tornado generated missiles. However, neither the design basis nor licensing basis requires ventilation systems to be designed for the differential pressures associated with a tornado. Units 1 and 2 were licensed before the issuance of Regulatory Guide 1.76 and are not committed to it. Based on interactions with the Entergy fleet, the licensee subsequently determined that it would be prudent to further evaluate the tornado depressurization event and its potential impact on the diesel generator rooms ventilation systems. The licensee initiated Condition Report CR-ANO-C-2007-1308 to facilitate this. The licensee determined that this evaluation would not become part of the stations licensing basis but instead would provide reasonable assurance that the emergency diesel generator ventilation systems would not be damaged to the extent to render the emergency diesel generators inoperable. The licensee performed subsequent calculations, based upon sound engineering principles, to evaluate the emergency diesel generator ductwork and emergency diesel generator inlet dampers in both units for effects of a tornado depressurization event. This calculation used the differential pressure in Regulatory Guide 1.76, Revision 1. The licensee concluded that initially closed emergency diesel generator inlet dampers would be rendered inoperable by the event and resulting deformations would prevent subsequent automatic opening. The licensee further concluded that the Unit 1 emergency diesel generator inlet ductwork to the combustion air filters would collapse and cut off airflow to the engines. Calculations also indicated that the suction ductwork to the exhaust fans in both units would also collapse and cut off airflow to the exhaust fans. Based on these results, station design engineering could not ensure with a high level of confidence that the emergency diesel generator combustion air and ventilation systems would remain functional after a tornado event. The inspectors reviewed this position and calculations and determined that this was contrary to the regulatory position taken by the NRC in Regulatory Information Summary 2006-023. As such, the inspectors questioned the diesel generator rooms ventilation systems capabilities of withstanding the rapid depressurization effects that can occur coincident with a tornado. Specifically, the inspectors concluded that the evaluations that had been performed to date did not provide a reasonable expectation of operability for the diesel generator rooms ventilation systems in a tornado event. The inspectors presented their concerns to the licensee and the licensee determined that further review was necessary to determine the acceptability of the identified issues. The licensee initiated Condition Report CR-ANO-C-2009-2296 to address these concerns. Subsequent evaluations identified compensatory measure was necessary to maintain the ventilation systems operable during a tornado event. The inspectors determined that the potential vulnerability to Units 1 and 2 emergency diesel generator ventilation ductwork will be treated as an unresolved item, pending further inspector review of the licensees analysis. An unresolved item is an issue requiring further information to determine if it is acceptable, if it is a finding, or if it constitutes a violation of NRC requirements. In this case, additional NRC inspection will be required to assess the ability of the Unit 1 emergency diesel generator combustion air intake ductwork to cope with the rapid depressurization associated with a tornado event. Additional information was needed to determine whether a violation of regulatory requirements occurred. Pending further review of additional information provided by the licensee, this issue is being treated as an Unresolved Item 05000313/2009005-07; 05000368/2009005-07, Diesel Generator Ventilation Systems Susceptibility to the Depressurization Effects of a Tornado.
05000313/FIN-2009005-08	2009Q4	Arkansas Nuclear	Failure to Appropriately Scope Floor Drains in the Stations Maintenance Rule Monitoring Program	The inspectors identified an unresolved item associated with the licensees failure to appropriately monitor nonsafety-related structures, systems, and components whose failure could prevent safety-related components from fulfilling their safety-related function. The inspectors reviewed Station Calculation CALC-01-EQ-1001-01, MFW Critical Crack HELB Analysis, Revision 0. During this review the inspectors noted that (1) Door 19, a high energy line break door, was credited with isolating the emergency feedwater pumps from a harsh environment in the event of a main feedwater critical crack high energy line break event, (2) the door was assumed to remain closed as long as the differential pressure across the door remained less than 1 psid during the event, (3) the atmospheric pressure calculated in the room during the high energy line break event was 0.8 psi, and (4) a water accumulation in the room during the event was predicted to be 6 inches. The inspectors questioned the predicted value for water accumulation based on the assumed geometry of the crack in the feedwater piping. Further review of the calculation and discussions with station design engineers revealed that this water accumulation value was based on the modeling assumption that the drains were 4-inch openings which connected to other rooms through the drain system. The premise of this assumption was that the larger drain size would model the potential effects of steam transmission to other rooms through the floor drain system, therefore, determining if another room would have a potentially harsh environment created during this event. The inspectors questioned the validity of this modeling assumption. While it would be conservative for predicting potentially harsh environments in adjacent rooms, it appeared to be nonconservative for predicting the amount of water that would pool in the room and apply pressure to Door 19. With the drains modeled as 4-inch openings, the results appeared to under estimate the amount of water that would pool in this room. The inspectors determined that the amount of water that would pool in the room was important to determining whether Door 19 would be forced open during a main feedwater critical crack high energy line break event. Specifically, the pressure applied to Door 19 from the atmospheric pressure change due to the high energy line break event, in conjunction with the pressure that would be felt by the door due to water accumulation could potentially exceed 1 psid, and this would cause the door to open and expose the emergency feedwater pumps to a harsh environment. The inspectors informed the licensee of their concerns. The licensee initiated Condition Report CR-ANO-1-2009-1421 to address these concerns. The inspectors determined that the potential vulnerability to the Unit 1 emergency feedwater pumps during a main feedwater critical crack high energy line break event will be treated as an unresolved item pending further inspector review of the licensees analysis. An unresolved item is an issue requiring further information to determine if it is acceptable, if it is a finding, or if it constitutes a violation of NRC requirements. In this case, additional NRC inspection will be required to assess the ability of high energy line break Door 19 to remain shut during a main feedwater critical crack event. Additional information was needed to determine whether a violation of regulatory requirements occurred. Pending further review of additional information provided by the licensee, this issue is being treated as an Unresolved Item 05000313/2009008-08, Failure to Appropriately Scope Floor Drains in the Stations Maintenance Rule Monitoring Program
05000313/FIN-2009005-09	2009Q4	Arkansas Nuclear	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, measures to be established to assure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, the licensee approved nonconservative engineering calculations for two safety-related, motor-operator valve actuators where the adapter plate bolts were not included in the seismic and weak link engineering calculation, and eight safety-related, motor-operated valve actuators whose adapter plate installed bolts were smaller than those included in the applicable seismic and weak link engineering calculations. Specifically on September 11, 2009, during the performance of a postmaintenance test valve stroke for valve 2CV-4821-1, the 5/16-inch yoke to adapter plate fasteners failed during an over thrust event. Extent of condition review identified that valve 2CV-4820-2 seismic and weak link analysis did not include the adapter plate fasteners, and the seismic and weak link analysis for the following valves included adapter plate fasteners that were larger than what was actually installed in the field: Valves 2CV-3850-2, 2CV-0711-2, 2CV-0716-1, 2CV-8233-1, 2CV-5672-1, 2CV-5673-1, CV-2806, and CV-5611. This was licensee identified because the initial failure was discovered during postmaintenance testing of valve 2CV-4821-1, and the subsequent extent of condition review from the licensees corrective action process identified additional valves that were affected. Using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, Mitigating Systems Cornerstone, the finding was determined to have very low safety significance because it did not represent an actual loss of safety function and did not screen as potentially risk significant due to a seismic initiating event. The issue was entered into the licensees corrective action program as Condition Report CR-ANO-2-2009-2554
05000313/FIN-2011003-01	2011Q2	Arkansas Nuclear	Failure to Follow Compliance Work Order for Corrective Maintenance on Safety Related Equipment	The inspectors indentified a Green noncited violation of Unit 1 Technical Specification 5.4.1.a for a failure to perform proper placekeeping and to revise a compliance work order for the replacement of the auto-manual pushbutton, PB-2613, emergency feedwater steam admission valve. Specifically, the electrician had completed critical steps in a compliance work order without following the work order as written as required in Section 5.15 of station Procedure EN-MA-101, Fundamentals of Maintenance, Revision 9. The electricians also failed to stop and revise the work order when encountering an unexpected wiring configuration that was different than was specified in the work order. This was also not in accordance with the station Procedure EN-MA-101. The licensee took immediate corrective action to restore compliance. This issue has been entered into the corrective action program as Condition Reports CR-ANO-C-2011-0284, CR-ANO-C-1695, and CR-ANO-C-2011-1673. The inspectors determined that the failure to follow and revise the compliance work order as required by station Procedure EN-MA-101, Fundamentals of Maintenance, Revision 9, was a performance deficiency because it was within the licensees ability to foresee and correct and is also a violation of technical specifications. The performance deficiency was determined to be more than minor because it was associated with the human performance attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Specifically, not following compliance work orders while working on safety related equipment could adversely affect the system or component if required to respond to an event. Furthermore, working on safety related equipment without proper procedural guidance could also adversely affect the system or component. Using Manual Chapter 0609, Exhibit 1, Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because it did not result in the loss of operability or functionality; did not represent a loss of system safety function; did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; did not represent an actual loss of safety function of any risk significant system for greater than 24 hours; and did not screen as potentially risk significant due to external events. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with work practices in that the licensee failed to use human error prevention techniques, such as self and peer checks, and questioning attitude, to ensure that the compliance work order was being followed and revised as required.
05000313/FIN-2011003-02	2011Q2	Arkansas Nuclear	Failure to Provide Adequate Procedural Guidance Results in Control Element Assembly Shaft Extension Damage	The inspectors documented a self-revealing noncited violation of Unit 2 Technical Specification 6.4.1.a for an inadequate procedure that resulted in damaging a control element assembly shaft extension. Specifically, station Procedure OP-2505.007, Unit 2 Upper Guide Structure Installation, Revision 18, failed to give adequate guidance on aligning the center control element assembly shaft extension with the in-core instrumentation thimble support plate lifting frame funnel. This misalignment resulted in damaging the shaft extension, and required additional inspection and analysis for possible damage to the control element assembly and reactor fuel. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-2-2011-1284. The inspectors determined that the failure to provide adequate procedural guidance for installing the thimble support plate into the Unit 2 reactor vessel was a performance deficiency because it was within the licensees ability to foresee and correct and also violated technical specifications. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that the physical design barriers will protect the public from radionuclide releases caused by accidents or events, and is therefore a finding. Specifically, inadequate procedural guidance resulted in the damaging of a control element assembly shaft extension and could have resulted in fuel cladding damage. Using Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the finding was determined to be of very low safety significance (Green), because the finding did not prevent or degrade core heat removal, inventory control, electrical power, containment control, or core reactivity capabilities. The finding was determined not to have a cross-cutting aspect because the performance deficiency occurred in 2002 and is not indicative of current plant performance.
05000313/FIN-2011004-01	2011Q3	Arkansas Nuclear	Failure to Take Timely Corrective Actions for Invalid Local Leak Rate Test	The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions for the licensees failure to take corrective action for an invalid local leak rate test performed on the Unit 2 escape hatch, 2C-2. Specifically, the licensee failed to take appropriate and timely corrective action to develop an appropriate testing method for the inner and outer escape hatch door seals. The issue was entered into the licensees corrective action program as Condition Report CR-ANO-2-2011-3198. The inspectors determined that the licensees failure to develop an adequate testing method that did not use the strong backs to precondition the escape hatch door seals prior to the 2R20 fall 2009 outage was a performance deficiency. Specifically, the licensee failed to provide timely corrective actions to a condition adverse to quality that had been identified in a previous NRC identified noncited violation and was within the licensees ability to foresee and correct. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events and is therefore a finding. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance, Green, because the finding does not represent a degradation of the radiological barrier, or the smoke and toxic gas barrier functions provided for the control room, or does not represent an actual open pathway in the physical integrity of the reactor containment or a heat removal component. The finding was determined to have a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program in that the licensee did not thoroughly evaluate the problem in a manner to make certain that the resolution addressed the causes and the extent of condition to ensure a new test method, that did not use preconditioning, would be completed in a timely manner to resolve the problem.
05000313/FIN-2011004-02	2011Q3	Arkansas Nuclear	Failure to Provide Adequate Work Instruction Results in a Main Feedwater Recirculation Valve Failing Open	The inspectors documented a self-revealing finding for inadequate work instructions that resulted in the failure of a Unit 2 main feedwater pump A recirculation valve. Specifically, the licensee failed to provide adequate work instructions for reassembling and testing of the Unit 2 main feedwater recirculation valve, 2CV-0731. This valve failed full open during full power operations resulting in exceeding licensed reactor power. The licensee has implemented corrective action to communicate the importance of the positioning of the feedback arm support bracket and has changed the work orders to verify angle and tension of the feedback arm following reassembly of the positioner. The licensee entered this issue into the corrective action program as Condition Report ANO-CR-2-2011-1782. The failure to provide adequate work instruction for the assembly and testing of the Unit 2 main feedwater pump A recirculation valve positioner was determined to be a performance deficiency, because it was within the licensees ability to foresee and correct and was a failure to meet station requirements to provide adequate maintenance work instruction to maintenance personnel. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone and affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety function during power operations. Specifically, the failure of the recirculation valve caused reactor power to exceed licensed reactor power. Using MC 0609, Exhibit 1, Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance because it did not contribute to both the likelihood of a reactor trip and that mitigation equipment or functions would not be available. The inspectors determined that the finding did not have a crosscutting aspect because the performance deficiency is not indicative of current plant performance.
05000313/FIN-2011004-03	2011Q3	Arkansas Nuclear	Failure to Provide Adequate Work Instruction Results in Failed Bearing on Motor Generator Set	The inspectors documented a self-revealing finding for an inadequate work instruction for the 2-02 control element motor generator set flywheel bearing replacement that resulted in a failure of that bearing. Specifically, the licensee failed to provide instructions to obtain flywheel shaft dimensions to ensure adequate interference fit between the bearing and the shaft during corrective maintenance. This bearing subsequently failed on April 6, 2011. The licensee placed the issue into the corrective action program as Condition Report ANO-CR-2-2011-1817. The licensee replaced the failed bearing and shaft assembly and the system was returned to service. The failure to provide adequate maintenance work instruction to verify dimensional fit up between the flywheel shaft and bearing for the Unit 2, 2-02 motor generator set prior to reassembly was determined to be a performance deficiency. Specifically, it was within the licensees ability to foresee and correct and was a failure to meet station requirements to provide adequate maintenance work instruction to maintenance personnel. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Initiating Event Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, due to both control element motor generator sets being in the same room, the failure of the motor generator flywheel bearing caused the failure of that motor generator shaft and could have affected the only operating motor generator set and resulted in a reactor trip. Using Manual Chapter 0609, Exhibit 1, Phase 1 Initial Screening and Characterization of Finding, the finding was determined to be of very low safety significance because it did not contribute to both the likelihood of a reactor trip and that mitigation equipment or function would not be available. The inspectors determined that the finding did not have a crosscutting aspect because the performance deficiency is not indicative of current plant performance as the cause of not developing adequate work instructions stems from the late 1990s.
05000313/FIN-2011004-04	2011Q3	Arkansas Nuclear	Failure to Verify the Adequacy of Design of Unit 1 Emergency Core Cooling Systems to Address Potential Voiding	The inspector identified a noncited violation of 10 CFR 50 Appendix B Criterion III for failure to verify and check the adequacy of design by performance of design reviews, alternate calculations, or a suitable testing program. Specifically, the licensee identified potential void locations during engineering evaluations of the Unit 1 High Pressure Injection, Decay Heat Removal / Low Pressure Injection, Core Flood, and Building Spray systems and did not verify the adequacy of the design of those systems to ensure continued operability. The licensee performed ultrasonic testing on these locations at the time of the identification, but did not install vents, determine an acceptable void size, or establish a program to periodically vent or monitor these locations. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-1-2011-1406. The failure to verify and check the adequacy of design of the Unit 1 High Pressure Injection, Decay Heat Removal/Low Pressure Injection, Core Flood, and Building Spray systems is a performance deficiency. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspector performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green), because the finding was confirmed not to result in a loss of operability. This finding has a crosscutting aspect in the area of problem identification and resolution in the corrective action component because the licensee did not takes appropriate corrective actions to address safety issues in a timely manner.
05000313/FIN-2011004-05	2011Q3	Arkansas Nuclear	Failure to Identify the Decay Heat Removal Coolers as Potential Void Locations	The inspector identified a noncited violation of 10 CFR 50 Appendix B Criterion III for failure to verify and check the adequacy of design by performance of design reviews, alternate calculations, or a suitable testing program. Specifically, when performing a design review, the licensee did not identify the Decay Heat Removal coolers as locations where gas could accumulate in the Decay Heat Removal system and establish methods to verify the adequacy of design to ensure operability. The licensee performed immediate inspection of the heat exchangers by ultrasonic testing and did not find any voids. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-1-2011-01306. The failure to identify the Decay Heat Removal heat exchangers as locations where gas could accumulate is a performance deficiency. The performance deficiency is more than minor because if uncorrected, it could lead to a more significant safety concern. Specifically, the licensee could be unaware of an unanalyzed void in the Decay Heat Removal system because they failed to consider the potential for gas accumulation and had no program in place to detect it. The inspector performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green), because the finding was confirmed not to result in a loss of operability. This finding has a crosscutting aspect in the area of human performance in the decision making component because the licensee did not use conservative assumptions in decision making or conduct effectiveness reviews of safety-significant decisions to verify the validity of the underlying assumptions.
05000313/FIN-2011004-06	2011Q3	Arkansas Nuclear	Failure to Adequately Evaluate the Required Minimum Level in the Borated Water Storage Tank	The inspectors identified a noncited violation of 10 CFR 50 Appendix B Criterion III for failure to verify and check the adequacy of design by performance of design reviews, alternate calculations, or a suitable testing program. Specifically, the licensee did not adequately evaluate the required minimum level in the Borated Water Storage Tank to ensure adequate net positive suction head for Emergency Core Cooling System pumps and prevent gas entrainment due to vortex formation. The licensee performed an immediate operability evaluation and concluded that there was sufficient margin in the level to maintain operability. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-1-2011-1407 and CR-ANO-1-2011-1440. The failure to adequately evaluate the minimum level in the Borated Water Storage Tanks to ensure adequate net positive suction head for Emergency Core Cooling System pumps and prevent vortex formation is a performance deficiency. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affects the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not adequately ensure that the design of the Borated Water Storage Tank was sufficient to avoid loss of net positive suction head and prevent air entrainment in the Emergency Core Cooling System pumps. The inspector performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green), because the finding was confirmed not to result in a loss of operability. The finding was determined to have no cross-cutting aspect because the performance deficiency occurred in 2004, and is not indicative of current plant performance.
05000313/FIN-2011004-07	2011Q3	Arkansas Nuclear	Licensee-Identified Violation	Title 10 of the Code of Federal Regulation Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Contrary to the above, the licensee failed to implement adequate corrective actions to prevent a recurrence of an oil leak previously discovered in July of 2004, on the Unit 1 B Reactor Building Spray Pump inboard bearing housing. Specifically, on April 28, 2011, a system engineer discovered oil leakage from the same Unit 1 Reactor Building Spray pump inboard bearing housing. The previous corrective action, which called for tightening of the bolts, failed to prevent the leak from reoccurring. The performance deficiency of failing to perform proper corrective actions to prevent the inboard bearing housing from leaking was determined to be more than minor because it was associated with the with the SSC and Barrier Performance attribute of the Containment Barrier Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events and is therefore a finding. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance because the finding does not represent a degradation of the radiological barrier, or the smoke and toxic gas barrier functions provided for the control room, or does not represent an actual open pathway in the physical integrity of the reactor containment or a heat removal component. The issue was placed into the corrective action program as Condition Report CR-ANO-1-2011-588.
05000313/FIN-2011004-08	2011Q3	Arkansas Nuclear	Licensee-Identified Violation	Title 10 of the Code of Federal Regulation Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly are promptly identified and corrected. Contrary to the above, the licensee failed to promptly identify and correct a condition adverse to quality, associated with protecting and ensuring service water piping integrity from microbiological induced corrosion issues. Specifically, the licensee was aware of microbiological induced corrosion issues but failed to take timely corrective action to repair numerous issues with the service water chemical injection system during the previous year. The licensee also failed to assess the long term, cumulative risk of deciding not to follow their service water pipe replacement program. Both actions constitute a failure to take timely corrective actions resulting in an increase in microbiological induced corrosion induced service water leaks. The inspectors determined this finding to be of very low safety significance, Green, because it: 1) is not a design or qualification deficiency that resulted in loss of operability or functionality; 2) did not represent an actual loss of safety function of a system or train; 3) did not result in the actual loss of one or more trains of non-technical specification equipment designated as risk significant by 10 CFR 50.65 for greater than 24 hours; and 4) did not screen as potentially more risk significant due to a seismic, flooding, or severe weather initiating event. This issue was entered into the corrective action program as CR-ANO-C-2011-0336.
05000313/FIN-2011004-09	2011Q3	Arkansas Nuclear	Licensee-Identified Violation	Title 10 of the Code of Federal Regulation Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, the licensee failed to provide adequate work instructions to place the high pressure safety injection valve, 2CV-5035-1, in a condition to support effective motor-operated valve contact burnishing. Specifically, the licensee failed to properly burnish the closed contacts for motor-operated valve 2CV-5035-1 during a preventative maintenance activity. The valve subsequently failed to fully close during a surveillance test. The inspectors determined this finding to be of very low safety significance, Green, because it: 1) is not a design or qualification deficiency that resulted in loss of operability or functionality; 2) did not represent an actual loss of safety function of a system or train; 3) did not result in the actual loss of one or more trains of non-technical specification equipment designated as risk significant by 10 CFR 50.65 for greater than 24 hours; and 4) did not screen as potentially more risk significant due to a seismic, flooding, or severe weather initiating event. This issue was entered into the corrective action program as CR-ANO-2-2011-1329.
05000313/FIN-2013003-01	2013Q2	Arkansas Nuclear	Failure to Evaluate and Correct Excessive Containment Isolation Valve Leakage	The inspectors identified a non-cited violation of Unit 2 Technical Specification 6.5.16, Containment Leakage Rate Testing Program, for the failure to evaluate and take appropriate corrective actions to achieve acceptable performance for containment isolation valves that exceed the local leak rate administrative limit. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-2-2013-01370. The failure to perform a cause determination and take appropriate corrective actions for containment isolation valves that exceed the local leak rate administrative limit was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and is therefore a finding. Specifically, the failure to perform a cause determination and take appropriate corrective actions adversely affected the licensee\'s ability to ensure containment isolation valves function properly. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, the finding is determined to have very low safety significance because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components, and the finding did not involve an actual reduction in function of hydrogen igniters in the reactor containment. Since the cause of the performance deficiency occurred more than three years ago, the inspectors concluded that the finding was not representative of current licensee performance and no cross-cutting aspect was assigned.
05000313/FIN-2013003-02	2013Q2	Arkansas Nuclear	Failure to Correctly Install Control Room Emergency Chiller Supply Breaker	Inspectors documented a Green self-revealing non-cited violation of Technical Specification 6.4.1.a for the licensees failure to implement procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, the licensee failed to follow procedures for the replacement of the supply breaker for control room emergency chiller 2VE-1A. As a result, the breaker was installed incorrectly and the chiller was inoperable for over two months. Immediate corrective actions included proper installation of the breaker and procedural requirements for visual verification of breaker configuration. The licensee documented the issue in their corrective action program as CR-ANO-2-2013-00233. Inspectors concluded that the failure to follow Procedure 1403.179 for replacement of the train A control room emergency chiller breaker is a performance deficiency. The performance deficiency is more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. Specifically, the loose breaker connection adversely affected the availability and reliability of the control room emergency chiller A. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding required a detailed risk evaluation because it represented an actual loss of function of a single train for longer than its technical specification allowed outage time. The senior reactor analyst performed a detailed risk evaluation using the Arkansas Nuclear One Standardized Plant Analysis Risk models. The dominant risk sequences include a seismically-induced loss of offsite power with the failure of control room emergency chiller A. The analyst assumed that the operators and control room instrumentation could survive a peak control room temperature of 120 F, and that chiller A was susceptible to failure during a seismic event for the 83 days. None of the core damage sequences affected by this performance deficiency were important to the large, early release frequency. Therefore, based on the combined internal and seismic ICCDP of 2.9 x 10-7, this finding was of very low safety significance (Green). The finding was determined to have a cross-cutting aspect in the area of human performance, associated with work practices, in that the licensee failed to use work practices that support human performance. Specifically, licensee personnel were aware of the possibility of misaligning the wire grip style lug, but failed to use adequate self and peer checking to ensure the lug was correctly installed.
05000313/FIN-2017003-01	2017Q3	Arkansas Nuclear	Failure to Maintain Service Water Train Separation	The inspectors identified a non- cited violation of Technical Specification 5.4.1.a for the licensees failure to maintain train separation between safety -related service water trains when swapping the swing high pressure injection (HPI) pump between trains. Specifically, by following procedure OP 1104.002, Makeup and Purification System Operation, Revision 89, operators cross -tied service water trains, placing the system in an unanalyzed condition. This condition resulted in the train A electrical equipment room emergency chiller and train B reactor building emergency cooling coils being inoperable for a maximum of 25 minutes per occurrence. Additionally, it was determined that service water temperatures over the past 3 years did not result in an actual loss of function associated with these components if a design basis accident would have occurred. The immediate corrective actions were to assess past operability for not maintaining service water train separation and to revise Operating Procedure 1104.002 with adequate work instructions to maintain service water train separation. The licensee entered this deficiency into the corrective action program as Condition Report CR -ANO -1-2017- 02518. The licensees failure to maintain safety -related service water train separation when swapping the swing HPI pump between trains was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensees failure to maintain service water train separation placed the system in an unanalyzed condition and was subsequently determined to cause the train A electrical equipment room emergency chiller and train B reactor building emergency cooling coils to be inoperable for a maximum of 25 minutes per occurrence . Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding s At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety -significant , non -technical specification train. Specifically, inspectors confirmed that service water temperatures were never high enough to result in an actual loss of function for either limiting component. The finding had 3 a cross -cutting aspect in the area of human performance associated with conservative bias because the licensee failed to determine whether the proposed action was safe to proceed, rather than unsafe in order to stop. Specifically, in December 2015 when this approach was revise d to declare only the non- protected service water train inoperable, the licensee did not ensure that the transition lineup was analyzed to be within safety analyses before adopting the revised steps. (H.14)
05000382/FIN-2007007-04	2007Q2	Waterford	Inadequate Acceptance Criteria for Battery Cell-to-Cell and Terminal Connection Resistance Value	The Team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure that the 125 Vdc safety-related batteries would remain operable if all the intercell and terminal connections were at the resistance value of 150 micro-ohms as allowed by Technical Specification Surveillance Requirement 4.8.2.1.b.2 and 4.8.2.1.c.3. The finding was greater than minor because it affected the mitigating systems cornerstone objective (design control attribute) to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be a design deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment, (Section 1R21.b.4)
05000382/FIN-2008002-01	2008Q1	Waterford	Failure to Re-Evaluate Previously Identified Boric Acid Leaks (Section 1R04)	The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to implement corrective actions for a condition adverse to quality. Specifically, the licensee developed a corrective action to evaluate the condition of existing boric acid leaks. However, the effort failed to identify and evaluate multiple existing boric acid leaks on safety-related components, including some that had deteriorated since initial discovery. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2007-3951. This finding was more than minor because, if left uncorrected, it would have become a more significant safety concern. Specifically, some unchecked boric acid leaks may have worsened and corroded safety-related equipment. Using the Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Screening Worksheet, the finding had very low risk significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, \"Operability Determination Process for Operability and Functional Assessments.\" This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because engineers failed to implement proper error prevention techniques when identifying boric acid leaks for additional review (H.4(a)) (Section 1R04)
05000382/FIN-2008002-02	2008Q1	Waterford	Failure to Correct Fuel Oil Receipt and Transfer Procedure (Section 1R19)	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to implement adequate corrective actions for a deficient emergency diesel generator fuel tank filling procedure (a condition adverse to quality). The licensee had identified the deficiency following a previous event when fuel oil leaked out of multiple fuel oil injectors during a diesel run. Procedural steps were needed to adequately vent the fill line following pressurization during fuel oil tank filling. However, the licensee only corrected the procedure in one section and, when a different section was used, the problem reoccurred. The fuel oil leak led to the emergency diesel generator being declared inoperable. In addition, the fuel oil created a potential fire hazard. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-1345. The finding was more than minor because it was similar to nonminor example 4.f in Inspection Manual Chapter 0612, \"Examples of Minor Issues,\" in that emergency diesel generator operability was affected. Further, the oil created a fire hazard. Using the Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event (Section 1R19)
05000382/FIN-2008002-03	2008Q1	Waterford	Essential Chiller AB Return Header B Isolation Valve CHW 786B Misposition (Section 1R22)	The inspectors documented a self-revealing noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to correctly position a valve during a surveillance. The procedure required operators to position the essential Chiller AB return Header B isolation Valve CHW-786B closed but operators left the valve in the open position. This resulted in cross connecting the essential services chilled water Loops A and B, which led to an unplanned entry into Technical Specifications 3.7.12 and 3.0.3. The violation was revealed through a control room alarm. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-0778. The finding was more than minor because, if left uncorrected, would have become a more significant safety concern. Specifically, with both loops of the essential services chilled water system cross connected, the system was no longer single-failure proof. A leak in one of the essential chilled water loops would have caused both units to become inoperable. Using the Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it was a qualification deficiency confirmed not to result in loss-of-operability in accordance with NRC Manual Chapter Part 9900, Technical Guidance, \"Operability Determination Process for Operability and Functional Assessments.\" This finding had a crosscutting aspect in the Human Performance area, Work Practices component, because operators failed to implement self-checking techniques when performing procedure steps (H.4(a)) (Section 1R22)
05000382/FIN-2008002-04	2008Q1	Waterford	ACCW Pump Failure Due to Inaccurate Operator Aid (Section 4OA2)	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to correct a condition adverse to quality (inadequate instructions that led to low fuel oil and the failure of auxiliary component cooling water pump bearing). Specifically, the licensee\'s corrective action for a previous event called for an operator aid (oil level label). However, the operator aid contained incorrect and confusing information. Consequently, another auxiliary component cooling water pump failed. The licensee entered this deficiency into their corrective action program as Condition Report CR-WF3-2008-0350. The finding was more than minor because it was similar to nonminor violation example 4.f in Inspection Manual Chapter 0612, Appendix E, \"Examples of Minor Issues,\" in that the problem affected auxiliary component cooling water Pump B operability. Using the Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Screening Worksheet, the issue screened as having very low safety significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had crosscutting aspects associated with Human Performance area, resources program component, because the licensee failed to have correct labeling on components (H.2.(c)) (Section 4OA2)
05000382/FIN-2008003-01	2008Q2	Waterford	Failure to Follow Procedural Guidance When Entering a Radiological Controlled Area	The inspectors reviewed two examples of a self-revealing, noncited violation of Technical Specification 6.8.1 because workers failed to follow procedural requirements when preparing to enter the radiological controlled area. The first example, on April 28, 2008, involved a contract employee who informed the radiation protection shift control technician he would be working on the reactor coolant pump 1B platform where dose rates were below 350 millirems per hour. Subsequently, the contract employee entered another area, one which had not been surveyed and on which the worker had not been briefed, and received a dose rate alarm measuring 553 millirems per hour. The second example, on April 30, 2008, involved a rigger who was assigned to help rig and lift a reactor coolant pump seal from the pump to the top of the D-ring. However, the rigger did not report to radiation protection personnel to receive a briefing on the dose rates in the area of Reactor Coolant Pump 1A. Before being reassigned, the rigger was briefed for an area with dose rates less than 180 millirems per hour, but during his work on the reactor coolant pump, the worker entered an area with dose rates as high as 628 millirems per hour and received a dose rate alarm. Radiation protection personnel counseled the workers and documented the occurrences in the corrective action program. The occurrence involved the program attributes of exposure control and affected the cornerstone objective, in that the failure of the workers to follow procedural guidance and inform radiation protection personnel of the workers intended activities work area resulted in the workers being unknowledgeable of 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 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 workers failed to use human error prevention techniques such as self and peer checking (H.4.a) (Section 2OS1.1)
05000382/FIN-2008003-02	2008Q2	Waterford	Failure to Post a Radiation Area	The inspectors identified a noncited violation of 10 CFR 20.1902 because the licensee failed to post a radiation area conspicuously. On May 14, 2008, the inspectors toured the hot machine shop and noted a box with high radiation area signs attached. Dose rates around the box ranged from 55 to 90 millirems per hour at 30 centimeters. The inspectors noted there was no posting to identify the radiation area. The nearest radiation area posting was on the entry door of the decontamination room, outside the hot machine shop. As a result of the inspectors finding, the licensee erected a rope barricade around the radiation area and posted it conspicuously. The finding was more than minor because it was associated with one of the cornerstone attributes and the finding affected the Occupational Radiation Safety cornerstone objective, in that, uninformed workers could unknowingly accrue additional radiation dose. Because the finding involved the potential for unplanned, unintended dose resulting from conditions that were contrary to NRC regulations, the finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The inspectors determined that the finding had very low safety significance because it was not: (1) an as low as reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose. The finding also had a crosscutting aspect in the area of human performance, resource component, because the licensee did not have complete procedures (H.2.c) (Section 2OS1.2)
05000382/FIN-2008003-03	2008Q2	Waterford	Licensee-Identified Violation	The following finding of very low significance was identified by the licensee and is a violation of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as a NCV. Technical Specification 6.12.2 requires entrance to areas with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrems, at 30 centimeters from the radiation source or any surface the radiation penetrates, be controlled by requiring issuance of a radiation work permit. On May 16, 2008, the licensee identified two workers who were placing major portions of their whole body in an area with dose rates greater than 1000 mrems, but were not working in accordance with a radiation work permit that allowed entry into the area. While pumping out the bowl of steam generator No. 1, two decontamination technicians reached across the locked high radiation area boundary at the plane of the manway. Dose rates at the plane of the manway were 1500 mrems per hour. The workers used Radiation Work Permit 2008-0509, which stated there was to be no entry into locked high radiation areas. The licensee documented the finding in Condition Report WF3-2008-02367. The inspectors determined that the finding had no more than very low safety significance because it was not: (1) an as low as reasonably achievable finding, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an inability to assess dose
05000382/FIN-2008004-01	2008Q3	Waterford	Intersystem loss of coolant event	Introduction. The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III (Design Control) for an inadequate \"pressure locking\" design calculation involving shutdown cooling Valves SI-405A and SI-405B. Plant engineers also used the calculation to support valve operability following a valve malfunction, which appeared to be caused by pressure locking. Entergy engineers had derived valve bonnet leakage rates (for pressure locking conditions) from local leak rate testing results. However, a national laboratory had already proven the Entergy theory invalid and plant engineers had taken no steps to validate the theory themselves. Finally, in response to an NRC generic letter concerning pressure locking and thermal binding of valves, the licensee\'s conclusions were based on incorrect facts and improper assumptions. Background: Valves SI-405A/B are opened to place shutdown cooling and lowtemperature/ over-pressure protection (Valves SI-406A/B) in service. Valves SI-405A/B are in series between Valves SI-401A/B and SI-407A/B, which are also opened for shutdown cooling operations. In the past refueling outage (RF-15, Spring 2008), plant craftsmen replaced Valves SI 405A/B hydraulic actuators with air-operated actuators. Plant engineers specified the modification to resolve performance problems with the hydraulic units. However, the behavior of the air-operated actuators was fundamentally different than that of the hydraulic model. Specifically, the hydraulic actuators provided up to the maximum amount of stem thrust starting at the beginning of the opening valve stroke, when thrust demands were greatest. In contrast, the air-operated actuators gradually increased valve stem thrust as air on the top of the actuator piston was vented off. If a load existed on the stem (typically the case, i.e. packing and unwedging), the valve would not initially move when commanded to open. It would remain closed until the actuator developed sufficient thrust to move the valve from its seat. Once off the seat, the valve disc would pop to a midposition and then continue to the open position at the normal speed. On May 14, 2008, during post-installation testing, a plant engineer observed that Valve SI-405A did not initially move on the first two attempts. As noted in CR WF3-2008-02326, approximately 2 minutes following the second attempt, the valve opened with a: . . . loud abrupt sound and shaking of the surrounding area. The valve coupling on the stem immediately moved approximately one half of the travel distance following the loud noise. Just prior to the valve operation, operators had pressurized the shutdown cooling piping system to approximately 1700 psig. The valve had pressure locked following the test. Pressure locking occurs when elevated residual pressure is trapped in a valve\'s bonnet and prevents the valve from opening. The NRC has issued several industry wide generic communications concerning pressure locking, most notably Generic Letter 95-07, \"Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves,\" dated August 22, 1995. Entergy was required to respond to the generic letter, documenting their plans to address these generic concerns. Entergy engineers had concluded that pressure locking of these valves would not occur. Specific deficiencies with the Entergy generic letter response are provided later in this report. Intersystem Loss-of-Coolant Event: On September 1, plant operators shut down Waterford-3 in preparation for Hurricane Gustav. During initiation of Train B shutdown cooling operations, control room operators identified and responded to an intersystem loss-of-coolant event that lasted approximately 4 minutes. About 800 gallons of reactor coolant was lost through low-temperature over-pressure protection Relief Valve SI-406B. The relief valve unexpectedly opened following an apparent malfunction of Valve SI-405B. While placing Train B shutdown cooling in service, Valve SI-405B had failed to initially move when the control room operator repositioned the valve\'s control room switch to open. The indication showed full closed for about 12 minutes. The valve indication then showed midposition, which was followed by full open indication. The air-operated valve normally had a 5 to 6 minute stroke time. Licensee personnel determined that the valve was stuck in its seat until the valve actuator developed sufficient thrust to move the valve disc. Then, the valve popped to almost the full open position. Entergy engineers determined that the sudden motion of Valve SI-405B created a pressure transient in the residual heat removal piping. Since system pressure was about 350 psig at the time, system pressure combined with the pressure transient that resulted from the opening of Valve SI-406B exceeded the setpoint of Relief Valve SI-406B, which was approximately 430 psia. Once lifted, relief valves will not normally seat at their set pressure but will close at a pressure somewhat below the setpoint. Since system pressure was already relatively high, the valve did not immediately reseat. Operator action was necessary to stop the leakage by closing Valve SI-401B. This was the preferred isolation valve because of the slower closing time of Valve SI-405B. At the end of the inspection period, the inspectors were still evaluating information related to the intersystem loss-of-coolant event. The inspectors provided questions to licensee personnel concerning the valve actuator design change package. This is an URI pending further evaluation by the NRC inspectors: URI 05000382/2008004-01, intersystem loss-of-coolant event
05000382/FIN-2008004-02	2008Q3	Waterford	Inadequate pressure locking calculation	The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion III (Design Control) for an inadequate \"pressure locking\" design calculation for shutdown cooling Valves SI-405A and SI-405B. Plant engineers also used the calculation to support valve operability following a valve malfunction, which appeared to be caused by pressure locking. Entergy engineers had derived valve bonnet leakage rates (for pressure locking conditions) from local leak rate testing results. However, a national laboratory had already proven the Entergy theory invalid and plant engineers had taken no steps to validate the theory themselves. Finally, in response to an NRC generic letter concerning pressure locking and thermal binding of valves, the licensee engineers\' conclusions were based on incorrect facts and improper assumptions. Licensee personnel entered the noncited violation into the corrective action program as Condition Report CR-WF3-2008-4292. The failures to perform: (1) an adequate engineering calculation and (2) a valid operability determination were performance deficiencies. This finding was more than minor because it was similar to nonminor 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 calculation was completed in 2008 and was indicative of current performance
05000382/FIN-2008004-03	2008Q3	Waterford	Operability of safety injection Valves SI-405A/B	The inspectors opened an unresolved item to address potential concerns associated with the licensee\'s second operability determination for safety injection Valves SI-405A/B. Discussion. In response to the inspectors\' continued operability concerns, Entergy personnel performed a second operability determination, as documented in CR-WF3-2008-04294. The licensee contracted with MPR Associates, an engineering firm, to assess the potential for pressure locking of Valves SI-405A/B. MPR produced a calculation that concluded that the air-operated actuators could overcome the thrust requirements for a bounding pressure locking event. The MPR calculation assumed that bonnet pressure was 2250 psig (normal plant operating pressure) and that no pressure existed in the upstream and downstream piping sections. However, the inspectors observed the following apparent inconsistencies with the calculation: Waterford-3 engineers used, and relied upon, the calculation in their operability assessment but neither Waterford-3 engineers nor MPR had performed a design verification of the calculation. The inspectors noted that Procedure EN-DC-126, \"Engineering Calculation Process,\" Revision 1, stated, in part: Calculations prepared in accordance with this procedure to support a reasonable expectation of operability do not need to be design verified. This procedure stipulation appeared inconsistent with the Waterford-3 \"Quality Assurance Program Manual,\" Revision 18, dated April 15, 2008, Section B.3.d (Design Verification) which stated, in part: Independent design verification is to be completed before design outputs are used by other organizations for design work. . . In all cases (emphasis added), the design verification is to be completed before relying on the item to perform its function. The conditions assumed in the calculation were substantially more demanding than could have been experienced during known pressure locking event on May 14, 2008, (Valve SI-405A) but the calculation concluded that the actuator had sufficient thrust to overcome even more demanding pressure locking conditions. Specifically, the calculation assumed a bounding 2250 psig in the bonnet, which exceeded the worst case valve bonnet pressure of 1700 psig during the May 14, 2008, pressure locking event. Therefore, it appeared that MPR had concluded that the May 14, 2008, event could not have occurred. Entergy and MPR engineers stated that they had not evaluated the May 14, 2008, event. Entergy and MPR engineers were working to resolve the apparent inconsistency at the close of the inspection period. Validation analysis performed to support the MPR pressure locking calculation method, \"Pressure Locking and Thermal binding Evaluation of EPRI MOVs (Motor-Operated Valves),\" showed that the method appeared conservative if a high valve friction coefficient (0.61) was used, but this high friction coefficient was not used in the Waterford-3 SI-405A/B analysis. The validation calculation also provided case studies involving lower valve friction coefficients (called the \"bestfit\" coefficients of friction), but the method under predicted required actuator thrust for 38 percent of the examples when these values were used. In addition, when valve bonnet pressures were greater than 500 psig, the method under predicted required valve thrust in 75 percent of the cases and it under predicted required valve thrust in all three instances where bonnet pressure was greater than 1000 psig. NOTE: EPRI did not assess pressure locking as part of their MOV validation testing program. However, six of the valves demonstrated pressure locking symptoms (retaining bonnet pressure). In addition, two of these valves had a repeated instance of bonnet pressure retention. MPR validated their calculation method against this data. Also concerning the validation analysis, the \"best-fit\" friction coefficients were generally in the midrange of friction coefficients observed for any given valve tested as part of the EPRI program. For example, for Valve MOV 15, MPR had assumed a valve friction coefficient of 0.2. The actual EPRI test results stated that the value was approximately 0.18 to 0.2 but the test data ranged from 0.17 to 0.3. A second example involved Valve MOV 24. The friction coefficient used in the validation analysis was 0.53. The EPRI report stated that coefficient of friction stabilized at about 0.44 but data ranged between 0.32 and about 0.6. Similar observations were made for all six test valves and all eight test cases. It was not clear how MPR had selected the \"best-fit\" friction coefficients from the EPRI test data. The MPR calculation, \"Pressure Locking Evaluation of SI-405A(B),\" dated September 11, 2008, provided unexpected results. MPR performed case studies for the Waterford-3 valves using different valve friction coefficients (0.2, 0.35, and 0.5). All other variables were held constant. The inspectors noted that the required thrust for the 0.5 valve friction coefficient (41,787 pounds of thrust) was less than that calculated for the 0.35 friction coefficient (42,218 pounds of thrust). Normally, as with a friction coefficient (with all other variables held constant) the required thrust would also be higher. MPR had not answered all questions related to this inconsistency at the close of the inspection. For all of the above potential concerns, the licensee was working to obtain additional information to support their conclusions. This is considered an URI pending NRC review of the licensee provided information: URI 05000382/2008004-03, operability of safety injection Valves SI-405A/B
05000382/FIN-2008004-04	2008Q3	Waterford	Failure to follow integrated EDG test procedure	The inspectors identified a noncited violation of Technical Specification 6.8.1.c (Procedures) for the failure to open the Train A low pressure safety injection pump suction valve prior to pump operation during a surveillance. The butterfly valve was installed 90 degrees out of position and was closed when operators believed it was open. After starting the pump, operators observed loud noises coming from the unit and secured it 8 minutes later. Pump operation without adequate net positive suction head could cause damage. The valve\'s postmaintenance test was scheduled after the noted surveillance test, and the surveillance was not intended to check the valve\'s function. The safety injection train was considered inoperable but available at the time. Licensee personnel entered the noncited violation into the corrective action program as Condition Reports CR-WF3-2008-2280 and CR-WF3-2008-3045. This finding was more than minor because it affected both the configuration control and the equipment performance attributes of the Mitigating Systems Cornerstone objective to ensure reliability of the low pressure safety injection system. In addition, this condition, if left uncorrected, would also become a more significant safety concern. Equipment could be damaged without adequate postmaintenance checks prior to operation. Using the NRC Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Screening Worksheet, the finding was of very low risk significance because it did not: (1) represent a loss of safety function; (2) represent an actual loss of a single train of equipment for more than its Technical Specification allowed outage time; or (3) 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 human performance, associated with the decision-making component, in that, the plant personnel used nonconservative assumptions and chose to use the pump suction valve for system operation prior to verifying that the valve was properly assembled (H.1(b)) (Section 1R19)

Semantic search

Query log

Navigation menu

Search