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	Quarter	Site	Title	Description
05000445/FIN-2018002-02	2018Q2	Comanche Peak	Unacceptable Preconditioning of Main Steam Isolation Valves	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Controls, for the licensees unacceptable preconditioning of the Unit 1 main steam isolations valves (MSIV) prior to performing as-found in-service stroke time testing. Specifically, the licensee raised accumulator pressure prior to stroke time testing and this potentially masked an issue with MSIV 1-01. The licensee entered this issue into the corrective action program as Condition Report CR-2018-002405.
05000445/FIN-2018002-01	2018Q2	Comanche Peak	Failure to Identify and Correct a Condition Adverse to Quality	The inspectors identified a Green,non-cited violation of 10CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to identify and correct a condition adverse to quality associated with unacceptable main steam isolation valve (MSIV) stroke times. Specifically, during stroke time testing of MSIV 2-02 the valves stroke time was outside of the acceptance limit and the licensee failed to determine why the stroke time was out of specification and correct the issue prior to declaring the valve operable and placing it in service. The licensee entered this issue into the corrective action program as Condition Report CR-2018-002189.
05000445/FIN-2018002-03	2018Q2	Comanche Peak	Failure to Incorporate Design Information Into System Test Procedures	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Controls, for the licensees failure to ensure that the stations in-service testing program for main steam isolation valves (MSIVs) incorporated the requirements and acceptance limits contained in applicable design documents. Specifically, the licensees in-service procedures did not direct testing of the valves be performed at the minimum required pressure and this resulted in the licensees failure to identify two degraded MSIVs during in-service testing. The licensee entered this issue into the corrective action program as Condition Report CR-2018-003229.
05000483/FIN-2017003-01	2017Q3	Callaway	Spurious Containment Spray Pump Start	The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to implement Preventative Maintenance Basis document IC-LSELS, Load Shed and Emergency Load Sequencer (LSELS), Revision 0. Specifically, the licensee failed to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23, a Consolidated Controls 6N232 relay driver card, within the scheduled periodicity. On June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours, of which all 44 hours w ere unplanned. As immediate corrective actions, the licensee replaced the circuit card under Job 17002747, completed post -maintenance testing, and restored the system to operable status on June 30, 2017. The licensee entered this issue into the corrective action program under Condition Report 20170 3433. The failure to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 within the scheduled periodicity was a performance deficiency. This performance deficiency was more than minor , and therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, o n June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours . Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At - Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; ( 3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non- technical specification trains of equipment designated as high safety -significant in accordance with the licensees maintenance rule program for greater than 24 hours. Specifically, the total duration of inoperability was 44 hours which is less 3 than the technical specification allowed completion time of 72 hours for this system. The finding had a cross -cutting aspect in the area of problem identification and resolution associated with resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee did not replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 prior to failure although this issue was documented in corrective actions ranging from April 2008 to January 2017 (P.3).
05000445/FIN-2017002-01	2017Q2	Comanche Peak	Failure to Control Transient Combustible Material in Accordance with a Fire Protection Procedure	Green. The inspectors identified a non- cited violation of Operating Licenses NPF -87 and NP F-89, License Condition 2.G, Fire Protection Program, for the failure to control transient combustibles in accordance with the station s fire protection report. Specifically, Fire Protection Report, Revision 29, Section 5.3.8, Fire Area EO Control Room, includes Deviation 3c -1, Control Room Missile Door, which requires, in part, that since the control room missile door in the west wall is not a 3 -hour rated fire door, the area of the turbine deck within 100 feet of the door is to be void of combustibles. Contrary to this, the licensee allowed storage of combustible materials in this area without required compensatory measures. This issue does not represent an immediate safety concern because the licensee removed the combustible materials upon identification. The licensee entered this issue into corrective action program as Condition Report CR -2017 -5564. The failure to control transient combustible material in accordance with the approved fire protection report is a performance deficiency. The performance deficiency was more than minor and therefore a finding because it was associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the introduction of transient combustible materials decreased the external event mitigation for fire prevention. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined that the finding pertained to a failure to adequately implement fire prevention and administrative controls for transient combustible materials. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, September 20, 2013. The inspectors evaluated the finding through Appendix F, Attachment 1, Fir e Protection Significance Determination Process Worksheet, September 20, 2013, and determined that the finding was of very low safety consequence (Green) because the Fire Prevention and Administrative Controls finding would not prevent the reactor from re aching and maintaining a safe shutdown condition. The finding has a problem identification and resolution cross -cutting aspect associated with resolution, in that, the licensee failed to take effective corrective actions to address issues in a timely manner. 3 Specifically, the licensee had previously identified this issue in Condition Report CR- 2014010224 but had failed to take corrective actions to address it (P.3)
05000445/FIN-2017002-03	2017Q2	Comanche Peak	Relays not Environmentally Qualified	Green. The inspectors identified a non- cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to assure that design changes were subject to design control measures commensurate with those applied to the original design. Specifically, the licensee changed internal components for safety -related, steam generator atmospheric relief valve booster relays but failed to verify that these new components could withstand the environment created during a high energy line break. This issue does not represent an immediate safety concern because the licensee performed an operability determination which established a reasonable expectation for operability, and implemented corrective actions to replace the relays with qualified relays. The licensee 4 entered this issue into the corrective action program for resolution as Condition Report CR- 2017- 006236. The failure to ensure that changes to the facility were subject to design control measures commensurate with those applied to the original design was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out -of-service for longer than their technical specification 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 for greater than 24 hours in accordance with the licensees maintenance rule program. The inspectors did not assign a cross -cutting aspect because the performance deficiency was not reflective of present performance
05000445/FIN-2017002-04	2017Q2	Comanche Peak	Failure to Adequately Assess Risk and Implement Risk Management Actions for Proposed Maintenane	Green. The inspectors identified a non- cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to adequately assess risk and implement required risk management actions for a planned maintenance activity. Specifically, the licensee failed to evaluate the risk and implement required risk management actions associated with disabling a hazard barrier and breeching the control room envelope when blocking open door E -40A. This issue did not represent an immediate safety concern because, at the time of identification, the licensee stopped the activity and secured the door. The licensee entered this issue into the corrective action program for resolution as Condition Report CR- 2017- 006019. The failure to adequately assess the risk and implement required risk management actions for proposed maintenance activities was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the configuration control attribute o f the Barrier Integrity Cornerstone and affected the associated objective to ensure physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, Flowchart 2, Assessment of Risk Management Actions, the inspectors determined the need to calculate the risk deficit to determine the significance of this issue. A senior reactor analyst determined the finding to have very low safety significance (Green) based on combining the effects of the degradation of the radiological barrier and tornado missile barrier functions. The analyst performed a qualitative review of the screening criteria in Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At -Power, for the degradation of the radiological barrier function for the control room and considered the short exposure time (2.9E -5 years) and the Comanche Peak specific high winds frequency (3.0E -4/year) for the tornado missile barrier function of the control room to determine that the incremental core damage probability deficit and the incremental large early release probability deficit were less than 1E -6 and 1E -7, respectively. The finding has a human performance cross -cutting aspect associated with procedure adherence, in that operations personnel failed to follow procedures when allowing door E -40A to be opened
05000445/FIN-2017002-05	2017Q2	Comanche Peak	Failure to Translate Design Requirements Into the As Built Facility	Green. The inspectors identified a non- cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to assure that applicable regulatory requirements and the design bases, as defined in 10 CFR 50.2 and as specified in the license application, for those structure, systems and components to which this appendix applies, were correctly translated into specifications, drawings, procedures, and instructions. Specifically, from initial construction through March 2017, the licensee failed to fully incorporate applicable moderate energy line break design requirements for fire protection piping located in the vicinity of the station service water pumps, the latter which are needed to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition following a moderate energy line break. This issue does not represent an immediate safety concern because when the lines were identified the licensee took prompt action to isolate and depressurize them, and the licensee has implemented plant modifications. The licensee entered this issue into the corrective action program as Condition Report CR -2016- 008147. The failure to incorporate applicable design requirements into specifications for moderate energy line break protection was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, from initial construction through March 2017, the licensee failed to fully incorporate applicable design requirements for components needed to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition following a moderate energy line break. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated July 1, 2012, and Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings At -Power , Exhibit 2, Mitigating Systems Screening Questions, dated 5 October 7, 2016, the inspectors determined the finding required a detailed risk evaluation because the finding involved a deficiency affecting the design and qualification of a mitigating structure, system, or component, and resulted in a loss of operability, and represented an actual loss of function of at least a single train for longer than its allowed outage time. A senior reactor analysts from Region IV performed a detailed risk evaluation and determined that the bounding increase in core damage frequency for this issue was 5.1E -8/year for Unit 1 and 2.9E -10/year for Unit 2, and was therefore of very low safety significance (Green ). The inspectors did not assign a cross -cutting aspect because the performance deficiency was not reflective of present performance
05000445/FIN-2017002-06	2017Q2	Comanche Peak	Unanalyzed Condition Involving Potential Moderate Energy Line Break	Inspection Scope On September 13, 2016, based on initial observations by NRC inspectors, the licensee determined that pressurized fire protection piping in the service water intake structure was not properly shielded for moderate energy line break protection of service water components which resulted in inoperability of one train of service water for both Unit 1 and Unit 2. During extent of condition walk downs conducted on October 6, 2016, October 10, 2016, November 17, 2016, December 5, 2016, and December 22, 2016, additional piping in the Unit 1 and Unit 2 safeguards and auxiliary buildings was found to not be shielded correctly as well, resulting in inoperability of one train of various safety related equipment for both units. The licensee determined the most likely cause of this event was that the methodology used to conduct the initial moderate energy line break walk downs was flawed and allowed some threats to be missed. The licensees corrective actions include shielding the affected piping, performing a 100 percent walk down of rooms containing moderate energy line break piping identified for shielding, and revising the systems interaction program maintenance procedure. These activities constituted completion of one event follow -up sample, as defined in Inspection Procedure 71153. b. Findings Introduction. The inspectors identified a non- cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to assure that applicable regulatory requirements and the design bases, as defined in 10 CFR 50.2 and as specified in the license application, for those structure, systems and components to which this appendix applies, were correctly translated into specifications, drawings, procedures, and instructions. Specifically, from initial construction through March 2017, the licensee failed to fully incorporate applicable design requirements for components needed to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition following a moderate energy line break. Description. On September 13, 2016, inspectors performed walkdowns in the service water intake structure and identified a vertical run of unshielded, pressurized fire protection piping that appeared to pose a moderate energy line break threat to the service water pumps. Inspectors determined that in the event of a moderate energy line break crack along any portion of the unshielded piping, the resultant spray had the potential to impact the function of any one of the four service water pumps. However, only one train would have been affected during the event due to the physical configuration/separation relative to the source line and target pumps and/or associated motor control centers that support pump operation. Inspectors informed the licensee of their concern. Engineering personnel performed a subsequent walkdown of the intake structure and determined that the identified piping was not correctly shielded and operability of the service water pumps was in question. The licensee took immediate action to isolate and depressurize the fire protection line in question which addressed the operability concern. The licensee entered this issue into the station corrective action program as Condition Report CR -2016 -008147 for resolution. Part of the licensees actions was to perform extent of condition walkdowns for unshielded moderate energy piping in the safeguards building for Unit 1 and 2. During the extent of condition walk downs conducted on October 6, 2016, October 10, 2016, November 17, 2016, December 5, 2016, and December 22, 2016, additional piping in the Unit 1 and Unit 2 safeguards and auxiliary buildings was found to not be appropriately shielded against a moderate energy line break, resulting in the inoperability of various safety related equipment for both units. Unit 2 Train B 480 VAC motor control center 2EB2- 1 (Unit 2 Train B emergency core cooling, battery charger, containment spray, and containment isolation valve equipment) Unit 1 Train B 480V MCC 1EB4- 2, and Unit 1 Train B Distribution Panel 1ED2- 2 (Unit 1 Train B safety -related pumps, panels, sequencer, and transformers) Unit 1 Train B 480V MCC 1 EB4- 1 (Unit 1 Train B safety -related pumps, valves, fans, battery chargers, and transformers) Unit 2 Train B 480V MCC 2E134- 1 (Unit 2 Train B safety -related pumps, valves, fans, battery chargers, and transformer) Unit 1, Train B 480V MCC 1E84- 1 (Unit 1 Train B safety -related pumps, valves, fans, battery chargers, and transformers) In each of these instances the licensee took prompt action to isolate and depressurize the identified moderate energy piping pending modification. The licensee subsequently determined that the most probable cause of the issue was the use of a flawed methodology during the initial moderate energy piping walkdowns conducted in 1989. The licensee reported this issue to NRC in Event Report 52239, and Licensee Event Report 16 -002- 00. Analyses. The failure to incorporate applicable design requirements into specifications for moderate energy line break protection was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, from initial construction through March 2017, the licensee failed to fully incorporate applicable design requirements for components needed to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition following a moderate energy line break. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated July 1, 2012, and Inspection Manual Chapter 0609, Appendix A , Significance Determination Process for Findings At -Power , Exhibit 2, Mitigating Systems Screening Questions, dated October 7, 2016, the inspectors determined the finding required a detailed risk evaluation because the finding involved a deficiency affecting the design and qualification of a mitigating structure, system, or component, and resulted in a loss of operability, and represented an actual loss of function of at least a single train for longer than its allowed out age time. A senior reactor analysts from Region IV performed a detailed risk evaluation and determined that the bounding increase in core damage frequency for this issue was 5.1E -8/year for Unit 1 and 2.9E -10/year for Unit 2 , and was therefore of very low safety significance (Green). Additional information is included in the detailed risk evaluation in Attachment 3 of this report. The inspectors did not assign a cross -cutting aspect because the performance deficiency was not reflective of present performance. Enforcement. Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that, measures shall be established to assure that applicable regulatory requirements and the design bases, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies, are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, measures established by the licensee did not assure that applicable regulatory requirements and the design bases, as defined in 10 CFR 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies, were correctly translated into specifications, drawings, procedures, and instructions. Specifically, from initial construction through March 2017, the licensee failed to fully incorporate applicable design requirements for components needed to ensure the capability to shut down the reactor and maintain it in a safe shutdown condition following a moderate energy line break. This issue does not represent an immediate safety concern because when the lines were identified the licensee took prompt action to isolate and depressurize them, and the licensee has implemented plant modifications. Since this violation was of very low safety significance (Green) and has been entered into the corrective action program as Condition Report CR- 2016- 008147, this violation is being treated as a non -cited violation consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000445/2017002 -05; 05000446/2017002- 05, Failure to Translate Design Requirements Into the As Built Facility)
05000446/FIN-2017002-02	2017Q2	Comanche Peak	Inadequate Operability Evaluation for Safety - related Pipe Supports	Green . The inspector s identified a non- cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, that occurred when the licensee failed on two occasions to perform an adequate operability determination associated with multiple safety -related pipe supports. Specifically, the operability determination of multiple carbon steel pipe support clamps exposed to boric acid and a bent sway strut pipe restraint lacked the engineering rigor necessary to provide a high degree of confidence to support the operability of the components. Subsequently, the inspector s concluded that the licensee established reasonable expectation for operability once engineering provided the control room with further analysis on the degraded conditions, and the new information was reviewed and accepted. This issue was entered into the licensees corrective action program as Condition Report CR -2017- 05418. The licensee's failure to perform adequate operability determinations per plant procedures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee: (1) failed to perform the required corrosion evaluation for a comparison of material wastage against design dimensions of the pipe support clamps; (2) failed to perform a visual inspection of the material condition of the pipe support clamps as required by the work order; ( 3) used non- seismic design tolerances for the qualification of a seismically qualified strut in the immediate operability determination; and (4) failed to consider that the bent condition of the strut occurred after the previously accepted visual examinations on the same pipe support. All these issues could have resulted in safety -related components failing to perform their specified safety function during accident conditions. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance (Green) because the finding: (1) it was not a design deficiency; (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; (4) and did not result in the loss of a high safety - significant non- technical specification train. This finding had a cross -cutting aspect in the area of problem identification and resolution associated with resolution because the licensee failed to adequately assess the degraded condition of the pipe supports in a complete and accurate manner to support a reasonable expectation of operability (P.1).
05000445/FIN-2016004-01	2016Q4	Comanche Peak	Failure to Evaluate Inservice Testing Results of Power Operated Relief Valve	Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, for the licensees failure to evaluate inservice testing results of a power operated relief valve (PORV). Specifically, the licensee restored a unit 1 PORV to service that did not meet its specified opening time, which resulted in the inoperability of the low temperature overpressure protection (LTOP) system. Following maintenance on PORV 1-PCV-455A during October 2014, the licensee performed stroke time testing on the valve, but failed to recognize that the valve exceeded its test acceptance criteria until it failed again in May 2016. The licensee entered this issue into the corrective action program as CR-2016-003920. The failure to evaluate test results to ensure they met test requirements is a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Reactor Coolant System Equipment and Barrier Performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated October 7, 2016, Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 9, 2014, and Appendix G Attachment 1, Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier Integrity Screening Questions, the inspectors determined the finding affected the Barrier Integrity cornerstone and required a detailed risk evaluation because the finding involved the unavailability of a PORV during LTOP operations. Using the assumption that the slow opening time prevents the PORV from fulfilling its LTOP system function, a senior reactor analyst performed a bounding qualitative assessment, using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The influential assumptions used by the senior reactor analyst included an exposure time of approximately 9 hours and that the licensee maintained the availability of a single additional relief valve with capability sufficient to mitigate an LTOP event as described in the final safety analysis report. Using these assumptions, the senior reactor analyst determined that a bounding increase in core damage frequency for this issue was 1.45E-8 per year and was therefore, of very low safety significance (Green). The finding has a human performance cross-cutting aspect associated with work management, in that, the licensee failed to ensure that the work process includes the need for coordination with different groups or job activities (H.5).
05000445/FIN-2016004-02	2016Q4	Comanche Peak	Failure to Scope the Containment Ventilation System in the Maintenance Rule Program	Green. The inspectors identified a non-cited violation of 10 CFR 50.65(b)(2) associated with the licensees failure to scope the containment ventilation system into the maintenance rule program. Specifically, the containment ventilation system, a non-safety related system that is relied upon to mitigate accidents or transients and used in emergency operating procedures, was not included in the scope of the monitoring program specified in 10 CFR 50.65(a)(1). In response to this issue the licensee scoped the system in the plants maintenance rule monitoring program, and placed the equipment under 10 CFR 50.65(a)(1) monitoring requirements pending further review. The licensee entered this issue into the corrective action program as CR-2016-008491. The failure to monitor the performance and condition of a system that meets the maintenance rule scoping criteria of 10 CFR 50.65(b)(2) is the performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated July 1, 2012, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated October 7, 2016, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding affected the Mitigating Systems cornerstone and was of very low safety significance (Green), because the finding did not represent a loss of system function and the system was not designated as high safety-significant in accordance with the licensees maintenance rule program. The finding has a human performance cross-cutting aspect associated with avoiding complacency, in that, the licensee failed to ensure that individuals recognized and planned for the possibility of mistakes and latent issues when re-evaluating the basis for excluding the system (H.12).
05000445/FIN-2016004-03	2016Q4	Comanche Peak	Licensee-Identified Violation	Title 10 CFR 50.65(a)(2), requires, in part, that monitoring of system performance under 10 CFR 50.65(a)(1) is not required where it has been demonstrated that performance of the system is being effectively controlled through appropriate preventive maintenance. Contrary to the above, from June 2014 to May 2016, the licensee failed to demonstrate that performance of the 480 Volt AC system, a system not being monitored under 10 CFR 50.65(a)(1), was being effectively controlled by preventive maintenance. Specifically, the 480 Volt AC system exceeded the established performance criteria in June 2014, and the licensee failed to evaluate its performance. The licensee discovered in May 2016 through an engineering review that the system had exceeded its criteria in 2014 and should have been placed in (a)(1) monitoring status. The licensee evaluated the system performance and ensured appropriate corrective action had been taken. The violation is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the violation is of very low safety significance (Green) because the finding did not represent a loss of system or function, and did not represent a loss of function of a single train for greater than its technical specification allowed outage time. The violation was entered into the licensees corrective action program as CR-2016-009963.
05000445/FIN-2016004-04	2016Q4	Comanche Peak	Licensee-Identified Violation	Title 10 CFR 50 Appendix B, Criterion V, requires, in part, that licensees shall perform activities affecting quality in accordance with instructions appropriate to the circumstances. Contrary to the above, on May 10, 2016, the licensee failed to perform safety chiller maintenance, a quality related activity, in accordance with the approved instructions. Specifically, licensee personnel failed to torque electrical connections on overload relays on the unit 1 train A safety chiller as required by the licensees work instructions. The inadequate torque was present until June 9, 2016, when the licensee performed thermography on the chiller electrical connections. The licensee discovered elevated temperatures, shut down the chiller, and replaced and torqued the affected components. The licensee determined that the chiller was inoperable from May 28, 2016, when it was required to be in service due to the unit entering Mode 4, until the chiller was restored on June 9. The violation is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the violation required a detailed risk evaluation (DRE) because the finding represented a loss of function of a single train for greater than its technical specification allowed outage time. A senior reactor analyst from Region IV performed the risk evaluation. The licensee provided an analysis demonstrating that the chiller would be able to perform its safety function for at least 24 hours. Based on that demonstration, the analyst was able to determine that the risk was of very low safety significance (Green). The violation was entered into the licensees corrective action program as CR-2016-005798.
05000416/FIN-2015301-01	2015Q4	Grand Gulf	Inadequate Plant Operating Procedures with Eight Examples	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to this, The licensees Off-Normal Procedure ONEP 05-1-02-I-1, Reactor Scram, Revision 125, does not provide all necessary guidance on how to scram the reactor. Once the immediate action of placing the mode switch in the shutdown position is completed, all additional guidance for shutting down the reactor using alternate methods is contained in EP-2A. However, the first backup method of using the scram pushbuttons is missing from both of these procedures. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensee is missing several off-normal procedures that are required by Technical Specifications based on commitments to NRC Regulatory Guide 1.33, Revision 2. Specifically, there are no off-normal procedures for 1) a total or partial loss of DC power, 2) electrical grounds, and 3) partial or total loss of all annunciators. The licensee is committed to revision 2 of this regulatory guide in its Technical Specifications. These procedure deficiencies were entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Emergency Procedure 05-1-02-II-1, Attachment III, Shutdown from the Remote Shutdown Panel, Revision 47, does not include all of the required steps to complete the attachment. Step 3.2.5a of this procedure requires an operator to obtain one key while two keys are actually required to complete the task. One key is required to open the protective box covering the switch and a different key is required to operate the switch. This procedure discrepancy led to delays and confusion during examination administration by applicants and during examination validation by licensed operators. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Emergency Procedure 05-S-1-EP-1, Attachment 6, Defeating Reactor Feed Pumps RPV Level 9 Trips, Revision 32, contains labeling discrepancies in that the relay nomenclature in the procedure does not match the nomenclature in the main control room cabinet 1H13-P612 Bay B. This caused confusion among both the applicants and licensed operators. The confusion delayed the completion of the task administered during the examination. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees System Operating Instruction 04-1-01-P41-1, Standby Service Water System, Revision 140, Section 4.2, contains labeling discrepancies in that the control board labeling for several switches do not match the nomenclature listed in the procedure for the associated switches. Specifically, steps 4.2.2A(4)(a), 4.2.2A(4)(b), and 4.2.2A(6) each have a discrepancy. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Alarm Response Instruction 04-02-1H13-P870-2A-E1, Revision 134, for the residual heat removal (RHR) alarm RHR A PMP RM FLOODED contains non-conservative guidance to close the suction valve (valve 1E12-F004A) for RHR pump A without regard to ensuring that the pump is secured first. This creates a condition where the safety-related residual heat removal pump is tripped on interlock only in order to prevent damage. The expectation provided to the NRC by the operations staff is that the operators should first trip the residual heat removal pump and then shut the suction valve. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensee was unable to locate any written guidance for placing a safetyrelated diesel generator in maintenance mode to prevent automatic start and subsequent overheat of the machine when cooling water is unavailable. According to the Updated Final Safety Analysis Report, Section 9.5, Revision LDC 05077, the diesel generator jacket cooling water system provides sufficient heat sink to permit the standby diesel engines to start and operate for 2 minutes without cooling water available. Procedures that were reviewed included SOI 04-1-01-P75-1, SOI 04-1-01-Y47, and ONEP 05-1-02-I-4. An additional NRC concern for this sequence is that there is no time critical action associated with securing these diesel generators when cooling water (standby service water) is not available. The licensee needs to review the risk management program and ensure that this is not assumed in the risk management profile or if it is assumed, then operators are trained and can implement the shutdown in the appropriate time to prevent equipment damage. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The licensees Equipment Performance Instruction 04-1-03D21-1, Monthly Area Radiation Monitors Functional Test, Revision 37, has confusing guidance which led several applicants in not being able to complete the task administered during the NRC initial license examination. Specifically the procedure has a limit and precaution stating that not all ARM module function switches spring return to OPERATE after being taken to ALARM. Some must be manually returned to OPERATE after being taken to ALARM while the specific steps in the procedure have the operator place and hold function switch in alarm and then release. No guidance is given within the step to return the switch to operate and this creates a situation where the observation of indication returning to normal does not occur. A precaution in the front matter in the procedure stating that the equipment may not function as the procedure is written is not sufficient to meet the quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2015-07209. The failure of these eight procedures to have the appropriate qualitative and/or quantitative criteria to complete these activities was a performance deficiency. The finding was more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, inadequate procedures could adversely affect the operating crews ability to take appropriate actions to ensure reactor safety is being maintained. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, the team determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program for greater than 24 hours. The finding has a cross-cutting aspect in the area of human performance associated with procedure adherence because individuals did not follow the processes to change or correct procedures that contained incorrect, missing, or non-conservative guidance (H.8).
05000482/FIN-2015004-01	2015Q4	Wolf Creek	Inadequate Measures to Assure SGK05A Issues Were Promptly Corrected	The inspectors identified a Green cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees inadequate measures to assure that corrective action was taken to preclude repetition of a significant condition adverse to quality. Specifically, measures to correct train A Class 1E electrical equipment air-conditioning system (SGK05A) issues following two trips of the unit on October 18, 2013, failed to preclude repetition, which resulted in the SGK05A unit tripping twice on May 15, 2015; the train A safety-related batteries, inverters, and alternating and direct current buses being declared inoperable due to the loss of area cooling; two separate Technical Specification 3.0.3 entries; and separate technical specification required reactor power reductions to 93 and 94.7 percent. The licensees immediate corrective actions included troubleshooting to determine the direct cause of the compressor trips, stationing a dedicated operator following the second trip on May 15, 2015, and subsequently implementing Temporary Modification 15-013-GK-00, which restored compliance. Actions to prevent recurrence following the May 15, 2015, SGK05A trips, documented in apparent cause evaluation 96392, included conducting a seminar with station managers to review lessons learned from the event, completing a change package to replace the SGK05A compressor that has been the source of residual contamination that has led to numerous trips of the unit, and tracking of the timely replacement of the SGK05A compressor with a due date of December 15, 2016. Wolf Creek entered this issue into its corrective action program as Condition Reports 96392 and 96397. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the train A safety-related batteries, inverters, and alternating and direct current buses became inoperable and their capability to respond to initiating events to prevent undesirable consequences was impacted as a result of the SGK05A unit tripping. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 3 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects a mitigating structure, system, and component. The performance deficiency does not affect the design or qualification of a mitigating structure, system, and component, and the structure, system, and component did not maintain its functionality. Additionally, the finding does not represent a loss of system and/or function, the finding does not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than their technical specification allowed outage time, and the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, senior managers did not ensure successful completion of the replacement of the SGK05A compressor in Refueling Outage 20, which was a missed opportunity that resulted in the SGK05A unit tripping twice on May 15, 2015, as a result of the same direct cause.
05000482/FIN-2015004-02	2015Q4	Wolf Creek	Failure to Ensure Essential Service Water Valves Were Adequately Protected from External Flooding Hazards	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures to assure that applicable regulatory requirements and the design basis, for applicable structures, systems, and components, are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to ensure that safety-related essential service water valves in the control building were adequately protected from external flooding hazards in the event of a design basis local intense precipitation event, which resulted in a reasonable doubt on the operability of safety-related essential service water valves. The stations immediate corrective actions included entering the condition into the corrective action program and performing a prompt operability evaluation that showed the essential service water valves remained operable. Additional corrective actions include accelerating three Fukushima project schedules that include a new sump pump in the turbine building area four cable vault, ground and surface water improvements for non-safety related electrical duct banks, and new sump pumps in electrical manholes near the turbine building. The violation was entered into the licensees corrective action program as Condition Report 102250. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, during design basis local intense precipitation events, the safety-related essential service water train A and B service water cross-connect motor-operated valves EFHV0023, EFHV0024, EFHV0025, and EFHV0026, and the essential service water train A and B to service water system valves EFHV0039, EFHV0040, EFHV0041, and EFHV0042 were susceptible to external flooding hazards, and there was a reasonable doubt on the operability of these essential service water valves; however, subsequent evaluation determined that the essential service water valves would not have been impacted in the event of a design basis local intense precipitation event, and the valves were determined to be operable. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 2 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects mitigating structures, systems, and components. The finding is a deficiency affecting the design or qualification of mitigating structures, systems, and components, and the structures, systems, and components maintained their operability and functionality. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross cutting aspect in the area of human performance, challenge the unknown, because Wolf Creek individuals did not stop when faced with uncertain conditions. Specifically, the licensee did not maintain a questioning attitude during flooding walk-downs performed in accordance with NEI 12-07 or during evaluation of Condition Report 59257 to identify and resolve unexpected conditions like the floor drain pathway from the communication corridor to the control building basement (room 3101), which was an opportunity for the station to identify the open pathway from the exterior of the plant.
05000482/FIN-2015004-03	2015Q4	Wolf Creek	Failure to Perform an Adequate Operability Determination and Consider Design Basis Events	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish activities affecting quality in accordance with Procedure AP 26C-004, Operability Determination and Functionality Assessment, Revision 31. Specifically, the licensee failed to document an operability determination of sufficient scope to address the capability of safety-related essential service water valves in the control building to perform their specified safety functions in the event of a design basis local intense precipitation event. Immediate corrective actions included completing a prompt operability determination and performing analyses that determined the valves remained operable. Additional corrective actions include accelerating three Fukushima project schedules that include a new sump pump in the turbine building area four cable vault, ground and surface water improvements for non-safety related electrical duct banks, and new sump pumps in electrical manholes near the turbine building. The violation was entered into the licensees corrective action program as Condition Report 100299. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, during design basis local intense precipitation events, the safety-related essential service water train A and B service water cross-connect motor-operated valves EFHV0023, EFHV0024, EFHV0025, and EFHV0026, and the essential service water train A and B to service water system valves EFHV0039, EFHV0040, EFHV0041, and EFHV0042 were susceptible to external flooding hazards, and there was a reasonable doubt on the operability of these essential service water valves; however, subsequent evaluation determined that the essential service water valves would not have been impacted in the event of a design basis local intense precipitation event, and the valves were determined to be operable. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 2 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects mitigating structures, systems, and components. The finding is not a deficiency affecting the design or qualification of mitigating structures, systems, and components; the finding does not represent a loss of system and/or function; the finding does not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than their allowed outage times; and the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, conservative bias, because Wolf Creek did not use decision making-practices that emphasize prudent choices over those that are simply allowable, and proposed action was not determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee did not consider long-term consequences or design basis events when determining how to resolve emergent concerns like the unexpected water in room 3101, which resulted in the licensees failure to thoroughly evaluate and assess impacts to the plant when Condition Report 96404 was entered into the corrective action program on May 17, 2015.
05000482/FIN-2015010-01	2015Q4	Wolf Creek	Incomplete and Inaccurate Medical Information Resulted in Issuance of a Renewed Operator License Without a Required Medical Restriction	Wolf Creek Nuclear Operating Corporation (Wolf Creek) identified an apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information. Specifically, on January 10, 2010, Wolf Creek submitted certified copies of an NRC operator license application that did not specify that the applicant required a restriction (to take medication as prescribed) in order to maintain medical qualifications. The NRC issued the renewed operators initial license on February 25, 2010, but without the necessary medical restriction. On July 15, 2015, the NRC issued the license amendment with the new restriction. This issue was entered into Wolf Creeks corrective action program. The inspector determined that Wolf Creeks failure to provide complete and accurate information to the NRC in the operator license application and to notify the NRC of a change in a licensed operators status for a condition was a performance deficiency. This performance deficiency was known by the licensee and within its ability to foresee and correct and should have been prevented. The inspector determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC relies upon Wolf Creek to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued a renewed operator license to the applicant based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of Inspection Manual Chapter 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute apparent violations in accordance with the NRCs Enforcement Policy and their final significance will be dispositioned in separate future correspondence.
05000458/FIN-2014302-01	2014Q4	River Bend	Inadequate System Operating Procedures with Two Examples	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to this, System Operating Procedure SOP-0049, 125 VDC SYSTEM (SYS # 305), Revision 29, did not have the necessary qualitative acceptance criteria (procedure steps) to accomplish the required activity of transferring the 125 VDC standby switchgear ENB-SWG01A to the backup charger using Section 5.7 of this procedure. During in-plant job performance measure validation for the initial exam, licensed operators were unable to simulate the transfer using System Operating Procedure SOP-0049. This procedure directed the operators to use an operator aid that, according to the procedure, was located inside panel BYS-TRS4. The operator aid was not inside the panel and was never found. Because of this, the job performance measure had to be rejected and another developed. To correct this issue, the licensee added the appropriate steps to System Operating Procedure SOP-0049 that were originally located in the missing operator aid and released it for use as Revision 30 on December 11, 2014. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-RBS-2014-05684. System Operating Procedure SOP-0071, ROD CONTROL AND INFORMATION SYSTEM (SYS # 500), Revision 29, did not have the necessary qualitative acceptance criteria (procedure steps) to accomplish the required activity of clearing a rod-block after pulling a control rod to raise reactor power during a start-up. During exam administration, an applicant for a senior reactor license could not get the rod block and associated alarm reset during a scenario using "Method 1" as described in System Operating Procedure SOP-0071. This procedure had incorrect guidance in Section 5.13 using "Method 1" in that the ROD SELECT CLEAR push button must be pressed several times to clear the rod block and this method only directed a single push of this button to reset the rod block and its associated alarm. Because of this, the applicant struggled to get through the reactivity change for the reactor during the scenario. To correct this issue, the licensee is working through the procedure change process for this procedure and has informed the licensed operator crews of the issue with "Method 1" until the appropriate steps are corrected within the procedure and it is released as Revision 30. This procedure deficiency was entered into the licensees corrective action program as Condition Report CR-RBS-2014-06331. The failure of these two procedures to have the appropriate qualitative criteria to complete these two activities was a performance deficiency. The finding was more than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, inadequate procedures could adversely affect the operating crews ability to take appropriate actions to ensure reactor safety is being maintained. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the team determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two s eparate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program for greater than 24 hours. The finding has a cross-cutting aspect in the area of human performance associated with documentation because the organization did not ensure that the procedures were accurate and up to date for these activities (H.7).
05000498/FIN-2014005-01	2014Q4	South Texas	Failure to Identify a Condition Adverse to Quality on Train A Emergency Diesel Generator	The inspectors documented a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality following an unexpected alarm on the train A emergency diesel generator. Specifically, after receiving the, E-5 Starting Air System Malfunction alarm, the licensee did not identify the correct cause of the alarm or take the necessary action to ensure the operability and reliability of the emergency diesel generator. As a result, the train A emergency diesel generator was degraded for 20 days, and was later rendered inoperable and non-functional for approximately 26 hours when operators removed the only air start subsystem that remained unaffected from service. This issue was entered into the corrective action program as Condition Report 14-18639, and the cause was corrected. Failure to identify the cause for the starting air system alarm and recognize that this degraded the starting function was a performance deficiency. This performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correctly identify and correct the cause of the E-5 Starting Air System Malfunction alarm resulted in the train A emergency diesel generator being degraded and later inoperable. Using NRC Inspection Manual 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance (Green) because it did not: 1) affect the design or qualification of a mitigating structure, system, or component; 2) represent a loss of system and/or function; 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as having high safety-significance. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with Evaluation because the licensee failed to thoroughly evaluate the issue to ensure that resolutions address the causes and extent of conditions commensurate with the safety significance. Specifically, the licensees failure to fully evaluate the cause of the starting air system alarm, and as a result, failed to recognize and correct the out-of-position valve before it rendered the system inoperable (P.2).
05000498/FIN-2014005-02	2014Q4	South Texas	Failure to Update the UFSAR for the Ultrasonic Feedwater Flow Measurement System	The inspectors identified a non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making Reports, for the failure to update the Updated Final Safety Analysis Report with information on the installation and use of the ultrasonic feedwater flow measurement system to control reactor power and calibrate nuclear instruments, which was installed in both units by the end of 1999. This violation was entered into the corrective action program as Condition Report 15-420. The failure to update the Updated Final Safety Analysis Report, as required by 10 CFR 50.71(e), with a description of the ultrasonic feedwater flow measurement system was a performance deficiency. The inspectors determined that this performance deficiency was not more than minor. However, because it had the potential to impact the NRCs ability to perform its regulatory oversight function, the inspectors assessed more the significance of the violation using traditional enforcement. Using the NRC Enforcement Policy to evaluate the significance, the violation was determined to be a Severity Level IV violation in accordance with Section 6.1.d.3, since the lack of information in the Updated Final Safety Analysis Report was not used to make an unacceptable change to the facility or procedures. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000368/FIN-2014004-03	2014Q3	Arkansas Nuclear	Failure to Implement Procedural Requirements for Axial Shape Index during a Rapid Power Reduction	The inspectors documented a Green self-revealing non-cited violation of Technical Specification 6.4.1.a for the failure to implement procedures for changing load recommended by Regulatory Guide 1.33, Revision 2, Appendix A, Section 2.f, dated February 1978. Specifically, the licensee did not maintain axial shape index within the limits of the core operating limits report during a rapid power reduction at the end of core life, resulting in an automatic reactor trip. The issue was documented in Condition Report CR-ANO-C-2014-01142. The inspectors determined that the failure to maintain axial shape index within the limits of the core operating limits report during a rapid power reduction was a performance deficiency. The performance deficiency is more than minor because it is associated with the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge the critical safety functions during shutdown as well as power operations. Specifically, the failure to maintain axial shape index caused an automatic reactor trip. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 1, Initiating Events Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding did cause a reactor trip but did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with training because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce. Specifically, the operators were not trained to understand the effects of the axial shape index during rapid power reductions with a core at an End-of-Life condition.
05000313/FIN-2014004-01	2014Q3	Arkansas Nuclear	Improper Maintenance on Circuit Breaker Caused Loss of Unit 1 Decay Heat Removal Pump	Inspectors documented a Green self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to ensure activities affecting quality were accomplished in accordance with documented instructions. Specifically, the licensee failed to follow Job Order JO-00968863 for replacement of a prop spring in circuit breaker MA137. As a result, the wrong prop spring was replaced, reducing the reliability of the Unit 1 train B decay heat removal pump P-34B and ultimately causing a failure of the pump to start. The licensee corrected the condition by replacing the breaker and returning the pump to service. The issue was documented in Condition Report CR-ANO-1-2013-00701. The inspectors determined that the failure to follow Job Order JO-00968863 in 1998 for replacement of a prop spring in circuit breaker MA137 was a performance deficiency. The performance deficiency was 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 was therefore a finding. Specifically, the failure to replace the appropriate prop spring in 1998 adversely affected the availability and reliability of Unit 1 decay heat removal pump P-34B and caused a failure to start in 2013. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 3, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding did not represent a loss of system safety function and did not represent an actual loss of safety function of at least one train for greater than its technical specification allowed outage time. The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current licensee performance.
05000368/FIN-2014004-02	2014Q3	Arkansas Nuclear	Failure to Establish Preventative Maintenance on Unit 2 Main Steam Isolation Valves	Inspectors documented a Green self-revealing non-cited violation of Technical Specification 6.4.1.a for the licensees failure to establish procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, Section 9, February 1978. Specifically, the licensee failed to establish preventative maintenance procedures for valve internal inspection and testing of the Unit 2 main steam isolation valves. On December 23, 2013, the train A main steam isolation valve (2CV-1010-1) was declared Inoperable due to the valve sticking at fifteen percent open on multiple stroke attempts. The licensees cause evaluation identified that mechanical binding and corrosion of the valve internals were results of a lack of preventive maintenance. The licensee repaired the 2CV-1010-1 valve and performed subsequent testing to demonstrate Operability. The issue was documented in Condition Report CR-ANO-2-2013-02502. The inspectors determined that the failure to establish preventative maintenance procedures for valve internal inspection and testing of the Unit 2 main steam isolation valves was a performance deficiency. The performance deficiency is more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. Specifically, the lack of preventative maintenance adversely affected the reliability of the main steam isolation valve 2CV-1010-1 to close within the time assumed in the accident analysis. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, the inspectors determined the finding to be of very low safety significance (Green) because the finding did not represent the loss of a system safety function and did not represent an actual loss of safety function of at least one train for greater than its technical specification allowed outage time. The finding was determined to have a cross-cutting aspect in the area of problem identification and resolution, in that the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes commensurate with their safety significance. Specifically, during a previous stroke test of the 2CV-1010-1 valve in 2011, the licensee identified that the valve experienced a sluggish or jerky motion and took longer than normal to open. The licensee entered this issue into the corrective action program but did not fully evaluate and troubleshoot the condition adverse to quality to ensure resolution of the cause.
05000313/FIN-2014004-04	2014Q3	Arkansas Nuclear	Licensee-Identified Violation	The following violation of very low safety significance (Green) and Severity Level IV wa identified by the licensee and is a violation of NRC requirements which meets the criteria of th NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Title 10 CFR 55.49, Integrity of Examinations, requires, in part, that facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Contrary to the above, on June 24, 2014, the licensee caused a compromise to examination integrity by violating an examination security agreement to not divulge information about examination content to unauthorized individuals. The failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV violation because it involved a non-willful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because there was not an actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Requalification. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-1-2014-01062.
05000313/FIN-2013005-01	2013Q4	Arkansas Nuclear	Failure to Maintain Fluorescent Light Fixture Above Emergency Feedwater Pump in Seismically Qualified Configuration	Inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to hang the fluorescent light fixture above the Unit 1 motor driven emergency feedwater pump in a seismically qualified design configuration. This was not an immediate safety concern because operability was adequately demonstrated when the misconfiguration was identified and because the licensee restored the light fixture to its seismically qualified configuration on November 12, 2013. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-1-2013-02830. Inspectors concluded that the licensees failure to hang the fluorescent light fixture above the Unit 1 motor driven emergency feedwater pump in accordance with Drawing E-2060 was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the mitigating system cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. Specifically, the licensee failed to ensure that, during a design basis seismic event, the light would not fall and adversely impact the safety-related pump below. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of mitigatin equipment, in which the equipment maintained its operability; and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event The finding was determined to have a cross-cutting aspect in the area of human performance, associated with resources, for the licensees failure to ensure that sufficient personnel were available for light inspections. Specifically, during the safety-related room inspections that were completed on August 27, 2013, the licensee failed to identify that the light above the motor driven emergency feedwater pump was inappropriately hung, due to the hurried nature of the inspections.
05000368/FIN-2013005-02	2013Q4	Arkansas Nuclear	Inadequate Operability Evaluation Due to Failure to Characterize Weld Flaw	Inspectors identified a non-cited violation of 10 CFR 50.55a(b)(5), In-Service Inspection Code Cases, for the licensees failure to implement ASME Code Case N-513-2, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or Piping, Section XI, Division 1. Specifically, when a service water weld developed a leak the licensee failed to characterize the flaw using a volumetric inspection method. Th licensee corrected the condition by performing volumetric inspections of the flawed weld and then repaired the weld. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-2-2013-01961. Inspectors concluded that the licensees failure to characterize a service water weld flaw was a performance eficiency. The performance deficiency was 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 licensee failed to ensure the reliability of the service water system wasnt adversely affected by a significant weld flaw. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined this finding was of very low safety significance (Green) because the degraded condition was not a design deficiency that affected system operability; did not represent an actual loss of function or a system; did not represent an actual loss of function of a single train or two separate trains for greater than its technical specification allowed outage time; did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant; and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with resources, for the licensees failure to ensure adequate training of personnel. Specifically, personnel performing the flaw inspection were not adequately trained in the non-destructive testing requirements of the code case.
05000483/FIN-2013004-02	2013Q3	Callaway	Licensee-Identified Violation	Technical Specification 3.7.10, Control Room Emergency Ventilation System (CREVS), requires that two control room emergency ventilation system trains shall be operable in Modes 1, 2, 3, and 4 and during movement of irradiated fuel assemblies. Contrary to the above, on April 18, 2013, with the plant in Mode 6 for Refueling Outage 19, Callaway workers impaired the control building envelope, causing the control room emergency ventilation system to be rendered inoperable while a fuel assembly was in movement in the fuel handling building. Specifically, licensee workers blocked open door DSK32013, breaching the control building ventilation system envelope, to run temporary power cables to the train B battery chargers. The inspectors evaluated the finding in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment as determined in Appendix G, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\'OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer. Corrective actions included coaching of operations and planning staff on the correct modes of applicability for Technical Specification 3.7.10 and enhancing procedures and forms to evaluate the technical specification appropriately. This violation was entered into the licensees corrective action program as Callaway Action Request 201302882.
05000483/FIN-2013004-03	2013Q3	Callaway	Licensee-Identified Violation	Title 10 of the Code of Federal Regulations, Section 55.49, Integrity of Examinations and Tests, requires that facility licensees shall not engage in any activity that compromises the integrity of an examination. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination. Individuals with knowledge of the content of requalification exams are required to sign an exam security agreement based on NUREG 1021, Form ES-601-1, which reads, in part, I understand that I am not to instruct, evaluate, or provide performance feedback to those operators scheduled to be administered these examinations . . . . Contrary to the above, in June 2013, an individual who had developed questions for the upcoming biennial requalification written exam also developed weekly cycle exams which were administered to licensed operators. This was a violation of the exam security agreement, as developing exams is a form of evaluating operators. This item was entered into the licensees corrective action program as Callaway Action Request 201305585. Using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process, the finding was determined to have very low safety significance because the potentially compromised questions were replaced before they were administered and, therefore, did not affect the equitable and consistent administration of the test.
05000483/FIN-2013004-04	2013Q3	Callaway	Licensee-Identified Violation	Technical Specification (TS) 3.8.1.a, requires two qualified electrical circuits between the offsite transmission network and the onsite for AC power system during Modes1, 2, 3, and 4. Required Action A.3 of this TS requires that with one offsite circuit inoperable, the licensee restore the circuit to operable status within 72 hours or be in Mode 3 within the next 6 hours. Contrary to the above, one offsite circuit was inoperable but was not restored within 72 hours, and the plant was not placed in Mode 3 within the next 6 hours. Specifically, on May 28, 2013, an oil leak on the startup transformer was discovered that had most likely begun during maintenance performed on the component on May 19, 2013, near the completion of a refueling outage. However, unaware of the leak at the time, the plant entered operating Mode 4, on May 22, 2013, and exceeded the 72-hour action. The licensee completed the leak repair on May 30, 2013. Using Inspection Manual Chapter 609, Appendix A, Exhibit 2, Mitigating Systems Cornerstone screening questions, Section A, the finding was determined to be of very low safety significance (Green) because within the first 24 hours, the startup transformer would have still supplied AC power to plant safety systems. The violation was entered into the licensees corrective action program as Callaway Action Request 201304347.
05000483/FIN-2013004-01	2013Q3	Callaway	Failure to Administer a Comprehensive Requalification Operating Test	The inspectors identified a non-cited violation of 10 CFR 55.59, Requalification, for failure to administer a comprehensive annual requalification operating test to one crew. After a quality review by NRC inspectors, it was determined that the job performance measure set administered in Week 2 of the testing cycle did not contain at least 40 percent alternate path job performance measures, as required by Procedure CTM-OPS, Callaway Training Manual: Operations Programs, Section 6.5.3.g.1.c. One of the job performance measures which the licensee had credited as an alternate path did not meet the criteria to be considered an alternate path, thereby leaving only one actual alternate path job performance measure in the set (20 percent). As an immediate corrective action, the licensee replaced one of the job performance measures from the Week 2 set with a new alternate path job performance measure which was administered to the affected operators, thereby ensuring that the 40 percent requirement was met prior to the completion of the 2-year requalification cycle. This issue was entered into the licensees corrective action program as Callaway Action Request 201306740. Failure to administer a comprehensive annual operating test containing at least 40 percent alternate path job performance measures to one crew is a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it adversely impacted the human performance attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the finding could have become more significant in that allowing licensed operators to return to the control room without a valid demonstration of appropriate knowledge on the annual operating test could be a precursor to a more significant event if latent knowledge deficiencies went unidentified. Using NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process, the finding was determined to have very low safety significance (Green) because, while it was related to annual operating test quality, less than 40 percent of the reviewed job performance measures and simulator scenarios were flawed (Manual Chapter 0609, Appendix I, Flowchart, Blocks 6, 7, and 8). This finding has a cross-cutting aspect in the area of resources associated with ensuring that work packages (in this case exam packages) are complete, accurate, and up-to-date such that industry standards for exam quality are met.
05000498/FIN-2012005-03	2012Q4	South Texas	Failure to Maintain Adequate Fire Penetration Seal Material Thickness	The inspectors identified a non-cited violation of Technical Specification 6.8.1.d, Fire Protection Program Implementation, for the failure to follow work order package instructions requiring the use of Drawing C012-00081-F7F, Detail E-1 Silicone Elastomer Typical Electrical Pen. Seals (Walls & Floors), to establish 6 inches of fire retardant sealant material for penetrations in Units 1 and 2. The inspectors noticed that Unit 1 train B safety-related 4160 Vac switchgear room electrical penetration F4476 had gaps around the edge. A design change installed new electrical cables that required the penetration be sealed using work order package 139376, that stated the penetration seal WILL BE IAW the Penetration Seal Permit and detail Drawing C012-00081-F7F. During the repair activities to correct the gaps, it was discovered that a portion of the seal was only 4.5 inches. The licensee captured this issue as Condition Report 12-28283. Corrective actions included restoring the seal to 6 inches, performing additional analysis to support a 3-hour fire barrier with just 5 inches, and performing extent of condition inspections. The finding was more than minor because it was associated with the Initiating Events Cornerstone attributes of Design Control and Procedure Quality, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions because it resulted in multiple fire penetration seals being declared nonfunctional as a result of being less than the design thickness. The inspectors used Manual Chapter 0609, Attachment 0609.04, to determine that fire protection issues are processed through Appendix F, Fire Protection Significance Determination Process, dated February 28, 2005. The inspectors used Appendix F, Attachment 1, to determine that the finding was of very low safety significance because it was a Moderate A fire confinement issue that screened out using Task 1.3.2 questions, since the seals would still have provided a 2-hour fire endurance rating or a 20 minute fire endurance rating without the seal being subject to direct flame impingement. In addition, this finding had human performance cross-cutting aspects associated with work practices because the licensee did not communicate human error prevention techniques such as self and peer checking, commensurate with the risk, such that the work activity was performed safely
05000498/FIN-2012005-01	2012Q4	South Texas	Failure to Perform Pressure Testing of the Reactor Vessel Flange LEAK-OFF Lines	Inspectors identified a non-cited violation of 10CFR50.55a(g)(4) involving the licensees failure to perform a system pressure test of the reactor vessel flange leak-off line of Units 1 and 2, in accordance with the applicable edition of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Contrary to the above, prior to November 1, 2012, the licensee failed to perform the required pressure test of the reactor vessel flange seal leak-off line for both units. Specifically, the licensee failed to implement the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, Class 2 requirements for pressure retaining components as provided by Article IWC 5220, System Leakage Test. The licensee entered the finding into their corrective action program as Condition Report 12-28600. The inspectors determined that the licensees failure to perform a pressure test of the reactor vessel flange leak-off line was a performance deficiency. This finding was more than minor because it affected the Initiating Events Cornerstone attribute of Equipment Reliability and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609, Attachment A, The Significant Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. This issue did not have a crosscutting aspect associated with it because it is not indicative of current performance
05000498/FIN-2012005-02	2012Q4	South Texas	Failure to Follow Procedure for the Control of Tools for Use on Stainless Steel	Inspectors identified a non-cited violation of very low safety significance of Technical Specification 6.8.1.a and Regulatory Guide 1.33, for the failure to follow procedures that ensured abrasive tools for use on stainless steel systems were not contaminated with carbon steel. Specifically, the inspectors determined that the licensee was not maintaining tools as required by Procedure 0PGP03-ZG-0001, Control of Materials and Products By User Groups, Revision 30, and Procedure 0PNP01-ZP-0032, Tools and Measuring &Test Equipment Control, Revision 6, because inspectors observed multiple instances of tools coded for use on stainless steel or aluminum bronze stored with tools marked for use on carbon steel, rust deposits on tools marked for use on stainless steel, and rust deposits on stainless steel components in the plant. This indicated that carbon steel contaminated tools may have been used on these systems. The licensee took corrective actions to segregate the coded tools and trained tool room attendants to properly store and mark abrasive tools designated for use on stainless steel, and evaluated the systems with indications of rust deposits. This issue was entered into the licensees corrective action program as Condition Report 12-28689. Inspectors determined the failure to assure that abrasive tools designated for exclusive use on stainless steel were stored separately from tools used on other materials was a performance deficiency. This finding was more than minor because it affected the Initiating Events Cornerstone attribute of Equipment Reliability and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609, Attachment A, The Significant Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. This finding had a cross-cutting aspect in the area of human performance work practices in that the licensee failed to effectively communicate expectations regarding procedural compliance, and personnel did not follow procedures. Specifically, the inspectors observed that although there were requirements to segregate tools, tools were not consistently segregated when returned to the storage locations as required by procedures
05000498/FIN-2012005-04	2012Q4	South Texas	Licensee-Identified Violation	Technical Specification 3.0.4 requires, in part, that entry into a mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation for an unlimited period of time, or after performance of a risk assessment addressing inoperable systems, or when specifically allowed by the specification. Contrary to the above, in April 2010 and November 2011, Unit 2 transitioned from Mode 4 to Mode 3 without all required equipment being operable, without performing a risk assessment, and when not allowed by the specification. Specifically, the turbine trip signal from the reactor trip breakers, the turbine trip signal from the reactor trip signal, and the turbine trip signal from a steam generator HI-HI level were all inoperable due to a jumper being installed for testing when the plant transitioned from Mode 4 to Mode 3. The inspectors used Manual Chapter 0609, Appendix A since the finding was identified after residual heat removal was secured, and determined that the finding was of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment. The licensee entered this issue into the corrective action program as Condition Report 11-27377.
05000285/FIN-2012012-07	2012Q4	Fort Calhoun	Licensee-Identified Violation	Title 10 of the Code of Federal Regulations, Part 50.47(b)(16), requires, in part, that licensee emergency planners are properly trained. Contrary to the above, two licensee emergency planners were not trained in accordance with station training requirements as described in EPDM 12, Emergency Planning Staff Training and Qualification Program, Revision 3. Specifically, one emergency planner was 36 months overdue on five required reading packages and 30 months overdue on four required reading packages, and another emergency planner was 36 months overdue on a required offsite training course. The finding is more than minor because if left uncorrected it could have led to a more significant safety concern and it impacted the Emergency Response Organization Performance attribute. The finding could have led to a more significant safety concern because an untrained licensee emergency planner could have failed to recognize and correct risk-significant emergency preparedness issues. The finding was evaluated using the EP Significance Determination Process and determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements and was not a lost or degraded planning standard function. The planning standard function was not degraded because the licensee had a formal program for training emergency preparedness department staff, the identified emergency planners had completed some required training activities, and three other emergency planners were current in their training activities. The finding was entered into the licensees CAP as CR 2012-10400.
05000285/FIN-2012012-04	2012Q4	Fort Calhoun	Failure to Adequately Implement the Maintenance Rule Program	The team identified a Green NCV of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants which states, in part, that the licensee shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industry-wide operating experience. Specifically, from March of 2012 until October of 2012, the licensee allowed the maintenance rule program to deteriorate by not performing initial screenings in a timely fashion. In some cases, the initial screenings were being done months later and the actual evaluation of the equipment status was not being performed at all for a period of eight months. Consequently, several components, including electrical relays and electrical load centers, were not being evaluated in accordance with program requirements. Additionally, the licensee was not implementing the operating experience program as required by this regulation. The licensee discontinued performance of level 1 and level 2 operating experience evaluations by direction from the senior management in August of 2012 based on resource concerns. Several examples where operating experience was not properly evaluated included the containment spray pump low oil issues (ACA 2008-5695), vendor manual updates, and loose fasteners (both electrical and mechanical) from San Onofre Nuclear Generating Station Licensee Event Reports (LER) 3612007005, 3612007006, and 3612008006. This finding was entered into the licensees Corrective Action Program as CR 2012-17572. The team determined that the failure to adequately implement the maintenance rule was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected it could lead to a more serious concern. Using Manual Chapter 0609, Attachment 4, Significance Determination Process router on Table 3, it sends the user to Appendix G for Shutdown Operations Significance Determination Process. Using Checklist 4 of Appendix G for the given plant conditions, the finding was determined to have very low safety significance (Green) because the finding did not 1) increase the likelihood of a loss of RCS inventory, or 2) degrade the licensees ability to terminate a leak path or add RCS inventory when needed, or 3) degrade the licensees ability to recover decay heat removal once it is lost. This finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee did not use conservative assumptions in decision making and did not identify the possible unintended consequences of suspending maintenance rule program activities and the corresponding impact on the program.
05000285/FIN-2012012-03	2012Q4	Fort Calhoun	Failure to Properly Manage the Functionality of the River Sluice Gates	The team identified a finding exemplified by multiple violations for the failure to manage the functionality of the river sluice gates. Specifically, the licensees preventive maintenance program requirements were not appropriately implemented for a period of 6 months and as a result, the functionality of the river sluice gates was improperly maintained. The examples were: 1. A licensee identified violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to perform preventive maintenance required to demonstrate the functionality of the river sluice gates. An NRC identified violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to accomplish activities affecting quality in accordance with prescribed instructions when in September 2012, the licensee failed to test the C and D river sluice gates in accordance with station procedure SAO-12-001, to properly maintain functionality of the river sluice gates. 2. An NRC identified violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to accomplish activities affecting quality in accordance with prescribed instructions when the licensee failed to test all six gates in October 2012, to maintain functionality of the river sluice gates in accordance with station procedure SAO-12-001. 3. An NRC identified violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to properly identify and timely enter conditions adverse to quality into the Corrective Action Program following multiple failures of the river sluice gates. 4. An NRC identified violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to demonstrate effective control of performance of the circulating water system river sluice gates and failure to place the system in (a)(1) when system performance deteriorated. 5. An NRC identified violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to accomplish activities affecting quality in accordance with prescribed instructions when the licensee failed to make the appropriate functionality assessment when the circulating water river sluice gates failed to close during the August 2012 monthly test. The licensee entered these issues into their Corrective Action Program under various CRs described in the body of this report. The team concluded that the failure to manage the functionality of the sluice gates was a performance deficiency that warranted further evaluation. Specifically, the licensees preventive maintenance program requirements were not appropriately implemented for a period of 6 months and as a result, the functionality of the sluice gates was improperly maintained. Using the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined this finding affected the Mitigating Systems cornerstone. The finding is greater than minor because it is associated with both of the Mitigating Systems Cornerstone attributes of Equipment Performance and Protection Against External Factors and, it adversely affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding is bounded by the significance of a related Yellow finding regarding the ability to mitigate an external flooding event (Inspection Report 05000285/2010008). The inspectors determined the finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not take appropriate corrective action to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
05000285/FIN-2012012-01	2012Q4	Fort Calhoun	Hot Work Procedures Allowed a Roving Fire Watch	The inspectors identified a Green non-cited violation (NCV) of Technical Specification 5.8.1.c for the failure to maintain written procedures covering fire protection program implementation. Specifically, the licensee changed the hot work procedure to allow a roving fire watch in lieu of the continuous fire watch required by the fire protection program. The licensee entered this issue into their Corrective Action Program as Condition Report (CR) 2012-19945. The failure to maintain adequate written procedures covering fire protection program implementation was a performance deficiency. This finding was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone and it adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the risk significance of this finding using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because the performance deficiency involved a failure to adequately implement fire prevention and administrative controls for hot work activities. A senior reactor analyst performed a limiting Phase 3 evaluation and determined this finding had very low risk significance (Green). The finding did not have a cross-cutting aspect since it was not indicative of present performance.
05000285/FIN-2012012-06	2012Q4	Fort Calhoun	Licensee-Identified Violation	LER 05000285/2012-005-01 described a failure to monthly verify the automatic start features of the diesel fuel oil pumps. This was a violation of Technical Specification 3.7(1)e and Table 3-2, Item 12. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. It was determined to be of very low safety significance since there was not an actual failure of the automatic start features of the diesel fuel oil pumps. This issue was entered into the CAP as CR 2012-01324. This violation is also discussed in Section 4OA3.1.
05000285/FIN-2012004-04	2012Q3	Fort Calhoun	Failure to Ensure Breaker Coordination of 480 VAC Electrical Power Distribution System Was Maintained	The team identified a violation of 10 CFR 50 Appendix B Criteria III, Design Control. Specifically, the design modification package for the 480 VAC breaker replacements failed to ensure the breaker coordination for the 480 VAC electrical buses was maintained. As a result, feeder breaker 1B3A tripped unexpectedly during the fire event in the 1B4A switchgear. This performance deficiency also resulted in the loss of multiple buses on both trains of 480 VAC, including ECCS systems, from a single fault on a 480 VAC bus. This finding and its corrective actions will be managed by the NRCs Inspection Manual Chapter 0350 Oversight Panel. This finding is associated with Enforcement Action 12-121. The failure to ensure that the 480 VAC electrical power distribution system design requirements were maintained was a performance deficiency that was within OPPDs ability to foresee and prevent. The performance deficiency was reviewed using NRC Inspection Manual Chapter 0612, Appendix B, Issue Screening, and the issue was determined to be more than minor because it affected the Initiating Events Cornerstone attributes of protection against external events (i.e., fire) and design control. The issue adversely affected the associated cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of this finding is bounded by the significance of the Red finding documented in Inspection Report 05000285/2012010. The licensee entered this issue into its corrective action program as CR 2011-6621. The performance deficiency had a cross-cutting aspect in the area of human performance associated with resources because OPPD failed to ensure that station procedures for engineering changes, plant modifications, inspections, installations, and maintenance contained sufficient details.
05000285/FIN-2012005-02	2012Q3	Fort Calhoun	Untimely Corrective Actions for 480 VAC Breaker Issues	The NRC identified a noncited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to take timely corrective actions with respect to nonconforming conditions in several circuit breakers. These conditions were determined to have been the cause of the 1B4A bus bar failure that initiated a fire on June 7, 2011. These conditions were not corrected in a timely manner and the licensee continued to operate with a degraded breaker for nine months after the breaker tripped unexpectedly during the June 7, 2011, fire event. The licensee entered this issue into their corrective action program as CRs 2012-01884 and 2011-5414. The violation was determined to be more than minor because it affected the Initiating Events Cornerstone attribute of protection against external events (i.e., fire). The issue adversely affected the associated cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations because the condition that contributed to the fire event was left uncorrected. The finding screened to Green in accordance with IMC 0609, Appendix G because RCS makeup capability was not degraded. The inspectors determined that the issue had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program.
05000285/FIN-2012005-01	2012Q3	Fort Calhoun	Failure to Update the Safety Analysis Report Solid Waste	The inspectors identified a cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, for the failure to update the Updated Safety Analysis Report with a detailed description of the Original Steam Generator Storage Facility. Specifically, since December 2006, the licensee stored a significant source of radioactivity in the Original Steam Generator Storage Facility, but failed to describe the volume of waste, the principal sources of radioactivity, the total quantity of radioactivity, and the estimated dose rate at the site boundary per curie of radioactivity in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2012-05725. This issue was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. This issue is being characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy. Cross-cutting aspects are not assigned to traditional enforcement violations. NOV summary: The inspectors identified a cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, for the failure to update the Updated Safety Analysis Report with a detailed description of the Original Steam Generator Storage Facility. Specifically, since December 2006, the licensee stored a significant source of radioactivity in the Original Steam Generator Storage Facility, but failed to describe the volume of waste, the principal sources of radioactivity, the total quantity of radioactivity, and the estimated dose rate at the site boundary per curie of radioactivity in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2012-05725. This issue was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. This issue is being characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000285/FIN-2012004-01	2012Q3	Fort Calhoun	Failure to report an event to the NRC within 60 days for an operation prohibited by Technical Specifications	The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a) for the failure to submit a Licensee Event Report within 60 days after the discovery of performing an operation prohibited by technical specifications. The licensee failed to report to the NRC that they moved fuel while the Spent Fuel Pool Area Charcoal Filtration System, VA-66, was not in operation, contrary to Technical Specification 2.8.3(4). The licensee discovered in September 2011 that the fuel movement in December 2009 was inappropriate based on technical specifications, but failed to submit Licensee Event Report, 2012-008-0 until July 27, 2012. This issue was entered into the licensees corrective action program and evaluated with an Apparent Cause Analysis under Condition Report, 2012-08521 and 2012-08386 The failure to make an official report to the NRC regarding an operation prohibited by the Technical Specifications is a performance deficiency. The issue was dispositioned using traditional enforcement because failing to submit the Licensee Event Report had the potential to adversely impact the NRCs ability to perform its regulatory function. The issue is characterized as a Severity Level IV violation in accordance with the NRC Enforcement Policy, Section 6.9.d.9. Since this issue was dispositioned using traditional enforcement, there is no cross-cutting aspect.
05000285/FIN-2012004-02	2012Q3	Fort Calhoun	Fuel Move with SFP Ventilation Inoperable a Condition Prohibited by Technical Specification 2.8.3(4)	The inspectors identified a non-cited violation of very low safety significance of Technical Specification 2.8.3(4), the limiting condition for refueling operations in the spent fuel pool. In December 2009, the licensee performed refueling operations with the Spent Fuel Pool Area Charcoal Filtration System, VA-66, declared inoperable. The failure to establish an operable Spent Fuel Pool Area Charcoal Filtration System, VA-66, before moving spent fuel was a performance deficiency and a violation of Technical Specification 2.8.3(4). The licensee entered this issue into the corrective action program as Condition Reports 2012-08521, 2012-0836 and Licensee Event Report 2012-008-0. The performance deficiency was determined to be more than minor because it adversely impacted the attribute of the Barrier Integrity Cornerstone objective to maintain radiological filtration functionality during operations in the spent fuel pool to protect the public from radionuclide releases caused by accidents or events. Using IMC 0609 Appendix A, Barrier Integrity Significance Determination Process, the inspectors determined this finding to be of very low safety significance (Green). Although fuel movements were contrary to the licensees technical specifications limiting condition for refueling operations, the finding represented a degradation of the radiological barrier function provided for the spent fuel pool fuel building. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not effectively incorporate internal operating experience and lessons learned from previous VA-66 ventilation system failures during spent fuel pool refueling operations and plant safety. Specifically, the licensee failed to systematically collect, evaluate, and communicate to affected internal stakeholders in a timely manner relevant internal and external operating experience.
05000285/FIN-2012004-03	2012Q3	Fort Calhoun	Failure to Establish and Implement Adequate Procedures for Meteorological Monitoring and the Off-Site Dose Calculation Manual	Inspectors identified two examples of a non-cited violation of very low safety significance of Technical Specification 5.8.1 for the failure to adequately establish, implement, and maintain procedures for: (1) the onsite meteorological monitoring systems; and (2) reporting meteorological data in accordance with the Offsite Dose Calculation Manual requirements. The licensee entered these issues into the corrective action program as Condition Reports 2012-05658, 2012-05724 and 2012-05777. The failure to establish, implement, and maintain procedures to ensure the meteorological monitoring equipment is operable and required meteorological data is reported was a performance deficiency. This finding is more than minor because it affected the Public Radiation Safety cornerstone attribute of program and process. The failure to have and use applicable procedures to ensure the operability of the meteorological monitoring system and the accuracy of the Annual Radiological Effluent Release Report has the potential to impair public dose assessments of routine and accidental radioactive effluent releases. Using IMC 0609 Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined this finding to be of very low safety significance because the finding did not represent a significant degradation of the ability to assess dose to members of the public and the actual releases were well below established limits for members of the public. This finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to ensure that personnel, procedures, and other resources were adequate for the operability of the meteorological monitoring system and implementation of Offsite Dose Calculation Manual requirements related to the annual effluent report.
05000361/FIN-2012003-03	2012Q2	San Onofre	Failure to Maintain Foreign Material Exclusion Controls During Maintenance	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of maintenance personnel to implement procedures associated with foreign material exclusion controls while performing maintenance activities on safety-related 120Vac inverter equipment. Specifically, on June 8, 2012, maintenance personnel failed to follow Procedure SO123-FO-1, Site Foreign Material - 4 - Exclusion Control Program, Revision 6, and Procedure SO123-I-1.18, Foreign Material Exclusion (FME) Control, Revision 18, when maintenance personnel failed to implement adequate foreign material exclusions controls during inverter 2Y004 troubleshooting activities. This issue was entered into the licensees corrective action program as Nuclear Notification NN 202016714. The performance deficiency is more than minor, and therefore a finding, because it is associated with the Mitigating Systems Cornerstone attribute for human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, maintenance personnel failed to implement adequate controls, as required, to prevent the introduction of foreign materials during troubleshooting and repair activities associated with electrical cabinet of inverter 2Y004. Using Checklist 4 from the Manual Chapter 0609, Appendix G, Shut-down Operations Significance Determination Process, Phase 1 guidance, the finding is determined to have very low safety significance because all safety function guidelines were met, and thus, the finding did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the expectations regarding procedural compliance, and that personnel follow procedures were not effectively communicated to maintenance personnel regarding foreign material exclusion controls for unattended and opened electrical components
05000285/FIN-2012301-02	2012Q2	Fort Calhoun	Failure to Periodically Update the Final Safety Analysis Report for the Emergency Diesel Generators	The team identified a Severity Level IV non-cited violation of Title 10 CFR Part 50.71(e), for failure to periodically update the final safety analysis report originally submitted as part of the application for the license to assure that the information included in the report contains the latest information developed. Specifically, the licensee failed to update the final safety analysis report to include the required continuous hour rating used for emergency diesel generator testing as referenced by Technical Specification Surveillance Requirement 3.7(1)(a)(ii). After identification, the licensee entered this issue in the corrective action program as Condition Report 2011- 06612. The failure to update the final safety analysis report to include information that is specifically referenced by the technical specifications is a performance deficiency. The inspectors considered this issue to be within the traditional enforcement process because it has the potential to impede or impact the NRC\'s ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The violation is more than minor because the omitted final safety analysis report update information had a potential impact on safety and licensed activities in that the licensee did not have the required reference for a surveillance test. Consistent with the guidance in Section 2.2.2 and Section 6.1.d.3 of the NRC Enforcement Policy, the inspectors concluded that the violation is a Severity Level IV because the omitted information did not result in an unacceptable change to the facility or procedures.
05000361/FIN-2012003-04	2012Q2	San Onofre	Failure to Follow Design Control Procedures	The inspectors identified a non-cited violation 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of engineering personnel to follow Procedure SO123-XXIV-10.1, Engineering Design Change Process NECPs, Revision 28, to change the design, through physical plant modifications, of a facility used to handle radioactive material. Specifically, on February 2, 2012, engineering personnel issued as-built engineering change package NECP 800841701 which physically modified the design of the fuel reconstitution gantry crane with no turnover when an issued for construction engineering change package with turnover was required. This issue was entered into the licensees corrective action program as Nuclear Notification NN 202026584. The performance deficiency is more than minor, and therefore a finding, because it would become a more significant safety concern if left uncorrected since handling fuel with improperly modified equipment could result in fuel barrier damage. This finding cannot be evaluated by the significance determination process because Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At- Power Situations, and Appendix G, Shut-down Operations Significance Determination Process, do not apply to the spent fuel pool. This finding affects the Barrier Integrity Cornerstone and is determined to be of very low safety significance by NRC management review because it was a deficiency that did not result in the actual degradation of spent fuel. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the expectations regarding procedural compliance, and that personnel follow procedures were not effectively communicated to Design Engineering, Nuclear Fuels Management, and Project Management Organization personnel

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