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	Title	Quarter	Description
05000458/FIN-2018003-01	Licensee-Identified Violation	2018Q3	This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the design basis for those structures, systems, and components to which Appendix B applies is correctly translated into specifications, drawings, procedures, and instructions. The design basis for the control building air conditioning system, as specified in the updated safety analysis report, requires that the system be capable of performing its safety function in the event of a single failure in any component. Contrary to the above, the licensee failed to assure that the design basis was correctly translated into specifications for the control building air conditioning system. Specifically, while reviewing the control logic for the control building air conditioning system, the licensee discovered that the control logic was designed such that a single failure in a component in the control logic could have prevented the system from performing its specified safety function.
05000458/FIN-2018301-01	Inadequate Procedure for Shutdown Operations Protection Plan	2018Q3	The team reviewed a self-revealed Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide accurate qualitative procedural guidance to determine Shutdown Cooling Safety Function color state in Modes 4 and 5, using OSP-0037, Rev 36, Shutdown Operations Protection Plan, Attachment 1, Shutdown Cooling Function Color States. Specifically, OSP-0037 Rev 36 defines the term Flooded Up as, Flooded Up condition requires greater than 23 ft in the Reactor Cavity and the Cavity Gate open. OSP-0037 Rev 36, Attachment 1 contains a table with 6 columns that list combinations of decay heat loads (high, medium, and low), and reactor cavity water inventory, which are used to determine Shutdown Cooling risk. Only one of the six columns uses the correctly-defined term, Flooded Up, for reactor cavity water inventory. Two of the six columns state Flooded, and three of the columns state Not FL, neither of which are defined terms in OSP-0037. This creates the potential for an operator to misinterpret the meaning of the column, and select a color code for Shutdown Cooling that represents a lower risk than is actually present. This potential was confirmed by erroneous applicant performance on the July 2018 NRC initial license exam. As an interim corrective action, the station issued a night order clarifying the typographical error, and initiated action to revise the procedure. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2018-04414 The failure to provide accurate qualitative procedural guidance to determine Shutdown Cooling Function Color State is a performance deficiency. The inspectors determined the performance deficiency was more than minor because it adversely affected the Procedure Quality attribute of the Mitigating Systems cornerstone, the objective of which is to ensure the availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the procedure errors could cause a crew to underestimate Shutdown Cooling risk, with an adverse effect on conservative implementation of defense in depth in the planning, scheduling, and implementation of outage activities. The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix G, Shutdown Operation Significance Determination Process, dated May 9, 2014. 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 safety function of any train or safety system for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program, with cavity either flooded or not flooded; (5) did not degrade a functional auto-isolation of RHR on low reactor vessel level; (6) did not screen as potentially risk significant due to an external event; (7) did not involve Fire Brigade training and qualification requirements, or brigade staffing; (8) did not involve the response time of the Fire Brigade to a fire; and (9) did not involve fire extinguishers, fire hoses, or fire hose stations. No Cross-Cutting Aspect is assigned, because the procedural errors were introduced in Revision 19, issued September 16, 2009, and are therefore not indicative of current licensee performance.
05000458/FIN-2018301-02	Exam Security Compromise While Administering Simulator JPM	2018Q3	The team reviewed a self-revealed Green non-cited violation of 10 CFR Part 55.49, Integrity of examinations and tests, for the licensees compromise of a simulator JPM during exam administration. Specifically, on June 8, 2018, during a review of the draft exam, the NRC identified that the draft examiner guide for simulator JPM S7 contained exam material for simulator JPM S8. The licensee removed the erroneous exam material from the examiner guide, but failed to evaluate the extent of condition to ensure that the applicant handout did not also contain exam material from JPM S8. On July 26, 2018, while performing JPM S7, an applicant was reviewing his handout and found that the last page was the cue sheet for JPM S8, which was intended to be administered the following day. As an immediate compensatory measure, JPM S8 was administered to all applicants the day of the compromise, to prevent any opportunity for the compromise to spread or for applicants to specifically prepare for the JPM. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2018-04141 The compromise of initial license exam material to an applicant before the intended date of administration is a performance deficiency. The inspectors determined the performance deficiency was more than minor, and therefore a finding, because it adversely affected the Human Performance attribute of the Mitigating Systems cornerstone, the objective of which is to ensure the availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Additionally, if left uncorrected the performance deficiency could have the potential to lead to a more significant safety concern, by causing license decisions to be made based on compromised exams which were not administered in an equitable or consistent manner. The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix I, Licensed Operator Requalification Significance Determination Process. The team determined that the finding was of very low safety significance (Green) because the finding was related to initial license exam security (block 10), but did not cause an actual negative effect on the equitable and consistent administration of the initial license exam (block 11), because the exam team took immediate compensatory action to rearrange the exam schedule and administer the compromised JPM that day, preventing any opportunity for applicants to unduly prepare for the compromised JPM, or for other applicants to learn of it. The finding was related to the cross-cutting aspect of Evaluation in the cross-cutting area of Problem Identification and Resolution because after the NRC identified that draft simulator JPM S7 erroneously contained exam material from simulator JPM S8, the licensee failed to evaluate the extent of condition of this error to ensure that other exam materials were not also affected.
05000458/FIN-2018012-07	Failure to Perform 10 CFR 50.59 Evaluation for Main Feedwater System Sparger Nozzle Damage	2018Q2	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59 , Changes, Tests, and Experiments, for the licensees failure to provide a written safety evaluation for the determination that operation with compensatory measures for damaged feedwater sparger nozzles did not require a license amendment pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit. Specifically, the licensee failed to recognize that compensatory measures prohibiting operation in single loop conditions required technical specification changes, and as such required prior NRC approval.
05000458/FIN-2018012-02	Failure to Identify and Correct a Broken Feedwater Chemistry Probe	2018Q2	Two examples of a self-revealed non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified for the licensees failure to identify that a broken chemistry probe in the feedwater system had the potential to cause an adverse impact on plant safety, and promptly implement appropriate measures to address that condition.
05000458/FIN-2018012-06	Failure to Provide Adequate Procedures for Post-Scram Recovery	2018Q2	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to establish, implement and maintain a procedure required by Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, Procedure OSP-0053, Emergency and Transient Response Support Procedure, Revision 22, which is required by Regulatory Guide 1.33, inappropriately directed operations personnel to establish feedwater flow to the reactor pressure vessel using the main feedwater regulating valve as part of the post-scram actions. This resulted in the main feedwater regulating valves being operated outside their design limits. This resulted in catastrophic failure of the main feedwater regulating valve variseals and subsequent damage to multiple fuel assemblies.
05000458/FIN-2018012-05	Failure to Develop an Adequate Operational Decision-Making Issue for Compensatory Measures Related to a Degraded Condition of the Feedwater System Sparger Nozzles	2018Q2	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to develop an adequate Operational Decision-Making Issue (ODMI) document per Procedure EN-OP-111, Operational Decision-Making Issue Process. Specifically, the licensee failed to develop an ODMI that provided adequate guidance to the operators for safely operating the plant with degraded feedwater sparger nozzles.
05000458/FIN-2018012-03	Failure to Establish Procedural Guidance for Determining Core Flow During Unanticipated Single Loop Operations	2018Q2	The inspectors reviewed a self-revealed,non-cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish appropriate instructions in the abnormal operating procedure for thermal hydraulic instabilities. Specifically, the procedural step for determining core flow when in single loop operations at low power did not provide appropriate instructions to operators. As a result, station personnel could not conclusively determine core flow and inserted a manual reactor scram.
05000458/FIN-2018012-01	Failure to Conduct Adequate Transient Snap Shot Assessment Following Recirculation Pump Trip	2018Q2	The inspectors identified a finding for the licensees failure to adequately validate simulator response during a transient snap shot assessment following an unexpected trip of reactor recirculation pump A on December 19, 2012.
05000458/FIN-2018012-04	Failure to Submit a Licensee Event Report for a Manual Scram	2018Q2	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to submit a required licensee event report (LER). Specifically, on February 1, 2018, after an unexpected trip of the recirculation pump B, the licensee initiated a manual scram of the reactor that was not part of a preplanned sequence and failed to submit an LER within 60 days.
05000458/FIN-2018002-01	Failure to Correct Inadequate Technical Specification Pressure Temperature Curves	2018Q2	The inspectors identified a Severity Level IV 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. Specifically, after receiving vendor information indicating that existing technical specification pressure temperature (PT) curves were inadequate, the licensee failed to promptly identify and correct the condition through the license amendment process.
05000458/FIN-2018002-02	Enforcement Action (EA)-18-053: Enforcement Discretion for Tornado-Generated Missile Protection Noncompliances	2018Q2	Title 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 2, Design Bases for Protection Against Natural Phenomena, states, in part, that systems, structures, and components (SSCs) important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes. Criterion 4, Environmental and Dynamic Effects Design Basis, states, in part, that SSCs important to safety shall be appropriately protected against dynamic effects including missiles that may result from events and conditions outside the nuclear power unit. Section 3.5.2, Structures, Systems, and Components to be Protected from Missiles, of the Updated Safety Analysis Report (USAR) details the structures that are designed to withstand tornado missile impact.On February 7, 2017, the NRC issued Enforcement Guidance Memorandum (EGM) 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance, Revision 1 (ADAMS Accession Number ML16355A286). The EGM referenced a bounding generic risk analysis performed by the NRC staff that concluded that tornado missile vulnerabilities pose a low risk significance to operating nuclear plants. Because of this, the EGM described the conditions under which the NRC staff may exercise enforcement discretion for noncompliance with the current licensing basis for tornado-generated missile protection. Specifically, if the licensee could not meet the technical specification required actions within the required completion time, the EGM allows the staff to exercise enforcement discretion provided the licensee implements initial compensatory measures prior to the expiration of the time allowed by the limiting condition for operation. The compensatory actions should provide additional protection such that the likelihood of tornado missile effects are lessened. The EGM then requires the licensee to implement more comprehensive compensatory measures within approximately 60 days of issue discovery. The compensatory measures must remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Because EGM 15-002 listed River Bend Station as a Group A plant, enforcement discretion expired on June 10, 2018. On May 10, 2018, River Bend Station submitted a request to extend the enforcement discretion period to June 10, 8 2020. On May 31, 2018, River Bend Station submitted asupplement to the May 10 request. On June 6, 2018, the NRC granted an extension to the enforcement discretion until June 10, 2020. The initial conditions of Design Basis Accident (DBA) and transient analyses in the USAR, Chapter 6 and Chapter 15, assume Engineered Safeguards Features (ESF) systems are operable. The AC, DC, and AC vital bus electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, reactor coolant system, and containment design limits are not exceeded.The onsite standby power source for each 4.16 kV ESF bus is a dedicated emergency diesel generator (EDG). An EDG starts automatically on a loss of coolant accident signal (i.e., low reactor water level signal or high drywell pressure signal) or on an ESF bus degraded voltage or under voltage signal. In the event of a loss of preferred power, the ESF electrical loads are automatically connected to the EDGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a DBA such as a loss of coolant accident. Standby service water (SSW) is required by Technical Specification 3.7.1. The ultimate heat sink (UHS) consists of one 200 percent capacity cooling tower and one 100 percent capacity water storage basin. The UHS basin capacity is required by Regulatory Guide 1.27 and USAR 9.2.5 to maintain a minimum of 30 days inventory to mitigate the consequences of a DBA without replenishment. The UHS is designed to perform its safety function assuming a single failure coincident with a loss of offsite power and with respect to the 30 day mission time assuming a single division of SSW is in service.The safety design bases of these SSCs includes ensuring the SSCs are protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).On May 4, 2018, the licensee identified vulnerabilities in the EDG building, the control building, and the SSW cooling tower where tornado-born missiles could potential render safety-related equipment contained in these buildings inoperable. Potentially affected equipment included all three EDGs, Division II DC electrical power distribution subsystem, residual heat removal (RHR) pumps B and C, SSW pumps A, B, C, and D, Division I standby cooling tower fans, and multiple Division I SSW motor operated valves. These vulnerabilities were identified as part of the licensees review of Regulatory Information Summary 2015-06, Tornado Missile Protection. These issues were entered into the corrective action program as Condition Reports CR-RBS-2018-02687, 02768, and 02775.Corrective Actions: As a result of these issues, the licensee declared all three EDGs, the Division II DC electrical power distribution subsystem, RHR pumps B and C, SSW pumps A, B, C, and D, Division I standby cooling tower fans, and multiple Division I SSW motor operated valves inoperable, complied with the applicable technical specification action statements, initiated Condition Reports CR-RBS-2018-02687, 02768, and 02775, invoked the EGM discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The licensee instituted compensatory measures intended to reduce the likelihood of tornado missile effects. These included verifying that guidance was in place for severe weather procedures, abnormal and emergency operating procedures, and FLEXsupport guidelines, verifying that training on these procedures was current, and verifying that a heightened level of awareness of the vulnerability was established.Corrective Action Reference(s) : CR-RBS-2018-02687, CR-RBS-2018-02768, and CR-RBS-2018-02775Enforcement:Violations: Technical Specification 3.8.1 requires, in part, that three diesel generators shall be operable in Modes 1, 2, and 3. Technical Specification 3.8.1.H requires entry into LimitingCondition for Operation 3.0.3 when three or more required AC sources are inoperable. Limiting Condition for Operation 3.0.3 requires that action shall be initiated within one hour to place the unit in Mode 2 within 7 hours, in Mode 3 within 13 hours, and in Mode 4 within 37 hours.Contrary to the above, prior to May 4, 2018, three diesel generators were not operable, and action was not initiated to place the unit in Mode 2 within 7 hours, in Mode 3 within 13 hours,and in Mode 4 within 37 hours. Specifically, the EDG building was not designed to withstand the effects of natural phenomena, such as tornadoes. The licensee initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02687. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Technical Specification 3.8.9 requires, in part, that the Division II AC and AC vital bus electrical power distribution subsystems shall be operable in Modes 1, 2, and 3. Technical Specification 3.8.9.D requires the station to take action to place the unit in Mode 3 within 12 hours when one or more AC or AC vital bus electrical power distribution subsystems have been inoperable for more than 8 hours. Contrary to the above, prior to May 4, 2018, the Division II AC and AC vital bus electrical power distribution subsystems were not operable for more than 8 hours, and action was not initiated to place the unit in Mode 3 within 12 hours. Specifically, the control building was not designed to withstand the effects of natural phenomena, such as tornadoes. The licensee initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02768. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Technical Specification 3.5.1 requires, in part, that each emergency core cooling system (ECCS) injection subsystem shall be operable in Modes 1, 2, and 3. Technical Specification 3.5.1.D requires the station to take action to place the unit in Mode 3 within 12 hours when two ECCS injection subsystems have been inoperable for more than 72 hours. Contrary to the above, prior to May 4, 2018, two required ECCS injection subsystems that included RHR pumps B and C were inoperable for more than 72 hours, and action was not initiated to place the unit in Mode 3 within 12 hours. Specifically, the control building was not designed to withstand the effects of natural phenomena, such as tornadoes.The licensee initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02768. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Technical Specification 3.7.1 requires, in part, that two SSW subsystems shall be operable in Modes 1, 2, and 3. Technical Specification 3.7.1. H requires the station to take action to place the unit in Mode 3 within 12 hours when both pumps associated with one SSW subsystem have been inoperable for more than 72 hours. Contrary to the above, prior to May 4, 2018, SSW pumps P2B and P2D, associated with SSWsubsystem B, were inoperable for more than 72 hours, and action was not initiated to place the unit in Mode 3 within 12 hours. Specifically, the SSW cooling tower was not designed to withstand the effects of natural phenomena, such as tornadoes. The licensee initiated a condition report, invoked the enforcement discretion guidance, implemented initial compensatory measures, and returned the SSCs to an operable- degraded/non-conforming status. The inspectors verified through inspection sampling that the EGM 15-002 criteria were met and that the issue was documented in Condition Report CR-RBS-2018-02775. Therefore, EGM 15-002 enforcement discretion was applied to the required shutdown actions associated with this technical specification.Severity/Significance: Not ApplicableBasis for Discretion: The NRC exercised enforcement discretion in accordance with EGM 15-00, Revision 1, because the licensee implemented initial compensatory measures in accordance with the EGM.
05000458/FIN-2018406-01	Security	2018Q2
05000458/FIN-2018406-02	Security	2018Q2
05000458/FIN-2018406-03	Licensee-Identified Violation	2018Q2
05000458/FIN-2018001-02	Installation of an Incorrectly Specified Relay Causes Plant Transient and Reactor Scram	2018Q1	The inspectors reviewed two examples of a self-revealed finding for the licensees installation of an incorrectly specified relay in 1) the control circuitry for the feedwater level control systemand 2) the turbine generator voltage regulator circuitry. In each instance, the incorrectly specifiedrelay failed in service, causing a plant transient and automatic reactor scram
05000458/FIN-2018001-01	Failure to Implement Procedure for Storage of Material in the Pools	2018Q1	The inspectors identified a non-cited violation of Technical Specification 5.4.1.a for the licensees failure to implement written procedures for activities referenced in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensee failed to implement radioactive material control Procedure ADM-0071, Fuel Pools Material Control, Revision 8, for the storage and movement of spent Tri-Nuke filters
05000458/FIN-2017003-02	Manual Reactor Scram Initiated in Response to Increase in Steam Pressure during Steam Leak Troubleshooting	2017Q3	The inspectors reviewed a self-revealed finding for the licensees failure to properly complete steps of an approved procedure during the installation of a modification to the turbine electro-hydraulic control system. Specifically, the licensee failed to properly install a tee connection in a steam supply line to turbine pressure transmitters in the system, creating conditions for an eventual steam leak that led to a reactor scram. Corrective actions included properly installing the tee connection and writing specific procedural guidance on compression fitting inspection, installation, remake, and repair (CR-RBS-2017-02405).The failure to properly complete steps of an approved procedure during the installation of a modification to the turbine electro-hydraulic control system 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 Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to properly install the tee connection caused a steam leak that led to a reactor scram. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The finding had a cross-cutting aspect in the area of human performance, work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority (H.5).
05000458/FIN-2017003-01	Failure to Account for Delayed Closure of Isolation Valves in the Ultimate Heat Sink Inventory Analysis	2017Q3	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in Section 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. Specifically, prior to September 28, 2017, the licensees current calculation for assuring adequate ultimate heat sink inventory did not support the acceptability of the timing of a critical operator action in the abnormal operating procedure for the loss of standby service water. The potential safety consequence is that sufficient ultimate heat sink inventory might not be available to safely shut down the plant and maintain it in a cold shutdown condition for a 30-day period with no external makeup water source available. In response to this finding, the licensee performed an initial analysis and determined that the ultimate heat sink had sufficient inventory to account for the losses associated with the delayed closure of the normal service water return isolation valves and that the losses would likely be less than those previously calculated. This finding was entered into the licensee's corrective action program as Condition Report CR-RBS-2017-06998.The inspector determined that the failure to account for delayed closure of isolation valves in the ultimate heat sink inventory analysis 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 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 performance deficiency resulted in a condition where the current analysis to determine the acceptability of the ultimate heat sink with respect to the 30-day inventory requirement needed to be re-performed to assure that accident analysis requirements were met. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with design margins because the failure to account for delayed closure of isolation valves in the 30-day ultimate heat sink inventory analysis resulted in a significant reduction in the available margin (H.6).
05000458/FIN-2017404-01	Licensee-Identified Violation	2017Q3
05000458/FIN-2017009-01	Failure to Obtain Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection Point	2017Q2	Green. The NRC identified a Severity Level IV violation for the licensees failure to restore compliance for a non-cited violation (NCV) associated with failure to obtain NRC approval prior to making a change to the reactor core isolation cooling injection point. Specifically, as of April 28, 2017, the licensee had not restored compliance with a violation the NRC identified on October 8, 2015. This violation described a previously made change to the facility without prior NRC approval in violation of 10 CFR 50.59, Changes, Tests, and Experiments. The team determined that the licensees failure to restore compliance within a reasonable amount of time was a performance deficiency. Title 10 CFR 50, Appendix B, Criterion XVI, requires in part that, measures shall be established to assure that conditions 3 adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2017-03505. The finding was more than minor because it is associated with the initiating events aspect of the reactor safety cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The finding is of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a human performance cross-cutting aspect associated with procedural adherence because individuals failed to follow the procedures delineated by the corrective action program (H.8). Originally, the licensee met the criteria for dispositioning the issue (50.59) as a NCV. However, based upon the fact that the condition report, which documented the NCV, was closed without restoring compliance, the licensee no longer met the criteria for a NCV and therefore, this violation is being cited in a notice of violation
05000458/FIN-2017007-02	Failure to Perform an Adequate Operability Determination for a Condition Identified During an NRC Walkdow	2017Q2	Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, between June 15, 2017, and June 28, 2017, the licensee failed to address the operability of a terminal block installed within an unsealed junction box. In response to this issue the licensee performed an operability determination to ensure that the terminal block would perform its design function in this condition. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2017-05084. The team determined that the failure to perform an adequate operability determination was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure operability of valve E51-AOVF054 and its associated circuits would impact the operability of the reactor core isolation cooling system. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This 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 failed to perform an adequate operability determination for an identified condition (P.3)
05000458/FIN-2017002-02	Single Component Failure Leads to Loss of Both Divisions of Control Building Air Conditioning	2017Q2	The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly translate the design basis into plant specifications. Specifically, the licensee implemented a breaker design in the control building air conditioning system that allowed a single failure of one train of the system to render the other train inoperable, contrary to the design basis. The licensee entered this condition into their corrective action program as Condition Report CR-RBS-2017-01740. The licensee restored compliance by implementing modifications to the affected breakers designed to eliminate the single failure vulnerability.The failure to correctly translate the design basis into plant specifications 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 adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to implement an appropriate design in the main control room and standby switchgear room air conditioning subsystems adversely affected the availability, reliability, and capability of safety-related components that rely on those subsystems for cooling. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The finding required a detailed risk evaluation because it involved a loss of system and/or function. A Region IV senior reactor analyst performed a detailed risk evaluation for the issue and determined the issue to be of very low safety significance (Green). No cross-cutting aspect was assigned because the finding did not reflect current performance.
05000458/FIN-2017007-01	Failure to Evaluate the Extent of Condition for a Degraded 4.16KV Magne Blast Safety-Related Circuit Breaker	2017Q2	Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, on October 28, 2014, the licensee failed to perform extent of condition on other safety-related 4.16KV Magne Blast circuit breakers due to the failure of 4.16KV Magne Blast circuit breaker ACB03 on bus E22-S003, with damaged brush and misaligned brush holder of the circuit breaker charging motor, in accordance with Procedure EN-OP-104, Operability Determination Process. Failure to perform this evaluation could adversely impact safety-related circuit breakers. In response to this issue the licensee reviewed their breaker performance records to assure that no additional failures had occurred and revised the procedure to assure that extent of condition is addressed. This finding was entered into the licensee's corrective action program as Condition Report CR-RBS-2017-05078. The team determined the failure to evaluate the impact of a damaged brush and misaligned brush holder of the charging motor of a safety-related 4.16KV Magne Blast breaker was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it related to the equipment 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. Specifically, failure to perform an extent of condition on other safety-related 4.16KV Magne Blast circuit breakers due to the failure of E22-S003 safety-related circuit breaker ACB03, 4.16KV Magne Blast breaker with damaged brush and misaligned brush holder could adversely affect the ability of these breakers to perform their safety functions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance 3 associated with conservative bias because the licensee failed to ensure that individuals used decision-making practices that emphasized prudent choices (H.1)
05000458/FIN-2017405-01	Security	2017Q2
05000458/FIN-2017002-01	Failure to Maintain Operability of the Division I Control Room Fresh Air System While Changing Reactor Modes	2017Q2	The inspectors reviewed multiple examples of a self-revealing, non-cited violation of Technical Specification 3.0.4, Limiting Condition for Operation Applicability, for the licensees failure to restore safety-related equipment to operable status prior to changing modes. Specifically, the licensee failed to restore Division I of the Control Room Fresh Air system to operable status prior to entering Mode 2 on March 8, 2017, and again on March 11, 2017. The licensee entered this condition into their corrective action program as Condition Report CR-RBS-2017-03082. The licensee restored compliance by properly positioning damper HVC-DMP4A and restoring the Division I Control Room Fresh Air system to operable.The failure to restore Division I of the Control Room Fresh Air system to operable status prior to entering Mode 2 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the structures, systems, and components (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the incorrect positioning of damper HVC-DMP4A resulted in inadequate air flow through Division I of the Control Room Fresh Air system and rendered it inoperable. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3 Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. This finding had a cross-cutting aspect in the area of human performance, challenge the unknown, because individuals did not stop when faced with uncertain conditions. Specifically, workers positioned damper HVC-DMP4A without work instructions or specified torque values (H.11).
05000458/FIN-2017001-02	Failure to Properly Pre - Plan and Perform Maintenance on the Control Building Chilled Water System	2017Q1	Green . The inspectors identified a non- cited violation of Technical Specification 5.4, Procedures, for the licensees failure to properly pre-plan and perform maintenance on safety -related components in accordance with documented instructions appropriate to the circumstances. Specifically, the licensee used work order instructions that did not contain sufficient detail for the reassembly of SWP -PVY32C, a safety -related valve in the control building ventilation system . As a result, SWP -PVY32C developed a refrigerant leak, and on November 17, 2015 , the valve failed. This in turn caused the control building ventilation system to fail , and the high pressure core spray system was consequently declared inoperable. The licensee entered this condition into their corrective action program as Condition Report CR- RBS -2017- 02364. Corrective actions included incorporating the torque values into the model work order instructions for future maintenance and reassembly . The failure to properly pre-plan and perform maintenance on safety -related components in accordance with documented instructions was a performance deficiency. The performance deficiency was more than minor , and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the control building ventilation system failed, it impact ed the operability of the high pressure core spray system. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect the design or qualification of a mitigating structure, system, or component (and the structure, system, or component maintained its operability), it did not represent a loss of safety function, it did not represent an actual loss of function of at least a single train for greater than its technical specification outage time, and it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees Maintenance Rule program for greater than 24 hours. This finding has a cross- cutting aspect in the area of human performance, challenge the unknown, because individuals did not stop when faced with uncertain conditions. Specifically, workers proceeded with assembling the valve when the torque values or torqueing sequence were not specified (H.11)
05000458/FIN-2017001-01	Failure to Follow Station Guidance on Control of Scaffolding	2017Q1	Green . The inspectors identified a non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to follow station maintenance procedures related to the control of scaffolding in the reactor building. Specifically, the licensee installed scaffolding less than two inches from safety -related containment unit cooler HVR -UC1 B without completing an engineering evaluation. The licensee entered this issue into their corrective action program as Condition Report CR- RBS -2016- 07963 . Corrective actions included removing the scaffolding. The licensees installation of scaffolding within two inches of a safety -related containment unit cooler , without completing an engineering evaluation, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, containment unit cooler HVR- UC1B was rendered inoperable by the incorrectly installed scaffolding and remained inoperable until the scaffolding was removed. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance ( Green) because the finding did not represent an actual loss of function of one or more trains of safety-related equipment for greater than its technical specification allowed outage time. This finding has a cross -cutting aspect in the area of human performance, avoid complacency , because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risks, even while expecting successful outcomes . Specifically, the station failed to implement appropriate error reduction tools when it did not perform and document Procedure EN -MA -133, Control of Scaffolding, Attachments 9.5 and 9.6 , which could have prevented the scaffolding construction error (H. 12).
05000458/FIN-2017001-03	Failure to Enter Applicable Technical Specification Action Statements When Control Building Chillers Were O ut of Service	2017Q1	Green . The inspectors identified a non- cited violation of Technical Specifications 3.8.4, DC Sources - Operating, 3.8.7, Inverters Operating, and 3.8.9, Distribution Systems Operating, for the licensees failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours. Specifically, electrical power systems required by the above limiting condition s for operation were inoperable due to the associated division of the control building chilled water system chillers being out of service and therefore unavailable to provide the technical specification support function of attendant cooling that is needed for the associated electrical systems to perform their specified safet y functions. As a result of this deficiency, the station reduced the reliability and availability of systems cooled by control building chilled water system chillers by allowing configurations that did not conform to the single failure criterion. The lic ensee entered this issue into their corrective action program as Condition Report CR- RBS -2015 -02525 . Corrective actions included entering the appropriate limiting conditions for operation of affected safety -related systems when the non -safety related support system were non -functional. 4 The failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours wa s a performance deficiency . Specifically, electrical power systems required by the above limiting condition s for operation were inoperable due to the associated division of the control building chilled water system chillers being out of service and therefore unavailable to provide the technical specification support function of attendant cooling that is needed for the associated electrical systems to perform their specified safety functions. The performance deficiency was more than minor, and therefore a finding, because it wa s associated with the configuration control attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that res pond to initiating events to prevent undesirable consequences. As a result of this deficiency, the station reduced the reliability and availability of systems cooled by control building chilled water system chillers by allowing configurations that did not conform to the single failure criterion. The inspectors performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to require a detailed risk evaluation because it represented a loss of system and/or function. A senior reactor analyst performed a det ailed risk evaluation for a previously identified performance deficiency associated with the licensees failure to account for a loss of all control building chilled water system cooling scenario, either quantitatively or qualitatively, which resulted in uncompensated impairment to all systems associated with the main control room (Agencywide Documents Access and Management System (ADAMS) Accession N o. ML16132A144). This previously performed detailed risk evaluation bounds the risk associated with the finding dispositioned in this write- up: the failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours. Therefore, the finding was determined to be of very low safety significance (Green). No cross -cutting aspect was assigned as the performance deficiency is not indicative of current licensee performance
05000458/FIN-2016404-04	Security	2016Q4
05000458/FIN-2016404-06	Licensee-Identified Violation	2016Q4
05000458/FIN-2016404-02	Security	2016Q4
05000458/FIN-2016404-05	Security	2016Q4
05000458/FIN-2016404-01	Security	2016Q4
05000458/FIN-2016404-03	Security	2016Q4
05000458/FIN-2016002-01	Failure to Follow Station Guidance on Use of Temporary Power Cables and Control of Transient Combustibles	2016Q2	The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to follow station maintenance procedures related to the use of temporary power cables and storage of transient combustible materials in the auxiliary building. Specifically, the licensee installed energized networking equipment and an associated power cable within one foot of a safety-related cable tray. The station did not initially correct the problem, but later resolved the deficiencies by removing the networking equipment and power cable. The failure to initially correct the issue is documented as a violation in Section 4OA2 of this report. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2016-02398. The licensees installation of energized networking equipment and an associated power cable within one foot of a safety-related cable tray was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a fire resulting from this energized equipment would impact the availability, reliability, and capability of the low pressure core spray system, residual heat removal system, component cooling primary system, and reactor core isolation cooling system. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings. Since the finding involved a failure to adequately implement fire prevention and administrative controls for transient combustibles, the inspectors dispositioned the finding using NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. In accordance with Manual Chapter 0609, Appendix F, Question 1.3.1.A, the inspectors determined that the finding was of very low safety significance (Green) because the reactor would be able to reach and maintain safe shutdown since the safe shutdown path was deemed independent of fire damage state scenarios for the given fire ignition source. The finding had a cross-cutting aspect in the area of human performance, work management, because the licensees work management processes failed to plan, control, and execute the work activity that included installation of temporary equipment such that impacts on nuclear safety were properly evaluated and addressed (H.5).
05000458/FIN-2016007-01	Inadequate Loop Flow Test Procedure	2016Q2	The team identified a non-cited violation of License Condition 2.C.(10) for the failure to implement and maintain in effect all provisions of their approved fire protection program. Specifically, the licensees fire protection program surveillance testing procedure for the fire main yard loop did not include appropriate guidance to properly flow test all portions of the underground fire main yard loop to buildings that contained fire safe shutdown equipment. The licensee entered this deficiency into their corrective action program as Condition Report CR-RBS-2016-03212 and initiated actions to correct the procedure and perform the flow testing. The failure to ensure that fire protection program Surveillance Test Procedure STP-251-3700, Fire System Yard Water Suppression Loop Flow Test, Revision 10, included requirements to functionally test all individual underground firewater flow paths to structures that contained fire safe shutdown components was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012. The team determined that an Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, review was required because the finding affected the fire water supply system. Using Inspection Manual Chapter 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013, the finding was screened as a Green finding of very low safety significance in accordance with Task 1.4.7, Fire Water Supply, Question A. Since the subject fire main yard loops had not been flow tested since initial testing, and nothing caused the licensee to reevaluate the testing procedure, the team determined that this failure did not reflect current performance, and no cross-cutting aspect was assigned.
05000458/FIN-2016007-02	Failure to Isolate Control Circuits for Safe Shutdown Equipment From the Effects of a Control Room Fire	2016Q2	The team identified a non-cited violation of License Condition 2.C.(10) for the failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the team identified two examples where the licensee failed to isolate control circuits for safe shutdown equipment to ensure independence from the effects of a fire in the control room. As immediate compensatory measures the licensee performed visual inspections of the affected cabinets for unacceptable fire hazards and issued Standing Order 323 to reinforce the need for operators to identify and prevent fire hazards while in the control room. The licensee entered this issue into their corrective action program as Condition Reports CR-RBS-2016-02953 and CR-RBS-2016-03264. The failure to isolate control circuits for safe shutdown equipment from the effects of a control room fire was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because it affected the ability to reach and maintain safe shutdown conditions in case of a fire. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a postulated control room fire that led to control room evacuation and determined the issue was of very low safety significance (Green). This finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago.
05000458/FIN-2016002-02	Failure to Conduct Common Cause Failure Evaluation in Response to Inoperable Emergency Diesel Generator	2016Q2	The inspectors identified a non-cited violation of Technical Specification 3.8.1, AC Sources Operating, for the licensees failure to take required actions for an inoperable emergency diesel generator. Specifically, after classifying the Division I emergency diesel generator as inoperable on the basis of a nonconforming condition discovered during an extended maintenance outage, and after failing to either verify that the Division II emergency diesel generator was not inoperable due to common cause failure within 24 hours or conduct a surveillance run on the Division II emergency diesel generator within 24 hours, the licensee failed to enter Mode 3 within 12 hours, as required by Actions C.3.1, C.3.2, and G.1 of Technical Specification 3.8.1, respectively. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2016-03978. Corrective actions included the scheduling of training to ensure that operations personnel fully understand the technical specification requirements for common cause evaluation as they relate to adverse conditions on emergency diesel generators. The failure to take required actions for an inoperable emergency diesel generator was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment reliability 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, the failure to follow technical specification requirements to ensure the availability, reliability, and capability of the operable emergency diesel generator directly affected the cornerstone objective. Using NRC Inspection Manual Chapter 0609, Significance Determination Process, Appendix A, Exhibit 2 -- Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding did not represent an actual loss of function of the Division II emergency diesel generator. The finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee failed to use a consistent, systematic approach to make decisions. Specifically, the licensee failed to review the required actions of the applicable technical specification so as to ensure that all of those actions would be properly carried out (H.13).
05000458/FIN-2016007-03	Failure to Demonstrate that Appendix R Emergency Lights Satisfied their Maintenance Rule Performance Criteria	2016Q2	The team identified a finding for the failure to provide an adequate monitoring and testing program to demonstrate that the required Appendix R emergency lights satisfied the licensees maintenance rule performance criteria. Specifically, the failure to provide an adequate monitoring and testing program could result in a large number of Appendix R emergency lights failing to last the required 8 hours without being detected. The team determined that, because the licensee had changed their program to a biennial replacement frequency for the 8-hour batteries, reasonable assurance existed that the lights would function long enough for operators to perform the time critical manual actions directed by their fire protection program. The licensee entered this finding into their corrective action program as Condition Report CR-RBS-2016-03177. The failure to establish an adequate monitoring and testing program to demonstrate that the required Appendix R emergency lights would satisfy the licensees maintenance rule performance criteria was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the failure to provide an adequate monitoring and testing program could result in a large number of Appendix R emergency lights failing to function for the required 8 hours without being detected through licensee monitoring and testing. The team determined this finding affected the Mitigating Systems Cornerstone. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated February 28, 2005, because it affected the ability to reach and maintain safe shutdown conditions in case of a fire. The team assigned the finding to the post-fire safe shutdown category since it impacted the remote shutdown and control room abandonment element. The team assigned the finding a low degradation rating since the ability to reach and maintain safe shutdown conditions in the event of a control room fire would be minimally impacted by the potential failure of the emergency lights to function for 8-hours. Because this finding had a low degradation rating, it screened as having very low safety significance (Green) in Task 1.3.1. The finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago. Specifically, the licensee began performing the 8-hour discharge test on a small sample of the batteries more than three years ago.
05000458/FIN-2016002-03	Failure to Identify and Correct Improperly Stowed Transient Combustibles	2016Q2	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, after writing a condition report identifying energized networking equipment and an associated power cable that had been installed within one foot of a safety-related cable tray, the licensee closed the condition report without removing the networking equipment and power cable. The licensee entered this issue into their corrective action program as Condition Reports CR-RBS-2016-02398 and CR-RBS-2016-03152. Corrective actions included removing the networking equipment and power cable and conducting a performance management review of the actions involved with correcting the condition and closing the condition report. The licensees failure to promptly identify and correct a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected 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. Specifically, the failure to correct a known deficient condition resulted in an extended period of vulnerability to a fire that could result from improperly installed energized equipment and challenge the availability, reliability, and capability of the low pressure core spray system, residual heat removal system, component cooling primary system, and reactor core isolation cooling system. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings. Since the finding involved a failure to adequately implement fire prevention and administrative controls for transient combustibles, the inspectors dispositioned the finding using NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. In accordance with Manual Chapter 0609, Appendix F, Question 1.3.1.A, the inspectors determined that the finding was of very low safety significance (Green) because the reactor would be able to reach and maintain safe shutdown since the safe shutdown path was deemed independent of fire damage state scenarios for the given fire ignition source. The finding had a cross-cutting aspect in the area of human performance, teamwork, because the licensee failed to properly communicate expectations to individuals performing work during the course of implementing corrective actions (H.4).
05000458/FIN-2016009-01	Failure to Follow Procedure While Installing Jumpers for Shutdown Cooling	2016Q1	The team reviewed a self-revealing, non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to correctly implement Procedure SOP-0031, Residual Heat Removal System, Revision 326. SOP-0031, Attachment 5, Step 5.4.1, required that a retractable sheathed banana jumper be used when bypassing the 135-psi SDC isolation. Instead, the licensee used a standard banana jumper, which resulted in a short circuit and inadvertent closure of Valves E12MOV-F008, Shutdown Cooling Suction Valve, and E12MOV-F053A, Shutdown Cooling Injection Valve. This caused a loss of decay heat removal. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2016-0210. Corrective actions included revising Procedure SOP-0031 to include actions to de-energize the applicable valves while bypassing the 135-psi shutdown cooling isolation. The failure to use the correct jumpers as specified in Procedure SOP-0031 was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the shorting of contacts resulting from the use of incorrect jumpers caused a loss of shutdown cooling and decay heat removal. The team evaluated the finding using NRC Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Screening and Characterization of Findings. When applying Exhibit 2 - Initiating Events Screening Questions, the team determined the loss of residual heat removal event did not occur when the refuel cavity was flooded, and therefore it required a risk evaluation using the Appendix G, Attachment 3, Phase 2 Significance Determination Process Template for Boiling Water Reactors during Shutdown. The analyst determined that a modified but still conservative Phase 2 quantitative estimate in combination with qualitative and deterministic insights led to a final conclusion that the finding was of very low safety significance (Green). The finding has a field presence cross-cutting aspect within the human performance area because the licensee failed to promptly correct deviations from standards and expectations. Specifically, the licensee failed to correct deviations from standards and expectations during the performance of the pre-job brief and ensure proper communication and oversight is maintained in the control room during risk significant evolutions (H.2).
05000458/FIN-2016201-01	Security	2016Q1
05000458/FIN-2015010-03	Failure to Identify and Correct Circuit Breaker Failure Mechanism	2016Q1	The team reviewed 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 related to Masterpact circuit breakers. Specifically, the licensee did not promptly identify and correct a Masterpact breaker failure mechanism, even though related industry operating experience was available. The licensee determined the failure mechanism caused nine breaker failures since 2007, and may have contributed to an additional six failures where the cause was not conclusively determined. In response to the NRCs conclusions, the licensee initiated Condition Report CR-RBS- 2015-03951. Further, the licensee modified all vulnerable Masterpact circuit breakers to remove this failure mechanism. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees untimely corrective action contributed to additional failures of Masterpact circuit breakers and decreased the reliability of Masterpact circuit breakers to respond during design basis events. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has an operating experience cross-cutting aspect within the problem identification and resolution area because the licensee failed to systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner.
05000458/FIN-2016009-03	Failure to Implement Corrective Actions to Prevent the Recurrence of a Reactor Scram Due to Grid Disturbances	2016Q1	The team reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish measures to assure that corrective action is taken to preclude repetition of a significant condition adverse to quality. Specifically, following a November 27, 2015, reactor scram, the licensee failed to implement corrective actions associated with the alternate power lineup of the reactor protection system buses to preclude repetition of a significant condition adverse to quality during the January 9, 2016, reactor scram. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2016-0180. Corrective actions included supplying reactor protection system bus A from the normal power source on January 12, 2016. The failure to assure corrective actions are promptly taken for a significant condition adverse to quality to preclude repetition of a reactor scram associated with both buses being affected by a switchyard voltage transient was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failed to implement corrective actions to address grid instabilities following the November 27, 2015, reactor scram to preclude the January 9, 2016, reactor scram. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, the team determined that this finding is of very low safety significance (Green) because it did not involve the loss of mitigation equipment or a support system. This finding has an evaluation cross-cutting aspect within the problem identification and resolution area because the licensee failed to thoroughly evaluate the cause of the November 27, 2015, reactor scram and ensure that the resolution addresses causes and extent of conditions commensurate with their safety significance (P.2).
05000458/FIN-2015010-02	Failure to Adequately Assess Risk During Chiller Unavailability	2016Q1	The NRC identified an apparent violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4) with preliminary white significance. Prior to March 30, 2015, before performing maintenance activities, the licensee failed to adequately assess the increase in risk that may result from proposed maintenance activities. Specifically, the risk assessment performed by the licensee for plant maintenance failed to account for certain safety significant structures, systems, and components that were concurrently out of service. On multiple occasions, the licensee failed to adequately assess the risk of operating the control building chilled water system (HVK) chillers in various single failure vulnerable configurations. As a result of this deficiency, the station reduced the reliability and availability of systems contained in the main control room and failed to account for the significant, uncompensated impairment of the safety functions of the associated systems. In response to the NRCs conclusions, the licensee initiated Condition Report CR-RBS-2016-00095. The licensee also completed engineering analyses to evaluate alternate cooling methods, including cross-connecting service water and the HVK chiller systems, in order to provide cooling to spaces housing electrical components. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to account for a loss of all HVK cooling scenario, either quantitatively or qualitatively, resulted in uncompensated impairment to all systems associated within the main control room. A loss of cooling to the control room could lead to multiple systems exceeding their equipment qualification temperatures and impact control room habitability. The finding was evaluated using Inspection Manual Chapter (IMC) 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Using Inspection Manual Chapter 0609, Appendix K, the finding was determined to require additional NRC management review using risk insights where possible because the quantitative probabilistic risk assessment (PRA) tools are not well suited to analyze failures from control room heat-up events. Thus, the analyst evaluated the safety significance posed by the heat-up of components cooled by the HVK chillers using Appendix K, Flowchart 1, Assessment of Risk Deficit, to the extent practical, with additional risk insights by internal NRC management review in accordance with Inspection Manual Chapter 0612, Power Reactor Inspection Reports. The significance of the finding was preliminarily determined to be White. The team determined the most significant contributing cause of the licensee failing to adequately assess the increase in risk from proposed maintenance activities was inadequate procedural guidance in Procedure ADM-0096, Risk Management Program Implementation and On-line Maintenance Risk Assessment, Revision 316. This finding has a resources cross-cutting aspect within the human performance area because leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety (H.1).
05000458/FIN-2016009-04	Failure to Adequately Assess Risk During Motor Generator Set Unavailability	2016Q1	The team identified a non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to adequately assess the increase in risk that may result from proposed maintenance activities. Specifically, the team identified that since 2012, the licensee failed to adequately assess the risk of simultaneously powering both reactor protection system buses from the alternate power sources, which resulted in an increased risk of a reactor scram due to grid instabilities. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2016-3176. Corrective actions included revising Procedure SOP-0079, Reactor Protection System, to include precautions to address the increased risk associated with supplying both reactor protection system buses from the alternate power source. The team determined that the licensees failure to adequately assess the increase in risk associated with simultaneously powering both reactor protection system buses from the alternate power sources was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Initiating Events Cornerstone and 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. Specifically, the performance deficiency resulted in an increased risk of a reactor scram due to grid instabilities. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, a detailed risk evaluation was required since the finding resulted in a reactor scram and main steam isolation valve closure. The finding was evaluated using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, Assessment of Risk Deficit, dated May 19, 2005, to assess the significance of the finding. A senior reactor analyst estimated the incremental core damage probability deficit to be 2.0E-7 and the incremental large early release probability deficit to be 4.0E-8. Since this incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability deficit was less than 1E-7, the analyst used Flowchart 1 to determine the finding was of very low safety significance (Green). This finding has a conservative bias cross-cutting aspect within human performance area because the licensee determined that powering both reactor protection system buses from the alternate source instead of the motor generator sets was safe even though the motor generator sets are the preferred source and provide protection against grid perturbations (H.14).
05000458/FIN-2015010-04	Failure to Accomplish an Operability Determination In Accordance With Procedures	2016Q1	The team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an operability determination in accordance with procedure EN-OP-104, Operability Determination Process, Revision 8. Specifically, the licensee referenced non-conservative data, contrary to steps 5.5 and 5.11 of procedure EN-OP-104, when assessing the reduced reliability of Masterpact circuit breakers as a degraded or nonconforming condition. The licensee restored compliance by completing a design modification to eliminate the failure mode and initiated Condition Report CR-RBS-2015-03952. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reliability of components powered by Masterpact circuit breakers was reduced and, by justifying operability using non-conservative data, the licensee did not recognize the actual unreliability. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has a conservative bias cross-cutting aspect within the human performance area because the licensee failed to use decision-making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee did not consider that the failure mechanism only occurs on a close command. Instead, the licensee included opening commands when summing the total demands and this resulted in a non-conservative failure rate (H.14).
05000458/FIN-2015010-05	Failure to Identify and Correct an Adverse Condition in a Timely Manner	2016Q1	The team identified a finding for the licensees failure to identify and correct an adverse condition in a timely manner as required by plant procedures. Specifically, the licensee did not recognize degrading trends associated with incorrect racking of Magne Blast circuit breakers and failures of the Magne Blast circuit breaker for the Reactor Feed Water Pump Motor 1B in a timely manner. For both cases, the licensee failed to initiate corrective action in a timely manner as required by procedure EN-LI-102, Corrective Action Program. In response to the NRCs conclusions, the licensee updated circuit breaker procedures, replaced the Magne Blast circuit breaker for the Reactor Feed Water Pump Motor 1B, and initiated Condition Reports CR-RBS-2015-04259 and CR-RBS-2015-03437. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees untimely corrective action contributed to the unreliability of the Magne Blast circuit breaker for Reactor Feed Water Pump Motor 1B and increased the potential for spurious trips of other Magne Blast circuit breakers during design basis events due to improper racking. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has an avoid complacency cross-cutting aspect within the human performance area because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee tolerated the adverse trends, did not plan for further degradation, and the latent conditions ultimately resulted in several Magne Blast circuit breaker failures in December 2014 before the trend was recognized (H.12).
05000458/FIN-2015010-01	Technical Specification Allowed Outage Time During Loss of Non-Technical Specification Supported Systems	2016Q1	The team identified an unresolved item related to the licensees treatment of the control building chilled water system (HVK) chillers as a non-technical specification support system for other technical specification systems. The team noted that when an entire division of HVK chillers is out of service, such as chillers 1A and 1C for division I, the licensee would only enter the Technical Specification (TS) 3.7.3, Control Room Air Conditioning (AC) System, action statement for the condition of one control room AC subsystem being inoperable (condition A). The licensee did not enter TS action statements associated with inoperability of other components cooled by HVK chillers, such as the AC switchgear, DC switchgear, and vital inverters. The licensee, instead, has incorporated a safety evaluation for the Perry Plant (ML020950074), dated April 5, 2002, into the bases for TS 3.0.6 and applied that document as guidance: ...no TS limits the duration of the non-TS support subsystem outage, even though the single-failure design requirement of the supported TS systems is not met. However, by assessing and managing risk in accordance with 10 CFR 50.65(a)(4), an appropriate duration for the maintenance activity can be determined. The NRC team questioned whether the Perry Plants safety evaluation could be applied generically, if the licensee improperly incorporated the safety evaluation via the 10 CFR 50.59 process, if the guidance conflicted with section 9.2.10.3 of the Updated Safety Analysis Report (USAR) for River Bend Station, and if the safety evaluation for the Perry Plant conflicted with guidance found in Generic Letter 80-30, Clarification of the Term Operable As It Applies to Single Failure Criterion For Safety Systems Required by TS. The aggregate impact of these decisions resulted in the River Bend Station placing TS systems cooled by HVK, such as the AC switchgear, DC switchgear, and vital inverters, in a single-failure vulnerable configuration for durations exceeding the allowed outage time specified in the TS. Pending further evaluation of the above issue by NRC Headquarters staff via a Technical Specification interpretation request (ML15231A111) and subsequent review by NRC inspectors, this issue will be tracked as URI 05000458/2015010-01, Technical Specification Allowed Outage Time During Loss of Non-Technical Specification Supported Systems. Further discussion of performance deficiencies associated with the HVK chiller system is included in Section 2.6.a of this report.

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