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	Site	Quarter	Title	Description
05000458/FIN-2018301-01	River Bend	2018Q3	Inadequate Procedure for Shutdown Operations Protection Plan	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.
05000416/FIN-2018002-07	Grand Gulf	2018Q2	Loss of Shutdown Cooling	A self-revealed,Green non-cited violation of Technical Specification 5.4, Procedures,for the licensees failure to follow written procedures was identified when the residual heat removal (RHR) system automatically isolated due to an inadvertent emergency core cooling system (ECCS) actuation. While the plant was shut down with the RHR system in decay heat removal mode, maintenance personnel inadvertently opened an incorrect valve during a transmitter calibration activity, which caused a false low reactor pressure vessel (RPV) water level signal, an ECCS actuation, and a loss of decay heat removal for approximately 31 minutes
05000369/FIN-2018010-02	Mcguire McGuire	2018Q1	Failure to Update the FSAR With Pertinent Design Information	The inspectors identified a Green finding and associated Severity Level IV Non-cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.71(e), for the licensees failure to update the final safety analysis report (FSAR) to include the design function of manually opening the residual heat removal (ND) system shutdown cooling suction valves. Consequently, the licensee failed to consider the design capability of the valves, time impacts on dose consequence analyses, and the implication of pressure locking.
05000373/FIN-2018001-01	LaSalle	2018Q1	Post-Maintenance Testing Failed to Demonstrate Testable Check Valve Function	A self-revealed Green finding of very low safety significance and an associated Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion XI, Test Control, was documented by the inspectors for the licensees failure to perform post-maintenance testing that would demonstrate that structures, systems and components (SSCs) would perform satisfactorily in service. Specifically, following maintenance on the Unit 2 B residual heat removal (RHR) shutdown cooling (SDC) return testable check valve, 2E12F050B, and the Unit 1 A RHR SDC return testable check valve, 1E12F050A, the post maintenance test performed failed to identify that they would not open fully when in service, resulting in the valves being unable to pass full flow during SDC mode of RHR operation.
05000282/FIN-2017004-04	Prairie Island	2017Q4	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to the above, on October 17, 2017, with Unit 2 in Mode 5, Cold Shutdown, the licensee failed to accomplish procedure 2C12.2, Purification and Chemical Addition Unit 2; Revision 34. Specifically, control room operators signed off steps as completed without validating that the procedure actions were performed in the field. These procedure steps that intended to close letdown valves and open purification valves, resulted in unintended transfer of primary coolant from the RCS to the chemical and volume control system hold-up tank instead of back to the RCS. In turn, this resulted in a reduction in RCS inventorywith reactor vessel level at approximately 1 foot below the flange (reduced inventory operations). Due to operators quickly recognizing a lack of letdown flow as discussed during a pre-job brief, the purification evolution was halted and actions were taken to restore reactor vessel level.Because the inspectors answered No to questions B.2 and B.3 under Exhibit 2, Initiating Events Screening Questions of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green). Specifically, the loss of inventory event was self-limiting such that the leakage would have stopped before impacting the operating method of decay heat removal (shutdown cooling via RHR in this case). The issue was entered into the licensees CAP as CAP 501000003923. Corrective actions included an operations department human performance clock reset to share the lessons learned from the event.
05000416/FIN-2017007-07	Grand Gulf	2017Q4	Licensee-Identified Violation	The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as non- cited violations. Technical Specification 5.4.1(a) requires written procedures to be established, implemented, and maintained as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Section 4.e recommends, in part, instructions for startup of shutdown cooling and reactor vessel head spray system be prepared. Contrary to the above, from about 2004 until September 1, 2017, the 04-1-01-E12-2 instruction failed to provide instruction for placing the alternate decay heat removal system in service. Specifically, Step 4.9.2a.7(d) instructs an operator to, Manually control component cooling water temperature by throttling P44-F010A(B)(C), PSW inlet to CCW HXs. However, the purpose of that step is to throttle plant service water flow through the alternate decay heat removal system and component cooling water system to ensure both systems have plant service water flow, which is not accomplished by the instruction step. The licensee identified this procedural violation before the system was credited for availability during an inservice demonstration on September 1, 2017, and entered it in the corrective action program as Condition Report CR-GGN-2017-08643. The violation is of very low safety significance (Green) because, although the procedure did delay placing the system in service due to the procedure error, the system was capable of performing its design function, consistent with Inspection Manual Chapter 0609, Appendix G, Attachment 1, Exhibit 3 screening.
05000335/FIN-2017004-03	Saint Lucie	2017Q4	Failure to Identify and Correct a Condition Adverse to Quality	The NRC-identified a Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for failure to identify and correct a condition adverse to quality. The licensee failed to identify that their procedures lacked actions to install control power jumpers that are required to defeat the reactor coolant systems (RCS) pressure interlocks for the shutdown cooling (SDC) suction line motor operated valves (MOVs) when aligning the plant for hot leg injection (HLI) and then correct the condition. Following the identification of this procedural vulnerability, the licensee fabricated control power jumpers and revised procedure 1-GME-100.03, Installation and Removal of Temporary Power Jumpers for MOV V3481, V3652, V3432 AND V3444, to provide direction for installation of power jumpers. In addition, the licensee performed a more detailed failure modes and effects analysis to ensure that the revised procedures accounted for all possible single failures. This issue has been entered into the licensees corrective action program (CAP) as CR 2217631.The PD was more than minor because it was associated with the Design Control attribute of the Mitigating System cornerstone objective of ensuring the capability of the low pressure safety injection (LPSI) system to perform its required long term cooling safety function (HLI). The condition was evaluated by a Regional Senior Reactor Analyst and determined to have very low safety significance (Green) based on the low likelihood of a loss of coolant accident (LOCA) and low likelihood of electrical failures requiring jumpers to be installed. This issue and corrective actions were documented in the licensees CAP as Action Request (AR) 2217631. This finding was not assigned a cross-cutting aspect because the underlying cause was a legacy issue and not indicative of current performance.
05000352/FIN-2017003-02	Limerick	2017Q3	Licensee-Identified Violation	LGS Unit 1 Renewed Facility Operating License, NPF- 39, and LGS Unit 2 Renewed Facility Operating License, NPF- 85, License Condition 2.C.(3) requires , in part, that Exelon Generation Company shall implement and maintain all provisions of the approved Fire Protection Program as described in the UFSAR. LGS Unit 1 and Unit 2 UFSAR Chapter 9A requires compliance with Branch Technical Position, Chemical Engineering Branch 9.5- 1, guideline C.5.b(1), to limit fire damage so that one train of systems necessary to achieve and maintain cold shutdown conditions from either the control room or emergency control station can be repaired within 72 hours. Contrary to the above, from July 2014 to December 2016, an unanalyzed condition existed in which an abnormal ESW system alignment placed two Fire Areas in noncompliance with the FSSD analysis described in the UFSAR. Specifically, in July 2014, ESW to RHRSW flow return valve, HV -011 -015A was de- energized and tagged closed following ESW system testing. With on ly one RHRSW return path available to the A ESW loop, a postulated fire in Fire Area 12 or Fire Area 18 could cause a single spurious valve operation of either spray pond bypass valves HV -012- 031A or HV -012 -031C, when the ESW system is aligned in the spray pond winter bypass mode. This condition would result in no return flow path for the A loop of ESW, which would in turn result in loss of cooling water to EDGs aligned to the A ESW cooling loop. The affected EDGs would be inoperable until the ESW system could be realigned to provide cooling water flow. This condition coupled with a loss of offsite power assumed in FSSD analysis would result in a loss of power to SRVs needed to transition both LGS units from hot shutdown conditions to cold shutdown conditions. Following the depletion of station batteries after 4 hours, until offsite power is assumed to be restored after 72 hours, direct current power would be lost to SRVs that are necessary to reduce plant pressure low enough to place the shutdown cooling system into service and establish cold shutdown plant temperatures. The failure to have a cold shutdown repair that could be implemented within 72 hours in accordance with the FSSD analysis described in the UFSAR, was a performance deficiency. 24 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 it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green ), based on IMC 0609, Appendix F, Fire Protection Significance Determination Process , Attachment 1, Part 1: Fire Protection Significance Determination Process Phase 1 Worksheet, dated September 2013. The finding screened to Green based upon task 1.3.1 screening question A, since the inspectors determined that for conditions evaluated by Appendix F the reactors were able to reach and maintain hot shutdown. Specifically, LGS Units 1 and 2 would have been able to achieve and maintain hot shutdown during the period the unanalyzed condition existed. This would have been accomplished by using HPCI and SRVs for pressure and level control. Both units would have been capable of maintaining hot shutdown conditions with postulated fire damage until offsite power could be restored. Because this issue was of very low safety significance (Green) and Exelon entered the issue into the corrective action program as IR 3955705, this finding is being treated as a licensee identified NCV , consistent with Section 2.3.2.a of the Enforcement Policy.
05000352/FIN-2017002-02	Limerick	2017Q2	Follow -Up of Events and Notices of Enforcement Discretion	Inspection Scope On March 20, 2016, Limerick Unit 1 was performing a planned shutdown to support a refueling outage. The drywell leak inspection team identified a 0.5 gallons per minute reactor coolant system (RCS) pressure boundary leak on the shutdown cooling equalizing line. The apparent cause evaluation determined that the 34 inch A RHR shutdown cooling return check valve equalizing line developed a crack at the toe of the weld due to high cyclic fatigue induced by vibration from the reactor recirculation system. This check valve was previously replaced in 2006, and the equalizing line came pre - fabricated to the valve body. The affected section of the piping was replaced with a new socket weld with a 2x1 overlay to improve the pipe stability and minimize stresses. The Unit 1 B RHR shutdown cooling return check valve equalizing line weld was also reworked using the 2x1 weld method during the Unit 1 refueling out age in April 2016. The similar Unit 2 welds on the equalizing lines were examined and reinforced during the May 2017 refueling outage. The LER and associated evaluations and follow -up actions were reviewed for accuracy, the appropriateness of corrective actions, violations of requirements, and potential generic issues. This LER is closed. b. Findings Description. On March 20, 2016, Limerick Unit 1 was performing a planned shutdown to support a refueling outage. The drywell leak inspection team identified a 0.5 gallons per minute RCS pressure boundary leak on the shutdown cooling equalizing line. Additionally, Exelon determined that this leakage constituted a violation of the Unit 1, TS 3.4.3.2. Operational Leakage that requires the RCS leakage to be limited to no pressure boundary leakage. The condition was reported in event notification 51809 as required by 10 CFR 50.72(b)(3)(ii)(A ) because it represented a degradation of a principal safety barrier. Exelon evaluated the flaw and determined the cause of the RCS pressure boundary leakage was that the 34 inch A RHR shutdown cooling return check valve equalizing line developed a crack at the toe of the weld due to high cyclic fatigue induced by vibration from the reactor recirculation system. The inspectors reviewed the LER and Exelons apparent cause evaluation of the event. The inspectors reviewed the event information and leakage data over the previous cycle and concluded that reactor pressure boundary leakage reasonably began on an unknown date that was more than 36 hours before March 20, 2016. However, the inspectors determined that the existence of R CS pressure boundary leakage was not within Exelons ability to foresee and correct and therefore was not a performance deficiency. In particular, the RHR shutdown cooling return check valve was replaced on the recommended periodicity, and the equalizing line that developed the crack came pre- fabricated to the valve body when replaced in 2006. For information, the inspectors screened the significance of the condition using IMC 0609, Appendix A, The Significance Determination Process For Findings At -Power , and determined that the condition represented very low safety significance (Green) because it would not result in exceeding the RCS leak rate for a small LOCA and would not have likely affected other systems used to mitigate a LOCA. 19 Enforcement. TS 3.4.3.2 requires, in part, that RCS operational leakage shall be limited to no pressure boundary leakage. If pressure boundary leakage exists, the TS 3.4.3.2 limiting condition for operation action statement requires Unit 1 to be in at least hot shutdown within 12 hours and in cold shutdown within the next 24 hours. Contrary to the above, for a period that began on an unknown date that was very likely more than 36 hours before March 20, 2016, and ending on March 20, 2016, RCS pressure boundary leakage existed, and Exelon did not place Unit 1 in at least hot shutdown within 12 hours and in cold shutdown within the next 24 hours. This issue is considered within the traditional enforcement process because there was no performance deficiency associated with the violation of NRC requirements. Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Section 03.22 states, in part, that traditional enforcement is used to disposition violations receiving enforcement discretion or violations without a performance deficiency. The NRC Enforcement Policy, Section 2.2.1 states, in part, that, whenever possible, the NRC uses risk information in assessing the safety significance of violations. Accordingly, after considering that the condition represented very low safety significance, the inspectors concluded that the violation would be best characterized as Severity Level IV under the traditional enforcement process. However, the NRC is exercising enforcement discretion (EA- 17- 076) in accordance with Section 3.10 of the NRC Enforcement Policy which states that the NRC may exercise discretion for violations of NRC requirements by reactor licensees for which there are no associated performance deficiencies. In reaching this decision, the NRC determined that the issue was not within the licensees ability to foresee and correct; the licensees actions did not contribute to the degraded condition; and the actions taken were reasonable to identify and address the condition. Furthermore, because the licensees actions did not contribute to this violation, it will not be considered in the assessment process or the NRCs Action Matrix.
05000382/FIN-2017002-05	Waterford	2017Q2	Failure to Perform Maintenance on the Correct Safety-Related Component	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, which occurred due to the licensees failure to perform field work on reactor coolant loop 2 shutdown cooling warm-up valve, SI-135A. Specifically, mechanical maintenance technicians, who were assigned work on safety injection train A, erroneously performed work on safety injection train B on reactor coolant loop 1 shutdown cooling warm-up valve, SI-135B. As a result, both trains of emergency core cooling systems were simultaneously inoperable, which placed the plant in a 1-hour technical specification shutdown action statement. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-01433. The licensees corrective actions included a revision of the model work order to require concurrent verification for component identification, and adding the valves to the protected equipment list for when the opposite train is inoperable.The performance deficiency was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the mechanics worked on valve SI-135B instead of valve SI-135A, they simultaneously made both trains of emergency core cooling systems inoperable. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix A, Significance Determination Process for Findings At-Power. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, and component; (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 greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with licensees maintenance rule program for greater than 24 hours.The finding had an avoid complacency cross-cutting aspect in the area of human performance because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, and did notimplement appropriate error reduction tools. Specifically, maintenance technicians repeatedly visited the incorrect work location and didnt properly verify the valve number to ensure they would work on the correct component (H.12).
05000275/FIN-2017002-04	Diablo Canyon	2017Q2	Failure to Follow Procedures Results in Partial Loss of Cooling Flow to Shutdown Cooling	Green . The inspectors reviewed a self -revealing, non- cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because PG&E personnel failed to follow the requirements of AD7.ID14, Assessment of Integrated Risk, Revision 11. Specifically, PG&E personnel failed to obtain shift manager permission, conduct a protected equipment briefing, and document shift manager approval prior to performing work on protected equipment. This resulted in a loss of flow of cooling water to one of two in- service shutdown cooling residual heat removal heat exchangers and subsequent perturbation in reactor coolant system temperature during refueling outage 1R20. The inspectors determined that PG&E s failure to follow AD7.ID14, Assessment of Integrated Risk, Section 5.14 Performing Work on Posted Protected Equipment, was a performance deficiency within PG&Es ability to foresee and correct. This performance deficiency was considered to be more than minor because it impacted the configuration control attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the loss of cooling flow to the RHR heat exchanger while in shutdown cooling mode resulted in a perturbation in RCS temperature of approximately 8 degrees Fahrenheit. The finding was evaluated in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined to be of very low safety significance (Green) since it did not represent a loss of system safety function of at least a single train for greater than four hours. The finding had a cross- cutting aspect in the area of human performance associated with conservative bias because PG&E personnel did not use decision- making practices that emphasize prudent choices over those that are simply allowable. Specifically, despite being authorized to close component cooling water cross connect valves by the work control process, PG&E personnel did not question the impact of their actions on shutdown cooling (H.14 ).
05000263/FIN-2017002-02	Monticello	2017Q2	Licensee-Identified Violation	The licensee identified a finding of very low safety significance (Green) and associated NCV of TS 3.7.1, Residual Heat Removal Service Water (RHRSW) System; which requires, in part, that two RHRSW subsystems shall be operable in Modes 1, 2, and 3 or per Condition A, One RHRSW subsystem inoperable; the RHRSW subsystem must be restored to OPERABLE status within 7 days or the applicable conditions and required actions of Limiting Condition forOperations 3.4.7, Residual Heat Removal Shutdown Cooling System Hot Shutdown, for RHR shutdown cooling made inoperable by RHRSW System must be entered. Contrary to the above, on March 27, 2017, the licensee exited the requirements in TS 3.7.1, with a Tag Section still hanging, rendering B RHRSW subsystem inoperable, while in Mode 1. This was identified by the licensee when the maintenance organization notified operations that work was complete, and the Tag Section was released. The licensee reentered TS 3.7.1, Condition A, entered the issue as CAP 1554105 and assigned a Human Performance Event Investigation. A crew clock reset was also taken as well as communicating lessons learned to the entire plant organization.This finding was more-than minor because the performance deficiency wasassociated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, RHRSW System is designed to provide cooling water for the RHR System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident or transient. Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post-accident heat loads, assuming the worst case single active failure occurs coincident with the loos of offsite power. The finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not involve an actual loss of safety system, did not represent actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not represent an actual loss of function of one or more non-Tech Spec Trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hours.
05000424/FIN-2017002-02	Vogtle	2017Q2	Failure to Follow Work Instructions for Implementation of Open Phase Protection System	(Green). A self -revealing, Green, non -cited violation of Technical Specifications 5.4.1.a, Procedures, was identified for the licensees failure to redline new wiring installation associated with an open phase protection system modification, as required by work instructions . As result, control circuit wires were not installed per wiring diagrams and caused a loss of the offsite power feed to the B train 4160- volt emergency power bus. The licensee's failure to redline new wiring installation associated with an open phase protection system modification installation, as required by work instruction SNC804606 and 3 maintenance procedure NMP -MA -017 was a performance deficiency. The licensee entered this issue into their corrective action program under condition reports 10343972 and 10344136 and restored offsite power to the emergency bus by correcting the wiring configuration . The performance deficiency was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because the in- service train of shutdown cooling (i.e. , 'A' train of the residual heat removal system ) was not affected. The finding was assigned a cross -cutting aspect of Procedure Adherence, in the Human Performance area becaus e individuals did not follow work instructions and redline procedures when installing new wiring for the open phase protection system (H.8)
05000416/FIN-2016008-02	Grand Gulf	2017Q2	Failure to Have Adequate Procedures	The team reviewed a self -revealed, non -cited violation of Technical Specification 3.4.10, Residual Heat Removal Shutdown Cooling System Cold Shutdown, for the licensees failure to verify an alternate method of decay heat removal was available when residual heat removal subsystem A was inoperable and unavailable due to a pump replacement. Specifically, the licensee inappropriately credited the alternate decay heat removal system as an available alternate method of decay heat removal. Credit for this system was inappropriate because, although the licensee believed the system had been aligned in standby, the alternate decay heat removal heat exchanger isolation valves had remained tagged closed, rendering the system unavailable to satisfy the technical specification requirement during the time period that residual heat removal subsystem A was unavailable. The licensee restored compliance by restoring residual heat removal subsystem A to available status . The licensee entered this issue into their corrective action program as Condition Report CR -GGN -2016 -07281. The failure to perform the required action to verify an alternate method of decay heat removal was available, when a residual heat removal shutdown cooling system was inoperable, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the 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. A detailed risk evaluation (Attachment 2) calculated an increase in core damage frequency of 3.2E -7/y ear and an increase in large early release frequency of 7.3E -8/year. Therefore, this violation is associated with a finding having very low safety significance (Green). The team determined the finding had a cross -cutting aspect within the human performance area, field presence, because leaders failed to reinforce standards and expectations in the work areas of the plant (H.2).
05000416/FIN-2016008-01	Grand Gulf	2017Q2	Failure to Have Alternate Decay Heat Removal Capability	The team identified two examples of a non -cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to have adequate procedures for activities affecting quality. Specifically, Grand Gulf Nuclear Station failed to have adequate procedures for feedwater, condensate, and shutdown cooling activities. The licensee implemented corrective actions to revise the procedures. The licensee entered this issue into their corrective action program as Condition Reports CR- GGN -2016- 08334, 08273, and 08290. The failure to have adequate procedures for activities affecting quality was a performance deficiency. Example (1) of this performance deficiency was more than minor, and therefore a finding, because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not having procedural guidance for the alternate decay heat removal system alignment resulted in misalignment of the system and its subsequent inability to perform its required function if needed. A detailed risk evaluation (Attachment 2) calculated an increase in core damage frequency of 3.2E -7/year and an increase in large early release frequency of 7.3E -8/year, which has a very low safety significance (Green) . Example (2) of this performance deficiency was more than minor, and therefore a finding, because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, not having procedural guidance for feedwater isolation valve operation resulted in inadvertent over fill of the reactor vessel. This violation is associated with a finding having very low safety significance (Green). The team did not assign a cross -cutting aspect because the performance deficiency was not reflective of current plant performance .
05000528/FIN-2017007-01	Palo Verde	2017Q1	Failure to Analyze Shutdown Cooling and Feedwater Lines for High-Energy Line Break Pipe Whip Effects	Green. The team identified a Green 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 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, from August 11, 1982, to March 3, 2017, the licensee did not analyze dynamic pipe whip effects of a main feedwater line for a high-energy line break of a shutdown cooling line. In response to this issue, the licensee performed immediate and prompt operability evaluations and determined that the piping systems remained operable and could withstand the effects of a high-energy line break. This finding was entered into the licensees corrective action program as Condition Report CR-17-02815. The team determined that the failure to perform an adequate analysis for shutdown cooling and feedwater lines for high-energy line break pipe whip effects was a performance deficiency. This finding was more-than-minor 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 failure to analyze the main feedwater piping for high-energy line break effects called the operability of the piping system into question. 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 hav ing 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. The team determined that this finding did not have a cross- cutting aspect because the most significant contributor to the performance deficiency did 3 not reflect current licensee performance. Specifically, the licensee performed the calculation in 1982 and revised it in 1991; therefore, the performance deficiency occurred outside of the nominal three-year period for present performance.
05000440/FIN-2017001-01	Perry	2017Q1	Failure to Implement Procedures for Combating a Loss of Shutdown Cooling	Green. A finding of very-low safety significance and associated NCV of TS 5.4, Procedures, was identified by the inspectors for the failure to implement procedures for combating a loss of shutdown cooling (SDC). Specifically, the licensee failed to implement its procedure for combating a loss of SDC resulting from emergency service water (ESW) inoperability and during high decay heat load. This finding was entered into the licensees Corrective Action Program to perform analyses for various conditions to identify available alternate methods of decay heat removal and provide associated procedural guidance. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as very-low safety significance (Green) because it was a design deficiency that did not impact the operability or Probabilistic Risk Assessment functionality of any mitigating structures, systems, and components. The inspectors did not identify a cross-cutting aspect associated with this finding because it did not reflect current performance due to the age of the performance deficiency
05000461/FIN-2017001-02	Clinton	2017Q1	Failed to Verify an Appropriate Alternate Method of Decay Heat Removal	Green. The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.36(c)(2)(i), Limiting conditions for operation, for failing to meet/follow the required actions for limiting condition for operation 3.9.9 and 3.4.10. Specifically, the operators failed to verify a credited alternate decay heat removal method that would satisfy the required action for the limiting condition for operation. The licensee entered this issue into their corrective action program as AR 03987440. The corrective actions in response to this violation were to identify appropriate alternate methods of decay heat removal and incorporate them into the shutdown safety management program utilized during plant outages. The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance 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, with the operators failing to identify a credited alternate method of decay heat removal and taking credit for the inoperable but in service RHR shutdown cooling train, the actual available methods that could have been credited were not verified to ensure their availability to provide the required function. The finding was screened against the Mitigating Systems Screening questions and determined to be of very low safety significance because the answer to all of the applicable screening questions was No. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the senior reactor operators at the station had historically credited inoperable RHR shutdown cooling subsystems as their own alternate decay heat remove method because they believed it was allowable without determining that it was safe in order to proceed. (H.14
05000298/FIN-2017001-05	Cooper	2017Q1	Loss of Shutdown Cooling due to Relay Maintenance	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to implement Maintenance Procedure 7.3.16, Low Voltage Relay Removal and Installation, Revision 22, for relay replacement work. Specifically, on October 28, 2016, the licensee failed to evaluate the potential impact of primary containment isolation system relay PCIS-REL-K27 work on shutdown cooling relay PCIS-REL-K30, which was mounted next to K27 and shared a common mounting rail. As a result, the licensee did not identify the potential of losing residual heat removal shutdown cooling, and while installing the K27 relay and snapping it into the mounting rail, workers caused a momentary actuation of relay K30 and a loss of residual heat removal shutdown cooling. Corrective actions to restore compliance included restoration of shutdown cooling, completion of the K27 relay maintenance with shutdown cooling out of service, and an outage risk management procedure change that prohibited work on or near shutdown cooling relays while the system was required to be in service. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-07645. The licensees failure to implement Maintenance Procedure 7.3.16, in violation of Technical Specification 5.4.1.a, 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 Initiating Events Cornerstone and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the inspectors determined that the finding did not require a quantitative assessment because the event occurred when the refuel canal/cavity was flooded. Therefore, the finding screened as very low safety significance (Green). The finding had a cross-cutting aspect in the area of human performance associated with work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the need for coordination with different work groups or job activities. Specifically, the licensee failed to control, execute, and coordinate safety-related primary containment isolation system relay work activities to ensure residual heat removal shutdown cooling was not adversely impacted (H.5).
05000354/FIN-2016004-01	Hope Creek	2016Q4	Trip of Protected RWCU Pump during Maintenance Activity	Green. A self-revealing very low safety significance (Green), non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4) was identified for inadequately assessing and managing risks associated with maintenance activities to prevent plant transients that upset plant stability. Specifically, because PSEG did not identify a conflict with the reactor water cleanup (RWCU) pump trip logic prior to conducting a planned breaker swap, the A RWCU pump tripped while it was credited to as a defense-in-depth system for decay heat removal (DHR). PSEG assigned a corrective action to perform a work group evaluation and address lessons learned from this event. The issue was more than minor because it was associated with the Equipment Performance (availability) attribute of the Initiating Event cornerstones and adversely affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown. Additionally, this issue was similar to IMC 0612, Appendix E, examples 7.e and 7.f, in that the resulting increased risk put the plant into a higher risk category. In this case, the plant risk would have been reclassified from Yellow to Orange when RWCU pump was unavailable during residual heat removal (RHR) shutdown cooling outage window. The inspectors evaluated the finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was Green because no quantitative phase 2 analysis was required, and RWCU system was not identified as a major system on Table G1 for Decay Heat Removal safety function. This finding had a cross-cutting aspect in the area of Human Performance, Work Management, because PSEG did not identify and appropriately manage risk associated with the breaker swap activity. Specifically, PSEGs work order to swap the breaker was not planned or scheduled during a RWCU system outage window where the plant shutdown safety risk would have been properly managed (H.5).
05000285/FIN-2016004-02	Fort Calhoun	2016Q4	Licensee-Identified Violation	Technical Specification 2.0.1 requires the unit to be shut down within 6 hours in the event a limiting condition for operation and/or associated action requirement cannot be satisfied because of circumstances in excess of those addressed in the specification. Contrary to the above, the licensee failed to enter Technical Specification 2.0.1 and take the prescribed actions on several occasions when shutdown cooling heat exchanger valves were opened which impacted component cooling water (CCW) flow to the containment air cooling units under certain accident conditions. On May 10, 2016, an unanalyzed condition was discovered during scheduled maintenance on the shutdown cooling heat exchanger valves. As part of the maintenance, HCV-484, Shutdown Heat Exchanger AC-4A Component Cooling Water Outlet Valve, and HCV-481, Shutdown Cooling Heat Exchanger AC-4B Component Cooling Water Inlet Valve, were failed open which rendered both valves inoperable. Under these conditions, with the assumed single failure loss of DC control power during a loss of coolant accident (LOCA), CCW would be allowed to flow through both shutdown cooling heat exchangers, effectively reducing CCW system flow to the containment air cooling units. These conditions are not assumed under plant design basis calculations and placed the plant in an unanalyzed condition. It has not been demonstrated that the CCW system would adequately perform its design function of providing a cooling medium for the containment atmosphere under LOCA conditions with CCW flow diverted through the shutdown cooling heat exchangers. With two containment air cooling units inoperable, Technical Specification 2.4, does not provide an associated action; therefore, Technical Specification 2.0.1 applies. Upon completion of the maintenance activity, both valves were returned to service which eliminated the condition. The licensee conducted an extent of condition review and identified that they had created this unanalyzed condition six times within the last 3 years and had exceeded the Technical Specification 2.0.1 6-hour shutdown action statement on March 8, 2016; April 21, 2016; and May 10, 2016. In addition, the licensee determined this condition was first identified on February 3, 2015, in Condition Report 2015-01388. Procedure TDB-VIII, Equipment Applicability Guidance, Revision 64, incorrectly stated the valves had a required safety function in the open direction. The licensee initiated procedure change EC-68088 on September 26, 2015, to correct the procedure; however, the proposed change did not accurately reflect the safety function of the valves to remain closed for all LOCA conditions. This procedure change was still under review on May 10, 2016. The failure to promptly correct Procedure TDB-VIII was a contributing cause of the violation. The violation is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone. On March 8, 2016; April 21, 2016; and May 10, 2016, the plant was placed in a condition prohibited by technical specifications and exceeded the Technical Specification 2.0.1, 6-hour shutdown action statement. This adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A senior reactor analyst qualitatively determined that this finding was of very low safety significance (Green) for increases in core damage frequency and large early release frequency because of the short exposure time of less than 3 days and because of the low frequency of events where a LOCA with an independent and coincidental loss of DC control power would occur. Therefore, this finding screens to Green. The licensee entered the issue into their corrective action program as Condition Reports 2016-05340 and 2016-04468.
05000461/FIN-2016009-05	Clinton	2016Q4	Failure to Promptly Identify that the Incapability of the RHR Design to Support TS Operability Requirements Was a CAQ	The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to promptly identify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit, RHR could not be realigned from shutdown cooling mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and AR 03948042, and planned to submit a License Amendment Request to align TS requirements with the design capabilities. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated 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 voluntarily declaring TS functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding: (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out-of-service for greater than their TS allowed outage time; (3) involved non-TS trains of equipment; (4) involved a degradation of a functional RHR auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross-cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago.
05000335/FIN-2016003-01	Saint Lucie	2016Q3	Reactor Coolant System Leakage Technical Specification Violation	An NRC-identified Green non-cited violation (NCV) of Unit 1 Technical Specification 3.4.6.2 Reactor Coolant System Leakage was identified. Specifically, the licensee failed to enter TS 3.4.6.2 Action c for reactor coolant system pressure isolation valve (V3217) when the valve experienced operational seat leakage of approximately 30 gpm during flushing and cooling the shutdown cooling system. Immediate corrective actions were not required since the valve was later determined to be inoperable and repaired. The licensee entered this issue into the licensees corrective action program. The licensees failure to recognize that gross seat leakage from check valve V3217 indicated of a major problem with valve seat alignment and that higher differential pressure would not help seat the valve was a performance deficiency (PD). The performance deficiency is more than minor because it is associated with the barrier integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers such as the containment, protected the public from radionuclide releases caused by accidents or events. The PD resulted in 46 additional hours of operation with V3217 seat leakage outside of TS acceptance criteria which required the unit to be in cold shutdown. The finding involved the cross-cutting area of human performance and specifically within that area was associated with conservative bias because the operability evaluation did not demonstrate it was safe to proceed with valve V3217 experiencing gross seat leakage (H.14).
05000354/FIN-2016003-03	Hope Creek	2016Q3	Inadequate Procedure Adherence Resulted in a Loss of Shutdown Cooling	A self-revealing non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, occurred when PSEG did not follow procedure during the transition from Cold Shutdown to refueling operations while filling up the reactor pressure vessel (RPV) to support RPV head cooling in preparation for reactor disassembly. This resulted in an automatic isolation of the operating residual heat removal (RHR) pump while it was providing decay heat removal in shutdown cooling. PSEG has entered this issue into their corrective action program (CAP) in notification (NOTF) 20684861, and corrective actions included performing a root cause evaluation for the event, revising the operating procedures to provide clarity, and conducting training with all operators on the lessons learned from the event. This issue was evaluated in accordance with IMC 0612, Appendix B, and determined to be more than minor since it was associated with the human performance attribute of the Initiating Events cornerstone and adversely affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations. The finding was evaluated using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process (SDP), and per Attachment 1, Exhibit 2, required a Phase 2 risk evaluation which determined the safety significance of this performance deficiency to be in the mid E-8 range, or of very low safety significance (Green). The inspectors determined this finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, in that the operator did not use decision-making practices that emphasized prudent choices over those that are simply allowable, and the operators proposed action was not determined to be safe prior to proceeding with the action. Specifically, the operator did not ensure his actions were safe prior to aligning and operating the feedwater system to fill the RPV during plant cooldown using an uncommon method.
05000315/FIN-2016002-02	Cook	2016Q2	Containment Closure Requirements during Unit 1 2016 Refueling Outage	The inspectors identified a finding of very low safety significance with an associated NCV of TS 5.4, Procedures, for the failure to implement all of the requirements of PMP4100SDR001, Plant Shutdown Safety and Risk Management, pertaining to the closure of containment airlocks in the event shutdown cooling is lost. Contrary to TS 5.4, the licensee failed to implement the procedure as demonstrated by lack of closure requirement knowledge by containment closure attendants, failure to include isolation valves on ice lines, missing a shiftly check, and lack of required anti-contamination clothing. The licensee corrected the issues and entered them into the CAP. The issue was greater than minor because it adversely affected the Human Performance attribute of the Barrier Integrity cornerstone, whose objective is to provide assurance that principal design barriers (e.g., containment) can protect the public from radionuclide releases. Additionally, the inspectors were informed by IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, because the issue was programmatic in nature and could lead to more significant issues if left uncorrected. The finding screened as Green per IMC 0609 Appendix H, Containment Integrity Significance Determination Process, because the inspectors determined despite the issues identified, containment closure could be achieved within the time-to-boil. The inspectors determined the finding had an associated cross-cutting aspect of H.1, Resources, because leaders did not ensure personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety.
05000220/FIN-2016001-03	Nine Mile Point	2016Q1	Inadequate Tagout Resulting in Reactor Building Closed-Loop Cooling Drain Down Event	A self-revealing Green non-cited violation (NCV) of Technical Specification (TS) 6.4.1, Procedures, was identified when a Unit 1 Exelon operator did not maintain proper configuration control of a plant system during a system tagout for planned maintenance. Specifically, on January 25, 2016, a Unit 1 non-licensed operator manipulated a reactor building closed-loop cooling (RBCLC) system drain valve out of sequence while performing a tagout for the #13 shutdown cooling (SDC) HX for planned maintenance. This resulted in unintentional draining of the operating RBCLC system, annunciation of multiple alarms in the main control room, and operators entering abnormal operating procedures to recover the RBCLC system. As part of corrective actions, proper configuration was promptly restored and the operator involved in the event was given a remediation plan for requalification and placed on an operations excellence plan. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences; and if left uncorrected, the event had potential to lead to a more significant safety concern. Specifically, the failure to quickly isolate the drain down of the RBCLC system would have required a manual reactor scram, a manual trip of all five reactor recirculation pumps (RRPs), a manual isolation of the reactor water cleanup system, a loss of cooling to the spent fuel pool (SFP) cooling system, instrument air compressors, and the control room emergency ventilation system. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not result in the loss of a support system, RBCLC, or affect mitigation equipment. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because the non-licensed operator failed to follow Exelons procedures and the instructions he received at the pre job brief stop when manipulating the drain valve. Specifically, the non-licensed operator rationalized, without being the designated performer of the tagout, that it was acceptable to perform a valve manipulation out of sequence with the tagout plan.
05000296/FIN-2016001-06	Browns Ferry	2016Q1	Failure to Identify Applicable Technical Specification Action Statement for a PCIV	An NRC identified non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures, for the licensees failure to implement OPDP-8, Operability Determinations and LCO Tracking. Specifically, the licensee failed to track the applicability of condition A of TS LCO 3.6.1.3 upon discovery of the equipment failure related to the Residual Heat Removal (RHR) Shutdown Cooling (SDC) inboard suction valve as described in LER 05000296/2014-003-00. As an immediate corrective action, the licensee entered the violation into the corrective action program as CR 1115172. The performance deficiency was more-than-minor because, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, this failure was indicative of a programmatic weakness with the licensees evaluation of certain logic circuit failures which can result in misapplication of the allowances of TS LCO 3.0.6 and inappropriate TS LCO entries. The inspectors determined that this type of error was likely to recur which could lead to worse errors if uncorrected. The inspectors determined the finding was Green because the error did not result in an actual open pathway in the physical integrity of reactor containment, containment isolation system or heat removal components. The inspectors determined that the finding had a crosscutting aspect of Training in the area of Human Performance because the finding was indicative of a knowledge gap among the operations department (H.9).
05000285/FIN-2016001-01	Fort Calhoun	2016Q1	Implementing a Procedure Change for Alternative Shutdown Cooling that would have Required NRC Approval	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, for the failure to recognize that a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the 10 CFR 50.59 evaluation revised a site procedure, without NRC approval, to substitute automatic flow control of shutdown cooling flow and temperature with manual control using the low pressure safety injection loop injection valves. The licensees corrective actions included revising the affected procedure to reflect the original automatic flow control. The licensee entered this issue in the corrective action program as Condition Report 2013-15342. The licensees failure to implement the requirements of 10 CFR 50.59 and adequately evaluate changes to determine if prior NRC approval is required was a performance deficiency. Because this violation had the potential to impact the NRCs ability to perform its regulatory function, the inspectors evaluated the violation using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the team evaluated this finding using the significance determination process to assess its significance. 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, dated July 1, 2012. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was determined to have 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. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy, the inspectors characterized this performance deficiency as a Severity Level IV violation. The inspectors determined that a cross-cutting aspect was not applicable because the issue involving the failure to perform an adequate 10 CFR 50.59 evaluation was strictly associated with a traditional enforcement violation.
05000247/FIN-2016001-02	Indian Point	2016Q1	Failure to Adequately Implement Risk Management Actions for the Containment Key Safety Function	The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4) because Entergy did not effectively manage the risk associated with refueling maintenance activities. Specifically, Entergy did not demonstrate they could implement their planned risk management action to restore the containment key safety function within the time-to-boil using the equipment hatch closure plug. Entergy wrote CRIP2- 2016-01503 and CR-IP2-2016-01883 to address this issue. This performance deficiency is more than minor because it impacted the barrier performance attribute of the Barrier Integrity cornerstone and affected the objective to provide reasonable assurance that containment protects the public from radionuclide releases caused by accidents or events. Specifically, Entergy did not demonstrate that they could install the hatch plug within the time-to-boil and that the plug would seal the equipment hatch opening, which affected the reliability of containment isolation in response to a loss of shutdown cooling or other event inside containment. The inspectors determined the finding could be evaluated using Attachment 0609.04, Initial Characterization of Findings. Because the finding degraded the ability to close or isolate the containment, it required review using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. Since containment status was not intact and the finding occurred when decay heat was relatively high, it required a phase two analysis. Since the leakage from containment to the environment was less than 100 percent containment volume per day, the finding screens as very low safety significance (Green). A subsequent demonstration showed that the hatch plug provided an adequate seal with the containment hatch opening. The inspectors concluded this finding had a cross-cutting aspect in the area of Human Performance, Documentation, because Entergy did not maintain complete, accurate, and up-to-date documentation related to the use of the hatch plug. Specifically, they tested the seal integrity without using a work order (WO), and made pen-and-ink changes to the procedure without processing a procedure change form.
05000333/FIN-2016001-02	FitzPatrick	2016Q1	Uncontrolled RPV Level Increase after Initiation of RHR Shutdown Cooling	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to take actions specified in the procedure for initiation of shutdown cooling. Specifically, prior to placing the A loop of the residual heat removal (RHR) system into shutdown cooling, an operator was not stationed to close the condensate transfer system cross-connect valve to the A RHR loop (10RHR-274), nor was the valve immediately closed after initiation of shutdown cooling, as specified by the operating procedure. This resulted in a significant loss of operational control, in that RPV level increased to the point of putting water down the main steam lines. As immediate corrective action, operators closed 10RHR-274, thus stopping the RPV inventory increase. The issue was entered into the CAP as CR-JAF-2016-00273. The finding was more than minor because it was associated with the human performance 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 resultant loss of RPV level control represented a significant loss of operational control that could have affected the operability of the HPCI and reactor core isolation cooling (RCIC) systems, as well as the S/RVs, had their use again been required in the near term. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because operators did not stop when faced with uncertain conditions. Specifically, without otherwise having maintained status control on the condensate transfer system cross-connect valve to the A RHR loop, operators did not stop to positively establish the condition of the valve when it appeared in a conditional step in the procedure (that is, if 10RHR-274 is open, then station an operator at 10RHR-274) (H.11).
05000263/FIN-2016008-01	Monticello	2016Q1	Failure to provide acceptable Alternate Methods of Decay Heat Removal	The inspectors identified an Unresolved Item associated with Technical Specification (TS) 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System Cold Shutdown. Specifically, the licensee failed to verify that the capability of the alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown Cooling, were adequate to combat a loss of shutdown cooling resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay heat load. The Limiting Condition for Operation (LCO) 3.4.8 of TS Residual Heat Removal Shutdown Cooling System Cold Shutdown, required in Mode 4, two RHR shutdown cooling subsystems shall be operable, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation. The TS Bases Section 3.4.8, indicated that an operable RHR shutdown cooling subsystem consisted of one operable RHR pump, one heat exchanger, the associated piping and valves, and the necessary portions of the RHR Service Water System System capable of providing cooling water to the heat exchanger. The TS Bases Section 3.4.8 further indicated that the two subsystems have a common suction source and were allowed to have a common heat exchanger and common discharge piping. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be operable. Since the piping and heat exchangers were passive components that were assumed not to fail, they were allowed to be common to both subsystems. When TS 3.4.8, LCO could not be met, Condition A, for one or two RHR shutdown cooling subsystems inoperable, the Required Action was to, verify an alternate method of decay heat removal was available for each inoperable RHR shutdown cooling subsystem. The completion time for the required action was 1 hour, and once per 24 hours thereafter. The TS Bases 3.4.8 for Condition A indicated that with one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem was capable of providing the required decay heat removal. However, the overall reliability was reduced, therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This was to ensure the re-establishment of backup decay heat removal capabilities, similar to the requirements of the LCO. The bases further stated that the required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Alternate methods that can be used included (but not limited to) the Reactor Water Cleanup System by itself or using feed and bleed in combination with Control Rod Drive System or Condensate/Feed Systems. Abnormal Procedure, Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown Cooling, provided instructions for establishing alternate methods for decay heat removal. The inspectors noticed that except for the alternate method as described below in the G-EK-1-45, the licensee was not able to show by calculation or demonstration that the systems and methods credited in this procedure would be capable of providing sufficient heat removal capability or appropriate levels of redundancy as required by TS 3.4.8. The G-EK-1-45 was a General Electric Letter to Northern States Power, Subject: Cold Shutdown Capability Report, dated April 22, 1981. This letter provided a report which described the capability of the Monticello Nuclear Generating Plant to achieve cold shutdown using only safety class systems and assuming the worst single failure. The alternate shutdown decay heat removal method used in the report credited combinations of the RHR pumps and heat exchangers in the suppression pool cooling mode of RHR to ensure suppression pool water temperatures were below the design limit. This method utilized the core spray system and safety relief valves to circulate reactor inventory to remove decay heat from the reactor. The inspectors noted that calculations supporting the above alternate strategy utilized an RHR subsystem that could be inoperable and/or unavailable and therefore may not be credited to comply with TS 3.4.8. Specifically, the inspectors were concerned that while the plant was in mode 4, with a credited one subsystem inoperable, the licensees credited alternate decay heat removal method that relied on an RHR subsystem, to perform the required suppression pool cooling function. The inspectors were concerned that relying on the only operable RHR subsystem for the alternate method did not meet the intent of the TS requirement as described in the TS Bases. Furthermore, the inspectors noticed for Mode 4 with two RHR subsystems inoperable, the licensee failed to verify by calculation or demonstrations that two additional redundant alternate decay heat removal methods existed with sufficient capacity to maintain the average reactor coolant temperature below 212 degrees Fahrenheit. During the inspection, the licensee indicated that the Boiling Reactor Owners Group was in the process of developing a draft TS Task Force Traveler to address the requirement of TS 3.4.8 and its Bases. Based on the information above, the inspectors were concerned that the plant Operations Manual was inadequate and failed to include alternate decay heat removal methods that would enable the licensee to comply with the requirement of TS 3.4.8. The Operations Manual was required per TS 5.4.1, Procedures, which required that written procedures shall be established, implemented, and maintained covering the emergency operating procedures. The inspectors determined that this issue was unresolved pending the actions by the licensee and the Boiling Reactor Owners Group and the NRC review of these actions. The licensee entered the inspectors concerns into their Corrective Action Program as AR 01516098.
05000313/FIN-2016007-06	Arkansas Nuclear	2016Q1	Failure to Correct Degraded Unit 2 Train B Emergency Diesel Generator Heat Exchangers Service Water Flow and Degraded Unit 1 Containment Coatings	The team identified two examples of a Green finding and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct conditions adverse to quality. Specifically, the licensee failed to correct long term degraded service water flow to the Unit 2 safety-related train B emergency diesel generator heat exchangers since 2008, and degraded Unit 1 reactor containment building coatings since 2009. The licensees corrective actions included performing an operability determination and determining that the service water system and the Unit 1 containment sump were operable and documenting the issue in the corrective action program as condition reports CR-ANO-C-2016-00946, and CR-ANO-1-2015-00200. The failure to correct conditions adverse to quality associated with Unit 2 service water flow to the B emergency diesel generator heat exchangers and the Unit 1 reactor containment building coatings was a performance deficiency. The performance deficiency was determined to be more than minor 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, the failure to correct long term degraded: 1) service water flow beyond the action limit in accordance with procedure EN-DC-159, Component and System Monitoring, to the B emergency diesel generator heat exchangers, which challenged the capability of emergency diesel generator response to design basis events; and 2) containment coatings which challenged the Unit 1 emergency core cooling system capacity. The finding was evaluated using Inspector Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The team determined the finding was of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of mitigating system, structure or component, but the system, structure or component maintained operability. This finding had a human performance cross-cutting aspect of Design Margins because the licensee failed to place special attention on maintaining margins in safety related equipment. Specifically the licensee has repeatedly: 1) throttled service water flow away from the safety-related shutdown cooling heat exchangers, reducing the shutdown cooling design margins to maintain minimally acceptable flow to the emergency diesel generator heat exchangers since 2008; and 2) reduced the available containment sump margin rather than correct containment coating deficiencies (H.6).
05000458/FIN-2016009-01	River Bend	2016Q1	Failure to Follow Procedure While Installing Jumpers for Shutdown Cooling	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).
05000456/FIN-2016001-03	Braidwood	2016Q1	Failure to Correct a Condition Adverse to Quality Leads to Loss of One Train of Shutdown Cooling in Mode 6	A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, was self-revealed when the licensee failed to ensure that a condition adverse to quality was promptly identified and corrected. Specifically, on October 8, 2015, valve 2RH606 failed to open and caused a loss of one train of shutdown cooling in Mode 6 and an unplanned orange risk condition. The reason for the failure was improper use of a lower strength carbon steel valve key instead of the specified high strength hardened steel valve key, which had been the subject of a vendor Part 21, Reports of Defects and Non Compliance, Report. This issue was entered into the licensees CAP as IR 2567811. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance 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 failure to correct a condition adverse to quality in the form of the improper use of a lower strength carbon steel key instead of the specified high strength hardened steel key in a safety-related valve ultimately led to a loss of one train of shutdown cooling in Mode 6. The inspectors determined that the finding was of very low safety significance based upon a detailed risk evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because the performance deficiency was greater than three years old and therefore was not indicative of current performance.
05000458/FIN-2016009-02	River Bend	2016Q1	Failure to Establish Adequate Procedural Guidance	The team reviewed a self-revealing, non-cited violation of Technical Specification 5.4, Procedures, for three examples of the licensees failure to establish sufficient procedural guidance. Specifically, the licensees operations and radiation protection procedures did not provide sufficient direction to plant personnel to expeditiously establish a reactor vessel vent path, restore from a loss of shutdown cooling, and perform time sensitive entries into radiologically controlled areas. This issue was entered into the licensees corrective action program as Condition Reports CR-RBS-2016-0210, CR-RBS-2016-0370, and CR-HQN-2016-0132. Corrective actions included revising the applicable procedures. The failure to establish adequate procedural guidance in accordance with Regulatory Guide 1.33 was a performance deficiency. Specifically, Procedures GOP-0002, Power Decrease/Plant Shutdown, Revision 72, and AOP-0051, Loss of Decay Heat Removal, Revision 313, failed to provide adequate direction to operations personnel to expeditiously establish a reactor vessel vent path and recover shutdown cooling following an isolation. Additionally, Procedure EN-RP-101, Access Control for Radiologically Controlled Areas, Revision 11, failed to provide adequate guidance to perform time sensitive entries into radiologically controlled areas. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that adequate procedural direction was provided to operations personnel following a loss of shutdown cooling. This resulted in a delay in the restoration of shutdown cooling and plant heatup. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Exhibit 3, Mitigating Systems Screening Questions, the team determined that the finding is of very low safety significance (Green) because it: (1) affected the design or qualification of a mitigating structure, system, or component, and (2) the structure, system, or component maintained its operability and functionality. A cross-cutting aspect is not being assigned to this finding due to the timing of the performance deficiency not being indicative of current licensee performance.
05000296/FIN-2016001-08	Browns Ferry	2016Q1	Licensee-Identified Violation	Licensee Event Report (LER) 05000296/2014-003-00 Primary Containment Isolation Valve Inoperable for Longer Than Allowed by Technical Specifications 10 CFR 50, Appendix B, Criterion 5 required, in part, that activities affecting quality be implemented in accordance with documented procedures and drawings. Contrary to the above, between March 7, 2014 and June 6, 2014, relay 3-RLY-074-10A-K98A was wired incorrectly as discussed in LER 05000296/2014-003-00. The licensee corrected the wiring and entered the issue into the licensee's corrective action program as CR 892500. Inspectors screened the violation using IMC 0609, Appendix G, Attachment 1, Exhibit 3 Mitigating Systems Screening Questions, dated May 9, 2014. Because the finding degraded a functional auto-isolation of RHR on low reactor water level, a Phase 2 screening was required. Using attachment 3, Phase 2 Significance Determination Process Template for BWR During Shutdown, dated February 28, 2005, inspectors completed Worksheet 1 for Loss of Inventory in Plant Operating State 1 (Head On) and determined the risk was approximately 1e-7/yr, which was less than the 1e-6/yr threshold for a greater than Green finding. The dominant core damage sequence was the failure to isolate a reactor coolant leak and subsequent failure by operators to open vent paths (e.g. a safety relief valve) to control RCS pressure to enable continued low pressure injection. In the evaluation, no operator recovery credit was given for leak isolation, but credit was given for the redundant isolation valve that was operable which could have satisfied the automatic isolation function. The Regional Senior Reactor Analyst performed a detailed risk review of the finding. The risk review considered both the outage related risk, and the risk associated with a trip from power that would have the plant in shutdown cooling during the recovery. A screening analysis using bounding assumptions and the risk models ISLRHR event tree was performed. The dominant cutsets involved failure of the redundant valve to operate, and operator actions to recover. Because of the short exposure time during the shutdown periods, the redundant valve with the automatic action available, and the availability of operator recovery, the Finding was determined to be Green. This violation is being treated as an NCV consistent with Section 2.3.2 of the Enforcement Policy.
05000397/FIN-2015004-02	Columbia	2015Q4	Licensee-Identified Violation	Technical Specification 5.4.1.a, Procedures, requires, in part, that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Paragraph 9.a of Regulatory Guide 1.33, Appendix A, requires, in part, written procedures for performing maintenance that can affect the performance of safety-related equipment. The licensee established Procedures ISP-MS-Q901, RPS, Reactor Water Level Low, Level 3 Div 1 CFT/CC, Revision 10, and PPM 10.24.34, PM Calibration Test Barton Differential Indicating Switch, Revision 13, to meet the Regulatory Guide 1.33 requirements when performing maintenance on safety-related Barton main steam level indicating switches. Contrary to the above, prior to June 25, 2015, the licensee failed to maintain written procedures for performing maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to include instructions in Procedures ISP-MS-Q901 or PPM 10.24.34 for setting the mechanical stop inside Barton main steam level indicating switches. Subsequently, the mechanical indicator in the switches for MS-LIS-24A and MS-LIS-24C became mechanically bound on the rubber stop within the switch when the level was raised off-scale high during the refueling outage. The licensee implemented corrective action by inserting a half scram signal to comply with technical specifications, calibrating the affected switches including steps to set the mechanical stop, and initiating a condition report. The finding represented a loss of safety system function for reactor water level low (level 3) scram signals and for shutdown cooling isolation logic. Because the finding affected mitigating equipment during at-power and shutdown operations, the inspectors assessed the finding in both the Inspection Manual Chapter (MC) 0609, Appendix A, Significance Determination Process for At-Power Findings, and MC 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Exhibit 2 of MC 0609, Appendix A, and Exhibit 3 of MC 0609, Appendix G, inspectors determined that the finding required a detailed risk evaluation for the at-power portion of the finding and a Phase 2 evaluation for the shutdown portion of the finding because the finding represented a loss of safety-system function. A Region IV senior reactor analyst determined the issue was of very low safety significance (Green) and represented a total change to the core damage frequency of 4.4E-7/year. The dominant sequences were anticipated transients without scram and shutdown loss of inventory. For the at-power exposure, risk was mitigated by the use of the standby liquid control system and recirculation pump trips for the anticipated transients without scram. For the shutdown exposure, risk was mitigated by automatic injection by an emergency core cooling system pump for the losses of inventory. This issue was entered into the licensees corrective action program as AR 332078.
05000457/FIN-2015004-01	Braidwood	2015Q4	Loss of Shutdown Cooling Train During Refueling Cavity Fill and Associated Reduced Inventory Operations	On October 8, 2015, the inspectors identified an Unresolved Item (URI) regarding the failure of valve 2RH606, which is the 2A RHR heat exchanger flow control valve. The valves failure to open caused a loss of one train of shutdown cooling, and an unplanned Orange risk configuration with Unit 2 in Mode 6, and the reactor refueling cavity level less than 23 feet above the vessel flange. At the closure of the inspection period, the licensees investigation on the cause of the failure was ongoing. Resolution of this issue will be based on the inspectors review of the licensees completed investigation. A function of the RHR system in Mode 6 is to remove decay heat and sensible heat from the reactor coolant system (RCS). Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to the component cooling water system. The coolant is then returned to the RCS via the RCS cold legs. On October 8, 2015, valve 2RH606 became mechanically bound while in the process of filling the Unit 2 reactor refueling cavity to greater than 23 feet. This was identified when the operators attempted to open the valve from the control room. The failure of the valve to open caused Unit 2 shutdown risk to change from a planned Yellow configuration to unplanned Orange condition. Additionally, the licensee entered Limiting Condition for Operation 3.9.6, Residual Heat Removal and Coolant Recirculation-Low Water Level, Condition A, for one train of RHR cooling inoperable. This action required the licensee to initiate actions immediately to either restore the affected RHR loop to operable status or to initiate actions to establish greater than or equal to 23 feet of water above the reactor vessel flange. The licensee accomplished this action by raising water level in the cavity to greater than 23 feet. Troubleshooting of the failed valve revealed that a shaft key sheared, which prevented the valve from opening. The valve had been previously manipulated during the outage without an issue. The malfunctioning part was sent offsite for failure analysis. The valve was repaired. At the conclusion of the inspection, an apparent cause investigation was in process. This URI will remain open until the investigation is complete and the inspectors review the report to determine whether a performance deficiency exists.
05000336/FIN-2015012-01	Millstone	2015Q4	Failure to Implement Procedural Guidance During a Loss of RCS Inventory	The NRC identified a Green NCV of Millstone Power Station Unit No. 2 Technical Specifications (TS) 6.8.1, Procedures involving Dominions failure to implement procedural steps when prompted by plant conditions to mitigate the event. Specifically, when pressurizer (PZR) level began to decrease while placing the shutdown cooling (SDC) system in service, the crew did not implement procedural guidance in OP-2207, Plant Cooldown, nor enter AOP 2568A, RCS Leak, Mode 4, 5, 6, and Defueled, as these procedures would have directed operators to locate the source of the leak. Later in the event, once the procedural guidance was implemented, action was taken to identify the location of the leak and it was isolated. After the event, selected crew members were removed from watch standing duties pending remediation. Dominion entered this issue into their corrective action program as CR1012358. The finding was more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, when entry conditions were met, operators did not implement procedural guidance that would have directed them to locate the source of the leak. The finding screened to very low safety significance (Green) using Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Screening and Characterization of Findings, Exhibit 3 - Mitigating Systems Screening Questions. Specifically, the finding did not represent a loss of system safety function. This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, in that licensed operators are expected to implement processes, procedures, and work instructions. Specifically, Dominion operators did not implement procedural guidance when prompted by plant conditions immediately after starting the A Low Pressure Safety Injection Pump (LPSI).
05000458/FIN-2015004-05	River Bend	2015Q4	Failure to Follow Procedure Results in Inadvertent Draindown of Reactor Pressure Vessel	The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to correctly implement procedure STP-200-0605, Remote Shutdown System Control Circuit Operability Test, Revision 307. The procedure was incorrectly performed leading to an unexpected configuration in which the reactor pressure vessel was aligned to the suppression pool, and approximately 360 gallons of reactor coolant were inadvertently transferred to the suppression pool. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2015-02354. The licensee restored compliance by restoring the system to a configuration that was consistent with plant operating procedures. Corrective actions included increased management oversight of remote shutdown system operation. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Initiating Events Cornerstone attribute of configuration control, 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, a loss of reactor pressure vessel inventory occurred due to the establishment of an unintended system configuration caused by the inadvertent repositioning of the reactor pressure vessel suction valve. The inspectors initially screened the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Exhibit 2 of NRC Inspection Manual Chapter 0609, Appendix G, Attachment 1, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding required a Phase 2 evaluation because the loss of inventory resulted in leakage to the suppression pool that if undetected or unmitigated in 24 hours or less would cause shutdown cooling to isolate. A Region IV senior reactor analyst performed a Phase 2 evaluation of this issue and determined the issue was of very low safety significance (Green) and represented a change to the core damage frequency of 3.8E-8/year. The event sequence was an actual loss of inventory which occurred after core refueling in the shutdown. Risk was mitigated by prompt operator recovery action to stop the loss of inventory along with the operating plant configuration, which had two residual heat removal pumps aligned for automatic injection, one control rod drive pump in operation at the time of the event, and all manual injection paths fully available to mitigate the event. This finding has a cross-cutting aspect in the area of human performance associated with avoid complacency because the licensee failed to ensure that individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
05000293/FIN-2015004-01	Pilgrim	2015Q4	Inadequate Implementation of Corrective Action following Winter Storm Juno	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, because Entergy did not adequately implement corrective actions for an identified condition adverse to quality. Specifically, Entergy did not implement all of the procedure changes needed to ensure shutdown cooling was placed in service in a timely manner after plant shutdown in preparation for or during a severe winter storm. Entergy entered this issue into the CAP as CR 2016-0120 and updated procedure 2.1.42 to meet the requirements of the corrective actions in CR 2015-0558. Inspectors verified that the new procedure revision included the required actions. The inspectors determined this performance deficiency is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity cornerstone, and adversely affected its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding is of very low safety significance (Green) in accordance with IMC 0609, Attachment 4 and Exhibit 3 of Appendix A, because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Entergy staff did not ensure procedure revisions were made in accordance with the requirements of EN-LI-102, Corrective Action Program.
05000263/FIN-2015003-05	Monticello	2015Q3	Failure to Provide Complete and Accurate Information in LER 05000263/2015-002-00	The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9 due to the licensees failure to provide information to the NRC that was complete and accurate in all material respects in accordance with the NRCs reporting requirements in 10 CFR 50.73(a)(1), Licensee Event Report (LER) System. Specifically, on June 29, 2015, the licensee failed to include an accurate assessment of the safety consequences and implications of a loss of shutdown cooling event when they issued LER 05000263/2015-002-00. This LER included an inaccurate assessment of safety implications, stating that engineering calculations show a potential worst case maximum temperature of 115 degrees Fahrenheit (F). The inspectors identified that engineering models actually showed potential worst case temperatures of 25-26 degrees F higher, which could have challenged or exceeded fuel pool cooling design specifications. Corrective actions included issuance of a revision to LER 2015-002-00 which contained the correct engineering modeling results and associated discussion of safety implications. The licensee entered this issue into its CAP (CAP 1484633). This issue was of more than minor significance under the Traditional Enforcement Process because the NRC relies on licensees to identify and correctly report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., loss of shutdown cooling) was evaluated separately and determined to be a finding of very low safety significance as documented in the 2015 2nd Quarter Integrated Inspection Report (05000263/2015002-01). In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was of more than minor concern with relatively inappreciable potential safety significance and is related to a finding that was determined to be a more than minor issue. Consistent with Example 6.9.d.1, this represented an example where the licensee submitted inaccurate information in a required report, which resulted in expansion of the scope of the next regularly scheduled inspection and required LER revision. Because there was no finding evaluated with this violation, the inspectors did not assign a cross-cutting aspect to this issue.
05000400/FIN-2015003-04	Harris	2015Q3	Loss of A ESW Train	A self-revealing green NCV of 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Criterion III, Design Control, was identified for failure to implement design control measures that verify adequacy of design. Specifically, EC 83681 involved the installation of a new pump bearing with different wear characteristics but the EC failed to evaluate the impact of the bearing replacement on alignment sensitivity of the pump shaft. The licensee took immediate action to align the Normal Service Water system to provide cooling to the heat loads affected by the loss of the A ESW pump. Failure to incorporate alignment requirements for the pump shaft in the work instructions associated with EC 83681 was a performance deficiency. The performance deficiency was related to the equipment performance attribute of the initiating events cornerstone. The performance deficiency was determined to be more than minor because the performance deficiency 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 failure of the ESW pump shaft resulted in a loss of service water which ultimately led to the loss of the A train of shutdown cooling for a period of twelve minutes. Inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 4 and Appendix G (June 19, 2012), Shutdown Operations Significance Determination Process. The inspectors determined the finding was associated with the Initiating Event cornerstone and required a detailed risk evaluation because the finding involved a loss of safety function. A detailed risk evaluation was completed by a regional Senior Reactor Analyst (SRA). The regional SRA performed a detailed risk review of the finding. The SRA performed the analysis by increasing the maintenance unavailability for the pump, and evaluating it versus the base case. This method was chosen because the pump was in standby service, and the dominant method of determining there was a failure would have been during testing, or operation under non accident conditions. The additional time for the unnecessary repair was used to adjust the base case maintenance unavailability. Online and shutdown risk were evaluated. The total impact was determined to be low enough for the finding to be GREEN for SDP purposes The finding had a cross-cutting aspect in the Human Performance area of Design Margin (H.6).
05000387/FIN-2015003-01	Susquehanna	2015Q3	RHR Shutdown Cooling Procedure Not Maintained Consistent with Technical Specification Requirements	Inspectors identified a finding of very low safety significance (Green) and associated NCV of SSES Unit 1 and 2 TS 5.4.1, Procedures, because Susquehanna did not maintain the procedure for operation of the residual heat removal (RHR) shutdown cooling (SDC) system consistent with the requirements in TS 3.4.8, RHR Shutdown Cooling- Hot Shutdown. As TS 3.4.8 requires two RHR SDC loops to be operable and, if no reactor recirculation pumps (RRPs) are running, one of the loops to be in-service in Mode 3 below the RHR cut in permissive pressure (98 psig), inspectors determined that OP-1(2)49-002, RHR Shutdown Cooling, was not maintained appropriately because a change to the procedure precluded operation of the system between 40 psig and the RHR cut in permissive pressure (98 psig). Susquehanna entered the issue into the corrective action program (CAP) as CR-2015-22882 and CR-2015-24137 and revised the procedure to remove the requirement that precluded operation of the SDC system between 40 psig and the RHR cut in permissive pressure This finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 40 psig procedural limit impacted the availability and capability of RHR to be placed in SDC between 98 psi, the cut-in permissive for the system, and 40 psig. In accordance with Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of Human Performance, Change Management because Susquehanna did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority (H.3). Specifically, implementation of Susquehannas procedure change process did not ensure that the RHR SDC procedure was maintained consistent with the requirements of plant TSs.
05000298/FIN-2015003-06	Cooper	2015Q3	Failure to Preclude Repetition for a Significant Condition Adverse to Quality	The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, associated with the inadequate extent of condition and extent of cause evaluations to preclude repetition for a significant condition adverse to quality identified in a 2012 root cause evaluation documented CR-CNS-2012- 07174 for the isolation of shutdown cooling system isolation in valves RHR-MOV-17 and RHR-MOV-18 due to localized pressure perturbations at the pressure sensors. Specifically, in 2012, the licensee failed to conduct an adequate extent of cause and condition evaluation to preclude repetition of this event from occurring on May 30, 2015 with the reactor plant in Mode 4. On May 30, 2015, isolation of shutdown cooling system isolation valves RHR-MOV-17 and RHR-MOV-18 due to localized pressure perturbations at the pressure sensors, led to the isolation of the shutdown cooling system for approximately 22 minutes. The station entered Station Procedure 2.4SDC, Shutdown Cooling Abnormal, Revision 14, and restored shutdown cooling. The reactor coolant system temperature increased approximately 20 degrees Fahrenheit but did not exceed 212 degrees Fahrenheit, maintaining the reactor plant in Mode 4. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2015-03188. The licensees failure to conduct an adequate extent of cause and condition evaluation to preclude repetition of a significant condition adverse to quality identified in a 2012 root cause evaluation documented in CR-CNS-2012-07174 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 Initiating Events Cornerstone, and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. Specifically, the failure to preclude repetition of the isolation of shutdown cooling system isolation valves RHR-MOV-17 and RHR-MOV-18 due to localized pressure perturbations at the pressure sensors led to the isolation of the shutdown cooling system for approximately 22 minutes when the reactor plant was in Mode 4 on May 30, 2015. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, inspectors determined that the finding did not require a quantitative assessment because adequate mitigating equipment remained available, and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as a very low safety significance (Green). The inspectors determined that the finding did not have a cross-cutting aspect because the most significant contributor of this finding occurred in 2012, and does not reflect current licensee performance.
05000458/FIN-2015002-01	River Bend	2015Q2	Inadequate Operating Margin for Reactor Protection System A Motor Generator Set for Overvoltage Protection Results in Loss of Shutdown Cooling	The inspectors reviewed a finding for the licensees failure to raise the overvoltage setpoint on the reactor protection system A motor generator set when the output of the generator was raised. This resulted in a reduction of the operating margin between the overvoltage trip setpoint and normal operating voltage. As a result, a spike in the output of the A motor generator on February 24, 2015, exceeded the overvoltage trip setpoint and caused the reactor protection system motor generator set output breaker to open which resulted in a loss of shutdown cooling while the reactor was shut down for refueling operations. With spent fuel in the reactor vessel, reactor coolant temperature increased 6.4 degrees until reactor protection system A was re-energized and shutdown cooling was restored. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2015-01216. The performance deficiency is more than minor, and therefore a finding, because it is associated with the Initiating Events Cornerstone attribute of configuration control, 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 tripping of the reactor protection system A motor generator set output breaker, resulted in a loss of power to the reactor protection system. This subsequently caused a loss of shutdown cooling and decay heat removal while the plant was shut down for a refueling outage. The inspectors initially screened the finding in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used NRC Inspection Manual 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 5, 2014, to evaluate the significance of the finding. The finding did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as Green. A cross-cutting aspect to this finding is not being assigned as this performance deficiency occurred in 1988 and therefore is not indicative of current licensee performance.
05000263/FIN-2015002-06	Monticello	2015Q2	Loss of Electrical Buses and Shutdown Cooling (SDC) Due to Inadequate Procedure Adherence	A self-revealed finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified due to the failure to properly implement Procedure 0304-01, Safeguard Bus Loss of Voltage Protection Relay Unit Calibration Safeguards Bus No. 15. Specifically, electrical maintenance workers failed to comply with Step 20 which directed the installation of a jumper between terminals ZX10 and ZX11 in an electrical panel, when they incorrectly installed the electrical jumper between terminals ZX11 and ZX12. This resulted in the loss of the Division I safety related 4160 Volts Alternating Current (Vac), 480 Vac, and 125 Volts Direct Current (Vdc) electrical buses, which subsequently led to the loss of shutdown cooling (SDC) for approximately 3 hours and 15 minutes. Initial corrective actions for this issue included immediately invoking strict plant status controls to focus efforts on recovery, restoring the electrical buses and SDC to operation, and reinforcing risk recognition and human performance tools. This issue was entered into the licensees CAP (CAP 1477351) and a root cause evaluation (RCE) was in progress at the time this inspection period concluded. The inspectors determined that the issue was more than minor because it adversely impacted the Initiating Events Cornerstone attribute of Human Performance and Configuration Control, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors utilized IMC 0609, Appendix G for shutdown operations and determined that the issue was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency aspect because of the failure of licensee individuals to implement error reduction tools and the failure of the organization to plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12).
05000368/FIN-2015008-02	Arkansas Nuclear	2015Q2	Failure to Correct containment Spray Pump Interlock to Shutdown Cooling Heat Exchanger Room Coolers	The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality. Specifically, the licensee failed to correct the containment spray pump interlock to automatically start the shutdown cooling heat exchanger room coolers. The licensees failure to promptly correct a condition adverse to quality as required by 10 CFR Part 50, Appendix B, Criterion XVI, was a performance deficiency. The licensee has identified in multiple instances since 1989 a degraded or nonconforming condition with shutdown cooling heat exchanger room cooler interlocks, but has failed to correct the condition. This finding was more than minor because it was associated with the design control and equipment performance attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to correct the interlock feature that automatically starts the room coolers when the pump starts. Using Inspection Manual Chapter 0609, Appendix A, the team determined that the finding was of very low safety significance (Green) because it did not result in the loss of operability or functionality of any system or train and did not screen as risk-significant in response to external events. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation because the licensee failed to thoroughly evaluate the issue to ensure that the resolution addressed the cause (P.2).
05000354/FIN-2015008-01	Hope Creek	2015Q1	Inadequate Preventive Maintenance for Safety-Related Optical Isolators in the Residual Heat Removal System	The inspectors identified a Green NCV of TS 6.8.1.a, Procedures and Programs, regarding PSEGs failure to adequately establish, implement, and justify a replacement frequency for the Residual Heat Removal (RHR) system optical isolators AT14 and AT18. These optical isolators were the most likely cause of an October 2013 RHR pump trip that resulted in a loss of shutdown cooling (SDC) during Hope Creeks R18 refueling outage. PSEG determined that the optical isolators did not have an established replacement frequency, and they had been installed since original plant construction. PSEG replaced the optical isolators and established a replacement preventive maintenance (PM) task going forward. The inspectors determined that PSEG had previous opportunity to identify the deficient PM strategy and replace the optical isolators prior to the October 2013 loss of SDC. In response to this finding, PSEG plans to conduct a causal evaluation and document the basis for their new PM frequency. This issue is more than minor because it was associated with the equipment performance 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 RHR optical isolators were determined to be the most likely cause of the B RHR pump trip and associated loss of SDC on October 17, 2013. The inspectors, with the assistance of a Region I Senior Reactor Analyst (SRA), used IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, to evaluate the safety significance of this issue. Based upon Appendix G, Attachment 1, Exhibit 2, this issue required a Phase 2 analysis, because the performance deficiency resulted in an actual loss of decay heat removal event. Using Attachment 3, Phase 2 Significance Determination Process Template for BWRs During Shutdown, Worksheet 5, the SRA determined this issue was of very low safety significance (Green). The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, which states that licensees thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. In this case, when the PCM template process was initially implemented in 2008, PSEG failed to evaluate AT14 and AT18 against the applicable PCM template (Signal Conditioner Electronic) and generate replacement PMs. Although this performance deficiency dates back to 2008, the inspectors determined the issue is reflective of current licensee performance, because PSEGs root cause evaluation (RCE) and the associated PM change request (PCR), conducted in 2013, constituted a second missed opportunity for PSEG to evaluate the applicable PCM template against the PM strategy for AT14 and AT18.
05000461/FIN-2015001-03	Clinton	2015Q1	Failure of the Division 3 Shutdown Service Water Pump due to an inadequate Bushing Design	A self-revealed finding, preliminarily determined to be of low to moderate safety significance (White) and an associated apparent violation (AV) of 10 CFR Part 50 Appendix B, Criterion III, Design Control was identified for the failure to verify the suitability of the replacement pump design for the Division 3 Shutdown Service Water system. Specifically, the licensee failed to verify the design of the suction bell bushing for the replacement pump would pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water, resulting in the failure of the pump. This finding was self-revealed on September 16, 2014, during a surveillance test to ensure operability of the Division 3 shutdown cooling water pump, after the pump failed to start due to a damaged bushing rendering the pump inoperable. This finding does not represent an immediate safety concern because the licensee replaced the pump in September of 2014 with a pump of similar design in combination with additional monitoring of pump performance and provided adequate documentation that assures the pump will remain operable until a different design can be installed by June of 2016. The inspectors determined that the licensees failure to verify the suitability of the design for the Division 3 SX replacement pump for conditions under which it was to be used, as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency. Specifically, the licensee failed to verify the design of the suction bell bushing for the replacement pump would pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water, resulting in the failure of the pump. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, 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. A Significance and Enforcement Review Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross-cutting aspect was identified with this finding. (Section 4OA3)

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