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	Site	Quarter	Title	Description
05000251/FIN-2018003-02	Turkey Point	2018Q3	Inoperable Auxiliary Feedwater Steam Supply Flow Path	A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion V, Procedures, was identified when FPL failed to ensure that the torque arm of the 4A steam generator (SG) auxiliary feedwater (AFW) steam supply valve, MOV-4-1403, remained engaged with its valve stem key. A disengaged torque arm subsequently caused the geared limit switch settings for the 4-1403 motor operator to become out of sync with the valve travel and rendered the AFW 4A SG supply flow path inoperable.
05000483/FIN-2018003-02	Callaway	2018Q3	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 6 of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for combating emergencies and other significant events. The licensee established Emergency Operating Procedure (EOP) ES-0.2, Natural Circulation Cooldown, Revision 9, in part, to meet the regulatory requirement. Figure 1 of ES-0.2 allowed cooldown rates that exceeded the values used in the license basis for radiological consequence analyses and exceeded the values used in the design of the nitrogen accumulators for atmospheric steam dumps and turbine-driven auxiliary feedwater system actuations. This issue was discussed in Licensee Event Report 2018-002-00, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown Contrary to the above, from April 29, 2008 through May 7, 2018, the licensee failed to maintain procedures for combating emergencies and other significant events. Specifically, the licensee failed to maintain EOPs for natural circulation cooldown. This performance deficiency resulted in atmospheric steam dumps and turbine-driven auxiliary feedwater systems being rendered inoperable due to depletion of the safety-related actuation nitrogen.
05000400/FIN-2018002-07	Harris	2018Q2	Minor Violation	A minor, self-revealing violation of TS 6.8.1.a, Procedures and Programs,was identified for failure to follow procedure AD-OP-ALL-0200, Clearance and Tagging. On April 7, 2018, while the plant was in Mode 3 at 0 percent power, the licensee isolated breaker DP-1A-1 circuit 28 in accordance with clearance OPS-1-18-5015-DEH MODS-0093. Isolating this breaker caused an unexpected auto start signal for both motor driven auxiliary feedwater (MDAFW) pumps for a loss of last running main feed pump despite the 1B main feedwater pump still being in operation. Both MDAFWs started and operators manually secured the 1B main feedwater pump to maintain proper feedwater flow to the steam generators. TS 6.8.1.a, requires, in part, that written procedures be implemented covering activities referenced in Regulatory Guide 1.33, Revision 2, dated February 1978, including safety-related activities carried out during operation of the reactor plant. Procedure AD-OP-ALL-0200, Section 5.5, step 4, states Clearance impacts must be evaluated to ensure that effects on systems and components outside of the boundary are identified and are acceptable, or properly dispositioned. Contrary to this requirement, the licensee did not identify that the isolation of breaker DP-1A-1 circuit 28 would cause the MDAFWs to auto start in Mode 3 when developing clearance OPS-1-18-5015-DEH MODS-0093. Screening: The violation is minor because the impact to the plant was minimal; the unit was in Mode 3 throughout the event, the reactor remained subcritical, and feedwater flow to the steam generators was not lost. Enforcement: Because the performance deficiency is minor, it will not be subject to enforcement action in accordance with the NRCs Enforcement Policy. The licensee entered this issue into their CAP as NCR 02196873. The associated LER is closed.
05000315/FIN-2018002-06	Cook	2018Q2	Minor Violation	Technical Specification (TS) 5.4, Procedures, requires that the applicable procedures recommended in Regulatory Guide 1.33 be established, implemented, and maintained. Regulatory Guide 1.33 states that maintenance that could affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with procedures appropriate to the circumstances. Contrary to this requirement, procedure 12EHP4030056218, Automatic Operation of Auxiliary Feedwater Pumps, was not performed as written in the procedure. Specifically, pages were skipped which resulted in the 2CD EDG inadvertently starting during the surveillance. Screening: The issue resulted in momentary loss of the T21C and T21D vital busses until the 2CD EDG reached rated speed and connected to the busses. The reactor was defueled at the time. One train of spent fuel pool cooling was lost for several minutes, but the other train stayed in service and there was no apparent change in spent fuel pool temperature. The issue screened as minor based on the guidance in IMC 0612 Appendix E because there were no safety consequences and there was no transient of any significance. Violation: This failure to comply with TS 5.4 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000315/FIN-2018002-05	Cook	2018Q2	Minor Violation	While there did appear to be a reduction in operational errors being made in the field while manipulating equipment (such as during clearance activities and in performing certain evolutions) the inspectors noted a trend in configuration control issues. Most of these dealt with some kind of operation department interface or coordination with another department. In one case, valves associated with feedwater heater level control were left closed following a project to replace some of the heaters, which contributed to a manual reactor trip due to high moisture-separator drain tank level when starting the plant following the Unit 2 refueling outage. Other examples were Chemistry and Operations department coordination on an non-essential service water (NESW) valve alignment which led to NESW being isolated to generator seal oil cooling during plant startup, poor coordination between Maintenance and Operations which resulted in a containment penetration being left open, a pressure gauge remaining isolated after the Projects department completed the heater drain pump replacements, and the failure to ensure that valve-closure tests were done following the feedwater heater replacements. Another identified trend was in the area of post-maintenance testing (PMT). During the refueling outage on Unit 2, both the NRC and the licensee identified instances of improper PMTs being scheduled for safety-related equipment. Inspectors identified work on an EDG fuel oil transfer pump that did not have an in-service test (IST) scheduled. The licensee identified the lack of a time response test following a motor-driven AFW pump motor replacement, was a repeat issue from the previous outage. The licensee also identified the lack of an IST following a seal replacement on a CCW pump. In each case, the issues were discovered and corrected before equipment was restored to fully operable status. In response to the trend, the licensee reviewed other work on safety-related equipment for the outage to confirm the proper PMTs would be done. No other issues were identified. Finally, early in the observation period, the inspectors noted a trend in procedure quality for maintenance activities on safety-related equipment. There were instances regarding Turbine-Driven Auxiliary Feedwater (TDAFW) pump linkages where better procedure direction could have precluded binding and governor-valve travel issues. Additionally, while replacing a TDAFW governor, a snap ring was inadvertently left out of a coupling due to insufficient procedure detail. Regarding the EDGs, the licensee discovered instructions for assembly of air start check valves did not contain the torque guidance that the vendor drawings stipulated. In response to this trend, the licensee started to perform deliberate reviews of OE before maintenance on some safety-related equipment, to verify maintenance instructions had up-to-date guidance before starting work. No violations or findings were identified by the inspectors. 12 Licensee management acknowledged the issues discussed by the inspectors.
05000348/FIN-2018002-08	Farley	2018Q2	Licensee-Identified Violation	Violation: Farley Nuclear Plant Unit 2 Technical Specifications (TS) limiting condition for operation (LCO) 3.7.5, Auxiliary Feedwater System, required all three auxiliary feedwater (AFW) trains shall be operable in modes 1, 2, and 3. For Condition A, one steam supply to turbine driven AFW pump inoperable, the required action A.1 was to restore the affected equipment to operable status within the required completion time of 7 days. If the required action and associated completion time is not met, action statement, Condition C required that the unit be in mode 3 within 6 hours and mode 4 within 12 hours. TS Surveillance Requirement (SR) 3.7.5.5 required verification that the turbine driven AFW pump steam admission valves open when air is supplied from their respective air accumulators. Contrary to the above, the licensee determined the steam admission valve (Q2N12HV3235B) was inoperable longer than the required action completion time of 7 days between May 6, 2016 and October 15, 2017, while Unit 2 was in modes 1, 2, and 3. Unit 2 was not placed in mode 3 or 4 as required by condition C of TS LCO 3.7.5. On October 31, 2017, a turbine-driven auxiliary feedwater (TDAFW) pump steam admission valve (Q2N12HV3235B) was tested with a flow scan analysis device during a refueling outage, while the plant was in Mode 6. This valve is the B-train steam admission valve that supplies steam to the TDAFW pump from the 2C steam generator. There is a redundant A-train steam admission valve that supplies steam from the 2B steam generator. During valve flow scan testing of the valve actuator it was discovered that air was leaking past the actuator piston o-ring seal inside the valve air actuator. Air leakage was measured greater than 10 psig per minute which was significant enough that the valve would not meet surveillance requirement (SR) 3.7.5.5 when instrument air was supplied solely from the valves associated air accumulator. Although the valve would stroke open with air supplied only from the accumulator, the SR 2-hour acceptance criteria to maintain the valve open could not be met. Each steam admission valve has an air accumulator associated with it. The air accumulator is designed to provide a sufficient quantity of air to ensure operation of the valve during a loss of power event or other failure of the normal instrument air supply for a period of two hours. Also, the inspectors determined that the licensee missed an opportunity to determine the cause of the o-ring failure since the o-ring was discarded during actuator rework. Procedure NMP-ES-001, Equipment Reliability Process Description, requires the preservation of physical evidence when failures occur.
05000266/FIN-2018002-02	Point Beach	2018Q2	Unanalyzed Condition for Tornado Generated Missiles	On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015, (ML15111A269) and revised on February 7, 2017, (ML16355A286). The EGM applies specifically to a SSC that is determined to be inoperable for tornado generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Point Beach, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states, in part, that SSCs, which are essential to the prevention and mitigation of nuclear accidents, shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon, such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice 53239 as an unanalyzed condition and potential loss of safety function. Enforcement discretion was previously authorized and documented in Inspection Report 05000266/2018001 (ADAMS Accession Number ML18128A229). Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The licensees longer term compensatory measure was to modify AOP13C, Severe Weather Conditions procedure, to include actions for removing potential airborne hazards and damage assessments for systems with a vulnerability to damage from tornado missiles. Corrective Action Reference: AR 2252240 Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies, was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS Accession Number ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented more comprehensive compensatory measures to address the nonconforming conditions within the required 60 days. These comprehensive actions are to remain in place until permanent repairs are completed, which, for Point Beach, were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC, such that discretion was no longer needed.On April 26, 2018, the licensee submitted a request to extend the enforcement discretion in letter titled Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15002 for Tornado-Generated Missile Protection Non-conformances Identified in Response to Regulatory Issues Summary 201506, Tornado Missile Protection. On May 21, 2018, the NRC approved this request and extended the enforcement discretion until June 10, 2020. The disposition of this enforcement discretion closes LER 201800100.
05000457/FIN-2018002-01	Braidwood	2018Q2	Inadequate Detail in Maintenance Work Instructions Resulted in Failed Gearbox Oil Cooler Head Gasket and Inoperable 2B Auxiliary Feedwater Pump	A self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to have adequate detail within their maintenance work instructions to enable proper reassembly of the 2B auxiliary feedwater (AF)pump gearbox oil cooler. Specifically, during the licensees 19th Unit 2 refueling outage in April 2017, the gearbox oil cooler closure head was reassembled following scheduled maintenance using an excessive amount of room temperature vulcanizing silicone (RTV) on the joint and an insufficient amount of torque on the closure head bolting. As a result, on March 16, 2018, the closure head joint failed causing several hours of unplanned inoperability and unavailability for the 2B AFPump.
05000455/FIN-2018002-01	Byron	2018Q2	Overspeed Trip of 2B Auxiliary Feedwater Pump During Surveillance	A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed when the 2B diesel-driven auxiliary feedwater (AF) pump tripped on overspeed during a quarterly inservice test (IST). Specifically, operators with portable instrumentation used an erroneous speed value to adjust pump speed beyond the range specified in the procedure resulting in a pump overspeed trip, entry into a 72-hour technical specification (TS) required action statement, and unplanned pump unavailability with an associated change in Unit 2 risk from green to yellow.
05000483/FIN-2018002-01	Callaway	2018Q2	Failure to Adequately Assess and Manage Risk Associated with Switchyard Work During a Planned Risk Significant Turbine-Driven Auxiliary Feedwater Pump Equipment Outage	The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at Nuclear Power Plants, for the licensees failure to adequately assess and manage risk associated with switchyard work during a planned risk significant turbine-driven auxiliary feedwater pump equipment outage. Specifically, the licensee failed to properly classify switchyard work and manage the risk as required by Procedures APA-ZZ-00322, Appendix F, Online Work Integrated Risk Management, Revision 16, and ODP-ZZ-00002, Appendix 2, Risk Management Actions for Planned Risk Significant Activities, Revision 13.
05000315/FIN-2018001-01	Cook	2018Q1	Failure of Unit 1 Turbine Driven Auxiliary Feedwater Pump to Reach Rated Speed	A self-revealed finding of very low safety significance with an associated Non-Cited Violation of Technical Specification 5.4 Procedures, occurred on December 21, 2017, when the Unit 1 Turbine-Driven Auxiliary Feedwater Pump failed to reach rated speed during a surveillance. Procedure 12MHP5021056008, Turbine-Driven Auxiliary Feedwater Pump Governor Valve Maintenance, was not appropriate for the circumstances in that direction was not given to check that the governor valve could fully open following maintenance on the governor valve.
05000369/FIN-2018010-01	Mcguire McGuire	2018Q1	Failure to Update Offsite Circuit Operability Limit for Single Busline Alignment	The inspectors identified a Green finding for the licensees failure to update calculations as required by procedure AD-EG-ALL-1117, Design Analyses and Calculations, Rev. 5. Specifically, the licensee revised calculation MCC-1381.05-00-0258, U1, 6.9kV, 4.16kV & 600V Auxiliary Power Systems Safety-Related Voltage Analysis, Rev. 6, to identify the effect of longer motor-driven auxiliary feedwater pump (CA pump) acceleration times on the switchyard voltage limits in place to ensure offsite power source operability. However, the licensee failed to update the previously analyzed condition of only one offsite circuit in service from the switchyard to the 4160V Class 1E buses via the unit step-up and unit auxiliary transformers (single busline alignment). As a result, there was no verification that the offsite circuit operability limit was adequate during single busline alignment
05000454/FIN-2018010-01	Byron	2018Q1	Failure to Prescribe Motor Driven Auxiliary Feedwater Pump Test Procedures that Accounted for the Allowed Emergency Diesel Generator Frequency Variation	The inspectors identified a Green finding and an associated Non-Cited Violation (NCV)of Title 10 of the Code of Federal Regulations (CFR),Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to prescribe motor driven auxiliary feedwater pump test procedures that accounted for the allowed emergency diesel generator frequency variation. Specifically, the motor driven auxiliary feedwater pump surveillance procedures would allow a pump with degraded and unacceptable performance to meet the test acceptance criteria based upon the test being performed at nominal frequency and not accounting for potentially lower, allowable, emergency diesel generator frequency.
05000454/FIN-2018010-04	Byron	2018Q1	Use of 10 CFR 50.54(x) for Unit AFW Cross-Tie	In 2008, the licensee added steps to Emergency Operating Procedure (EOP) 1/2BFR-H.1, Response to Loss of Secondary Heat Sink, to use the MDAFW train of a non-accident unit to combat a loss of all feedwater event in the opposite unit by using a recently installed unit cross-tie. The EOPs also directed operators to enter the technical specification LCO action statement for the unit donating the MDAFW train because the MDAFW trains were not designed and licensed to be shared between the reactor units.In 2011, the resident inspectors noted that the EOP change resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the Updated Final Safety Analysis Report because the Updated Final Safety Analysis Report described the MDAFW trains as non-shared systems. However, the licensee implemented this change without prior NRC approval. As a result, the inspectors documented a Severity Level IV NCV of 10 CFR 50.59 in Inspection Report 05000454/2011004; 05000455/2011004 as NCV 05000454/2011004-02; 05000455/2011004-02, Modification of the Auxiliary Feedwater System Without Prior NRC Approval (REF: Accession No. ML 113070678).As corrective actions to this NCV, the licensee removed the steps in the EOPs that directed the unit cross-tie to be used and removed credit for the cross-tie in the stations Probabilistic Risk Assessment model. However, on August 8, 2017, the licensee added direction in EOP1/2BFR-H.1 to use the Unit Auxillary Feedwater cross-tie by invoking 10 CFR 50.54(x). Specifically, the change added a note and a caution that provided direction to initiate the MDAFW unit cross-tie before bleed and feed. The note stated:If at any time it has been determined that restoration of feed flow to any SG is untimely or may be ineffective in heat sink restoration, then the AF crosstie should be implemented per Step 5 (Page 8). The caution stated: The AF crosstie should be implemented per Step 5 if other attempts to restore feed flow to the SG(s) will not prevent the initiation of feed and bleed. Step 5 provided instructions on how to perform the cross-tie and did not include instructions on when to initiate it. The caution also stated Use of the AF crosstie requires invoking 50.54(x).During this inspection period, the inspectors challenged the use of 10 CFR 50.54(x) to implement this permanent change. In addition, the inspectors noted that the licensees 10 CFR 50.59 screening for the procedure change did not include in its review the added note and caution statements. Because the added note and caution were the only procedure provisions that provided direction on when to use the MDAFW cross-tie, the 10 CFR 50.59 screening did not review the instructions about when to use the MDAFW cross-tie. As a result, the screening failed to determine that the change may have required a technical specification change and, thus, a license amendment as originally planned.At the end of the inspection, the NRC continued to evaluateif a performance deficiency and or violation occurred. This Unresolved Item will remain open pending the outcome of this ongoing review.
05000445/FIN-2018001-05	Comanche Peak	2018Q1	Failure to Correct a Significant Condition Adverse to Quality	The inspectors identified a Green,non-cited violation of 10CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to take corrective action for the identified cause of a significant condition adverse to quality. Specifically, a feedwater bypass control valve vibrated open resulting in a turbine trip and initiation of auxiliary feedwater. The licensee determined that the cause was an inadequate procedure for performing maintenance on the feedwater bypass control valves, but failed to correct the inadequate procedure after identifying it as the cause of a control valve failure and a turbine trip. This finding was entered into the licensees corrective action program as Condition Report CR-2018-000959.
05000483/FIN-2018001-01	Callaway	2018Q1	Failure to Maintain Emergency Operating Procedures	The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, "Procedures," for the licensee's failure to maintain emergency operating procedures for aligning auxiliary feedwater suction sources. Specifically, the licensee added continuous action steps to emergency operating procedures that placed both motor-driven auxiliary feedwater pumps in pull-to-lock and isolated their associated recirculation lines after depleting the two non-safety-related suction sources. These actions cause two of the three safety-related auxiliary feedwater pumps to be rendered inoperable prior to aligning the safety-related suction source of essential service water which is credited in accident analysis.
05000335/FIN-2018001-01	Saint Lucie	2018Q1	Improper Evaluation of LCV-9005 position setpoints Leads to AFAS	On November 19, 2013, during reactor startup activities, feedwater bypass valves, A (LCV-9005) and B (LCV-9006), were found to be operating at different throttle positions while maintaining their respective steam generator water levels. Valves LCV-9005 and 9006 were both originally installed in April 1978. LCV-9005 was replaced in 1994, with an equivalent valve, due to obsolescence. The original valve had a full open stroke length of 1.5 inches (in.), while the new equivalent valve had a full open stroke length of 2 in. to provide the same flow as the original valve. When installed, LCV-9005 was set up to limit its stroke length to 1.5 in., matching the replaced valve, and the associated drawings were never revised to show that the new valve had a full 2 in. open stroke length. In 2009, the distributed control system (DCS) was installed utilizing these drawings and was setup under the assumption that both valves, LCV-9005 and LCV-9006, were the same model valves and stroke lengths.The DCS system was designed to provide a signal to throttle the feedwater bypass valves following a reactor trip to 20 percent open to provide approximately 5 percent feed flow in order to recover steam generator water levels utilizing main feedwater. During Unit 2 startup activities in November 2013, the licensee noted a discrepancy in the valve positions for LCV-9006 and LCV-9005 when they were providing steam generator water level control. The licensee placed the issue in the corrective action program under Action Request (AR) 1921720 and determined that it was necessary to evaluate a revision of the LCV-9005 DCS setpoint, which was accomplished by an engineering condition evaluation under AR 1925428. The engineering condition evaluation was inadequate in that it failed to recognize the differences in the two different model valves, and therefore failed to provide adequate corrective actions to address performance issues associated with these differences.The final recommendation from AR 1925428 was that the current LCV-9005 setting did not impose any risk to the plant operation, as the 2A steam generator level had been within acceptable range with no control room alarm observed. Therefore, no setpoint change was required at that point.On October 26, 2017, following a Unit 2 trip, LCV-9005 was sent a digital DCS demand signal to be 20 percent open. Since the valve was locally set to have a maximum stroke of 1.5 in. instead of 2 in. open, the actual flow through the valve was less than 5 percent. This resulted in flow lower than needed to maintain 2A steam generator level, and caused level to lower, which eventually resulted in an actuation of the A train auxiliary feedwater actuation system (AFAS). Corrective Action(s):The licensee implemented corrective actions to: 1) properly set up LCV-9005 in order for it to have a full stroke length of 2 inches so that it could provide the required feedwater flow and, 2) update associated drawings to include correct stroke lengths.Corrective Action Reference(s): This issue was entered into the licensees CAP as AR 2232869
05000247/FIN-2018001-01	Indian Point	2018Q1	Failure to Incorporate Adequate Test Controls for Quarterly Stroke Close Testing of the Steam Supply Valves to Turbine-Driven Auxiliary Feedwater Pump	The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when Entergy did not assure that surveillance tests required to demonstrate that structures, systems, and components will perform satisfactorily in service are identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, during quarterly stroke testing of the steam isolation valves to the 22 turbine-driven auxiliary feedwater pump, PCV-1310A and PCV-1310B, Entergy did not ensure that these valves traveled to the closed position as required to verify that the safety function was met.
05000266/FIN-2018001-04	Point Beach	2018Q1	Enforcement Action: EA18030: Unanalyzed Condition for Tornado Generated Missiles	On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015, (ML15111A269) and revised on February 7, 2017, (ML16355A286). The EGM applies specifically to an SSC that is determined to be inoperable for tornado generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of 12 Point Beach, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states in part that SSCs which are essential to the prevention and mitigation of nuclear accidents shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice (EN) 53239 as an unanalyzed condition and potential loss of safety function. Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. Corrective Action Reference: AR 2252240 Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources - Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS Accession No. ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15-002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions are to remain in place until permanent repairs are completed, which for Point Beach were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed
05000266/FIN-2018001-05	Point Beach	2018Q1	Licensee-Identified Violation	Violation: Technical Specification (TS) 3.0.4 states in part that entry into a MODE or other specified condition in the Applicability of a limiting condition for operation (LCO) shall only be made when the LCOs Surveillances have been met... TS 3.7.5 Auxiliary Feedwater (AFW) Limiting Condition SR 3.7.5.1 required in part Verify each AFW manual, power operated, and automatic valve in each water path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. Contrary to the above, at 1500 on October 29, 2017, Unit 1 entered MODE 3 and the licensee failed to verify that AFW (System required for MODE 3) turbine driven (TD) AFW steam supply valves 1MS235 and 1MS237 were in the correct (open) position. These valves were in fact shut rendering the TDAFW pump inoperable until the licensee identified this error and opened these valves at 1610 on October 29, 2017(reference; Licensee Event Report 05000266/201700200, Operation or Condition Prohibited by Technical Specifications). Significance/Severity: This licensee identified finding, affected the Mitigating Systems Cornerstone and was screened in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, issued June 19, 2012. Because of the short duration (~1 hour) that the TDAFW pump was not operable, the inspectors determined that this finding is of very low safety significance (Green) because: the performance deficiency was not a design or qualification issue; it did not represent a loss of the system function; the train was neither inoperable for greater than its allowed outage time nor was it inoperable for greater than 24 hours; and was not part of an external event mitigating system. Corrective Action Reference: AR 02233500 Made Mode Change With Inoperable TDAFW
05000244/FIN-2018011-01	Ginna	2018Q1	Potential Preconditioning of Turbine Driven Auxiliary Feedwater Surveillance Testing	The NRC identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XI, Test Control, because Exelon established unevaluated preconditioning, with a reasonable doubt of whether the preconditioning was acceptable, prior to testing of the turbine driven auxiliary feedwater pump. This results in the loss of as-found conditions which challenge the capability of the test to assure that the turbine driven auxiliary feedwater pump will perform satisfactorily in service.
05000346/FIN-2017004-03	Davis Besse	2017Q4	Failure to Prescribe Appropriate Work Instructions for an Activity Affecting Quality	A self-revealed finding with an Apparent Violation (AV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and an associated violation of technical specification (TS) 3.7.5, Emergency Feedwater (EFW), was identified on September 13, 2017, due to the licensees apparent failure to prescribe appropriate work instructions for an activity affecting quality of the safety-related auxiliary feedwater (AFW) system. Specifically, the licensee apparently did not provide appropriate instructions to maintain an adequate amount of oil in the AFW turbine bearing oil sumps, resulting in the failure of AFW 1 on September 13, 2017. The licensee entered this issue into the CAP as CR201709443 and CR201709857, immediately replaced the damaged bearing, and updated the lubrication manual data sheets to include sight glass marking dimensions per vendor guidance. The apparent performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and potentially adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that respond to initiating events. Specifically, the apparent performance deficiency resulted in the failure of the AFW 1. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609 Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the mitigating systems cornerstone. The inspectors determined the finding represented an apparent actual loss of function of at least a single train for greater than its technical specification allowed outage time. Therefore, a detailed risk evaluation will be performed by a regional senior reactor analyst. Because the safety characterization of this finding is not yet finalized, it is being documented with a significance of to be determined (TBD). The inspectors determined this finding affected the cross-cutting aspect of challenge the unknown in the area of Human Performance, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, licensee personnel apparently did not stop when faced with uncertain conditions in the preventive maintenance procedure for replacing the AFPT sight glasses. Although the replacement of the AFPT 1 inboard bearing sight glass occurred in 1997, the licensee had the opportunity to challenge the lack of detail in the work instructions in late 2014 when the AFPT 2 outboard bearing sight glass was replaced. (H.11)
05000346/FIN-2017004-04	Davis Besse	2017Q4	Failure to Document a Degraded Condition on the AFPT 1 Outboard Bearing	The inspectors identified a finding of very low safety significance for the licensees failure to document a degraded condition of a safety-related system in the corrective action program (CAP), as required by licensee procedure, NOPLP2001. Specifically, during planned maintenance on auxiliary feedwater pump turbine (AFPT) 1, the licensee identified scoring on the outboard turbine bearing and failed to generate a condition report detailing the issue. The licensee entered this issue into the CAP as condition report (CR) 201712487 for evaluation. The inspectors determined the performance deficiency was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, the failure to document a degraded condition in the CAP did not allow the organization to properly assess the issue. Therefore, the underlying cause may not have been appropriately addressed. Using IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014, the inspectors determined the finding to be of very low safety significance (Green) because the inspectors answered no to all questions in Exhibit 3 of Appendix G, Attachment 1. The inspectors determined this finding affected the cross-cutting aspect of identification in the area of Problem Identification and Resolution, where the organization implements a corrective action program with a low threshold for identifying issues and individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to completely identify the degraded condition, resulting in the failure to document the issue. (P.2)
05000348/FIN-2017004-03	Farley	2017Q4	Failure to Follow Procedure Resulted in Inoperable TDAFW pump	A self -revealing NCV of Technical Specification (TS) 5.4.1.a, Procedures, was identified when the Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) uninterruptible power supplies (UPS) swapped to a bypass power source during maintenance on November 5, 2017. As a result, the TDAFW pump was rendered inoperable. Failure to follow licensee procedure FNP-1-EMP-1352.01, TDAFW UPS Battery Weekly Battery Inspection, Version 19, as written was a performance deficiency. The operability of the TDAFW pump UPS was restored after approximately 3 hours. The licensee entered this issue into their Corrective Action Program (CAP) as Condition Report (CR) 10427370.The finding was more than minor because it was associated with the equipment performance attribute of the mitigating system cornerstone and adversely affected that cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences since the TDAFW pump was rendered inoperable. The significance of this finding was evaluated using IMC 0609, Appendix A, The Significance Determination Process (SDP) for findings at Power, dated June 19, 2012. This finding was determined to be of very low safety significance (Green) because all of the mitigating systems screening questions were answered NO. The inspectors determined the finding had a cross-cutting aspect of Avoid Complacency in the Human Performance area because the individuals involved in this maintenance did not recognize or plan for the possibility of mistakes and appropriate error reduction tools were not implemented. (H.12)
05000247/FIN-2017010-01	Indian Point	2017Q4	Inadequate Diesel Fuel Oil Temperature Protection	The NRC identified a finding for the failure to assure that diesel powered Diverse and Flexible Coping Strategies (FLEX) equipment would be reliable to mitigate postulated beyond-design basis external events during very low temperature conditions. Specifically, at temperatures below 21F, portable FLEX equipment, such as emergency diesels, steam generator and reactor makeup pumps, and transfer pumps, were susceptible to conditions in which they would not have been capable of starting and operating due to fuel crystalizing or gelling. (CR-IP2-2017-04902/IP3-2017-05574)The failure to ensure that the portable diesel equipment could function within the required temperature range was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the finding was evaluated using NRC Inspection Manual Chapter 0609, Appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA-12-049 and EA-12-051), dated October 7, 2016, and Appendix M, Significance Determination Process Using Qualitative Criteria, dated April 12, 2012. The event of concern was determined to be a seismic event greater than 0.3g resulting in a loss of offsite power during extreme cold weather events. A bounding evaluation was performed in accordance with Step 4.1.1 of Appendix M. Indian Point declared full compliance with the order on August 12, 2016. The preliminary review of available weather conditions for the site, from the time of full compliance, shows that the temperature was below the cloud point of the fuel for over 200 hours. The Indian Point Unit 3 External Initiator Risk Informed Notebook was utilized to estimate the risk and was determined to adequately model the risk of both units. Utilizing Table 5.3.2, sequences that included emergency power, auxiliary feedwater, and high pressure makeup were evaluated. Assuming a 200 hour exposure and the unavailability of all diesel driven FLEX equipment the risk was determined to less than 1E-7/yr. Therefore, the finding was determined to have a very low risk significance. The finding had a cross-cutting aspect in the Avoiding Complacency of the Human Performance area because the licensee failed to ensure that all susceptible elements of the mitigation strategies were designed, maintained, or operated in such a manner that they could reliably function over then entire temperature spectrum for beyond-design basis external events. (H.12)
05000389/FIN-2017004-02	Saint Lucie	2017Q4	Failure to Follow Surveillance Maintenance Procedure Resulting in a Condition Prohibited by Technical Specifications	A Green, self-revealing, NCV of TS 6.8.1 was identified for the licensees failure to adequately implement a maintenance procedure during a monthly flow channel check for the 2C Auxiliary Feedwater (AFW) pump. Specifically, the licensee failed to implement as-written surveillance maintenance procedure 2-SMI-09.05C, 2C Auxiliary Feedwater Pump Flow Channel Check, when performing the channel checks for both 2C AFW pump flow transmitters. The licensees failure to follow surveillance maintenance procedure 2-SMI-09.05C, was a PD. Upon discovery, the flow transmitters were declared inoperable and subsequently, the condition was promptly restored to normal. The PD was more than minor because it was associated with the Human Performance 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 PD adversely affected the licensees ability to monitor 2C AFW flow during a design basis accident. The inspectors determined that the finding was not greater than Green because it did not represent a deficiency affecting the design or qualification of a mitigating system; it did not represent a loss of system and/or function; it did not represent an actual loss of function for at least a single train for more than its TS allowed outage time; and it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding involved the cross-cutting area of human performance, with an aspect of avoiding complacency (H.12), in that, the licensee failed to ensure that personnel effectively used human performance tools during the AFW pump flow channel check to ensure procedure steps were completed as required.
05000454/FIN-2017007-01	Byron	2017Q3	Fai lure to Perform Maintenance in Accordance with Performance Centered Maintenance Template	The inspectors identified a finding of very low safety significance and an associated NCV of TS 5.4.1, Procedures, when licensee personnel failed to perform maintenance in accordance with written procedures as required by Regulatory Guide 1.33. Specifically, from February 3, 2014, through August 25, 2017, the licensee failed to develop and execute work instructions of sufficient scope to accomplish the 3 preventive maintenance to replace flexible hoses on the essential service water (SX) makeup pumps and the diesel driven auxiliary feedwater (AFW) pumps and did not have a technical justification for a deviation from the Exelon Corporate Performance Centered Maintenance (PCM) template. The licensee entered this issue into their CAP as Action Request (AR) 03961955, AR 03971962, and AR 04045769 and planned to replace the flexible hoses at the next available opportunity. The inspectors determined that failure to perform maintenance in accordance with written procedures as required by TS 5.4.1, Procedures, and Regulatory Guide 1.33 was a performance deficiency. The performance deficiency was more than minor because it was associated with 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, failing to replace flexible hoses on the SX makeup pumps and the Unit 1 and Unit 2 diesel -driven AFW pumps at a pre - established frequency could allow hose degradation to remain unidentified and lead to the unplanned inoperability of these safety-related systems. Since the finding is a deficiency affecting the design or qualification of mitigating systems, structures and components (SSC s) and the SSC s remained operable and functional, the finding screened as having very low safety significance. This finding affected the C ross -Cutting area of Human Performance in the aspect of Work Management because the licensee failed to perform required maintenance in accordance with their associated maintenance strategy as well as the corporate PCM template (H.5) .
05000336/FIN-2017007-01	Millstone	2017Q3	Failure to Replace Auxiliary Feedwater Solenoid Valves within the Required Frequency	The inspection team identified a Green non-cited violation of Technical Specification 6.8.1.a, Procedures, because Dominion did not implement procedures as required by Regulatory Guide 1.33, Revision 2, Appendix A.9, Procedures for Performing Maintenance, to properly maintain the environmental qualification of safety-related auxiliary feedwater solenoid valves 2-FW-43AS and 2-FW-43BS. Specifically, Dominion failed to implement the recurring work event task and associated work order to ensure that these auxiliary feedwater solenoid valves were replaced prior to exceeding the qualified life of the solenoid coil and elastomer components. Dominion entered this issue into their corrective action program as condition report 1076005, planned replacement of the solenoid valves, and calculated an alternate ambient temperature for use in determining the qualified life of the solenoid valves. Dominion re-performed the qualified life calculation using this revised ambient temperature and extended the qualified life to support operability. The inspection team determined that this issue was more than minor because it adversely impacted the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This issue is also similar to more- than-minor examples 3.j and 3.k presented in IMC 0612, Appendix E, Examples of Minor Issues. Specifically, this performance deficiency resulted in a condition where there was reasonable doubt as to the operability and reliability of the solenoid valves for both auxiliary feedwater regulating valves, and thus, both trains of auxiliary feedwater. As such, Dominion needed to conduct additional engineering evaluation to extend the service life of the solenoid valves, thus justifying that the valves would continue to perform their safety function. The inspection team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the reliability of a mitigating structure, system, or component, and the structure, system, or component maintained its operability or functionality. The inspection team determined that no cross-cutting aspect was applicable because the finding was not indicative of current performance.
05000266/FIN-2017007-01	Point Beach	2017Q3	Failure to Correct a Condition Adverse to Quality Associated with a Seismic Interaction of the Motor-Driven Auxiliary Feedwater Piping	The NRC identified a finding of very-low safety significance (Green) and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to correct a Condition Adverse to Quality (CAQ) associated with a seismic piping interaction affecting the Motor Driven Auxiliary Feedwater (MDAFW) system. Specifically, the licensee identified a flange clearance to the Unit 1 MDAFW suction piping was nonconforming and captured it in the Corrective Action Program (CAP) as Action Request (AR) 01684524. However, the licensee closed the AR without correcting the CAQ. The licensee captured the inspectors concern in the CAP as AR 02212810 and performed an evaluation that reasonably concluded the MDAFW remained operable.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability determination which concluded the stresses resulting from the seismic interaction would reasonably be bounded by the applicable stress operability limits. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance because the performance deficiency occurred more than 3 years ago. Specifically, the licensee closed AR 01684524 without correcting this CAQ on September 20, 2011.
05000247/FIN-2017003-01	Indian Point	2017Q3	Component Misalignments for Nuclear Instrumentation P6 Permissive and AFW Flow Transmitter FI-1201 Following Scheduled Maintenance	A self-revealing Green NCV of Technical Specification (TS) 5.4.1, Procedures, with two examples was identified when Entergy failed to implement procedures to ensure correct system alignment for the nuclear instrumentation permissive interlock, P6, and auxiliary feedwater (AFW) flow transmitter, FI-1201. Entergy promptly corrected the alignment issues and entered them into their corrective action program (CAP) as condition report (CR)-IP2-2017-02193 for the P6 permissive interlock and CR-IP2-2017-02150 for the AFW flow transmitter. This performance deficiency is more than minor because it affects the configuration control attribute of the Mitigating System cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, in both cases, the instrumentation was left disabled following maintenance such that they could not perform their safety functions required by TSs. Additionally, the first example was similar to IMC 0612, Appendix E, example 2.g, because Entergy changed plant modes from Mode 5 to Mode 2 without ensuring P6 was operable. The second example was similar to IMC 0612, Appendix E, examples 5.a and 5.b, because Entergy failed to return the AFW flow transmitter to service after the refueling outage. The inspectors assigned a cross-cutting aspect in the area of Human Performance, Work Management, because both examples demonstrated a failure in the planning, control, and execution of work, and a lack of coordination between work groups to ensure quality.(H.5)
05000364/FIN-2017003-01	Farley	2017Q3	Failure to perform adequate corrective maintenance on the 2B EDG	The NRC identified a non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a, Procedures, for the licensees failure to implement corrective maintenance work order instructions to identify and replace piping as necessary for a degraded threaded joint on the 2B emergency diesel generator (EDG) jacket water keep warm system piping. As a result, a leak occurred at this threaded pipe joint during surveillance testing which rendered the 2B EDG inoperable. The inspectors determined that the failure to follow work order instructions to replace degraded jacket water system piping during corrective maintenance on the 2B EDG on March 3, 2017, was a performance deficiency (PD). The finding was more than minor because it was associated with the equipment reliability attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding was evaluated using IMC 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. Initial screening by the resident inspectors using the Saphire Farley 1 & 2 SPAR Model resulted in a potentially greater-than-green significance. Therefore, a detailed risk analysis was performed by a regional senior reactor analyst (SRA). The NRC Farley SPAR model was used for internal events, seismic and tornado/high winds risk estimates and the licensees Farley fire probabilistic risk assessment model was used for fire risk estimation. The major analysis assumptions included: a 51-day exposure period, EDG 2B operation at nominal failure to run probability until 8 hours when EDG assumed to fail due to the PD, PD treated as having common cause failure to run potential, no recovery of the 2B EDG was assumed, and no credit for FLEX equipment was assumed. The operation of the EDG for 8 hours prior to failure and remaining mitigating equipment limited the risk. The dominant sequence was a station blackout sequence consisting of a site-wide weather-related loss of offsite power, successful reactor shutdown, random failure to run of the 1/2A and 1C EDGs, failure of the 2B EDG due to the performance deficiency, failure to manually operate the turbine driven auxiliary feedwater pump long term, and failure to recover offsite power or an EDG leading to loss of core heat removal and core damage. The detailed risk evaluation (DRE) determined that the increase in core damage frequency due to the PD was <1.0 E-6 per year, a Green finding of very low safety significance. The finding had a cross-cutting aspect of Conservative Bias in the Human Performance area, because the decision to leave the diesel in a degraded condition following maintenance on March 3, 2017 was neither conservative nor prudent when additional action could have been taken to adequately repair or evaluate the threaded pipe joint (H.14).
05000346/FIN-2017003-02	Davis Besse	2017Q3	Auxiliary Feedwater Pump 1 Bearing Failure	A URI was identified by the inspectors relating to the final determination of the cause of the AFP 1 turbine inboard bearing failure. On September 13, 2017, the licensee performed the scheduled quarterlysurveillance test on AFP 1. This test requires the pump to run loaded with full flow of water, whereas the monthly test runs the pump only lightly loaded with water being 10 pumped through a minimum recirculation line. Within three minutes after the full flow adjustments were completed, the AFP 1 turbine inboard bearing high temperature alarm (>220 oF) actuated. The licensee verified the alarm was valid and manually tripped the AFP 1 turbine approximately 30 minutes after the alarm was received. Oil samples indicated bearing damage. The licensee disassembled the AFP 1 turbine bearing and observed bearing failure.Initial evaluation of the bearing by the licensee revealed that the damage was due to insufficient lubrication caused by low oil level. The oil level at the time of failure was within the indicated acceptable band of the oil sight glass, however, indicated band was significantly larger than the vendor recommended 3/8 inch and not at the correct height.The oil level in the sump was too low to sufficiently wet the oil slinger ring. This condition was determined to have existed since the previous pump quarterly test on June 21, 2017. After that test, a technician removed an oil sample, but did not replenish the oil. The oil level indicated low to mid band, but within the (incorrectly marked) acceptable range on the sight glass at the time. The licensee entered this issue into their CAP as CRs 201709443, 201709817, 201709527, and 201709857. Because the licensee had yet to complete their investigation and analysis of the event by the end of this inspection period, the issue is being treated as a URI pending the inspectors review of the licensees completed root cause evaluation. (URI 05000346/201700302, Final Cause Determination of Auxiliary Feedwater Turbine Bearing Failure)
05000482/FIN-2017003-02	Wolf Creek	2017Q3	Failure to Ensure the Design Basis was Adequately Represented in the Technical Specification Bases	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish adequate measures to ensure that the design bases are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee did not ensure the auxiliary feedwater system design basis was adequately represented in the Technical Specification Bases; as a result, the Technical Specification Bases and other station procedures allowed for one train of essential service water supply to the turbine-driven auxiliary feedwater pump to be removed from service without recognition that auxiliary feedwater operability was impacted. Immediate corrective actions included entering Condition Reports 113304 and 116852 into the corrective action program and incorporating a note on operations turnover documents to temporarily postpone applicable portions of the operations quarterly tasks.The licensee also completed a past operability review, and created actions to develop a license amendment request to add a specific Technical Specification condition and submit for NRC approval.The failure to ensure the auxiliary feedwater system design basis was adequately represented in the Technical Specification Bases was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is 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 (i.e., core damage). The inspectors evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, and determined this finding was of very low safety significance (Green). The inspectors determined that the finding has a problem identification and resolution cross-cutting aspect in the area of evaluation because the organization did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. This issue is indicative of current performance because the evaluation of Condition Report 111808 in May 2017 was a reasonable opportunity for the licensee to identify that the Technical Specification Bases was inadequate (P.2).
05000282/FIN-2017003-04	Prairie Island	2017Q3	Licensee-Identified Violation	Prairie Island Technical Specification 3.0.6 requires, in part, that an evaluation shall be performed in accordance with Technical Specification 5.5.13, Safety Function Determination Program, when a supported system LCO is not met solely due to a support system LCO not being met. Specifically, if a loss of safety function is determined to exist by the Safety Function Determination Program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.Contrary to this TS requirement, between August 18 and 22, 2017, control room operators did not evaluate Unit 2 A Component Cooling, Auxiliary Feedwater, and Cooling Water supported system LCOs while the 121 Safeguards Chilled Water support system LCO was not met. As a result, the appropriate Conditions and Required Actions were not entered during Unit 2 B Component Cooling and Auxiliary Feedwater supported system maintenance and testing activities for which a loss of safety function existed. Because the inspectors answered No to all questions under Exhibit 2.A of IMC 0609, Appendix A, The Significance Determination Process for Findings at-Power, the finding screened as very low safety significance (Green). Specifically, the finding did not represent (result in) an actual loss of function of two separate safety systems out-of-service for greater than their TS-allowed outage times. The above issues were documented in the licensees CAP as CAP 501000001929. Corrective actions included revisions to applicable station procedures for implementing TS 3.0.6 and the Safety Function Determination Program.
05000498/FIN-2017002-02	South Texas	2017Q2	Failure to Establish Procedures for Control of High - Energy Line Break Barriers	Green . The inspectors identified a non -cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to establish adequate procedures for the control of high -energy line break barriers. Specifically, on July 21, 2016, the inspectors identified that Procedure 0PGP03 -ZA -0514, Controlled System or Barrier Impairment, Revision 14, did not have any guidance on the control of barriers used for high -energy line breaks, despite the fact that the auxiliary feedwater pump room watertight doors are credited in the safety analyses for protection against such breaks. After discussing the acceptability of having both doors open simultaneously, the licensee shut the watertight door to auxiliary feedwater pump room for train A, and entered this condition into the licensees corrective action program as Condition Report 2016 -9006. The failure to prescribe procedures for the control of high -energy line break doors was a performance deficiency. This finding was more than minor because it was associated with the procedure quality 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, Procedure 0PGP03 -ZA -0514, Controlled System or Barrier Impairment, Revision 14, did not provide adequate procedures for the control of hazard barriers, which called the operability of the train A auxiliary feedwater 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 Quest ions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non- technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The NRC determined that this finding did not have a cross -cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance. Specifically, the auxiliary feedwater pump evaluation was performed in 2000; therefore, the performance deficiency occurred outside of the nominal 3- year period for present performance.
05000282/FIN-2017002-03	Prairie Island	2017Q2	Failure to Make an 8Hour Report Required by05000306/201700203 10 CFR 50.72(b)(3)(ii)(B)	The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.72(b)(3)(ii)(B) due to the licensees failure on March 20, 2017, to report an unanalyzed condition within eight hours of discovery. Specifically, removing the lower latch assembly of a transom above Door 225, a steam exclusion barrier, during maintenance resulted in the inoperability of the Units 1 and 2 safeguards batteries and Auxiliary Feed Water (AFW) systems, and Unit 1 safeguards bus as determined by CAP 1549724.The inspectors determined that the failure to submit a report required by 10 CFR 50.72 for the unanalyzed condition described above was a performance deficiency. The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the information that 10 CFR 50.72 reporting serves. Since the issue impacted the regulatory process, it was dispositioned through the Traditional Enforcement process. The inspectors determined that this issue was a SL IV violation based on Example 6.9.d.9 in the NRC Enforcement Policy. Example 6.9.d.9 specifically states, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. Because the issue has been evaluated under the Traditional Enforcement process, there was no cross-cutting aspect associated with this violation.
05000336/FIN-2017001-01	Millstone	2017Q1	Failure to Maintain CST Temperature in Accordance with Procedural Requirements	Green. The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement Operating Procedure (OP) 2319B, Condensate Storage and Surge System. Specifically, Dominion failed to maintain the Millstone Unit 2 condensate storage tank (CST) temperature above procedural requirements. Dominion has documented this condition within their corrective action program (CAP) as condition report (CR) 1066291. The inspectors determined this finding was more than minor as it adversely affected the protection from external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The reliability of the mitigating systems heat removal function was challenged based upon the reasonable doubt of lost operability of the CST to provide a sufficient supply of water to the auxiliary feedwater (AFW) system. There was reasonable doubt of lost operability due to indications of CST water temperature below OP 2319B prescribed limitations, winter temperatures falling, and an inability to restore CST recirculation system in a timely manner. The finding was determined to be of very low safety significance (Green), when all screening questions were answered No as the conditions discussed in the Dominion engineering evaluation, approved on January 7, 2017, were capable of showing that no safety systems or functions were lost. This finding has a crosscutting aspect in the Problem Identification and Resolution, Resolution, in that Dominion did not take effective corrective actions or corrective maintenance to address CST recirculation pump degradation in a timely manner, prior to the onset of winter, commensurate with their safety significance such that operations could maintain CST water temperature above procedurally defined limitations. (P.3)
05000529/FIN-2017001-01	Palo Verde	2017Q1	Failure to establish station procedure instructions for denial work authorizations	The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the failure to establish procedure instructions for work authorization denials or deferrals. Specifically, this led to a 60 day extended unavailability of the diverse auxiliary feedwater actuation system when corrective maintenance was inappropriately deferred by the operations department. Failure to provide adequate procedural guidance in the event of a denied work authorization, a circumstance anticipated to occur, is a performance deficiency. The performance deficiency is more than minor, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability and reliability of equipment that responds to an initiating event. Specifically, because the corrective maintenance was not performed in a timely manner, both trains of the diverse auxiliary feedwater actuation system remained in bypass for an additional 60 days whereby the system was not capable of performing its required safety function. The inspectors evaluated the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, Section A, Question 2, which required a detailed risk evaluation because the finding involved a loss of system safety function. A Region IV senior reactor analyst performed a detailed risk assessment of the finding and determined that the finding was of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect in the human performance area of Work Management. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the Unit Operations Managers decision to deny the work authorization was based on conservative but faulty assumptions, and if other work groups with greater specific technical knowledge had been involved, the corrective maintenance likely would have proceeded (H.5)
05000424/FIN-2016007-06	Vogtle	2016Q4	Turbine Driven Auxiliary Feedwater (TDAFW) Pumps 1/2-1302- P4-001 and Motor Driven Auxiliary Feedwater (MDAFW) Pumps 1/2-1302-P4-002/003	The NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control for the licensee's failure to translate the Auxiliary Feedwater (AFW) pumps design bases into adequate acceptance criteria for technical specifications SR 3.5.7.2 and for the failure to verify the adequacy of the design of the same AFW pumps. The licensee entered the violation into the corrective action program as condition reports 10293456 and 10294168. As an immediate corrective action, the licensee evaluated the operability of the Unit 1 and 2 AFW pumps, modify the allowed diesel frequency acceptance criteria, and initiated corrective action to develop new acceptance criteria and monitor pump performance for degradation. The performance deficiencies were more-than-minor because they were 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 (i.e., core damage). Specifically, when the quality of the established surveillance criteria was considered, there was a reasonable doubt on the operability of the Unit 1 and 2 turbine driven AFW and 2A and 1B motor driven AFW pumps. The team determined the finding to be of very low safety significance (Green) because it did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time. The team determined that the finding had a crosscutting aspect in the Human Performance area of Design Margins (H.6), because engineers did not demonstrate the characteristic of ensuring that design margins were guarded and changed only through a systematic and rigorous process.
05000315/FIN-2016009-02	Cook	2016Q3	Licensee-Identified Violation	The licensee identified a finding of very-low safety significance (Green) and associated NCV of License Conditions 2.C.4 and 2.C.3.o for Units 1 and 2 respectively for the licensees failure to establish an appropriate Monitoring Program in accordance with NFPA 805, Section 2.6. Section 2.6 of NFPA 805 required, in part, that monitoring shall ensure that the assumptions in the engineering analysis remain valid. Contrary to the above, the licensee failed to ensure that the assumptions in the engineering analysis remained valid for the availability and reliability of the auxiliary feedwater pumps in the Monitoring Program. The licensee used Maintenance Rule availability criteria to monitor the auxiliary feedwater pumps which did not bound the Fire PRA assumptions for the unavailability of the components. The performance deficiency was determined to be more-than-minor because the issue adversely impacted the Mitigating Systems cornerstone objective to ensure the capability of systems that respond to initiating events and prevent undesirable consequences due to external events such as fire. Specifically, the failure to adequately monitor plant equipment credited for post-fire SSD could result in that equipment being unavailable for longer periods of time than had been analyzed. The inspectors screened the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Fire Protection Significance Determination Process. Since the reactor was still able to reach and maintain a SSD condition, the finding screened as very-low safety significance (Green). The licensee entered the issue into the CAP as AR 2016-7239, NFPA 805 Monitoring Program FPRA\Maintenance Performance, and revised Maintenance Rule administrative procedures to consider the unavailability criteria of components and the impact on the fire PRA.
05000391/FIN-2016011-04	Watts Bar	2016Q3	Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 2 AFW Pumps	The NRC identified a SL IV NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate the available net positive suction head to the Unit 2 auxiliary feedwater pumps. These issues were entered into the licensees corrective action program as condition report 1196925. The licensee confirmed current operability and had determined that likely corrective actions will include revisions to the net positive suction head calculation. The performance deficiency was determined to be more than minor because it represented an inadequate quality oversight function that, if left uncorrected, could adversely affect the quality of the analysis of a safety related SSC. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 57%. The team determined this finding to be of very low safety significance, SL IV, because it represented a failure to meet a regulatory requirement, including one or more Quality Assurance criteria that had more than minor safety significance. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance.
05000390/FIN-2016011-02	Watts Bar	2016Q3	Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 1 AFW Pumps	The NRC identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate the available net positive suction head to the Unit 1 auxiliary feedwater pumps. These issues were entered into the licensees corrective action program as condition reports 1196925 and 1201623. The licensee confirmed current operability and had determined that likely corrective actions will include revisions to the net positive suction head calculation. The performance deficiency was determined to be 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 safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 74%. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating SSC that maintained its operability. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance.
05000369/FIN-2016003-01	McGuire	2016Q3	Licensee-Identified Violation	Technical Specifications 5.4.1.a, Procedures, requires, in part, that procedures for certain activities recommended in Regulatory Guide 1.33, Rev. 2, Appendix A, be established, implemented, and maintained. Administrative procedures for shift and relief turnover is one of the identified activities. Administrative procedure AD-OP-ALL-1000, Conduct of Operations, Rev. 4, implements the licensees shift and relief turnover standards. This procedure requires shift turnovers to contain detailed information on equipment and system status, alignments, and activities, to ensure watchstanders have a complete understanding of plant status. Contrary to the above, from August 10 to August 13, 2015, operators were not aware of the required nuclear service water system alignment which required a continuous vent (passing water flow) to be maintained in the condenser cooling water (RC) suction supply to the Unit 1 turbine driven auxiliary feedwater pump. The continuous vent mitigates the potential for air entrainment in the RC piping high point and is needed in order for the standby shutdown system to be functional during an Appendix R fire event when the suction of the turbine driven auxiliary feedwater pump is transferred from the auxiliary feedwater storage tank to the long term water supply provided by the RC system. This lack of operator awareness stemmed from a misunderstanding in the operator turnovers that the nuclear service water system was in a standby nuclear service water pond cooling alignment, which does not require the continuous vent to be maintained. The discrepancy was subsequently identified by oncoming shift operations personnel and the continuous vent was re-established on August 16, 2015, after removing material that obstructed the continuous vent line. As a result of not maintaining the continuous vent at the suction of the turbine driven auxiliary feedwater pump, the standby shutdown system was rendered non-functional for a period of eleven days, which was in excess of the 7-day limit allowed by Selected Licensee Commitments 16.9.7. This violation was determined to be of very low safety significance (Green) because it only affected the non-safety related Appendix R water supply to the turbine driven auxiliary feedwater pump. This violation was entered into the licensees corrective action program as NCR 01943414.
05000348/FIN-2016002-04	Farley	2016Q2	Condition Prohibited by Technical Specifications Due to Turbine Driven Auxiliary Feedwater Design Issue	On November 20, 2015, the Unit 1 TDAFW pump over sped and tripped on startup during surveillance testing. The licensee determined a previous design change that adjusted the MPU override timer setting from 10 seconds to 600 seconds resulted in a governor controller speed set point conflict that revealed itself during the performance of the test. The licensee implemented the set point change after consulting with the vendor as a corrective action to address previous issues with the setting of the MPU override timer. This new failure mode was not anticipated when the TDAFW pump governor controller MPU timer setpoint was changed in April 2015 on Unit 1, and January 2015 on Unit 2. After the modification was made on each unit, several successful starts were performed to validate the setpoint adjustment before the pumps were returned to service. Additionally, the TDAFW pumps had been successfully started 19 times on Unit 1 and 15 times on Unit 2 for surveillances, post-maintenance testing, and troubleshooting while the condition existed. Once discovered, the licensee implemented another design change to adjust the low idle speed setpoint to minimize the potential for turbine speed overshoot on startup. Enforcement: Farley Unit 1 and 2 Technical Specification (TS) limiting condition for operation (LCO) 3.7.5, Auxiliary Feedwater (AFW) System, required three operable AFW trains while the Unit is in modes 1, 2 or 3. With one AFW pump train inoperable, LCO 3.7.5. Condition B required restoration of the AFW train to operable status within 72 hours and within 10 days from discovery of failure to meet the LCO. Contrary to this requirement, Unit 1 operated from May 3, 2015, until November 22, 2015, with the Unit 1 turbine driven AFW inoperable. Unit 2 operated from January 10, 2015, until November 22, 2015, with the Unit 2 turbine driven AFW inoperable. A regional senior reactor analyst (SRA) performed a detailed risk evaluation to evaluate the risk increase associated with the condition. No failures occurred on Unit 2, therefore the condition did not result in a risk increase for Unit 2. The evaluation for Unit 1 was performed using the NRC Farley SPAR model with input from the licensees NFPA 805 Fire PRA model for the fire external event risk. The major analysis assumptions for Unit 1 included a 200 day exposure interval, recovery credit for local manual overspeed trip reset evaluated using the NRC SPAR-H human reliability analysis method, and an overspeed trip startup failure probability determined from plant specific data. The dominant risk sequence was a total loss of service water resulting in a plant trip and failure of the motor driven auxiliary feedwater (MDAFW) pumps, with failure of the turbine driven AFW pump due to the overspeed trip condition on startup with failure of the operator to accomplis overspeed trip reset which would lead to loss of core heat removal and core damage. The result of the detailed risk evaluation was an increase in risk due to the condition of <1.0 E-6/ year. The inspectors concluded that the violation was of very low safety significance (Green) and consistent with a Severity Level IV violation. The NRC exercised enforcement discretion (Enforcement Action EA-16-159) for this violation in accordance with sections 2.2.4.d and 3.5 of the NRCs Enforcement Policy because the impact of the design change was not within the licensees ability to foresee and correct beforehand. The inspectors reached this conclusion due to the number of successful TDAFW pump starts following implementation of the design change and the specific vendor recommendation to adjust the MPU override timer setting to greater than 30 seconds. This issue was entered into the licensees corrective action program as CR 10149716.
05000275/FIN-2016301-01	Diablo Canyon	2016Q2	Insufficient Procedural Direction Contained Within Procedure EOP E-2, Faulted Steam Generator Isolation	The examiners identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, Procedure EOP E-2, Faulted Steam Generator Isolation, does not contain sufficient procedural direction for isolating auxiliary feedwater flow to a faulted steam generator in the event that auxiliary feedwater control valves cannot be closed from the control room. Procedure EOP E-2, Appendix HH, Isolated Faulted Steam Generator, Step 1.d, and its associated column, Response Not Obtained, does not ensure that a faulted steam generator would remain isolated under all conditions. The Response Not Obtained column permits operators to either locally close auxiliary feedwater control valves OR secure the auxiliary feedwater pump feeding the faulted steam generator. However, due to the absence of pull-to-lock or hard stop switches for the auxiliary feedwater pumps, the possibility exists for an automatic restart of an auxiliary feedwater pump and a re-initiation of feedwater to a faulted steam generator. The failure to ensure that Procedure EOP E-2 contained sufficient direction to isolate a faulted steam generator when auxiliary feedwater flow control valves cannot be closed from the control room was a performance deficiency. This performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute of the Barrier Integrity cornerstone (reactor coolant system and containment) and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the re-initiation of feedwater to an isolated, faulted steam generator has the potential to adversely affect the reactor coolant system barrier by causing an additional unintended cooldown of the reactor coolant system, increased potential for pressurized thermal shock, and thermal stress to the steam generator u-tubes. Additionally, the containment barrier would be affected by the reinitiation of feedwater to a steam line break within containment. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the team determined that the finding required a detailed risk evaluation due to the potential to affect the reactor coolant system boundary. A senior reactor analyst performed a bounding detailed risk evaluation and estimated the maximum increase in core damage frequency to be 5.9E-8/year, and therefore the finding was determined to be of very low safety significance (Green). This increase in core damage frequency was mitigated by the low probability of multiple equipment failures in the auxiliary feedwater system when combined with the low initiating event frequency of a faulted steam generator. Because the violation was of very low safety significance (Green) and the issue was entered into the licensees corrective action program as Notification 50847218, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the Enforcement Policy: NCV 05000275/2016301; 05000323/2016301-01, Insufficient Procedural Direction Contained Within E-2, Faulted Steam Generator Isolation. This finding has a crosscutting aspect in the area of human performance associated with resources because the organization did not ensure procedures are available and adequate to support nuclear safety (H.1).
05000346/FIN-2016002-02	Davis Besse	2016Q2	Failure to Use the Corrective Action Program to Evaluate and Document Degraded Condition with Auxiliary Feedwater Train 2	An NRC-identified finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified for the licensees failure to have entered a degraded condition associated with Auxiliary Feedwater (AFW) Train No. 2 into their CAP until challenged by the inspectors. Specifically, a flow transient that occurred on May 7, 2016, and that caused damage to components in the AFW recirculation line during AFW Train No. 2 testing was not entered into the licensees CAP until May 8, 2016, following challenges from the inspectors. This omission on the part of the licensees staff had the effect of bypassing certain features of the licensees CAP associated with evaluating and documenting the operability of safety-related equipment. The physical event and equipment issues were entered into the licensees CAP as CR 201606515 on May 8, 2016, following prompting by the inspectors. Corrective actions taken by the licensee included repairs to all damaged equipment, detailed inspections of AFW Train No. 2, and an engineering analysis into why the event occurred. The matter of the licensees failure to enter the event into their CAP in a timely manner was documented as CR 201606516, with corrective actions including the coaching and counseling of personnel involved regarding the proper use of the CAP. This finding was of more than minor safety significance because it affected the equipment performance attribute of the Mitigating Systems cornerstone of reactor safety and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the units AFW system. The finding was determined to be of very low safety significance because it did not represent a deficiency affecting the design or qualification of a mitigating system, structure, or component (SSC); it did not, in and of itself, represent a loss of system and/or function; it did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, or two separate safety systems being out-of-service for greater than their TS allowed outage times; and it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution. The inspectors assigned the cross-cutting aspect of Identification to the finding because the licensees staff failed to identify the issue with AFW Train No. 2 within their CAP completely, accurately, and in a timely manner in accordance with program requirements. (P.1)
05000368/FIN-2016002-01	Arkansas Nuclear	2016Q2	Failure to Incorporate Vendor Guidance in Work Order	The inspectors identified a finding for the failure to incorporate vendor instructions in a work order. Specifically, the licensee exceeded the vendor specified torque values and performed the work with the component in service, contrary to vendor cautions, breaking the glass, wetting the auxiliary feedwater pump, and necessitating the unplanned shutdown of the main feedwater pump. The licensee replaced the ruptured sight glass and repaired and tested the wetted components. The licensee documented the issue in Condition Report CR-ANO-2-2015-04832. The failure to incorporate vendor instructions in a work order is a performance deficiency. The finding is more than minor because it adversely affected the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the performance deficiency resulted in the Unit 2 auxiliary feedwater pump and main feedwater pump B being rendered unavailable. The inspectors evaluated the finding with NRC 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 inspectors determined that the finding required a detailed risk evaluation because the finding involved an actual loss of function of auxiliary feedwater and one train of main feedwater, designated as having high safety significance in accordance with the licensees maintenance rule program, for greater than 24 hours. A senior reactor analyst performed a detailed risk evaluation and determined that the increase in core damage frequency was 1.3E-7/year (Green). The analyst assumed that all feedwater pumps were available until the time of the leak and that any increase in core damage frequency resulted from the unavailability of the pumps after the leak. The emergency feedwater system remained available to mitigate the increase in core damage frequency of this finding. The inspectors determined this finding has a cross-cutting aspect in the human performance area of Work Management because the primary cause of the performance deficiency involved the failure to identify and manage risk commensurate to the work and the need for coordination with different groups or job activities (Section 1R12). (H.5)
05000456/FIN-2016002-04	Braidwood	2016Q2	Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance	On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection, focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced enforcement guidance memorandum (EGM) 15002, which was also issued on June 10, 2015. The EGM provided guidance to allow the NRC staff to exercise enforcement discretion when an operating power plant licensee did not in comply with the current license basis for tornado-generated missile protection. Specifically, the discretion would be applied to structure system and components (SSCs) declared inoperable resulting in TS LCOs that would require a reactor shutdown or mode change if the licensee could not meet the required actions within the TS completion time. The discretion allowed the licensee to reestablish operability through compensatory measures and established criteria for continued operation of the facility as longer term corrective actions were implemented. This allows the licensee to continue operating until final corrective actions are taken in the timelines established in the EGM. The EGM stated that the bounding risk analysis performed for this issue concluded that this issue was of low risk significance and, in Braidwoods case, provided for enforcement discretion of up to 3 years from the date of issue of the EGM. However, the EGM does not provide the licensees enforcement discretion for any related underlying technical violations; and moreover, the EGM specifically requires that any associated underlying technical violation be assessed through the enforcement process. Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC), Criterion 4, Environmental and Dynamic Effects Design Basis, stated in part that SSCs important to safety shall be adequately protected against dynamic effects including missiles. On May 25, 2016, the licensee initiated IR 02673854, identifying a nonconforming condition of Criterion 4. Specifically, multiple locations were identified in the refueling water storage tank (RWST) roof hatches and in the L-line wall above the 451 elevation (separating the turbine building from the Class I auxiliary building) where SSCs were not adequately protected from tornado-generated missiles. The licensee declared multiple SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The inspectors reviewed the licensees compensatory measures that included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados/high winds, and potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The condition was reported to the NRC as Event Notice 51959 as an unanalyzed condition and potential loss of safety function. The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); 19 TS 3.3.7, Control Room Ventilation (VC) Filtration System Actuation Instrumentation; TS 3.5.2, ECCS Operating; TS 3.5.4, Refueling Water Storage Tank (RWST); TS 3.6.6, Containment Spray and Cooling Systems; TS 3.7.5, Auxiliary Feedwater System; TS 3.7.10, Control Room Ventilation (VC) Filtration System; TS 3.7.11, Control Room Ventilation (VC) Temperature Control System; TS 3.8.4, DC Sources Operating; TS 3.8.7, Inverters Operating; and TS 3.8.9, Distribution Systems Operating. The inspectors review addressed the material issues in the plant, and whether the measures were implemented in accordance with the guidance documentation for the EGM. The inspectors also evaluated whether the measures as implemented would function as intended and were properly controlled. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions were to remain in place until permanent repairs were completed, which for Braidwood were required to be completed in 3 years, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed. The inspectors did not review the underlying circumstances that resulted in the TS violations. As stated in the EGM guidance, violations of other requirements, including 10 CFR 50 Appendix A Criterion 4, that may have contributed to the TS violations would be evaluated independently of the EGM implementation. This operability inspection constituted a partial sample as defined in IP 71111.1505 since all the corrective actions to support continued operability and resolution of the nonconforming conditions had not been identified. These actions and any underlying technical violations will be addressed with the completion of this inspection sample.
05000446/FIN-2016002-01	Comanche Peak	2016Q2	Failure to Correct Conditions Adverse to Quality	The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to correct a condition adverse to quality in safety-related equipment. Specifically, following an in-service testing failure of auxiliary feedwater check valve 2FW-091 in November 2012, the licensee performed an operability evaluation of the auxiliary feedwater system. However, the inspectors identified that the licensee failed to take corrective action to address the condition adverse to quality that resulted in the valve failing to seat properly. Consequently, the same valve failed a subsequent inservice test in November 2015. Following discovery of this issue, the licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The licensee entered this issue into corrective action program as CR-2015-10961. The licensees failure to correct a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to correct auxiliary feedwater check valve 2FW-0191 failure to seat in November 2012 resulting in an additional failure in November 2015. Using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation, in that, the licensee failed to thoroughly evaluate issues to ensure that resolutions address extent of conditions. Specifically, the licensee failed to appropriately classify the issue of the check valve not seating and recognize this as a degraded condition (P.2).
05000391/FIN-2016002-08	Watts Bar	2016Q2	Failure to Follow Maintenance Procedure Results in overspeed trip of the 2C-S Turbine Driven Auxiliary Feedwater Pump	A self-revealed Severity Level (SL) IV non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified at Watts Bar Unit 2 for the licensees failure to follow procedure 0-MI-1.003, Disassembly, Inspection, and Reassembly of Auxiliary Feedwater Pump Turbine. Specifically, the valve stem spring coil gap was not set in accordance with procedure, causing the turbine-driven auxiliary feedwater (TDAFW) pump to trip on electrical overspeed when the level control valves (LCVs) were closed. This issue was corrected on May 30, 2016, when the proper spring coil gap was set and verified and the post maintenance test was performed satisfactorily. The issue was entered into the licensees corrective action program as CR 1175968. The performance deficiency was more than minor because it represented an improper or uncontrolled work practice that could impact quality or safety involving safety-related structures, systems, and components (SSCs). The finding was a SL IV violation because it represented a failure to meet a regulatory requirement, specifically a quality assurance (QA) criteria to follow quality-related procedures, which had more than minor safety significance. The finding was assigned a crosscutting aspect of resources in the Human Performance area because the licensee failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the procedure that set the coil spring gap lacked sufficient detail and rigor to ensure that the coil gap would be set appropriately by the technicians.

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