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05000277/FIN-2018003-03Peach Bottom2018Q3Reactor Core Isolation Cooling System Pressure Switch Failure Results in Condition Prohibited by TS - EA-18-108On April 22, 2018, during a routine surveillance test of the RCIC system, the RCIC turbine tripped approximately 28 seconds after startup, prior to the system reaching rated flow and pressure. Concurrent with the RCIC trip, an alarm was received for RCIC turbine high exhaust pressure; however, local indications did not indicate a true high pressure in the exhaust line. Therefore, the RCIC system was declared inoperable and TS 3.5.3, Condition A was entered, which requires the RCIC system to be restored to operable within 14 days. Troubleshooting determined that the B RCIC exhaust pressure switch (PS-3-13-72b) had prematurely tripped at normal operating pressure due to an age-related failure of the instrument diaphragm and O-ring. The RCIC system had been previously verified as operable during its last surveillance run on January 16, 2018. Corrective Actions: The failed pressure switch was replaced and the station performed an extent of condition review/inspection of similar pressure switch instruments. Following replacement of the switch, RCIC was retested and restored to operable on April 23, 2018. Furthermore, actions were established to modify the turbine trip logic to remove the single point trip vulnerability. Corrective Action Reference: IR 4129583 Enforcement:Violation: Peach Bottom Unit 3 TS 3.5.3 requires that the RCIC system shall be operable in Mode 1, and if RCIC becomes inoperable, it shall be returned to operable status within 14 days or the plant shall be placed in Mode 3 within the next 12 hours. Contrary to the above, based on relevant causal information, Unit 3 RCIC was likely inoperable prior to April 22, 2018, for a period greater than the TS allowed outage time of 14 days, and Unit 3 had not been placed in Mode 3. Specifically, on April 22, 2018, the Unit 3 RCIC turbine tripped during startup for a routine surveillance test due to a degraded turbine exhaust pressure switch which resulted in an inoperability time of greater than 14days. Internal inspection on the switch identified that it failed due to corrosion from water intrusion which had existed for an extended period of time. Severity/Significance: For violations warranting enforcement discretion, IMC 0612 does not require a detailed risk evaluation; however, safety significance characterization is appropriate. A Region I SRA performed a best estimate analysis of the safety significance using the Peach Bottom Unit 3 Standardized Plant Analysis Risk (SPAR) model, Version 8.51 and Systems Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE), Version 8.1.8. This model was used to evaluate the internal events increase in core damage frequency (CDF) per year. The SRA performed a site visit to review Exelons fire model output to estimate the external risk contributor of the issue. The final risk evaluation estimated the total (internal and external events risk) increase in CDF to be in the mid E-6/yr range, or of low to moderate safety significance. The SRA evaluated the internal and external events risk contribution due to the inoperability of the RCIC system for an assumed 47 day exposure time. 16 The analyst used the guidance in the Risk Assessment Standardization Project (RASP) Handbook, Volume I, Section 2.4, Revision 2.0, to estimate an exposure time using a time divided by two (t/2) approach. This would represent the time from the last successful surveillance test divided by two. The approach is appropriate for periodically operated components that fail due to a degradation mechanism that gradually could affect the component during the standby period. Given this approach, the internal event contribution was calculated to estimate the internal event risk increase due to the conditional failure of the RCIC pump to successfully start. The increase for internal events was estimated at 2.5E-6/yr increase in CDF. The dominant sequence involved a loss of condenser heat sink, with operator action failure to depressurize, and HPCI system failures. The SRA noted from discussions with Exelon staff that the RCIC system was assumed to be non-recoverable given the nature of the failure. To estimate the external risk contribution, the SRA had several discussions and a site visit to review Exelons preliminary fire model outputs for the conditional failure of the RCIC system for the 47 days. The 47 days included a conservative additional day for repair time. The SRA reviewed Exelons fire risk analysis and noted that one of the dominant risk increase contributors was fire within the 13kV switchgear room. Several other fire areas were reviewed and the SRA noted that the core damage sequences appeared technically reasonable given the plant areas and values assumed for mitigating equipment. Exelons preliminary results showed an increase in external event CDF/yr for the conditional failure of RCIC for 47 days to be approximately 4.5E-6/yr. The SRA determined the results to be reasonable. Exelons model for internal events resulted in an increase in CDF/yr of 1.05E-6/yr which was considered to compare well with the NRC SPAR model. Exelon performed a review of the large early release frequency (LERF) impact and determined an overall increase in LERF due to both external and internal events for the RCIC failure for 47 days to be a nominal 6E-8/yr. Therefore, the SRA review of the dominant sequences and Exelons LERF results affirmed that LERF did not increase the risk over that determined from the increase in CDF. Basis for Discretion: The inspectors determined that the maintenance strategy for these switches was consistent with requirements and standards that existed at the time and that there was no relevant operating experience that would have reasonably necessitated consideration of additional maintenance actions. As a result, no performance deficiency was identified. The inspectors assessment considered: The industry, regulatory, and Exelon service life standards were reviewed for static O-ring pressure switches. Exelons assessment of the pressure switch service condition (critical, mild conditions, low-duty cycle) required a preventive maintenance task to perform periodic calibration and to replace the switch as-required. There was no time-based replacement task prescribed by any standard for this switch. The inspectors determined that Exelons assessment was adequate and the corresponding preventive maintenance activities met applicable standards. The subject pressure switch was installed during original construction and the calibration results of the pressure switch had been satisfactory from 2003 until the 2018 failure. The inspectors reviewed the maintenance and calibration history on the pressure switch and did not identify any adverse trends or conditions adverse to 17 quality that would have required further evaluation or replacement of the pressure switch. Industry operating experience information available to Exelon did not identify the potential for the age-related failure mode of the pressure switch o-ring and diaphragm that occurred at Peach Bottom. The NRC determined that it was not reasonable for Exelon to have been able to foresee and prevent this violation of NRC requirements, and as such, no performance deficiency existed. Therefore, the NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.10 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TSs (EA-18-108). Further, because Exelons actions did not contribute to this violation, it will not be considered in the assessment process or the NRC Action Matrix
05000416/FIN-2018002-03Grand Gulf2018Q2Failure to Adequately Test NUS Temperature SwitchA self-revealed,Green non-cited violationof 10CFRPart50, AppendixB, CriterionIII, Design Control, was identified when the reactor core isolation cooling (RCIC) system automatically isolated due to an inadvertent high temperature input from the leakage detection system. Specifically, the licensee failed to fully test a modification that installed a new type of temperature switches, and the system inappropriately isolated the RCIC system when a loss and subsequent restoration of power occurred.
05000220/FIN-2018001-01Nine Mile Point2018Q1Potential Failure to Submit an 8-Hour Event Notification for a Valid Actuation of HPCOn March 18, 2018,at 1:18 a.m., during the Unit 1maintenance outage while the unit was in cold shutdown, operators received multiple low level alarms on the GEMAC 11 and 12 level indications. Operators responded by adjusting reactor water cleanup reject flow and the feedwater minimum flow control valve to raise reactor water level. Upon the operators making the adjustment to reactor water level, the feedwater low flow control valve was slow to respond, but eventually opened more rapidly, and the increased flow from feedwater resulted in a rapid rise in reactor water level. At 1:28 a.m., indicated reactor water level rose to the 95-inch trip setpoint for the 11 and 12 Yarway level indication instruments, resulting in a turbine trip and HPCI initiation signal. The HPCI pumps were tagged out and thus did not inject, and the turbine was offline for the shutdown. The 11 and 12 Yarway level indication instruments provide reactor protection system logic inputs for reactor vessel water level; however, the Yarway level indication instruments are not density compensated. Therefore, under cold shutdown conditions, actual reactor vessel water level was lower than indicated water level on the 11 and 12 Yarways. During cold shutdown conditions, the GEMAC level instruments, which are calibrated to cold shutdown conditions, provide an accurate indication of actual reactor vessel water level. The GEMAC instruments both indicated well below the trip setpoint of 95 inches (indicated ~72 inches) when the turbine trip and HPCI initiation signal were received. Exelon determined that this event was not reportable under 10 CFR 50.72.Title 10 CFR 50.72(b)(3)(iv)(A) states, Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. (B) The systems to which the requirements of paragraph (b)(3)(iv)(A) of this section apply are: 10 (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system. Planned Closure Action(s): The inspectors requested the 10 CFR 50.72 subject matter experts at the Office of Nuclear Reactor Regulation (NRR) and Office of General Council (OGC) to review whether this was a valid actuation and thus reportable. The inspectors are opening an unresolved item (URI) to determine if a performance deficiency exists.Licensee Action(s): Licensee entered the concern into their corrective action program, and communicated with NRC Region I and NRR Staff. Exelons position is that the event was not reportable. Corrective Action Reference:IR 04116336 NRC Tracking Number: 05000220/2018001-01
05000461/FIN-2018001-01Clinton2018Q1Failure to Follow Procedure Results in Unplanned Reactor Core Isolation Cooling UnavailabilityA self-revealed Green finding and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified when the licensee failed to follow station procedure Clinton Power Station (CPS) 9030.01C034, RCIC (Reactor Core Isolation Cooling) Steam Line Flow E31N683A(B), E31N684A(B), Checklist. Specifically, the licensee failed to reset the isolation logic for the RCIC steam line outboard isolation valve prior to turning on the breaker for this valve. This resulted in the isolation of the steam supply to RCIC causing RCIC to become unavailable,and elevating the plant risk to Yellow.
05000440/FIN-2017008-01Perry2017Q4Failure to Address the Susceptibility of the Condensate Storage TankLow Level Instrument Lines to FreezeThe team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations(CFR),Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.63, Loss of All Alternating Current Power, for the licensees failure to evaluate the capability to transfer the high pressure core spray (HPCS)and the reactor core isolation cooling (RCIC) pumps suction source from the condensate storage tank (CST)to the suppression pool during cold weather conditions. Specifically, (1) monitoring of the CST level instrument lines heat tracing was inadequate to detect a credible common mode failure before the instrument lines would freeze and be rendered inoperable during normal operation, (2)the licensee did not address the condensate (CST) level instrument lines susceptibility to freeze during a cold weather loss of off-site power (LOOP) event with or without a design basis transient or accident, and (3)the licensee incorrectly evaluated the capability to transfer the HPCS pump suction source from the CST to the suppression pool during a cold weather station blackout (SBO) event. The licensee captured the issues within their Corrective Action Program (CAP) as Condition Report(CR) CR-2017-08685, CR-2017-08930, and CR-2017-09006. Corrective actions implemented included: increased the CST level instrument line heat tracing circuit monitoring frequency, revised the affected procedures ensured HPCS and RCIC are adequately aligned to the suppression pool during LOOP design basis events, and ensured a timely transfer of the HPCS and RCIC pump suctions to the suppression pool during a SBO. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability,reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.Specifically, the failure of the HPCS and or RCIC pumps to automatically transfer their suction source from the CST to the suppression pool upon reaching a low CST water level condition could damage the pump(s) thus preventing them to be used to mitigate a transient or accident. A detailed risk evaluation was performed and determined that the finding was of very-low safety-significance (Green). The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the CST instrument lines were designed and the SBO coping strategy during cold weather was established more than 3 years ago.
05000461/FIN-2017007-01Clinton2017Q3Failure to Evaluate Defeating Reactor Core Isolation Cooling System Interlocks and Trips before Adding Them to an Emergency Operating Support ProcedureThe inspectors identified a Severity Level IV NCV of Title 10 Code of Federal Regulations (CFR) 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the change to an emergency operating procedure (EOP) support procedure to incorporate three reactor core isolation cooling (RCIC) system interlock defeats and did not provide a basis for the determination that this change would not create a possibility for a malfunction of a structure, system or component (SSC ) important to safety with a different result than any previously evaluated in the updated safety analysis report. The licensee entered this issue into the CAP as action request ( AR ) 04056394 and planned to perform a screening for the procedure change. 3 This performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the change did not ensure RCIC system reliability and availability during and following design basis accidents because it introduced a new failure mode and added reliance on monitoring activities and manual actions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it 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. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV significance because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The team determined that this finding had a cross -cutting aspect of Resources in the area of Human Performance because the licensee did not ensure that procedures, and other resources were available and adequate to support nuclear safety. Specifically, the procedure which required a 50.59 screening for changes to EOP support procedures, was not explicit in requiring the screening. (H.1)
05000416/FIN-2017003-01Grand Gulf2017Q3Isolation of Reactor Core Isolation Cooling System during Surveillance TestingThe inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish quality related activities in accordance with Surveillance Procedure 06-IC-1E31-A-1004, RCIC Equipment Room High Temperature Calibration Channel A, Revision 106. Specifically, on August 21, 2017, the licensee did not follow Step 5.15.4, which states, Identify and disconnect field lead located at Terminal EE-50 in 1H13-P632. This step was not performed correctly; therefore, the reactor core isolation cooling (RCIC) system isolation feature was not bypassed. When performing the next step, an inadvertent isolation of the RCIC system occurred. On August 21, 2017, the licensee restored compliance by performing actions to restore the leads to the correct location and performing the surveillance test satisfactorily. This issue has been entered into the licensees corrective action program as Condition Report CR-GGN-2017-08246.The failure to follow Surveillance Procedure 06-IC-1E31-A-1004 was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to follow Surveillance Procedure 06-IC-1E31-A-1004 resulted in unplanned inoperability and unavailability of the reactor core isolation cooling system. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating System Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; did not represent a loss of safety function; did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time, and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined that the finding had a field presence cross-cutting aspect within the human performance area because licensee management failed to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, the performer in the field was a supplemental worker that was observed by a licensee instrumentation and controls technician. The technician telephoned the supervisor to ensure that they were performing the steps correctly, and the supervisor did not go into the field to verify the step was performed correctly (H.2).
05000353/FIN-2017003-01Limerick2017Q3Operational Condition Mode Change from Startup to Run was Made with RCIC InoperableThe inspectors identified a Green NCV of Unit 2 technical specification (TS) 3.0.4, when Exelon changed the operating condition of Unit 2 from mode 2 (startup) to mode 1 (run) with reactor core isolation cooling ( RCIC ) inoperable for surveillance testing. Specifically, the TS 3.7.3 limiting condition for operation (LCO) for RCIC was not met, a mode change from startup to run was made, and none of the allowances, TS 3.0.4.a, TS 3.0.4.b, or TS 3.0.4.c, were met to allow the mode change in that condition. Exelon entered this issue into the corrective action program with issue report (IR) 4057128. The inspectors determined that the change in operating condition of LGS Unit 2 from startup to run with RCIC inoperable was reasonably within Exelons ability to foresee and correct and should have been prevented and therefore was a performance deficiency. This finding is more than minor because it adversely affected the configuration control attribute of the mitigating systems cornerstone to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, RCIC was inoperable during the time it was required to be operable, i.e. the mode change from startup to run. Additionally, this finding was similar to example 2.g of IMC 0612, Appendix E, in that a mode change was made without all required equipment being operable. Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the finding did not represent a loss of function and did not represent the loss of a single train for greater than technical specification allowed outage times or greater than 24 hours. The inspectors determined that this finding has a cross - cutting aspect in the area of Human Performance, Documentation, because with respect to TS 3/4.7.3 Exelon did not create and maintain complete and accurate documentation of the correct usage of TS 3.0.4 that was more fully explained in the applicable safety evaluation. (H.7)
05000458/FIN-2017007-02River Bend2017Q2Failure to Perform an Adequate Operability Determination for a Condition Identified During an NRC WalkdowGreen. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, between June 15, 2017, and June 28, 2017, the licensee failed to address the operability of a terminal block installed within an unsealed junction box. In response to this issue the licensee performed an operability determination to ensure that the terminal block would perform its design function in this condition. This finding was entered into the licensees corrective action program as Condition Report CR-RBS-2017-05084. The team determined that the failure to perform an adequate operability determination was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure operability of valve E51-AOVF054 and its associated circuits would impact the operability of the reactor core isolation cooling system. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee failed to perform an adequate operability determination for an identified condition (P.3)
05000278/FIN-2017002-01Peach Bottom2017Q2Corrective Action Not Implemented Correctly for Replacement of RCIC RCR ContactsA self-revealing non-cited violation (NCV) of 10 Code of Federal Regulation(CFR)Part 50, Appendix B, Criterion XVI, Corrective Actions, of very low safety significance (Green) was identified for Exelon not correcting a condition adverse to quality concerning reverse control relay (RCR) contacts for valves associated with the reactor core isolation cooling (RCIC) system. Specifically, Exelon specified a corrective action (CA) from an October 18, 2013, Unit 3 RCIC equipment apparent cause evaluation (EACE) to replace RCR contacts after 12 years of service, however, the CA was not correctly implemented. As a result, on January 12, 2017, an RCR contact associated with the Unit 3 RCIC suppression pool suction valve remained in service for 15 years, exhibited a high resistance failure during a surveillance which resulted in Unit 3 RCIC being inoperable. Following the failure, Exelon initiated issue reports (IRs) 03962563 and 03977949, implemented corrective actions to replace the RCR contact, restored Unit 3 RCIC operability, and risk-informed their corrective maintenance schedule for replacing all RCR contacts that currently exceeded the recommended 12-year service life.Exelons failure to recognize and correct a condition adverse to quality associated with certain RCR contacts in their Unit 3 RCIC system that had exceeded their 12-year service life, was a performance deficiency (PD) that was within their ability to foresee and correct and should have been prevented. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstones objective to ensure the reliability of systems to respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, not recognizing that existing RCR contacts were installed in safety-related equipment beyond their 12-year service life, resulted in the failure of the Unit 3 RCIC suppression pool suction valve. The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, SDP for Findings At-Power, and determined the finding was of very low safety significance (Green) because it did not represent a loss of system function or represent an actual loss of function of at least a single train for longer than its technical specification (TS) allowed outage time of 14 days. The inspectors determined that the finding has a cross-cutting aspect in Human Performance, Procedure Adherence, because Exelon did not validate that the correct revision of procedure WC-AA-120, Attachment 2, Preventive Maintenance (PM) Change Review Form, was used when creating a new PM to replace RCR contacts. (H.8)
05000254/FIN-2017002-04Quad Cities2017Q2Failure to have Adequate Guidance in the Fire/Explosion Response ProcedureThe inspectors identified a finding of very low safety significance and an associated non-cited violation of TS Section 5.4.1.c, Procedures, for the licensees failure to establish and maintain the fire response procedure. Specifically, Procedure QCOA 001012 Fire/Explosion, Revision 47, failed to provide adequate instructions to ensure that the reactor core isolation cooling (RCIC) system would not be potentially affected by a single spurious operation of any of its associated valves in the event of a fire in Fire Area TBII. The licensee entered the issue into their CAP as IR 2595878 and planned to revise the affected procedures.The performance deficiency was determined to be more-than-minor because it impacted the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire), and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the lack of adequate procedural guidance in the fire response procedure did not ensure a single spurious operation would not potentially impair the operation of RCIC system in the event of a fire in TBII. The finding was screened using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet. The inspectors determined the finding required a detailed risk evaluation by a Senior Reactor Analyst. The finding screened as very low safety significance because the calculated total Delta Core Damage Frequency (CDF) was 9.5E7/yr per the detailed risk evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency.
05000254/FIN-2017002-01Quad Cities2017Q2Failure to Establish aProcedure Appropriate for Calibration of RCIC GovernorA finding of very low safety significance and an associated non-cited violation of Technical Specification (TS) Section 5.4.1 was self-revealed for the licensees failure to establish a procedure as governed by Regulatory Guide 1.33, Revision 2, Appendix A that was appropriate for performing adjustments to the governing control system for the Unit 1 reactor core isolation cooling (RCIC) system. Specifically, on April 14, 2017, the licensee failed to ensure procedure QCIPM 130004, RCIC Woodward Governor EGM Control Box and Ramp Generator/Signal Convertor in Field Calibration, was appropriate for the accurate calibration of the RCIC system turbine governor actuator such that the system would be capable supplying its TS required flowrate of 400 gallons per minute (gpm). Immediate corrective actions included the licensee declaring the Unit 1 RCIC system inoperable and performing required calibrations at normal operating temperatures and pressures. Additional corrective actions included the licensee making procedural revisions to QCIPM 130004 to include specific guidance on performing turbine governor calibration adjustments and providing training to maintenance control system technicians on performing the procedure tasks and other related tasks that led to the inadequate adjustment. The issue was entered into the licensees CAP as IR 3998478.The performance deficiency was determined to be more than minor, and a finding,because it impacted the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective because the failure to properly calibrate the RCIC governor led to the system becoming inoperable. The inspectors determined the finding could be evaluated using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, and determined that the finding required a detailed risk evaluation by a senior reactor analyst (SRA) because it resulted in the loss of the RCIC system function. A SRA performed a detailed risk evaluation of the performance deficiency using the Quad Cities SPAR Model and determined the total Delta Core Damage Frequency (CDF) was 7E9 (Green). The inspectors determined this finding affected the cross-cutting area of Human Performance, in the aspect of Training, because the licensee failed to ensure the technicians performing the calibration understood null voltage adjustments to the RCIC turbine governor could only be performed when the system was at a specified rated speed and pressure (H.9).
05000458/FIN-2017009-01River Bend2017Q2Failure to Obtain Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection PointGreen. The NRC identified a Severity Level IV violation for the licensees failure to restore compliance for a non-cited violation (NCV) associated with failure to obtain NRC approval prior to making a change to the reactor core isolation cooling injection point. Specifically, as of April 28, 2017, the licensee had not restored compliance with a violation the NRC identified on October 8, 2015. This violation described a previously made change to the facility without prior NRC approval in violation of 10 CFR 50.59, Changes, Tests, and Experiments. The team determined that the licensees failure to restore compliance within a reasonable amount of time was a performance deficiency. Title 10 CFR 50, Appendix B, Criterion XVI, requires in part that, measures shall be established to assure that conditions 3 adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2017-03505. The finding was more than minor because it is associated with the initiating events aspect of the reactor safety cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The finding is of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a human performance cross-cutting aspect associated with procedural adherence because individuals failed to follow the procedures delineated by the corrective action program (H.8). Originally, the licensee met the criteria for dispositioning the issue (50.59) as a NCV. However, based upon the fact that the condition report, which documented the NCV, was closed without restoring compliance, the licensee no longer met the criteria for a NCV and therefore, this violation is being cited in a notice of violation
05000298/FIN-2017001-06Cooper2017Q1Failure to Install Correct Mechanical Stop and Verify Proper OperationThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 3.0.4 for the licensees failure to install the correct reactor core isolation cooling pressure control valve, RCIC-AOV-PCV23, mechanical stop and verify proper operation of the system prior to entering a mode of applicability for Technical Specification 3.5.3. This condition resulted in RCIC-AOV-PCV23 going fully open during surveillance testing following Refueling Outage 29, causing a pressure transient. This transient caused a failure of the reactor core isolation cooling turbine lube oil cooler gasket, lifting of a pressure relief valve, and a water leak. The licensee immediately shut down the reactor core isolation cooling system and declared it inoperable. The immediate corrective actions were to restore RCIC-AOV-PCV23 from the closed mechanical stop to the required open mechanical stop and to replace the turbine lube oil cooler gasket to restore operability of the system. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-08122 and initiated a root cause evaluation to investigate this condition. The licensees failure to install the correct reactor core isolation cooling pressure control valve, RCIC-AOV-PCV23, mechanical stop and verify proper operation of the system prior to entering a mode of applicability for Technical Specification 3.5.3, in violation of Technical Specification 3.0.4, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee installed RCIC-AOV-PCV23 with the incorrect mechanical stop, and proper valve operation was not verified after installation during Refueling Outage 29, which caused the reactor core isolation cooling system to lose function during surveillance testing. This transient caused a failure of the reactor core isolation cooling turbine lube oil cooler gasket and an associated water leak. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding required a detailed risk evaluation because it represented a loss of system and/or function. In the detailed risk evaluation, the analyst assumed the reactor core isolation cooling system was unavailable for 50 hours. The analyst used the Test/Limited Use Version COOPER-DEESE-HCI03 of the Cooper SPAR model run on SAPHIRE, Version 8.1.5. The analyst updated the initiating event frequencies for transients, losses of condenser heat sink, losses of main feed water, grid related losses of offsite power, and switchyard centered losses of offsite power to the more recent values from the 2014 update to the industry data found in INL/EXT-14-31428, Initiating Event Rates at U.S. Nuclear Power Plants, 1998-2013, Revision 1. From this, the finding was determined to have an increase in core damage frequency of 8.4E-8/year and to be of very low safety significance (Green). Transients, losses of condenser heat sink, and losses of main feed water were the dominant core damage sequences. The automatic depressurization system and the reactor protection system remained to mitigate these sequences. The finding had a cross-cutting aspect in the area of human performance associated with documentation because the licensee failed to create and maintain complete, accurate, and up-to-date documentation associated with RCIC-AOV-PCV23 design drawings and the maintenance procedure for setting and testing the mechanical stop (H.7).
05000387/FIN-2016007-01Susquehanna2016Q4Failure to Specify and Maintain Safety-Related Quality Standards and Materials Essential for Reactor Core Isolation CoolingThe team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the failure to classify and maintain reactor core isolation cooling (RCIC) system components as safety-related as specified by Updated Final Safety Analysis Report Table 3.2-1 and Section 7.1.1. Specifically, although Talen, the operator of Susquehanna Steam Electric Station, classified the RCIC system as safety-related, this classification did not extend to the Unit 1 and Unit 2 RCIC barometric condenser relief valves. The team determined failure of the non-safety related barometric condenser relief valves could result in a loss of RCIC lube oil cooling and failure of RCIC to perform its design basis safety function. Talen entered the issue into the corrective action program as condition report 2016-23615 and performed an immediate operability determination, which concluded RCIC remained operable. The finding was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2 - Mitigating System Screening Questions. The team determined the finding screened as very low safety significance (Green), because the finding was a design deficiency which did not result in an actual loss of functionality of the RCIC system. This finding was not assigned a cross-cutting aspect because the performance deficiency occurred during original plant design and did not reflect current licensee performance.
05000373/FIN-2016004-03LaSalle2016Q4Failure to Perform Preventive Maintenance Resulting in Two Subsequent Unit 1 RCIC Turbine Trips During Surveillance TestingA finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to perform preventive maintenance on the Unit 1 reactor core isolation cooling (RCIC) electronic governor-remote (EGR) actuator. Specifically, from June 4, 1993, to November 17, 2016, the licensees processes for the control and administration of preventive maintenance failed to ensure that the Unit 1 RCIC EGR actuator was replaced or refurbished on an interval that would prevent internal fouling of the EGR actuator from adversely affecting governor performance. As a result, contaminates and degradation accumulated in the EGR actuator from January 16, 2004, to November 17, 2016, ultimately causing the RCIC turbine to trip during quarterly surveillance testing on October 18, 2016, and again on November 17, 2016. The licensee entered this issue into the CAP as ARs 02729757 and 02742254. Corrective actions planned and completed included replacing the Unit 1 and Unit 2 RCIC EGRs and performing a root cause evaluation of the degraded condition. The performance deficiency was more than minor, and thus a finding 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 (i.e., core damage). Specifically, the failure to perform preventive maintenance on the Unit 1 RCIC EGR resulted in a degraded condition which adversely affected the reliability of the system to respond to an initiating event. A detailed risk evaluation determined that the finding screened as having very low safety significance (Green). This finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current licensee performance.
05000397/FIN-2016004-02Columbia2016Q4RCIC Trips After SurveillancesGreen. The inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design of the reactor core isolation cooling (RCIC) system. Specifically, in 2001, the licensee implemented a design change to the keep-fill pump, RCIC-P-3, that changed its operation from continuous to intermittent, and did not verify the adequacy of the design for all methods of operation, including surveillance testing. Placing the RCIC-P-3 pressure switch downstream of the steam-driven RCIC pumps discharge check valve allows a subsequent hydraulic transient to depressurize RCIC piping below the systems low pressure trip set point. This failure to provide design control measures resulted in RCIC tripping three separate times when RCIC-P-3 was unable to keep up with hydraulic transients. In response to this condition, the licensee changed their operation of the keep-fill pump to running continuously and initiated Action Request 352594 to address long-term issues such as procedure revisions and system design changes. The failure to verify the adequacy of design of the RCIC system was a performance deficiency. Specifically, in 2001, the licensee implemented a design change to the keep-fill pump, RCIC-P-3, that changed its operation from continuous to intermittent and did not verify the adequacy of the design for all methods of operation, including surveillance testing. The performance deficiency was more than minor, and therefore a finding, because it affected the design control attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this modification was inadequate, resulted in RCIC tripping three separate times when RCIC-P-3 was unable to keep up with hydraulic transients, and required compensatory measures to prevent future trips. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors did not identify a cross-cutting aspect for this issue. Specifically, the design change occurred approximately 15 years ago and does not represent current licensee performance.
05000331/FIN-2016003-01Duane Arnold2016Q3Failure to Satisfy 10 CFR 50.72 and 10 CFR 50.73 Reporting Requirements for a Condition that Could Have Prevented Fulfillment of a Safety FunctionThe inspectors identified a Severity Level IV NCV of 10 CFR Part 50.72(a)(1) and 10 CFR Part 50.73(a)(1) due to the licensees failure to make a required 8hour non-emergency notification and a 60 day Licensee Event Report to the NRC after discovering a loss of safety function for the reactor core isolation cooling (RCIC) system. The licensee documented this issue in the CAP as CR 02156273 and planned to perform a causal evaluation for the failure to recognize the reportable condition. The inspectors previously evaluated the RCIC systems loss of safety function under the SDP as a finding of very low safety significance (Green) as documented in Section 1R22.b of NRC Integrated Inspection Report 05000331/201600201 (ML16221A619). Violations of the NRCs reporting requirements are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy and determined the violation associated with the failure to report was a Severity Level IV Violation because the previously evaluated loss of safety function was determined to be a Green finding under the SDP. No cross cutting aspect was assigned to this issue due to the issue being a traditional enforcement violation.
05000387/FIN-2016008-01Susquehanna2016Q3Failure to Write a Condition Report for Degraded Conditions Which Challenged Operability of Safety Related EquipmentThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Susquehanna failing to identify and correct conditions adverse to quality in a timely manner. Specifically, between April 16, 2016 and April 22, 2016, condition reports for potential or suspected degraded or non-conforming conditions related to the High Pressure Coolant Injection System (HPCI) and Reactor Core Isolation Cooling System (RCIC) were not written and operability determinations performed. In both cases, the equipment was subsequently declared inoperable due to the conditions. The issues were entered into the CAP and the equipment was taken out of service, repaired, and retested satisfactorily. The inspectors determined that there were two examples of the same performance deficiency and violation. In accordance with NRC Enforcement Manual Section 1.3.4, Documenting Multiple Examples of a Violation, multiple examples of a single violation are allowed to be documented as a single violation bounded by the characterization of the most significant example. The RCIC example is considered the most significant due to the longer exposure time in a required mode and number of mode changes that occurred during the exposure period. The finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to identify and correct degraded conditions associated with a RCIC system lube oil leak which rendered that system inoperable. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that this finding screened to Green because the safety function was not lost, and the finding did not represent an actual loss of function of at least a single train for greater than its Tech Spec Allowed Outage Time or two separate safety systems out-of-service for greater than its Tech Spec Allowed Outage Time. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork, because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, in both examples, individuals were aware of potential degraded conditions but actions were not taken to communicate the activity to other groups, such as the control room operators, to allow for the issues to be evaluated for operability and determine if proposed actions were timely and/or appropriate. (H.4) (Section 4OA2.1.c(1))
05000354/FIN-2016007-02Hope Creek2016Q2Inadequate Testing of the Remote Shutdown Panel RCIC Flow Control Circuit (The team identified a finding of very low safety significance, involving a noncited violation of Hope Creek Technical Specification (TS) Surveillance Requirement (SR) 4.3.7.4.2, "Remote Shutdown System Instrumentation and Controls." Specifically, PSEG did not adequately test all components of the Reactor Core Isolation Cooling (RCIC) flow control circuit on the RSP to demonstrate operability. This finding was more than minor because it was similar to example 3.k of Inspection Manual Chapter (IMC) 0612, Appendix E, and was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the RCIC system. The inspectors evaluated this finding using IMC 0609.04, "Initial Characterization of Findings," and IMC 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions." This issue was determined to be of very low safety significance (Green) because it did not represent an actual loss of function of a single train mitigating system for greater than its TS Allowed Outage Time. The finding did not have a cross-cutting aspect because it was a legacy issue and was not considered indicative of current licensee performance.
05000354/FIN-2016002-01Hope Creek2016Q2Inadequate Maintenance Rule Monitoring of Multiple Systems, including the Effluent Radiation Monitoring System and the Reactor Core Isolation Cooling SystemThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(2) due to an inadequate maintenance rule (MR) monitoring of the effluent radiation monitoring system (RMS) and the reactor core isolation cooling (RCIC) system. Specifically, PSEG did not properly evaluate maintenance rule functional failures (MRFFs) for both systems in accordance with its Maintenance Rule Program (MRP). Consequently, unaccounted for maintenance preventable functional failures (MPFFs) in both the effluent RMS and RCIC systems caused each system to exceed their MR performance criteria, requiring (a)(1) evaluations. PSEGs corrective actions (CAs) include placing the effluent RMS system in (a)(1) monitoring status and establishing monitoring goals, evaluating the RCIC system for (a)(1) monitoring status, and performing procedure revisions of affected procedures. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with both the Plant Facilities/Equipment and Instrumentation attribute of the Public Radiation Safety cornerstone (effluent RMS) and the Equipment Performance attribute of the Mitigating Systems cornerstone (RCIC). The inspectors determined that this finding was of very low safety significance (Green) using: IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008; and, Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. This finding was associated with a cross-cutting aspect of Human Performance, Consistent Process, which states that individuals use a consistent, systematic approach to make decisions. Specifically, PSEG did not to properly evaluate the impact of equipment failures in the effluent RMS and RCIC system when making MRFF determinations. (H.13)
05000458/FIN-2016002-01River Bend2016Q2Failure to Follow Station Guidance on Use of Temporary Power Cables and Control of Transient CombustiblesThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to follow station maintenance procedures related to the use of temporary power cables and storage of transient combustible materials in the auxiliary building. Specifically, the licensee installed energized networking equipment and an associated power cable within one foot of a safety-related cable tray. The station did not initially correct the problem, but later resolved the deficiencies by removing the networking equipment and power cable. The failure to initially correct the issue is documented as a violation in Section 4OA2 of this report. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2016-02398. The licensees installation of energized networking equipment and an associated power cable within one foot of a safety-related cable tray was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a fire resulting from this energized equipment would impact the availability, reliability, and capability of the low pressure core spray system, residual heat removal system, component cooling primary system, and reactor core isolation cooling system. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings. Since the finding involved a failure to adequately implement fire prevention and administrative controls for transient combustibles, the inspectors dispositioned the finding using NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. In accordance with Manual Chapter 0609, Appendix F, Question 1.3.1.A, the inspectors determined that the finding was of very low safety significance (Green) because the reactor would be able to reach and maintain safe shutdown since the safe shutdown path was deemed independent of fire damage state scenarios for the given fire ignition source. The finding had a cross-cutting aspect in the area of human performance, work management, because the licensees work management processes failed to plan, control, and execute the work activity that included installation of temporary equipment such that impacts on nuclear safety were properly evaluated and addressed (H.5).
05000410/FIN-2016002-02Nine Mile Point2016Q2Failure to Identify Wide Range Level Indication Impacts Operability of HPCS and RCICThe inspectors identified a Green NCV of Unit 2 Technical Specification (TS) 3.5.1, Emergency Core Cooling (ECCS) Systems-Operating, and TS 3.5.3, Reactor Core Isolation Cooling (RCIC) System, for failure to ensure all necessary attendant instrumentation required for the systems to perform their specified safety functions were capable of performing their related support function in all require modes of applicability. Specifically, the inspectors identified the Unit 2 wide range level indication to be inaccurate during Mode 2 and at 200 pounds per square inch gauge (psig) reactor pressure, a mode of applicability requiring both high-pressure core spray (HPCS) and RCIC to be operable. This resulted in a high level trip signal being locked preventing HPCS or RCIC from auto initiating, rendering the systems inoperable. Upon identification, Exelon generated issue report (IR) 02667837 to address the inspectors concern regarding the wide range level indication. An action was created to evaluate the impact of the wide range level discrepancy with regard to its impact on safety-related functions to supply water in the TS Mode of Applicability. Exelon also plans to assess the need for a TS amendment. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelon failed to recognize that the wide range level indication did not provide accurate indication at low reactor pressures and temperatures, preventing automatic safety-related functions associated with high drywell pressure automatic initiation signals and manual start functions. This would require operators to manually open the HPCS and RCIC injection valves during these conditions should a loss of offsite power or loss-of-coolant accident occur. 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, the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification. Exelon personnel had many opportunities, including during the reactor startup in May of 2016, to question operability of the instrumentation that provides input for automatic initiation and isolation signals. As a result, Exelon personnel failed to identify that the wide range level indication did not support operability of the HPCS and RCIC systems during reactor startup on May 5, 2016. (P.1)
05000458/FIN-2016002-03River Bend2016Q2Failure to Identify and Correct Improperly Stowed Transient CombustiblesThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, after writing a condition report identifying energized networking equipment and an associated power cable that had been installed within one foot of a safety-related cable tray, the licensee closed the condition report without removing the networking equipment and power cable. The licensee entered this issue into their corrective action program as Condition Reports CR-RBS-2016-02398 and CR-RBS-2016-03152. Corrective actions included removing the networking equipment and power cable and conducting a performance management review of the actions involved with correcting the condition and closing the condition report. The licensees failure to promptly identify and correct a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct a known deficient condition resulted in an extended period of vulnerability to a fire that could result from improperly installed energized equipment and challenge the availability, reliability, and capability of the low pressure core spray system, residual heat removal system, component cooling primary system, and reactor core isolation cooling system. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings. Since the finding involved a failure to adequately implement fire prevention and administrative controls for transient combustibles, the inspectors dispositioned the finding using NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. In accordance with Manual Chapter 0609, Appendix F, Question 1.3.1.A, the inspectors determined that the finding was of very low safety significance (Green) because the reactor would be able to reach and maintain safe shutdown since the safe shutdown path was deemed independent of fire damage state scenarios for the given fire ignition source. The finding had a cross-cutting aspect in the area of human performance, teamwork, because the licensee failed to properly communicate expectations to individuals performing work during the course of implementing corrective actions (H.4).
05000373/FIN-2016007-04LaSalle2016Q2Alternate Shutdown Procedures Failed to Ensure RCIC MOVs Supply Breakers Were ClosedThe team identified a finding of very-low safety significance (Green) and associated NCV of the LaSalle County Station Operating License for the failure to ensure that procedures were in effect to implement the alternate shutdown capability. Specifically, the abnormal operating procedures (AOPs) established to respond to a fire at the main control room did not include instructions for verifying that supply breakers for three reactor core isolation cooling motor-operated valves (MOVs) were closed to ensure they could be operated from the remote shutdown panel. Fire-induced failures could result in tripping MOV power supply breakers prior to tripping the MOV control power fuses. The licensee captured the team concerns in their CAP as AR 02668854 and established compensatory actions to reset the affected breakers, if required The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external events (fire), 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 was assigned a low degradation factor. Specifically, the procedural deficiencies could be compensated by operator experience/familiarity and the fact that the AOPs included steps to verify other breakers at the same locations were closed would likely prompt operators to close the remaining breakers. The team determined that this finding had a cross cutting aspect in the area of problem identification and resolution because the licensee failed to take effective corrective actions for a similar issue identified in 2014. Specifically, the resolution of this issue included actions to revise the affected AOPs to include verifying all the reactor core isolation cooling MOVs supplied breakers were closed. However, the licensee failed to include all of the MOVs in the revised AOPs. (P.3)
05000341/FIN-2016001-09Fermi2016Q1Inadvertent Reactor Water Low Level Reactor Protection System Actuation Due to Operator ErrorA finding of very low safety significance with an associated NCV of Technical Specification (TS) 5.4, Procedures, was self-revealed when a valid automatic reactor scram signal and isolation signal for multiple primary containment isolation valves was actuated. A reactor operator, who was maintaining RPV water level and reactor pressure following a plant scram, did not initiate reactor core isolation cooling (RCIC) system flow in time to maintain level above the Level 3 reactor protection system actuation setpoint. As an immediate corrective action, control room operators promptly restored RPV level by manual operation of the RCIC system. The licensee entered this issue into the corrective action program and provided remedial training for the reactor operator in the simulator, communicated lessons learned from this event with other licensed operators, and subsequently implemented improvements for licensed operator training and procedure changes to incorporate a revised strategy for manual control of RPV level and pressure control with main steam line isolation valves closed. The performance deficiency was of more than minor safety significance because it was associated with the Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the human performance error unnecessarily challenged a plant protection feature, which resulted in a valid automatic reactor scram signal and isolation signal for multiple primary containment isolation valves. In addition, the finding was sufficiently similar to Example 4(b) in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," for not of minor safety significance since the error resulted in a valid automatic reactor scram signal and isolation signal for multiple primary containment isolation valves. The finding was determined to be of very low safety significance since it did not cause a reactor scram and a loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition (e.g., loss of condenser, loss of feedwater). The inspectors concluded this finding affected the cross-cutting aspect of resources in the human performance area. Specifically, the licensees evaluation identified the reactor operator had been performing a complicated task for a long period of time without adequate rest/recovery periods (IMC 0310, H.1).
05000260/FIN-2016001-01Browns Ferry2016Q1Unacceptable Preconditioning of RCIC Valve Prior to ASME In-Service TestingAn NRC identified finding (FIN) for failure to meet TVA procedure NETP-116.3, Inservice Testing Program Preconditioning Guidelines, because unacceptable preconditioning of the Unit 2 Reactor Core Isolation Cooling (RCIC) steam supply valve occurred prior to quarterly In-Service Test (IST). Specifically, the preconditioning was unacceptable because the testing sequence was avoidable, it masked the actual asfound condition of the valve, and it could possibly result in an inability to verify the operability of the valve. As an immediate corrective action, the licensee performed an evaluation that determined the valve remained operable. The finding was entered into the licensee's corrective action program as CR 1159463 . The performance deficiency was more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Additionally, if left uncorrected, the performance deficiency could lead to a more significant safety concern. Specifically, the licensees justification of this particular preconditioning event could be applied to justify additional, avoidable, preconditioning events and possibly result in an inability to verify the operability of components. This finding was evaluated in accordance with NRC IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined the finding was Green because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, did not result in a loss of function of a single train for greater than its TS allowable outage time, did not result in a loss of function of non-TS equipment, and did not involve the loss of equipment or function specifically designed to mitigate an external event. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance area of Consistent Process (H.13), because individuals did not complete the required preconditioning evaluation forms described in licensee procedure NETP-116.3, which would have challenged the validity of the licensees original determination of acceptability.
05000333/FIN-2016001-02FitzPatrick2016Q1Uncontrolled RPV Level Increase after Initiation of RHR Shutdown CoolingThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to take actions specified in the procedure for initiation of shutdown cooling. Specifically, prior to placing the A loop of the residual heat removal (RHR) system into shutdown cooling, an operator was not stationed to close the condensate transfer system cross-connect valve to the A RHR loop (10RHR-274), nor was the valve immediately closed after initiation of shutdown cooling, as specified by the operating procedure. This resulted in a significant loss of operational control, in that RPV level increased to the point of putting water down the main steam lines. As immediate corrective action, operators closed 10RHR-274, thus stopping the RPV inventory increase. The issue was entered into the CAP as CR-JAF-2016-00273. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the resultant loss of RPV level control represented a significant loss of operational control that could have affected the operability of the HPCI and reactor core isolation cooling (RCIC) systems, as well as the S/RVs, had their use again been required in the near term. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because operators did not stop when faced with uncertain conditions. Specifically, without otherwise having maintained status control on the condensate transfer system cross-connect valve to the A RHR loop, operators did not stop to positively establish the condition of the valve when it appeared in a conditional step in the procedure (that is, if 10RHR-274 is open, then station an operator at 10RHR-274) (H.11).
05000254/FIN-2016001-01Quad Cities2016Q1Failure to Control Deviation from EQ Standard Results in Limit Switch SubmergenceA finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed on February 2, 2016, when the operators received an alarm due to a steam leak in the Unit 1 main steam isolation valve room which resulted in the limit switch compartment for Unit 1 reactor core isolation cooling (RCIC) system motor-operated valve (MOV), MO 1130117 (outboard primary containment steam isolation valve), becoming submerged with water. Specifically, the licensee failed to ensure that deviations from design standard, Environmental Qualification Standard 74Q (EQ74Q), were controlled during original installation of MO 1130117 such that the valve would not be subjected to a spray or submergence environment. The licensee documented the issue in their corrective action program under Issue Report 2625523. Corrective actions included a temporary repair of the steam leak, removal of water from the limit switch compartment, and compensatory measures that included daily monitoring for steam leaks in the Unit 1 main steam isolation valve room. In addition, the licensee performed an extent of condition review of other valves in the main steam isolation valve room. Planned corrective actions included installing t-drains or weep holes in MOVs that the licensee deemed susceptible to spray or submergence. The performance deficiency was determined to be more than minor and a finding because it was associated with the Barrier Integrity Cornerstone attribute of Design Control and affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to control any environmental qualification design deviations had the potential to impact the ability of MO 1130117 to close on an isolation signal and prevent radioactive releases to the environment. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012. The inspectors determined the finding to be of very low safety significance (Green) in accordance with Exhibit 3, Barrier Integrity Screening Questions, because the inspectors answered No to all questions in Section B of Exhibit 3. This finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current performance.
05000331/FIN-2015004-02Duane Arnold2015Q4Failure to Declare High Pressure Coolant Injection and Reactor Core Isolation Cooling Inoperable when the High Pressure Coolant Injection and Reactor Core Isolation Cooling Pump Suction Swap Logic was InoperableThe inspectors identified a finding of very low safety significance, with two examples, and an associated NCV of Technical Specifications (TS) Sections 3.3.5.1, Condition D and 3.3.5.2, Condition D, for failure to initiate required TS action statements 3.3.5.1.D.1 and 3.3.5.2.D.1. Specifically, the licensee failed to declare the high pressure coolant injection (HPCI) and the reactor core isolation cooling (RCIC) systems inoperable when the automatic HPCI/RCIC pump suction swap function on low condensate storage tank (CST) level was revealed to be inoperable during surveillance testing. The licensee entered the inspectors concerns into the CAP as CR 2080489 and replaced the failed time delay relay. The inspectors determined the failure to declare the HPCI/RCIC systems inoperable when the pump suction swap function on low CST level failed during surveillance testing was a performance deficiency because it resulted in the licensees failure to implement TS required actions and the cause was reasonably within the licensees ability to foresee and should have been prevented. The performance deficiency was determined to be more than minor and a finding because if left uncorrected, failing to implement TS required actions reduced the margin of safety and had the potential to lead to significant safety concerns. The finding was determined to be of very low significance because the CST was assumed to contain sufficient inventory for HPCI and RCIC to perform their function for most scenarios. This finding was associated with the cross-cutting aspect of conservative bias in the area of human performance because the licensee failed to use decision-making practices that emphasize prudent choices over those that are simply allowable when the licensee failed to conservatively evaluate unexpected surveillance test results. (H.14)
05000366/FIN-2015007-05Hatch2015Q4Failure to Classify RCIC Sub-components as Safety-RelatedThe NRC identified a Non-cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to classify components in accordance with Regulatory Guide 1.26 as specified by the Unit 2 Updated Final Safety Analysis Report, Section 3.2.2. As an immediate corrective action, the licensee performed an operability evaluation, and determined that the reactor core isolation cooling (RCIC) was operable. In addition, the licensee planned to reclassify the relief valve as safety-related, and entered this issue into their Corrective Action Program as Condition Reports 10132353, 10136685, and 10141965. 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 affected the cornerstone objective, in that inadequate classification of the relief valves affected the reliability of safety-related function of the RCIC system. The finding was determined to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance (Section 1R21.2.b.5).
05000458/FIN-2015007-02River Bend2015Q4Failure to Obtain Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection PointThe team identified a Severity Level IV, Green, non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section (c)(2) which states, in part, A licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the final safety analysis report (as updated). Specifically, prior to October 8, 2015, the licensee failed to correctly evaluate that a spurious reactor core isolation cooling actuation injecting into the feedwater line resulted in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the updated final safety analysis report. In response to this issue, the licensee initiated a condition report to document completion of a new evaluation under current regulatory guidelines. This finding was entered into the licensees corrective action progam as Condition Report CR-RBS-2015-7259. The team determined that the failure to perform an adequate evaluation of a design change was a performance deficiency. This finding was also evaluated using traditional enforcement because it had the potential to impact the NRCs ability to perform its regulatory function. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, the licensee failed to correctly evaluate that a spurious reactor core isolation cooling actuation injecting into the feedwater line resulted in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the updated final safety analysis report. In accordance with Inspection Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency where the mitigating structure, system, or component maintained its operability or functionality. Since the violation is associated with a Green reactor oversight process violation, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. There is no cross-cutting aspect assigned to this performance deficiency because the performance deficiency is not indicative of current performance and also because cross-cutting aspects are not assigned to traditional enforcement violations.
05000458/FIN-2015007-03River Bend2015Q4Licensee-Identified ViolationThe licensee failed to meet the requirements of 10 CFR 50.71(e) to update the final safety analysis report. The final safety analysis report sections contained prior information and data that did not reflect the current plant configuration. The examples are: Chapter 3.9B, Mechanical Systems and Components (GE Scope Supply) Section 3.9.5.1.1.12B, Vent and Head Spray Nozzle described the head spray function of the reactor core isolation cooling system. However, the head spray function of the reactor core isolation cooling system had been removed from the final safety analysis report in 1998 via License Amendment Request 98-01 in which this section was erroneously omitted. Sections 2.4.2.3.1 and 12.3.2.4.5 of the updated safety analysis report describe the roofing systems on site to be sloped with built-up roofing material for construction. In 1996, a condition report was written documenting the discrepancy between design documents which indicate built-up roofing and insulation, and the actual condition of the roofs on the Auxiliary Building, Standby Cooling Tower Pump house, Radwaste Building, and Diesel Generator Building, which have neither of these materials. Title 10 CFR 50.71, Maintenance of Records, Making of Reports, Section (e) which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. This submittal shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the applicant or licensee or prepared by the applicant or licensee pursuant to Commission requirement since the submittal of the original or the last update to the final safety analysis report. Contrary to the above, the licensee did not update the final safety analysis report to assure that the information included in the report contains the latest information developed. Specifically, the licensee failed to ensure the final safety analysis report reflected the current plant configuration, as evidenced by the above examples. The NRCs significance determination process considers the safety significance of findings by evaluating their potential safety consequences. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this performance deficiency because it involved a violation that impacted the regulatory process. Assessing the violation in accordance with the NRC Enforcement Policy, the team determined it to be a Severity Level IV violation because the lack of up-to-date information in the final safety analysis report has not resulted in any unacceptable change to the facility or procedures (NRC Enforcement Policy example 6.1.d.4). This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2015-7259.
05000277/FIN-2015003-01Peach Bottom2015Q3Incomplete Testing of Components from the Remote Shutdown PanelsThe inspectors identified a Green NCV of Technical Specification (TS) 5.4.1.a after Exelon did not establish and implement procedures to adequately test the Unit 2 and Unit 3 remote shutdown panels (RSPs). Specifically, Exelons surveillance procedure did not test all the control circuits, as required by Surveillance Requirement (SR) 3.3.3.2.1, for the Unit 2 and Unit 3 RSPs. Exelons corrective actions included entering this issue into their CAP, the development of RSP testing procedures for the reactor core isolation cooling (RCIC), control rod drive (CRD), and emergency service water (ESW) system components, and a revision to the bases for TS 3.3.3.2 The performance deficiency (PD) was determined to be more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, examples 1.c, 4.l, and 4.m from IMC 0612, Appendix E, detail that a PD was more than minor if required TS surveillance testing is not performed and subsequent testing reveals that the equipment is out of specification or otherwise unable to perform a safety-related function. A detailed risk evaluation concluded that the issue was of very low safety significance (Green). This finding had a cross-cutting aspect in Human Performance, Avoid Complacency, because Exelon failed to recognize and plan for the possibility of latent problems.
05000440/FIN-2015003-02Perry2015Q3Failure to Properly Implement Steps Outlined in a Technical Specification Surveillance ProceA finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1., Procedures, was self-revealed on August 5, 2015, when an unexpected isolation of the reactor core isolation cooling (RCIC) system occurred as a result of the licensees failure to properly implement the steps outlined in TS Surveillance Procedure, SVIE31T5395B, RCIC Steam Line Flow High Channel Functional for 1E31N684B. Specifically, during performance of the surveillance, several steps were marked as not applicable that were applicable to prevent the isolation of the RCIC system. As a result, the licensee failed to lift leads as required by the procedure and the RCIC steam supply inboard isolation valve then closed when the isolation trip signal was applied during the test. The licensee took immediate actions to restore system operability and availability and conducted a human performance event response investigation. A standing order for both Operations and Instrumentation and Controls personnel was initiated addressing interim actions for control room surveillance performance and to reinforce maintenance fundamentals and human performance behaviors. The licensees failure to properly implement the steps in the procedure was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance, avoid complacency, for failing to recognize and plan for the possibility of mistakes, and for failure to implement appropriate error reduction tools, such as proper self-checks and peer checks, which resulted in an isolation of the RCIC system.
05000374/FIN-2015002-02LaSalle2015Q2Inadvertent Operation of Circuit Breaker Affecting Unit 2 Train A Residual Heat Removal Suppression Chamber Spray Isolation Valve (235Y-2 C3)A finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.4.1, Procedures, was self-revealed when the licensee failed to properly preplan and perform maintenance in accordance with written procedures and instructions appropriate to the circumstances. Specifically, on May 14, 2015, the Work Order (WO 1643222) for testing of the motor for the Unit 2 reactor core isolation cooling (RCIC) water leg pump and involving operation of the motors breaker did not include precautions or restrictions to prevent the inadvertent operation, by bumping, of the adjacent breaker for the safety-related Unit 2 A residual heat removal (RHR) suppression chamber spray isolation valve. Workers inadvertently bumped and opened the breaker for the RHR valve and rendered the system inoperable. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to provide a work order appropriate to the circumstances of the juxtaposed breakers. The subsequent, inadvertent opening of the 2A RHR suppression chamber spray isolation valve breaker, unexpectedly rendered the valve inoperable. This negatively impacted the RHR suppression chamber spray systems ability to reduce suppression chamber pressure by removing one of the required two spray paths. The inspectors determined the finding to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because configuration control and error prevention techniques (robust barriers) in an existing licensee procedure were not appropriately implemented due to the failure of individuals to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12). Specifically, licensee staff failed to implement the guidance found in procedure HU-AA-101, Human Performance Tools and Verification Practices.
05000461/FIN-2015002-04Clinton2015Q2Failure to Translate Sufficient Gland Stress to Packing Gland Nuts Resulted in Valve Packing Failure and Plant ShutdownA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was self-revealed on January 19, 2015, when a steam leak developed from the Reactor Core Isolation Cooling (RCIC) system inboard steam isolation valve (1E51F0063) stem packing. Specifically, the licensee failed to identify and implement a torque value for the gland packing to overcome service induced consolidation and prevent packing leakage. This resulted in a plant down power to 83 percent and subsequent plant shutdown due to increasing unidentified reactor coolant system leakage. The licensee documented the issue in the CAP as action request (AR) 02439437. The licensee repacked the valve utilizing the Procedure Clinton Power Station (CPS) 8120.37, Valve Packing Installation, and the applicable SealPro data sheet. A four ring set of A.P. Services graphite packing was installed with a new live load assembly sized to a new torque value of 59 ft-lbs. and the valve packing was tested to verify no leakage. The inspectors determined that the failure to apply sufficient packing gland torque to overcome service induced consolidation and prevent packing leakage on the RCIC system inboard steam isolation valve was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations, and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident, cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused an initiating event and did not affect mitigation equipment. The inspectors determined that no cross-cutting aspect would be associated with this finding since the performance deficiency occurred in 2010 and was not representative of current licensee performance. (Section 4OA2)
05000254/FIN-2015002-01Quad Cities2015Q2Failure to Conduct Post-Maintenance Testing Following Manual Operation of RCIC MOVA finding of very low safety significance and associated NCV of Technical Specification 5.4, Procedures, was self-revealed on March 22, 2015, for the licensees failure to conduct procedurally required post-maintenance testing on reactor core isolation cooling (RCIC) motor operated valve (MOV) MO 1130161, following operation of the valve in the manual mode. Immediate corrective actions included manually engaging the motor clutch and functionally stroking the valve from the control room to verify operation. The licensee captured this condition in their CAP as Issue Report 2472416. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee was not able to ensure the operability of the RCIC system when they failed to conduct post-maintenance testing (PMT) on RCIC 1130161. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors answered No to all questions in Section A of Exhibit 2 and the finding screened as Green, or very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance, Documentation, because the licensee did not maintain complete, accurate, and up-to-date documentation. Specifically, the licensee failed to document the status of the RCIC valve after placing it in the manual mode of operation to ensure that the required PMT was performed.
05000416/FIN-2015002-01Grand Gulf2015Q2Failure to Have Appropriate Instructions Resulted in the Unplanned Unavailability of the Reactor Core Isolation Cooling SystemThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, for failure to establish appropriate work instructions to properly preplan and perform maintenance that affected the performance of the reactor core isolation cooling system. Specifically, the work instructions failed to ensure that a steam supply drain pot drain alignment path was maintained while replacing valve packing 1-E51-F026. As a result, the drain path was isolated causing a group 4 isolation, which rendered the reactor core isolation cooling system unavailable. Operations personnel returned the reactor core isolation cooling system to operable status approximately 19 hours after the isolation occurred. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2015-01677. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to have an adequate maintenance work instruction resulted in the unplanned unavailability of the reactor core isolation cooling system. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. In addition, this finding has an avoid complacency cross-cutting aspect within the human performance area because the licensee failed to recognize and plan for the possibility of mistakes, inherent risks, and properly implement appropriate error reduction tools. Specifically, the licensee failed to recognize the importance of having a drain path during the entire maintenance activity to properly plan the activity using appropriate configuration control and work instructions.
05000331/FIN-2015007-02Duane Arnold2015Q2Failure to Correctly Update the Updated Final Safety Analysis ReportThe inspectors identified a Severity Level IV NCV of 10 CFR 50.71(e) for failure to assure that the information included in the last update of the updated final safety analysis (UFSAR) report contained the latest information developed. The licensee implemented a change to the UFSAR, in preparation for License Amendment 243 that did not contain the latest information developed. Specifically, Section 5.4.6.1 (page 5.430 of Revision 17) was updated with a note that stated the reactor core isolation cooling system was not safety-related. In fact, the reactor core isolation cooling system had always been designated as safety-related. The licensee entered this issue into the CAP as CR 01974995 and prepared an updated final safety analysis report (UFSAR) change that removed the statement that the reactor core isolation cooling system was not safety-related. The inspectors determined that the update to the UFSAR with incorrect information was a performance deficiency in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued on September 7, 2012. The inspectors concluded that traditional enforcement applied because the failure to correctly update the UFSAR impacted the regulatory process. The Enforcement Policy, dated February 4, 2015, Section 6.1.d.3, gave the example that if, a licensee fails to UFSAR as required by 10 CFR 50.71(e) but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures; then this was a Severity Level IV violation. In this case, the UFSAR was updated incorrectly and did not, result in any unacceptable change to the facility or procedures. The inspectors determined this to be a similar example and therefore was more than minor and a Severity Level IV violation. This violation was not associated with a finding that was evaluated by the significance determination process. Therefore, a cross-cutting aspect was not assigned to this traditional enforcement violation.
05000458/FIN-2015009-04River Bend2015Q2Failure to Identify High Reactor Water Level as a Condition Adverse to QualityThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure a condition adverse to quality was promptly identified. Specifically, the licensee failed to identify, that reaching the reactor pressure vessel water Level 8 (high) setpoint, on December 25, 2014, was an adverse condition, and as a result, failed to enter it into the corrective action program. To restore compliance, the licensee entered this issue into their corrective action program as Condition Report CR-RBS-2015-00620 and commenced a causal analysis for Level 8 (high) trips. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to identify Level 8 (high) conditions and unplanned automatic actuations as conditions adverse to quality, would continue to result in the undesired isolation of mitigating equipment including reactor feedwater pumps, the high pressure core spray pump, and the reactor core isolation cooling pump. The team performed an initial screening of the finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the team determined that the finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has an avoid complacency cross-cutting aspect within the human performance area because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee tolerated leakage past the feedwater regulating valves, did not plan for further degradation, and the condition ultimately resulted in the Level 8 (high) trip of the running reactor feedwater pump on December 25, 2014 (H.12).
05000293/FIN-2015007-04Pilgrim2015Q1Failure to Follow RCIC System Manual Restart ProcedureA self-revealing Green NCV of TS 5.4.1, Procedures, was identified because the operating crew failed to implement a procedure step to open the reactor core isolation cooling (RCIC) system cooling water supply valve during a manual startup of the system. As a result, the RCIC system was operated for over 2 12 hours with no cooling water being supplied to the lubricating oil cooler or to the barometric condenser. Entergy entered the issue into the CAP as CR-PNP-2015-0566, CR-PNP-2015-0570, and CR-PNP-2015-0952 and conducted a human performance review of the Control Room operators involved with the issue. The inspectors determined that the failure to implement Procedure 5.3.35.1, Attachment 29, RCIC Injection Manual Alignment Checklist, and the Vacuum Tank Pressure Hi Alarm, C904L-F3, alarm response procedure was a performance deficiency and was reasonably within the ability of Entergy personnel to foresee and prevent. This self-revealing finding was more than minor because it was associated with the human performance attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, on January 27, 2015, reactor operators failed to open MO-1301-62, cooling water supply valve, during a manual restart of the RCIC system in accordance with procedure 5.3.35.1, RCIC Injection Manual Alignment Checklist. Additionally, the operating crew failed to identify the valve was out of position even after the Vacuum Tank Pressure Hi Alarm, C904L-F3, was received two minutes after the system was re-started and the alarm response procedure identified Improper Valve Lineup as a probable cause. The team evaluated the finding using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The team determined this finding was not a design or qualification deficiency and was not a potential or actual loss of system or safety function, and is therefore of very low safety significance (Green). During the period when the RCIC system was operated in this condition, no temperature limits were exceeded. The inspectors noted that in the event of a RCIC system automatic start, the cooling water supply valve would have opened automatically. This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Entergy licensed personnel did not implement procedure 5.3.35.1, RCIC Injection Manual Alignment Checklist , to open MO-1301-62. Additionally, Entergy licensed personnel did not implement the Vacuum Tank Pressure Hi Alarm, C904L-F3, response procedure to check for an improper valve line-up.
05000373/FIN-2014008-01LaSalle2015Q1Failure to Ensure Circuits associated with Alternate Shutdown Capability Free of Fire-induced DaThe inspectors identified a finding of very-low safety significance (Green) and associated NCV of the LaSalle County Station Operating License for the licensees failure to ensure that the alternate shutdown capability was independent of the fire area. Specifically, in the event of a fire in the control room, the alternate shutdown capability for 16 motor operated valves (MOVs) associated with the Reactor Core Isolation Cooling (RCIC) may be affected, and may not be available due to lack of breaker fuse coordination. Fire-induced failures could result in tripping valve power supply breakers prior to tripping the control power fuses for several motor operated valves, thereby, potentially imparing the operation of RCIC from the Remote Shutdown Panel (RSP). The licensee entered this issue into their Corrective Action Program and established compensatory measures, and added steps to the safe shutdown procedures to reset the affected breakers if needed. In addition, the licensee intended to perform plant modifications to replace or revise existing breakers settings to correct the issue. The inspectors determined that the issue was more than minor, because fire induced circuits could impair the operation of RCIC and complicated shutdown of the plant in the event of a fire in the control room. The finding affected the Mitigating Systems Cornerstone. The finding was determined to be of very-low safety significance based on a detailed risk-evaluation. This finding was not associated with a cross-cutting aspect because the finding was not representative of the licensees current performance.
05000298/FIN-2015001-01Cooper2015Q1Inadequate Operations ProcedureThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the inadequate Operations Procedure 2.2.7, Condensate Storage and Transfer System, Revision 56. Specifically, the procedure did not require that the affected system, either the high pressure coolant injection system or the reactor core isolation cooling system, be declared inoperable when one or more of the high pressure coolant injection or reactor core isolation cooling test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, or RCIC-MOV-33, were moved off of their closed (passive safety function position) seats. The license entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-00274. The failure to establish and maintain a correct filling procedure required by Technical Specification 5.4.1.a. was a performance deficiency and resulted in the licensees failure to declare the high pressure coolant injection and reactor core isolation cooling systems inoperable when required to do so. The performance deficiency is more than minor, and therefore a finding, because it is associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the high pressure coolant injection and reactor core isolation cooling systems were not declared inoperable when their test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, and RCIC-MOV-33, were taken off their normally closed (passive safety function position) seats. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-ofservice for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction techniques. Specifically, licensee personnel fell into a pattern of acceptance regarding Procedure 2.2.7. This resulted in a failure to question the lack of an operability caution statement, even though there was other guidance in the inservice inspection program to that effect (H.12).
05000416/FIN-2015001-02Grand Gulf2015Q1Failure to Follow a Procedure Resulting in the Unplanned Inoperability of the Reactor Core Isolation Cooling SystemThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, for failure to follow a procedure which resulted in the unplanned inoperability of the reactor core isolation cooling system. This occurred when licensee technicians tested for continuity between incorrect points, while performing surveillance activities related to the residual heat removal system. This resulted in an invalid group 4 isolation signal and an isolation of the reactor core isolation cooling steam supply. The licensee entered this issue into the corrective action program as Condition Report CR-GGN- 2015-01532, and took immediate corrective actions to stop the residual heat removal system surveillance activity and restore the reactor core isolation cooling system to service. The failure to properly follow the surveillance procedure, which resulted in the unplanned inoperability of the reactor core isolation cooling system, was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Mitigating Systems Cornerstone. Specifically, the licensees failure to properly follow the surveillance procedure resulted in the unplanned inoperability of the reactor core isolation cooling system, which adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) in that the issue did not affect the design or qualification of the reactor core isolation cooling system; did not represent a loss of the reactor core isolation cooling system function (in that the isolation could have been promptly reset by procedures, had the system operation been required); and did not represent loss of function for greater than the Technical Specification allowed outage time. The inspectors determined this finding had cross-cutting aspect in the area of human performance associated with avoiding complacency, in that the I&C technicians did not implement appropriate error reduction tools to ensure the meter was connected to the correct points, which resulted in the invalid group 4 isolation signal, and inoperability of the reactor core isolation cooling system (H.12).
05000354/FIN-2015001-01Hope Creek2015Q1Failure to Identify and Correct a Condition Adverse to Quality Associated with the Reactor Core Isolation Cooling System Insulation and OilA self-revealing finding of very low safety significance (Green) and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, was identified because PSEG did not promptly identify and correct a condition adverse to quality (CAQ). Specifically, PSEG failed to identify a deficiency with the reactor core isolation cooling (RCIC) turbine thermal insulation on July 28, August 19, and November 18, 2014; and failed to initiate a notification (NOTF) identifying an adverse trend in RCIC oil moisture content and level on November 18, 2014 and in January 2015. The failure to identify and correct a CAQ resulted in high moisture content in the RCIC oil. PSEGs corrective actions included replacing the RCIC system oil on February 19, 2015 and repairing the non-conforming turbine insulation on February 25, 2015. The performance deficiency (PD) was determined to be more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). This PD was also similar to examples 3.j and 3.k of NRC IMC 0612, Appendix E, in that the increased moisture content in the RCIC oil created a reasonable doubt of operability of the RCIC system. The inspectors determined the finding to be of very low safety significance (Green) in accordance with Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, because: it was not a deficiency affecting the design or qualification of the mitigating system; it did not represent a loss of system function; it did not represent the loss of function for any TS system, train, or component beyond the allowed TS outage time; and it did not represent an actual loss of function of any non TS trains of equipment designated as high safety significance in accordance with PSEGs maintenance rule program. The inspectors determined the finding had a cross-cutting aspect in the area of Problem Identification and Resolution (PI&R), Trending, because PSEG did not periodically analyze information from the corrective action program and other assessments in the aggregate to identify programmatic and other common cause issues. Specifically, PSEG did not analyze multiple RCIC system oil sample results or RCIC system NOTFs in the aggregate to identify a CAQ.
05000333/FIN-2014005-03FitzPatrick2014Q4Licensee-Identified ViolationTS 3.3.5.2, Reactor Core Isolation Cooling (RCIC) System Instrumentation, requires that the RCIC system instrumentation for all 4 channels of low CST water level be operable while in Modes 1, 2, or 3 with reactor steam dome pressure greater than 150 psig. With any level switch inoperable, Condition D requires that the channel be placed in trip. When this condition is not met, Condition E requires that RCIC be declared inoperable. LCO 3.5.3, RCIC System, further requires that RCIC be restored to operable status within 14 days or be placed in Mode 3. Contrary to TS 3.3.5.2, with two RCIC CST level switches, 13LS-76B and -77B, inoperable from July 16, 2013 to August 19, 2013, Entergy did not place the channels in trip or declare RCIC inoperable, or place the reactor in Mode 3 per TS 3.5.3. The cause o the inoperability was corrosion buildup on the level switches caused by water intrusion through a junction box common to the switches. Entergy entered this issue into the CAP as CR-JAF-2013-04311. The inspectors determined, through a review of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, that the finding was of very low safety significance (Green) because the finding was not related to a design or qualification deficiency, did not represent a loss of a mitigating system safety function, and did not screen as potentially risk significant due to external initiating events. The Senior Reactor Analyst (SRA) used the Systems Analysis Programs for Hands-On Evaluatio (SAPHIRE), Revision 8.1.2, and the Standardized Plant Analysis Risk (SPAR) Model for Fitzpatrick, Model Version 8.1.17, to confirm that no loss of safety function occurred. The SRA determined that the RCIC pump suction is assumed to remain on the CST for the duration of operation to complete its safety function and therefore this issue was determined to be of very low safety significance (Green). The CST inventory is modeled to be sufficient because the function of RCIC is to respond to transient events to provide makeup coolant to the reactor.
05000416/FIN-2014005-01Grand Gulf2014Q4Failure to Assure Quality Installation on RCIC Steam LineThe inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to assure quality installation of the steam line tubing of the reactor core isolation cooling (RCIC) system. Specifically, the licensee failed to assure that rated performance limits of the ferrule connection, installed at the tee between the steam line and the pressure transmitter tube line, were met during initial installation. This failure resulted in an unplanned inoperability of the RCIC system. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2014-06792. As an immediate corrective action, the licensee replaced the tubing, the failed transmitter, and recalibrated the instruments. Furthermore, the licensee revised their system operation procedure for the RCIC system. This revision requires all steam isolation valves to be closed during this test, and that system recovery starts by opening Valve 1E51F076 (warming bypass valve around the 1E51F063) to allow adequate warming of the steam lines after isolation. The inspectors determined that the failure to assure quality installation of the ferrule connection on the steam line flow Transmitter 1E31N083B was a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the design control attribute of the Mitigating Systems Cornerstone. Specifically, failure to assure steam lines in the RCIC system meet rated performance limits, may result in the unavailability and unreliability of a system that is relied upon to respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that the issue required a detailed risk evaluation by the regional senior reactor analyst. This was because the finding represented an actual loss of a safety function due to the RCIC system being a single train system that was out of service for approximately 40 hours for repairs. The senior reactor analyst determined the change to the core damage frequency was 8.7E-8/year, and since the change to core damage frequency was less than E-7, no evaluation of external events or the large early release frequency was required. The finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect, as the performance deficiency is not reflective of current plant performance.
05000333/FIN-2014004-02FitzPatrick2014Q3Licensee-Identified ViolationTS 3.3.5.2, Reactor Core Isolation Cooling System Instrumentation, requires that the RCIC system instrumentation for all four channels of low CST water level be operable while in Modes 1, 2, or 3 with reactor steam dome pressure greater than 150 psig. With one level switch inoperable, Condition D requires that the channel be placed in trip. When this condition is not met, Condition E requires that RCIC be declared inoperable. TS 3.5.3, RCIC System, further requires that RCIC be restored to operable status within 14 days or be in Mode 3. Contrary to TS 3.3.5.2, with one RCIC CST level switch, 13LS-76B, inoperable from September 17, 2013 until November 4, 2013, Entergy did not place the channel in trip or declare RCIC inoperable, or place the reactor in Mode 3 per TS 3.5.3. The cause of the inoperability was the failure to align the microswitch in accordance with vendor manual instructions when the switch was replaced in September. Entergy entered this issue into the CAP as CR-JAF-2013-5576. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, because the finding was not a design or qualification deficiency, did not involve the actual loss of safety function, did not represent the actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen to potentially risk significant due to a seismic, flooding, or severe weather initiating event.
05000461/FIN-2014004-03Clinton2014Q3Failure to Establish a Surveillance Procedure for Reactor Core Isolation Cooling Pump due to Unacceptable PreconditioningThe inspectors identified a Green finding and an associated non-citied violation of Technical Specification 5.4.1.a, Procedures, for the failure to establish a surveillanc procedure to test the Reactor Core Isolation Cooling (RCIC) system without unacceptable preconditioning. Specifically, procedure CPS 9054.01C002, RCIC High Pressure Operability Checks, Revision 8, allows draining of the RCIC exhaust drain pot prior to the surveillance run. This action constitutes unacceptable preconditioning because it could make it difficult to determine whether the system would perform its intended function during an event in which the system might be needed. The licensee documented this issue in the CAP as IR 02386704 and made changes to the procedure to ensure inadequate preconditioning does not occur. The inspectors determined that the failure to establish a surveillance procedure to test the RCIC system without unacceptable preconditioning is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability to response to initiating events to prevent undesirable consequences and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safetysignifican for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee considered the impact of the operating experience for surveillance testing, but did not consider its impact during normal plant operation (P.5).
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