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05000237/FIN-2014003-012014Q2DresdenFailure to Take Appropriate Corrective Action When a Maintenance Rule Performance Goal for the Standby Coolant System was Not MetThe inspectors identified a finding of very low safety significance and non-cited violation of 10 CFR 50.65(a)(1), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to take corrective actions by performing an (a)(1) determination when the standby coolant supply system preventative maintenance (a)(2) demonstration was failed. Specifically, in November 2013, the standby coolant supply system exceeded its maintenance rule performance criteria when it experienced an additional maintenance preventable functional failure. The licensee failed to appropriately account for this failure in their Maintenance Rule Program and, as a result, the site failed to perform appropriate corrective action, by failing to perform an (a)(1) determination in accordance with Procedures ERAA310, Implementation of the Maintenance Rule, and ERAA-3101005, Maintenance RuleDispositioning Between (a)(1) and (a)(2), Revision 6. Corrective actions taken by the licensee to address this issue included performing a maintenance rule (a)(1) determination and placing the system into (a)(1) status. The issue was entered into the licensees corrective action program as issue report (IR) 1644740, NRC Questions D2R23 Performance of DOS 390001, and IR 1650033, MRule A1 Determination Needed for Missed MRFF Z391. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstones attribute of Equipment Performance and affected the cornerstones objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to identify a functional failure during a periodic (a)(2) demonstration purposed to provide reasonable assurance that the structures, systems, and components (SSCs), the standby coolant injection valve MO 23902, was capable of performing its intended function as specified in licensee emergency operating procedure DEOP 050003, Alternate Water Injection Systems, Revision 22. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 for the Mitigating Systems cornerstone. The inspectors answered Yes to the question Does the finding represent a loss of system and/or function and determined that a Detailed Risk Evaluation was required. The Senior Reactor Analysts (SRAs) evaluated the finding using the Dresden Standardized Plant Analysis Risk (SPAR) model version 8.18 and Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.0.9.0 software. The exposure time for the unavailability of the Standby Coolant Supply Valve 23902 was assumed to be the maximum value of one year. The result was a delta core damage frequency (CDF) of 6.6E8/yr. The dominant sequence was a medium loss of coolant accident initiating event with a failure of suppression pool cooling, a failure of power conversion system recovery, and a failure of late injection. Based on the Detailed Risk Evaluation, the SRAs determined that the finding was of very low safety significance (Green). This finding had a crosscutting aspect in Human Performance, Procedure Adherence, because the licensee failed to appropriately document the failure of a standby coolant supply valve in accordance with periodic test procedure DOS 390001, Standby Coolant Supply Functional Test. (H.8)
05000237/FIN-2014003-022014Q2DresdenLicensee-Identified ViolationTS 5.4.1 requires that written procedures shall be established, implemented, and maintained for procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Section 7.e.2 includes implementation of the Radiation Survey Program. Station Procedure RPAA503, unconditional release survey method requires, in part, that materials have no detectable radioactivity for unconditional release from the site. Contrary to the above, on June 9, 2014, a contracted sewage truck transporting contaminated sewage from Unit 1 ejector pit to the licensees sewage treatment plant was unconditionally released after the truck was emptied. Specifically, the sewage truck was unconditionally released to the contractors facility without the proper authorization from RP Management. On the following day, the truck was returned to the licensees facility for survey and decontamination by the RP staff. The empty sewage truck contained traced amount of radioactivity of Co60 and tritium above minimal detectable activities. The licensee investigation determined that the empty sewage truck did not leak or cause contamination during transit on the public road. This event was entered into the licensees CAP as CR 01673475. The Radiation Protection Department immediately stopped work. Future transport of sewage between the licensed facility and the licensees sewage treatment plant will be escorted by radiation protection personnel to ensure that drivers follow licensee direction. The significance of the finding was determined by using Inspection Manual Chapter 0609, Appendix D, "Public Radiation Safety SDP." The issue is of very low safety significance (Green) because it involved radioactive material control, was not a finding involving transportation, and did not result in public exposure greater than 0.005 rem.
05000237/FIN-2014007-012014Q1DresdenFailure to Adequately Incorporate GE Operating Experience into Vendor ManualThe inspectors identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to ensure that operating experience provided via a vendor Service Information Letter (SIL) was properly evaluated and incorporated into the vendor manual contrary to the requirements of procedure RSAA115, Operating Experience. The failure to properly assess operating experience for alternating current (AC) Motors resulted in a condition where specific deficiencies could go unrealized under the licensees conditioned based monitoring program. The licensee initiated action request (AR) 1633528 and 1635766 to document and develop corrective actions for the issue. The finding 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 to adequately evaluate and document the basis for the use or rejection of 9 out of 10 experiences presented in General Electric (GE) SIL 484, Supplement 6, could cause a reduction in reliability for safety related systems that use AC motors. 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 (SDP) for Findings At-Power. The finding was screened against the Mitigating Systems Cornerstone, Exhibit 2 of Appendix A, and determined to be of very low safety significance because the answer was no to all of the screening questions. This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency (H.12), because individuals failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
05000306/FIN-2013005-012013Q4Prairie IslandFailure to Evaluate Corrosive Effects of Boric Acid on the 22 Residual Heat Removal PumpThe inspectors identified a finding of very low safety significance on October 7, 2013, due to the failure to perform an adequate boric acid evaluation in accordance with Procedure H2, Boric Acid Corrosion Control Program. Specifically, the licensee failed to properly evaluate the impact of a boric acid leak following the leak coming into contact with carbon steel components on the 22 residual heat removal pump casing. Corrective actions included moving a carbon steel bolt for visual inspection and completing a technically adequate boric acid corrosion evaluation. The inspectors determined that this issue was more than minor because if left uncorrected the failure to complete technically adequate boric acid corrosion evaluations could result in components with questionable structural integrity being left in service. The inspectors determined that this issue was of very low safety significance because each of the questions provided in IMC 0609, Attachment 0609.04, Appendix A, Exhibit 2, were answered no. The inspectors concluded that this issue was cross-cutting in the Human Performance, Decision Making area because the licensee failed to use conservative assumptions while determining the applicability of a previously completed boric acid evaluation to a current plant condition. No violation was identified since all NRC requirements were met.
05000282/FIN-2013005-022013Q4Prairie IslandFailure to Establish Appropriate Design Control Measures for Selection of Replacement PartsThe inspectors identified a finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, on October 8, 2013, due to the failure to establish measures for the selection of parts that are essential to the safety-related functions of structures, systems, or components (SSCs). Specifically, the licensee failed to properly evaluate the specifications and quality of replacement parts such as gaskets, o-rings, packing materials, and diaphragms to ensure that these parts were suitable for installation in safety-related systems. As a result, multiple systems were required to be declared operable but non-conforming. Corrective actions for this issue included ensuring personnel understood the requirements regarding parts selection, determining the correct parts to be used and initiating work orders to ensure that parts were replaced in the future if required. The inspectors determined that this issue was more than minor because if left uncorrected, the installation of parts/materials which failed to meet requirements could lead to subsequent part failure. This failure would adversely impact the ability of safety-related equipment to perform its safety function. The inspectors determined that this issue was of very low safety significance because Question 1 in IMC 0609, Attachment 0609.04, Attachment A, Exhibit 2, was answered yes. The inspectors concluded that this issue was cross-cutting in the Human Performance, Resources area because the licensees parts specification and quality level documentation was not complete, accurate and/or up to date.
05000282/FIN-2013005-032013Q4Prairie IslandFailure to Promptly Correct Condition Adverse to Quality on D1 EDGA self-revealing finding of very low safety significance (Green) and an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, was identified on October 15, 2013, due to the failure to correct a condition adverse to quality. Specifically, the licensee failed to correct a D1 emergency diesel generator (EDG) lube oil cooler leak prior to the EDG being rendered inoperable. Corrective actions for this issue included reviewing the engineering departments equipment monitoring program, ensuring the lube oil cooler end bell was adequately torqued and repairing the lube oil cooler. The inspectors determined that this issue was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The failure to correct the lube oil cooler leak resulted in the licensee accruing unplanned unavailability on the D1 EDG during this emergent repair. The inspectors determined that this issue was of very low safety significance because each of the questions provided in IMC 0609, Attachment 0609.04, Appendix A, Exhibit 2, were answered no. The inspectors concluded that this issue was cross-cutting in the Problem Identification and Resolution, Corrective Action Program (CAP) area because the licensee failed to thoroughly evaluate the condition of the leaking lube oil cooler to ensure that repairs were properly prioritized.
05000282/FIN-2013005-042013Q4Prairie IslandLicensee-Identified ViolationTechnical Specification 3.7.10 requires that two trains of the control room special ventilation system (CRSVS) must be operable when the reactor is operating in Modes 1 through 4 or during the movement of irradiated fuel assemblies. Contrary to the above, the licensee identified on August 9, 2013, that one or more CRSVS trains had been inoperable since December 10, 2010, due to the failure to properly perform control room envelope unfiltered air in-leakage testing as required by TS Surveillance Requirement 3.7.10.5. The inspectors determined that this issue was more than minor because it was associated with the Barrier Integrity cornerstones attribute of barrier performance and affected the cornerstones objective of maintaining the barrier functional integrity of the Control Room. Specifically, the licensee failed to ensure that the measured control room envelope unfiltered air in-leakage remained less than or equal to the in-leakage rate assumed in the accident analysis. As a result, the licensee was not able to demonstrate that operations personnel located within the control room would have been adequately protected from the radiological consequences of a design basis accident. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Barrier Integrity cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3 for the Barrier Integrity cornerstone. For the control room envelope (CRE) finding, Section C of Exhibit 3 was applicable. The inspectors answered no to the first screening question and yes to the second screening question. As a result, a detailed risk evaluation was required to be performed by a Senior Reactor Analyst (SRA). The exposure time of the finding was assumed to be the maximum of one year. To evaluate the effects of the degradation of the CRE against a toxic atmosphere, the SRA determined that there was no delta risk from toxic gas events. The outside air damper which was identified to be leakage during subsequent testing did not auto-close and was not intended to preclude toxic gas from entering the Control Room from a design basis perspective. To evaluate the effects of the degradation of the CRE against smoke, the SRA identified that the Electric Power Research Institute (EPRI) Fire Probabilistic Risk Assessment Implementation Guide (report number EPRI TR-105928), Appendix B, stated that while actions performed from the control room should generally be unaffected by a fire outside the control room, a January 1989 event at an eastern nuclear plant indicated that fires outside the control room may increase the stress on the operators if smoke reaches the control room. In addition, NUREG/CR-6883, The SPAR-H Human Reliability Method, indicated that physical stress (such as that imposed by difficult environmental factors) may increase the stress and impede the operator from easily completing a task. Based on this, the SRA conservatively assumed that with the degradation of the CRE, any fire in the plant would result in smoke entering the control room and represent a risk increase due to the performance deficiency with the following exceptions: A fire initiated in the control room (since a fire in the control room would not represent additional risk given the performance deficiency); and A fire in containment (since smoke from a fire in containment would not reach the control room). This evaluation was considered bounding because of the tortuous path that would be required for smoke from a fire in the plant to enter the CRE through the leaking outside air damper. The delta risk due for the finding was attributed to high stress on the control room operators (e.g., the control room operators may need to wear self-contained breathing apparatus (SCBAs)). Using SPAR-H, a Performance Shaping Factor (PSF) with a multiplier of 2 was used due to High Stress while performing the affected control room actions. From the Prairie Island IPEEE, Appendix B, Revision 2, Internal Fires Analysis, the core damage frequency (CDF) for fires was spread across five accident classes: (TEH) - early core melt with the reactor at high pressure; (TLH) - late core melt with the reactor at high pressure; (SEH) - early core melt with the reactor at high pressure in conjunction with a small loss-of-coolant-accident (SLOCA); (SLH) - late core melt with the reactor at high pressure in conjunction with a SLOCA; and (BEH) - early core melt with the reactor at high pressure in conjunction with a station blackout. In the Prairie Island IPEEE, Appendix B, Attachment 8, Fire PRA Dominant Cutsets, the dominant cutsets for each of the five accident classes is given. The SRA's evaluation of the ?CDF associated with the finding for each of the five accident classes is provided below: Accident class TEH - of the top 100 cutsets for this accident class, cutsets 12, 58, 80, and 85 were found to contribute to a ?CDF associated with the finding. The ?CDF associated with this accident class was evaluated to be 8.8E-8/yr due to high stress. Accident class SEH - of the top 100 cutsets for this accident class, cutset 93 was found to contribute to a ?CDF associated with the finding. The ?CDF associated with this accident class was evaluated to be 7.1E-9/yr due to high stress. Accident class TLH - of the top 100 cutsets for this accident class, cutsets 17 and 70 were found to contribute to a ?CDF associated with the finding. The ?CDF associated with this accident class was evaluated to be 5.9E-9/yr due to high stress. Accident class BEH - involves fires that cause a loss-of-offsite-power (LOOP). In the IPEEE, Appendix B, Revision 2, it states that only one fire was determined to lead to a LOOP and involved a large fire in the control room G control panel. Since a control room fire does not represent a ?CDF associated with the finding, accident class BEH was eliminated from further consideration. Accident class SLH - of the top 100 cutsets for this accident class, 98 were associated with control room fires. The other two fires (cutsets 85 and 86 on the list of 100) involved relay room fires that did not contribute to the ?CDF associated with the finding because these cutsets did not contain basic events involving operator actions. The total ?CDF associated with the effects of the degradation of the CRE against smoke is the sum of the ?CDF for each of the five accident classes or 1.0E-7/yr. Taking into account the exposure time of the finding, the ?CDF associated with the effects of the degradation of the CRE against smoke was 1.0E-7/yr. Since the total estimated change in core damage frequency was greater than or equal to 1.0E-7/yr, the potential risk contribution from large early release frequency (LERF) was evaluated for risk significance. Appendix H, to IMC 0609, Containment Integrity Significance Determination Process was used to determine the potential risk contribution due to LERF. Prairie Island is a 2-loop Westinghouse Pressurized Water Reactor (PWR) with a large dry containment. Sequences important to LERF include steam generator tube rupture events and inter-system loss-of-coolant-accident events. These were not the dominant core damage sequences for this finding. Based on the Detailed Risk Evaluation, the Senior Reactor Analysts determined that the finding was of very low safety significance (Green).
05000456/FIN-2013007-012013Q3BraidwoodFailure to Adequately Evaluate SAT Overcurrent Relay Settings in Design CalculationsThe inspectors identified a finding of very low safety significance for the licensees failure to ensure the system auxiliary transformer (SAT) 242-1 overcurrent relay provided protection coordination with upstream and downstream protective devices as required by Institute of Electrical and Electronics Engineers (IEEE)-242 and Design Document RPS-TG-3. Specifically, the licensee failed to demonstrate the relays would have provided upstream directional discrimination to allow the offsite power to clear a system fault before disconnecting the plant from the grid. The licensee entered this issue into their corrective action program and after further evaluation concluded the SAT overcurrent relay settings were still acceptable. The inspectors determined the performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, it would have increased the likelihood of events that upset plant stability and affected the availability and reliability of the preferred alternating current (AC) power supply. The inspectors determined the finding was 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 inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000456/FIN-2013007-022013Q3BraidwoodFailure to Ensure Six Component Cooling (CC) System Manual Valves Were in the Correct Position as Required by Technical Specification (TS) Surveillance Requirement (SR) 3.7.7.1.The inspectors identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification Surveillance Requirement 3.7.7.1, for the licensees failure to ensure six component cooling (CC) system manual valves in the flow path servicing safety-related equipment, that were not locked, sealed, or otherwise secured in position, were verified in the correct position every 31 days. The licensee entered this finding into their Correction Action Program, verified the correct position of the six CC system manual valves, and revised surveillance procedures to include the requirement to periodically verify the correct position of these valves. The performance deficiency was determined to be more than minor because it was similar to IMC 0612, Appendix E, Example 3.c, since more than one valve was in the required position, but not locked, sealed, or otherwise secured in the correct position, and it impacted the Mitigating Systems cornerstones objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences, (i.e., core damage). Since the finding did not represent an actual loss of safety function, the inspectors screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000456/FIN-2013007-032013Q3BraidwoodFailure to Incorporate Accident Flows in Component Cooling Water Pump Net Positive Suction Head Calculations.The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to incorporate accident flows in component cooling water (CCW) pump net positive suction head (NPSH) available calculations. Specifically, the licensee failed to calculate the NPSH for the CCW pumps using the run-out flows, which would have resulted in much lower available NPSH. The licensee entered this issue into their Corrective Action Program and recalculated the CCW pump available NPSH and determined that margin remained. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of the CCW system to respond to an initiating event to prevent undesirable consequences. Specifically, by failing to consider the accident loads in the CCW pumps NPSH calculations there was reasonable doubt as to whether the CCW pumps would have been operable during accident conditions. The inspectors determined that the finding was of very low safety significance (Green) because it did not result in the loss of operability or an actual loss of the CCW system. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000456/FIN-2013007-042013Q3BraidwoodFailure to Consider Adequate Tornado Missile Protection in SX Discharge PipeThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to consider design control measures commensurate with those applied to the original essential service water (SX) design related to tornado missile protection. Specifically, the licensee processed a physical modification to the SX discharge pipe and failed to protect or evaluate the exposed portion from potential tornado missiles. The licensee entered this issue into their Corrective Action Program and showed by calculation that the modified SX pipe would shear off upon impact from the design basis tornado missile and the safety-related portion would be unharmed and capable of performing its intended function. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of the SX system to respond to an initiating event to prevent undesirable consequences. Specifically, by failing to consider tornado missile protection in the SX design, there was reasonable doubt as to whether the SX pumps would have been operable during accident conditions. Since the finding would degrade two or more trains of a multi-train system or function, the inspectors determined a Detailed Risk-Evaluation was required. Based on the Detailed Risk-Evaluation, the Senior Reactor Analysts determined the delta core damage frequency for the finding was 6.66E-7/yr and was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000456/FIN-2013007-052013Q3BraidwoodFailure to Consider Multiple Failures in the Emergency Operating Procedures (EOPs) 1(2)BwEP ES 1.3, Transfer to Cold leg Recirculation as Required by Technical Specification. (TS) (Section 5.4.1b)The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to ensure abnormal operating Procedures (AOPs) 1(2)BwOA S/D-2, Shutdown LOCA, Revision 104 (105 for Unit 2) contained the necessary actions to immediately terminate Safety Injection (SI) flow if reactor coolant system (RCS) leakage was isolated. Specifically, the licensee failed to update 1(2)BwOA S/D-2, Shutdown LOCA to Revision 2 of the Westinghouse Owners Group (WOG) Abnormal Response Guideline (ARG)-2, Shutdown LOCA, that resulted in a CAUTION not added to terminate SI flow in a timely manner to prevent RCS over-pressurization, if RCS leakage was isolated. The licensee entered this finding into their Correction Action Program to add the CAUTION statement in the procedure. The finding was determined to be more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedure quality and affected the cornerstones objective of providing reasonable assurance that physical design barriers protect the public from radioactive releases caused by accidents or events. Operations in accordance with the procedure may have challenged the RCS barrier during a shutdown LOCA event. Specifically, the licensee failed to update Procedure 1(2)BwOA S/D-2, Shutdown LOCA to Revision 2 of the WOG ARG-2, Shutdown LOCA guideline that resulted in a CAUTION that was not added to terminate SI flow in a timely manner to prevent RCS over-pressurization, if RCS leakage was isolated. The inspectors conducted an assessment of the risk significance of the issue in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors determined the finding did not require a Phase II assessment and was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000456/FIN-2013007-062013Q3BraidwoodProcedures for Shutdown Loss of Coolant Accident (LOCA) Not Appropriate If RCS Leakage Is IsolatedThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification, Section 5.4.1b for the licensees failure to establish the necessary actions as required in Emergency Operating Procedures (EOPs) 1(2)BwEP ES-1.3, Transfer to Cold Leg Recirculation, Revision 201. Specifically, the licensee failed to ensure EOPs 1(2)BwEP ES-1.3 contained the necessary actions for transition to 1(2)BwCA-1.1, Loss of Emergency Coolant Recirculation for a small loss of coolant accident (SLOCA) or medium loss of coolant accident (MLOCA) with a concurrent failure of residual heat removal (RHR) heat exchanger (HX) to safety injection (SI) and centrifugal charging pump (CCP) isolation valves. The licensee entered this finding into their Correction Action Program to revise the subject procedures. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and affected the cornerstones 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 licensee failed to ensure the procedure for establishing containment sump recirculation for a SLOCA or MLOCA contained the necessary actions for potential equipment failures. Since the finding resulted in the potential for a loss of the containment sump recirculation function during a SLOCA or MLOCA for certain equipment failures when transferring to containment sump recirculation, the inspectors determined a Detailed Risk-Evaluation was required. Based on the Detailed Risk-Evaluation, the Senior Reactor Analysts determined the delta core damage frequency for the finding was 1.0E-8/yr. and was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000456/FIN-2013007-072013Q3BraidwoodLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, through an operating experience review for the failure to verify the Component Cooling Water System was capable of withstanding a reactor coolant pump (RCP) thermal barrier break. Specifically, the licensee failed to evaluate the impact of a failure of valve CC685 to automatically isolate during a postulated RCP thermal barrier rupture event. 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 (i.e., core damage). The inspectors determined whether finding was of very low safety significance (Green) using because it did not result in a loss of operability or function. The licensee entered this issue into their Corrective Action Program as AR 01452558 and determined design pressures or temperatures would not be exceeded in the event of a thermal barrier break.
05000305/FIN-2013003-012013Q2KewauneeInadequate Procedure for Testing of the Diesel Room Ventilation Damper Actuator Back Up Air SystemA finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors, for the failure to have procedures appropriate to the circumstances for an activity that affects quality. Specifically, Operating Surveillance Procedure OP-KW-OSP-TAV-001A (and B), Diesel Generator A (or B) Back Up Air Supply Leak Rate Test, allowed the performer to open safety related (SR) pressure boundary valves and install non-safety related (NSR) test equipment on both back up air bottle sets without declaring the respective Emergency Diesel Generator (EDG) inoperable. The licensee initiated a condition report and revised both procedures to prevent both bottle sets from being tested at the same time while maintaining the respective diesel operable. The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to lead to a more significant safety concern. Specifically, the licensee concluded that procedures OP-KW-OSP-TAV-001A (and B) allowed unqualified test equipment to be relied upon as the SR pressure boundary for both back up air bottle sets without declaring the respective EDG inoperable. The inspectors concluded this finding was associated with the Mitigating Systems Cornerstone. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The licensee evaluated the installed test equipment and hose connections and concluded their pressure rating exceeded that necessary to function as a pressure boundary; therefore, the inspectors answered Yes to Mitigating Systems Screening question number 1, and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, Resources, because the licensee did not assure that procedures were adequate to assure nuclear safety.
05000305/FIN-2013003-022013Q2KewauneeLicensee-Identified Violation10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, procedure OP-KW-NOP-FH-003, Reactor Cavity Draining With Fuel or Upper Internals Installed, prescribed instructions for draining of the refueling cavity, in order to refill the refueling water storage tank in preparation for making the emergency core cooling system operable. However, the procedure was not appropriate to the circumstances because it did not prescribe that if procedure NSP-SI-004, Monitoring SI System for Void After System Refill, was performed prior to the cavity draining to the RWST, that procedure NSP-SI-004 shall be re-performed to ensure the emergency core cooling system was free from voids, which could impact availability of the system. Consequently, on April 24, 2012, the licensee performed NSP-SI-004 and ensured the system was free of voids to ensure SI system availability and then completed draining of the refueling cavity to the refueling water storage tank in the emergency core cooling system. The system was not monitored for voids following the refueling water storage tank refill and on June 27, 2012, during the normally scheduled surveillance to perform testing to ensure the system was free of voids, the licensee discovered a 2.8 cubic-foot void near the common suction of the SI pumps. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of configuration control and affected the Cornerstone objective of ensuring the reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The risk associated with this issue was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating system and the system was operable but nonconforming. The licensee conducted full scale tests of the system, which demonstrated that the system was available with the gas void present for all accident conditions. The licensee entered this issue into the CAP as CR480150, Gas Void in SI Pump Suction Piping, conducted and apparent cause evaluation and shared the operating experience with the industry.
05000315/FIN-2013010-012013Q2CookFailure to Maintain Emergency Operating Procedures for Mitigating the Consequences of a SGTR per TS Section 5.4.1, Procedures.The inspectors identified a finding of very low safety significance, with two associated NCVs of Technical Specification (TS), Section 5.4.1, Procedures, and TS 3.7.4, Steam Generator (SG) Power Operated Relief Valves (PORVs), for the failure to implement design measures which were consistent with the licensing bases for a Steam Generator Tube Rupture (SGTR) concurrent with a Loss of Offsite Power (LOOP) to the station. Specifically, the licensees emergency operating procedures (EOPs) 1(2) OHP-4023-E-3, Steam Generator Tube Rupture, failed to provide adequate actions to mitigate the consequences of a SGTR, coincident with a LOOP, in sufficient time to prevent overfilling the ruptured steam generator. Additionally, the licensee failed to declare the affected units SG PORVs inoperable and complete the required actions when the non-safety-related control air compressor (CAC) was made unavailable and incapable of providing its required support function. With the units CAC unavailable, the SG PORVs would not be capable of being remotely operated from the control room during a SGTR concurrent with the LOOP. The licensee entered this issue into their corrective action program and completed modifications to establish Nitrogen as another motive force to support SG PORV operability. This 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using the SDP in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power. Based on the Detailed Risk Evaluation required, the inspectors determined the finding was of very low safety significance (Green) because the resulting change in the Core Damage Frequency (CDF) was equal to 2.4E-8/yr. The inspectors determined the cause of this finding involved the crosscutting area of human performance, the component of decision making, and the aspect of conservative assumptions, H.1(b) in that the licensee did not adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather licensee incorrectly assumed the unaffected units plant air system (not backed by the emergency diesel generators) would be available during the SGTR scenario to supply motive power to the affected units SG PORVs. This assumption failed to take into account the licensing basis requirement of considering a SGTR and a loss of offsite power to the station (both units).
05000315/FIN-2013010-022013Q2CookFailure to Enter the Limiting Condition for Operations and Perform Required Actions per TS 3.7.4, SG PORVs.The inspectors identified a finding of very low safety significance, with two associated NCVs of Technical Specification (TS), Section 5.4.1, Procedures, and TS 3.7.4, Steam Generator (SG) Power Operated Relief Valves (PORVs), for the failure to implement design measures which were consistent with the licensing bases for a Steam Generator Tube Rupture (SGTR) concurrent with a Loss of Offsite Power (LOOP) to the station. Specifically, the licensees emergency operating procedures (EOPs) 1(2) OHP-4023-E-3, Steam Generator Tube Rupture, failed to provide adequate actions to mitigate the consequences of a SGTR, coincident with a LOOP, in sufficient time to prevent overfilling the ruptured steam generator. Additionally, the licensee failed to declare the affected units SG PORVs inoperable and complete the required actions when the non-safety-related control air compressor (CAC) was made unavailable and incapable of providing its required support function. With the units CAC unavailable, the SG PORVs would not be capable of being remotely operated from the control room during a SGTR concurrent with the LOOP. The licensee entered this issue into their corrective action program and completed modifications to establish Nitrogen as another motive force to support SG PORV operability. This 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using the SDP in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power. Based on the Detailed Risk Evaluation required, the inspectors determined the finding was of very low safety significance (Green) because the resulting change in the Core Damage Frequency (CDF) was equal to 2.4E-8/yr. The inspectors determined the cause of this finding involved the crosscutting area of human performance, the component of decision making, and the aspect of conservative assumptions, H.1(b) in that the licensee did not adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather licensee incorrectly assumed the unaffected units plant air system (not backed by the emergency diesel generators) would be available during the SGTR scenario to supply motive power to the affected units SG PORVs. This assumption failed to take into account the licensing basis requirement of considering a SGTR and a loss of offsite power to the station (both units).
05000315/FIN-2012007-012012Q4CookNon-conservative Condensate Storage Tank (CST) Cross-Tie NPSH CalculationThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure sufficient water volume in the condensate storage tank when both units auxiliary feedwater (AFW) pumps are aligned to a single condensate storage tank (CST.) Specifically, the licensee failed to perform a calculation to demonstrate sufficient volume and level to prevent net positive suction head and vortex issues when a single CST is providing water to all six AFW pumps as allowed by procedures. The licensees corrective action included performing a formal calculation and increasing the available water volume in the CST when both units AFW pumps are cross-tied. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System 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. The finding screened as of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability. Specifically, the licensee performed an operability determination which concluded the actual useable tank level during the previous 12 months had been sufficient. The inspectors determined the cause of this finding did not represent current licensee performance and, thus, no cross-cutting aspect was assigned.
05000305/FIN-2012005-052012Q4KewauneeRelay Room Carbon Dioxide Fire Suppression System and Control Room Envelope Potentially Affected by HELBDuring the inspectors review of the relay room carbon dioxide (CO2) fire suppression system actuation failure, the inspectors identified that control cabinets for both the CO2 fire suppression actuation system and the relay room ventilation damper ACC-22, are mounted on the outside of the SR east wall of the relay room and exposed to the turbine building. The control cabinets could be exposed to steam from a nearby 30-inch steam header if a crack were to develop. The inspectors were concerned that the potential existed to actuate the relay room CO2 fire suppression system and/or cause one of the relay room ventilation dampers to open, complicating the control room response to the small steam break. At the conclusion of this inspection period, the licensee was evaluating the issue; the inspectors needed additional information to determine if a performance deficiency existed. As a result, this item was considered unresolved (URI 05000305/2012005-05, Relay Room Carbon Dioxide Fire Suppression System and Control Room Envelope Potentially Affected by HELB).
05000315/FIN-2012007-022012Q4CookQualification Basis for Safety-Related Relays and Motor-Starter ContactorsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure sufficient water volume in the condensate storage tank when both units auxiliary feedwater (AFW) pumps are aligned to a single condensate storage tank (CST.) Specifically, the licensee failed to perform a calculation to demonstrate sufficient volume and level to prevent net positive suction head and vortex issues when a single CST is providing water to all six AFW pumps as allowed by procedures. The licensees corrective action included performing a formal calculation and increasing the available water volume in the CST when both units AFW pumps are cross-tied. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System 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. The finding screened as of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability. Specifically, the licensee performed an operability determination which concluded the actual useable tank level during the previous 12 months had been sufficient. The inspectors determined the cause of this finding did not represent current licensee performance and, thus, no cross-cutting aspect was assigned.
05000315/FIN-2012007-032012Q4CookConcerns with Periodic Design Basis Testing of Installed Relays and Motor-Starter ContactorsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure sufficient water volume in the condensate storage tank when both units auxiliary feedwater (AFW) pumps are aligned to a single condensate storage tank (CST.) Specifically, the licensee failed to perform a calculation to demonstrate sufficient volume and level to prevent net positive suction head and vortex issues when a single CST is providing water to all six AFW pumps as allowed by procedures. The licensees corrective action included performing a formal calculation and increasing the available water volume in the CST when both units AFW pumps are cross-tied. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System 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. The finding screened as of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability. Specifically, the licensee performed an operability determination which concluded the actual useable tank level during the previous 12 months had been sufficient. The inspectors determined the cause of this finding did not represent current licensee performance and, thus, no cross-cutting aspect was assigned.
05000315/FIN-2012007-042012Q4CookConcerns with Ensuring Margin to Overfill in a Ruptured SGDuring the week of July 23, 2012, the inspectors reviewed the licensing bases and plant response to a SGTR event. During the postulated design basis event, operators prevent overfill of the ruptured SG using the SG PORVs. During a SGTR, the normal source of motive force for the SG PORVs is air supplied by the compressed air system (CAS). The CAS compressors are powered from the offsite power supply and are non-safety-related. Assuming a concurrent loss of offsite power (LOOP) to the station, the only available source of immediate and remote (from the control room) motive force is air supplied by the control air compressor (CAC) since it could be powered from an EDG. The CAC is also non-safety-related. The inspectors also noted the backup Nitrogen system could be manually aligned to provide motive force for the SG PORVs. The facilitys EOPs 1(2) OHP-4023-E-3, Steam Generator Tube Rupture, Step 7, directs the operators to perform a rapid RCS cooldown by fully opening the SG PORVs on the 3 intact SGs. The margin to overfill (MTO) analysis assumes this step occurs on demand (the SGTR PORVs open immediately as soon as the operators manipulate the valves from the control room). Once the SG PORVs are full open, the MTO analysis assumes it takes 12 minutes to complete the RCS cooldown. This is a calculated (modeled) value which takes into account plant specific characteristics. If the SG PORVs do not open, the EOPs direct operators to establish backup Nitrogen. Because aligning Nitrogen to the SG PORVs is a manual operation requiring multiple manipulations outside the control room, the inspectors were concerned the additional time to complete these actions would result in a longer time to complete the RCS cooldown than assumed in the MTO analysis and the affected SG could overfill. On Friday, July 27, 2012, the inspectors requested the licensee to provide a non-licensed operator to perform a walkthrough of the procedures for establishing backup Nitrogen to locally open the SG PORVs. Although no significant issues were discovered during performance of the evolution, the operator required 13 minutes to establish backup Nitrogen to the auxiliary building and 14 minutes to place the local control stations in service and open the SG PORVs, a total of 27 minutes. The inspectors were concerned because this additional 27 minutes was not accounted for in the MTO analysis. Specifically, the MTO analysis assumes a total of 12 minutes for RCS cooldown to occur from the moment the operators get direction to open the SG PORVs (assumed to occur immediately) until commencement of RCS depressurization. The MTO analysis calculates it will take the operators and the plant a total of 52 minutes to mitigate the SGTR accident (no more RCS flow through the ruptured SG) and thus prevent overfilling the affect SG with a calculated MTO of approximated 8ft3. The MTO analysis has no margin to accommodate the additional 27 minutes for RCS cooldown and therefore the procedure fails to potentially prevent overfilling the ruptured SG. In response to the inspectors concerns, the licensee implemented immediate compensatory actions to ensure (1) backup Nitrogen was continuously available to the auxiliary building and (2) an operator would be immediately dispatched to commence lining up backup Nitrogen to the SG PORVs if a SGTR with a concurrent LOOP occurred while the EDG or CAC was unavailable. The licensee disagreed with the inspectors assumption with respect to the initial conditions of a LOOP affecting both units. The licensee believes their licensing basis for a SGTR is a concurrent LOOP in the affected unit and the unaffected unit maintains an intact source of offsite power. In the licensees scenario, the affected units SG PORVs would still have a source of immediate remote (from the control room) motive force via the compressed air system. On the contrary, the inspectors believe the licensees design bases accident is a SGTR coincident with a LOOP to the station (both units). To resolve this issue, the inspectors requested support from NRR through a concurrence Task Interface Agreement (TIA) to determine the licensing bases for this event. In a TIA memorandum dated December 7, 2012, NRR concluded the licensing bases included a LOOP for both units (ML13011A382). Although the inspectors received a response from NRR, the issue remained open awaiting additional licensing bases information from the licensee. This issue is considered an unresolved item (URI) pending further review this information (URI 05000315/2012007-04, 05000316/2012007-04; Concerns with Ensuring Margin to Overfill in a Ruptured SG).
05000315/FIN-2012007-052012Q4CookConcerns with Operability of SG PORVs with Control Air Compressor UnavailableThe inspectors identified an issue related to the definition of operability of the SG PORVs as specified by the TS. The most limiting design bases accident for the SG PORVs, is a tube rupture event. During this accident, the operators prevent overfill of the ruptured steam generator using the SG PORVs. Assuming a coincident loss of offsite power (LOOP) to the station (affecting both units), the only readily available source of pneumatic motive force for the SG PORVs is the unit-specific CAC, which has the capability of being powered from onsite emergency power (EDGs). The TS bases for the SG PORVs (B 3.7.4) states the Control Air System (the system composed of the CAC) provides the normal air supply for pneumatic control. Each unit-specific CAC is not safety-related and is not subject to a TS limiting condition for operation. Therefore, these compressors could be unavailable (i.e., for maintenance) for an indeterminate length of time, consistent with the performance goals established for the maintenance rule, regardless of the units current mode of operation. With the exception of emergent repair maintenance, all preventive maintenance is performed on the CACs when the units are online, in Mode 1. As defined in the facilitys Unit 1 and Unit 2 TS, A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function are also capable of performing their related support function(s). The electrical and control air appurtenances for the SG PORVs are non-safety grade and do not have an associated TS operability requirement. However, since the electrical control power and control air system are credited to ensure the SG PORVs will operate to perform the mitigating function of cooling the RCS during the SGTR accident with a LOOP, the inspectors concluded this equipment is required to be functional for the SG PORVs to be considered OPERABLE. The inspectors indentified several occasions when the CAC was non-functional; however, the licensee did not declare the PORVs inoperable. Specifically, a review of the facilitys unavailability records for the Unit 1 and Unit 2 CACs from January 1, 2000 to August 20, 2012 identified 13 instances (5 associated with Unit 1 and 8 associated Unit 2) when the units CAC was unavailable for greater than 24 hours, the TS allowed outage time for two or more SG PORVs being inoperable. In three of those instances for Unit 1 (April 18, 2001, for 79.8 hrs, November 23, 2003, for 133.2 hrs, and April 7, 2008, for 108.9 hrs) and two of those instances for Unit 2 (February 12, 2003, for 71.8 hrs and January 16, 2006, for 103.5 hrs) the unavailability time of the CAC was in excess of the total TS allowed outage times of 54 hours to place the unit in a mode where the limiting condition of operation does not apply. As stated in Section 1R21.6(b)(1), the licensee disagreed with the inspectors initial assumption with respect to the initial conditions of a LOOP affecting both units. The licensee believes their licensing basis for a SGTR is a concurrent LOOP in the affected unit and the unaffected unit maintains an intact source of offsite power. In the licensees scenario, the affected units SG PORVs would still have a source of immediate remote.
05000315/FIN-2012008-062012Q3Cook10 CFR 50.59 evaluation for modification of RHR pump casing drain lines was not performedThe inspectors identified a finding of very low safety significance and associated Severity Level IV violation of 10 CFR 50.59, Changes, Tests, and Experiments, for the 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 failed to provide a basis for not applying for a license amendment associated with a modification of the residual heat removal pump casing drain lines. The finding was entered into the licensees corrective action program to: (1) stage a hose and pipe couplings to support venting at the residual heat removal pump casing vent; (2) create a work order request to flush flow through the abandoned drain lines that were cut from the pump casing vent to show the lines could still pass water; (3) develop an alternate means to accomplish this activity; and (4) evaluate the change. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability and availability of systems that respond to initiating events to prevent undesirable consequences. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process because they are considered to be violations that potentially impede or impact the regulatory process. The finding screened as of very low safety significance (Green) using a Phase II evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not confirmed to reflect current performance due to the age of the performance deficiency.
05000316/FIN-2012008-012012Q3CookVortexing was not evaluated for the volume control, containment spray additive, refueling water storage tanksThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to evaluate vortexing in the volume control, containment spray additive, and refueling water storage tanks. Consequently, the minimum allowable submergence for the suction piping of these tanks did not consider the potential for air entrainment due to vortices. This finding was entered into the licensees corrective action program to evaluate the potential for vortexing at these tanks and revise the affected calculations The performance deficiency was determined to be more than minor because it was associated with the Mitigating System 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. In addition, the finding was associated with the Containment Barrier cornerstone attribute of structure, system, component, and barrier performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as of very low safety significance (Green) because: (1) the finding examples associated with the volume control and refueling water storage tanks were deficiencies confirmed not to result in loss of operability in that the licensee performed an evaluation that reasonably concluded the current limit setpoints prevent vortexing in these tanks; and (2) the finding example associated with the containment spray (CTA) additive tank was a design deficiency of the physical integrity of the reactor containment that did not affect the barrier function of the control room against smoke or a toxic atmosphere, represent an actual open pathway in the physical integrity of reactor containment, or involve an actual reduction in function of hydrogen igniters in the reactor containment. This finding did not have an associated cross-cutting aspect because it was not confirmed to reflect current performance due to the age of the performance deficiency.
05000316/FIN-2012008-022012Q3CookIncomplete methodology for developing acceptance criteria for suction voidsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to correctly incorporate the interim methodology for developing acceptance criteria for suction voids in Emergency Core Cooling Systems, Decay Heat Removal, and Containment Spray Systems pumps into procedures. Specifically, the licensee did not translate the limitations of the acceptance criteria with respect to rated performance of pump operation. This finding was entered into the licensees corrective action program to revise the affected procedure. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System 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. The finding screened as of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability. Specifically, a review of recent monitoring results determined that identified voids did not exceed the applicable acceptance criteria. The inspectors did not find an applicable cross-cutting aspect which represented the underlying cause of this performance deficiency; therefore, no cross-cutting aspect was assigned.
05000316/FIN-2012008-032012Q3CookProcedures were not developed for performance monitoring of plant parametersThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to establish appropriate procedures to implement the requirements of performance monitoring of plant parameters for gas accumulation. Specifically, the licensee had not established instructions in procedures to control important aspects such as frequency of monitoring and acceptance criteria. This finding was entered into the licensees corrective action program to determine the size of the limiting voids at the affected locations and establish appropriate acceptance criteria. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding screened as of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability. Specifically, a review of a sample of recent plant parameter trends determined that unacceptable void formation had not occurred. This finding had a cross-cutting aspect in the area of human performance because the licensee did not make safety significant decisions using a systematic process. Specifically, the licensee decided to use informal trending mechanisms to track the critical plant parameters instead of creating a formal and systematic approach to programmatically control the activity.
05000316/FIN-2012008-042012Q3CookMinimum flowrates and time requirements for dynamic flushing were not establishedThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to establish minimum flowrate and time required in procedures used to perform dynamic flushing activities affecting Emergency Core Cooling Systems, Decay Heat Removal, and Containment Spray Systems pumps. This finding was entered into the licensees corrective action program to revise the affected procedures. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to establish an appropriate procedure for flushing would have the potential of not removing voids to ensure system operability. The finding screened as of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability. Specifically, a historical review of previous dynamic flushing activities determined that sufficient flowrates and time values were achieved at the appropriate sequences. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not confirmed to reflect current performance due to the age of the performance deficiency.
05000316/FIN-2012008-052012Q3CookInadequate Procedure for Responding to a MODE 4 LOCAThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to include adequate venting instructions in the procedure use to respond to a MODE 4 loss-of-coolant accident. Specifically, the procedure did not include instructions to address all of the affected residual heat removal system high points, including the discharge piping. The finding was entered into the licensees corrective action program to leave one train of the system idle while the other train cools down the reactor coolant system below 200F to ensure that the discharge side of one train of residual heat removal system is not vulnerable to steam formation. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding screened as of very low safety significance (Green) using a Phase II evaluation. The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate relevant external operating experience. Specifically, the licensees evaluation of Information Notice 2010-11 incorrectly concluded that procedures contained sufficient direction to preclude flashing.
05000316/FIN-2012008-072012Q3Cook10 CFR 50.59 evaluation for modification of RHR pump casing drain lines was not performedThe inspectors identified a finding of very low safety significance and associated Severity Level IV violation of 10 CFR 50.59, Changes, Tests, and Experiments, for the 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 failed to provide a basis for not applying for a license amendment associated with a modification of the residual heat removal pump casing drain lines. The finding was entered into the licensees corrective action program to: (1) stage a hose and pipe couplings to support venting at the residual heat removal pump casing vent; (2) create a work order request to flush flow through the abandoned drain lines that were cut from the pump casing vent to show the lines could still pass water; (3) develop an alternate means to accomplish this activity; and (4) evaluate the change. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability and availability of systems that respond to initiating events to prevent undesirable consequences. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process because they are considered to be violations that potentially impede or impact the regulatory process. The finding screened as of very low safety significance (Green) using a Phase II evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not confirmed to reflect current performance due to the age of the performance deficiency.
05000316/FIN-2012008-082012Q3CookInadequate Procedure for RCS Vacuum Fill During Reduced Inventory OperationsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to establish procedures for reduced inventory operations that were appropriate to preclude air entrainment into Residual Heat Removal (RHR) and Reactor Coolant Systems (RCS). Specifically, a procedure allowed operation of RHR while in reduced inventory operations with a minimum RCS level and maximum pump flowrate combination that was determined to result in air-entrainment vortices. The finding was entered into the licensees corrective action program to place an administrative hold to the procedure until proper documentation is revised and updated and to revise the procedure to require stricter use of high accuracy level instrumentation. The performance deficiency was determined to be more than minor because it was associated with the initiating event cornerstone attribute of procedure quality and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The finding screened as of very low safety significance (Green) using a Phase II evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not confirmed to reflect current performance due to the age of the performance deficiency.
05000316/FIN-2012008-092012Q3CookComputer Program Used for Operability Evaluation Was Not BenchmarkedThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to establish procedures for reduced inventory operations that were appropriate to preclude air entrainment into Residual Heat Removal (RHR) and Reactor Coolant Systems (RCS). Specifically, a procedure allowed operation of RHR while in reduced inventory operations with a minimum RCS level and maximum pump flowrate combination that was determined to result in air-entrainment vortices. The finding was entered into the licensee s corrective action program to place an administrative hold to the procedure until proper documentation is revised and updated and to revise the procedure to require stricter use of high accuracy level instrumentation. The performance deficiency was determined to be more than minor because it was associated with the initiating event cornerstone attribute of procedure quality and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. The finding screened as of very low safety significance (Green) using a Phase II evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not confirmed to reflect current performance due to the age of the performance deficiency. The licensee used Flow-3D Version 9.0 for a past-operability evaluation of a gas void found in the containment recirculation sump suction piping. Specifically: Calculation ALION-CAL-AEP-4462-04, D.C. Cook Recirculation Containment Sump Hydraulic Analysis Task 4 Results, was performed in the early development of the GSI-191 response effort to postulate an air pocket venting from the CS and RHR suction lines into the recirculation sump. This analysis was performed to evaluate the necessary transition period to avoid air entrainment in the pumps and the pressure transients as they impact the recirculation sump strainers. This analysis was not used for any design basis or engineering modification; however, it was used as a tool for a past-operability evaluation of a gas void found in the sump suction piping. Calculation ALION-CAL-AEP-7354-02, D.C. Cook Unit 1 Operability Analysis to Evaluate Gas Void in ECCS Sump Suction Piping, was performed for a void evaluated in the same location as the one evaluated by ALION-CAL-AEP-4462-04. However, this analysis was performed to evaluate the past-operability of the void volume found in January 2009 as oppose to a postulated volume. Specifically, an air pocket was discovered downstream of the Unit 1 sump isolation valve (i.e., 1-ICM-306) in the B train suction pipe. This discovery was captured in the CAP as AR 00844125. The licensee did not own or maintain a license for Flow-3D. These analyses were performed by an outside contractor under the contractor s Quality Assurance Program and were accepted by the licensee through their Owner s Acceptance Review Process. However, when the inspectors questioned if the computer code was verified against test data, the contractor informed the licensee that no benchmark flow modeling was conducted for these software applications. The licensee did not own or maintain a license for Flow-3D. These analyses were performed by an outside contractor under the contractor s Quality Assurance Program and were accepted by the licensee through their Owner s Acceptance Review Process. However, when the inspectors questioned if the computer code was verified against test data, the contractor informed the licensee that no benchmark flow modeling was conducted for these software applications. NRC Regulatory Issue Summary (RIS) 2005-20, Revision 1, Revision to NRC Inspection Manual Part 9900 Technical Guidance, Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety, informed licensees that the NRC had revised NRC Inspection Manual Part 9900. Guidance provided in Appendix C, Section C.4 of the inspection manual stated the use of any analytical method must be technically appropriate to characterize the SSCs involved, the nature of the degraded or nonconforming condition, and specific facility design. It further stated that general considerations for establishing this adequacy include, in part, acceptable alternative methods such as the use of best estimate codes, methods, and techniques. The inspection guidance also stated that in these cases, the evaluation should ensure that the SSC s performance is not over-predicted by performing a benchmark comparison of the non-current licensing basis (CLB) analysis methods to the applicable CLB analysis methods. The inspectors consulted with NRR and determined the operability evaluation relied on a computer code that has not been demonstrated to be technically appropriate for the intended applications. Specifically, the computer code had not been qualified by benchmarking against test or plant data to demonstrate its applicability to the type of analyses being conducted. This issue is unresolved pending licensee past-operability evaluation and determination of NRC courses of action for resolution of the issue.
05000316/FIN-2012008-102012Q3CookLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements, which meets the criteria of Section 2.3.2 of the NRC Enforcement Policy for being dispositioned as an NCV. Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed and accomplished by procedures appropriate to the circumstances. Contrary to the above, on or about December 14, 2011, the licensee identified that Procedure 1-OHP-SP-238, Venting Pressure from the SI Discharge Header, was not appropriate to the circumstances. Specifically, on December 14, 2011, the licensee found a void at the affected location during a periodic gas monitoring surveillance and determined a previous depressurization evolution led to void formation because the depressurization procedure did not include instructions to verify that the piping was left full of water. The corrective action being considered at the time of the inspection was to add a step to Procedure 1-OHP-SP-238 to perform ultrasonic examinations. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very low safety significance because the identified void was determined not to result in loss of operability of the safety injection system.
05000263/FIN-2011010-012011Q4MonticelloFailure to Follow Fire Water Aging Management Program Implementing ProcedureThe inspectors identified a finding of very low safety significance (Green) involving the licensees failure to accomplish activities affecting quality in accordance with procedures. Specifically, the licensee failed to incorporate operating experience in accordance with procedures. This impacted the licensees ability to implement an effective aging management program for the fire protection system. No violation of NRC requirements was identified. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (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). Using IMC 0609, Appendix F, Fire Protection SDP, and the Monticello SPAR model, the inspectors determined that this finding had very low safety significance. The inspectors did not identify an associated crosscutting aspect for this finding.
05000263/FIN-2011010-022011Q4MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of License Condition 2.C.4 through a planned surveillance test for the failure to implement and maintain in effect all provisions of their approved fire protection program. Specifically, the installation of the intake structure pre-action sprinkler system did not comply with NFPA 13 (1983) section 3-11.1.1, which requires that all sprinkler pipe and fittings shall be so installed that the system may be drained and resulted in the plugging of the sprinkler system. This prevented water from flowing through sprinkler heads and caused the system to be non-functional. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (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). The Region III Senior Risk Analyst (SRA) used the risk assessment tools of IMC 0609, Appendix F, Fire Protection SDP, and performed bounding analyses using the Monticello Standard Plant Analysis Risk (SPAR Model), Version 8.15. The SRA also reviewed and discussed the licensees bounding risk assessment documented in PRA Memo 11-01-, Revisions 0 and 1, Risk Assessment of Intake Fire Suppression System Plugging. The finding was determined to be of very low safety significance (green) because the risk increase using bounding assumptions was below 1E-6. The licensee entered this issue into their corrective action program as AR 01305183, Intake Fire Sprinkler Configuration Discrepancy, and restored the functionality of the sprinkler system by flushing the piping and replacing system components. The licensee further planned to modify the system to allow proper drainage in accordance with the design requirements.
05000263/FIN-2011010-032011Q4MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of License Condition 2.C.4 for the failure to implement and maintain in effect all provisions of their approved fire protection program. This includes adhering to the 10 CFR 50, Appendix B Quality Assurance Program requirements for the design, procurement, installation, testing and administrative controls for the fire protection program. Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as deficiencies are promptly identified and corrected. Contrary to the above, from August 21, 2007 until August 26, 2011, the licensee failed to promptly identify and correct a condition adverse to quality that resulted in the plugging of the intake structure sprinkler system. Specifically, the licensee failed to perform corrective actions (work order 342675-02) to flush the intake structure sprinkler system following a blockage event in the EDG rooms in 2007. The performance deficiency was determined to be more than minor because the plugging in the intake structure pre-action sprinkler system was left uncorrected for four years and became a more significant safety concern. The inspectors concluded that this finding was associated with the Mitigating Systems cornerstone. The Region III SRA used the risk assessment tools of IMC 0609, Appendix F, Fire Protection SDP, and performed bounding analyses using the Monticello Standard Plant Analysis Risk (SPAR Model), Version 8.15. The SRA also reviewed and discussed the licensees bounding risk assessment documented in PRA Memo 11-01-, Revisions 0 and 1, Risk Assessment of Intake Fire Suppression System Plugging. The finding was determined to be of very low safety significance (Green) because the risk increase using bounding assumptions was below 1E-6. The licensee flushed the system, restored functionality, and wrote AR 01303860 to document the multiple rescheduling.
05000263/FIN-2011010-042011Q4MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and associated NCV of License Condition 2.C.4 for the failure to implement and maintain in effect all provisions of their approved fire protection program. This includes adhering to the 10 CFR 50, Appendix B Quality Assurance Program requirements for the design, procurement, installation, testing and administrative controls for the fire protection program. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions procedures, or drawings. Contrary to the above, on April 30, 2009, the license failed to follow procedure FP-OP-OL-01 Operability/Functionality Determination, when assessing identified blockage in the intake structure fire protection sprinkler piping. Specifically, the assessor failed to justify assumptions, perform an extent of condition, and obtain additional condition bounding information to ensure an accurate assessment of the condition. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (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). Using IMC 0609, Appendix F, Fire Protection SDP, and the Monticello SPAR model, the inspectors determined that this finding had very low safety significance. The licensee entered this issue into their corrective action program as AR 01304353, Inaccurate functionality assessment for CAP 1180222, in order to perform further evaluation of the deficiency.
05000263/FIN-2011010-052011Q4MonticelloLicensee-Identified Violation\ The licensee identified a finding of very low safety significance (Green) and associated NCV of License condition 2.C.4 for the failure to implement and maintain in effect all provisions of their approved fire protection program. This includes adhering to the 10 CFR 50, Appendix B Quality Assurance Program requirements for the design, procurement, installation, testing and administrative controls for the fire protection program. Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that test results shall be documented and evaluated to assure that test requirements have been satisfied. Contrary to this requirement, on April 30, 2009, the licensee failed to document and evaluate the results of a PMT that did not meet all of its acceptance criteria. Specifically, when a step in the PMT required flow through the inspector test valve was not accomplished, the PMT was not annotated as failure and the PMT work order was signed off as complete without further evaluation. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (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). Using IMC 0609, Appendix F, Fire Protection SDP, and the Monticello SPAR model, the inspectors determined that this finding had very low safety significance. The licensee entered this issue into their corrective action program as AR 01304348, Failed PMT results not captured in PMT WO, in order to perform further evaluation of the deficiency.
05000440/FIN-2011008-042011Q4PerryFailure to Report Unanalyzed Condition Related to Internal FloodingThe inspectors identified a Severity Level IV violation of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Reactors, for failure to report within eight hours an unanalyzed condition that significantly degrades plant safety. Specifically, the licensee failed to notify NRC upon discovery of a postulated internal flood in the control complex could result in loss of single failure capability of safety-related equipment. This violation was entered into the licensees corrective action program. The performance deficiency was determined to involve a traditional enforcement violation because it potentially impeded or impacted the regulatory process. The traditional enforcement violation was determined to be more than minor because the information that was not provided through the event notification had a material impact on safety and licensed activities. The traditional enforcement violation was determined to be a Severity Level IV violation because the failure to report within eight hours an unanalyzed condition did not result in an unacceptable change to the facility or procedures. An evaluation for cross-cutting aspect was not applicable because this was a traditional enforcement violation.
05000440/FIN-2011008-012011Q4PerryFailure to Adequately Protect Safety Related Equipment from Internal FloodingThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to ensure safety-related equipment would be adequately protected from internal flooding. Specifically, the licensee failed to adequately evaluate the volume of water originating from a postulated crack in service water (SW) piping within the control complex. This finding was entered into the licensees corrective action program. The corrective actions included performing additional analyses, establishing compensatory measures, issuing procedure orders, and revising operating procedures. The performance deficiency was determined to be more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance and 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. Based on the Phase III Analysis, the inspectors determined the finding was of very low safety significance (Green). The inspectors determined the cause of this finding did not represent current licensee performance and no cross-cutting aspect was assigned.
05000440/FIN-2011008-022011Q4PerryInadequate Control Circuit Voltage Calculation for Safety- Related Motor Starter ContactorsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control , for failure to adequately evaluate the capability of motor control starter contactors to operate during design basis degraded voltage conditions. Specifically, the licensee did not analyze all circuit elements of resistance and failed to incorporate the latest results of calculated plant bus voltages. The performance deficiency was determined to be more than minor because it was associated with the Mitigating System 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. The finding screened as of very low safety significance (Green) because the finding involved a design or qualification deficiency that did not result in a loss of operability. Specifically, the licensee performed an operability evaluation taking into account all resistances in the circuit, the latest load flow analysis and test data and concluded there was sufficient voltage available. This finding has a cross-cutting aspect in the area of Resources for failure to ensure complete, accurate, and up-to-date design documentation, procedures, work packages and correct labeling of components.
05000440/FIN-2011008-032011Q4PerryFailure to Test Safety-Related Contactors at Degraded Voltage ConditionsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to test safety-related motor starter contactors at design basis conditions. Specifically, the licensee failed to demonstrate the ability of ESW Pump A discharge valve 1P45F0130A motor starter contactor to operate at minimum pickup voltage during design basis degraded voltage conditions. This finding was entered into the licensees corrective action program. The performance deficiency was determined to be more than minor because if left uncorrected it would have the potential to lead to a more significant safety concern. The finding screened as of very low safety significance (Green) because the finding involved a design or qualification deficiency that did not result in a loss of operability. Specifically, after further evaluation, the licensees engineering staff concluded the issue did not impact current operability because periodic testing for other type of contactors provided validation the valve motor contactor would operate when required for the postulated degraded voltage conditions. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution Corrective Action Program for failure to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
05000266/FIN-2011009-012011Q3Point BeachFailure to Monitor outside Air TemperatureThe inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to correctly translate design basis assumptions into procedures or instructions. Specifically, the licensee failed to monitor average outside air temperature which was one of the design input criteria for the temperature heat-up calculation associated with rooms which housed safety-related equipment. This finding was entered into the licensees corrective action program. The performance deficiency was associated with Mitigating System Cornerstone and determined to be more than minor because, if left uncorrected, it could lead to a more significant safety concern. The finding screened as 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 a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of human performance, resources because the licensee did not ensure adequate training and qualification of personnel. Specifically, the licensee failed to adequately train licensed operators to ensure adequate knowledge with respect to the interface between functionality of a non-safety system component and the impact of a failure on the operability of safety-related equipment.
05000266/FIN-2011009-022011Q3Point BeachFailure to Incorporate Minimum AFW Flow Requirement into Emergency ProceduresThe inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure a minimum AFW flow of 275 gpm as specified in the accident analysis for the Loss of Normal Feedwater event. This finding was entered into the licensees corrective action program. The performance deficiency was associated with the Mitigating Systems Cornerstone attribute of design control and was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, an AFW flow rate of less than 275 gpm as specified in the procedures did not ensure the pressurizer would not become water solid and cause an over-pressure condition within the Reactor Coolant System during the Loss of Normal Feedwater. The finding screened as 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 a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of human performance, resources because the licensee did not maintain design documentation in a complete and accurate manner. Specifically, the licensee failed to maintain Emergency Procedures consistent with the design basis analysis for LONF.
05000266/FIN-2011009-032011Q3Point BeachContainment Spray Pipe Support DeficienciesThe inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to ensure the Containment Spray Pipe Support 2S-249 and Containment Spray Pipe Anchor 2A-35 meet Seismic Category I requirements. This finding was entered into the licensees corrective action program. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. This finding is of very low safety significance (Green) because there was no actual barrier degradation. The inspectors did not identify a cross-cutting aspect associated with this finding because this was a legacy design issue; and therefore, was not reflective of current performance.
05000266/FIN-2011009-042011Q3Point BeachTurbine Building Structural Steel Floor Beams Did Not Meet AISC RequirementsThe inspectors identified a finding of very low safety significance involving the licensees failure to meet the requirements of the American Institute of Steel Construction (AISC) Specification. Specifically, the licensees design basis calculation failed to ensure the turbine building structural steel floor beams met the AISC specification. This finding was entered into the licensees corrective action program. No violation of NRC requirements was identified. The performance deficiency was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of those events that upset the plants stability and challenged critical safety functions during shutdown, as well as power operations. The finding screened as very low safety significance (Green), because the transient initiator would not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. This finding had a cross-cutting aspect in human performance and work practice because the licensee did not ensure effective supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to have adequate oversight of design calculation and documentation for establishing structural adequacy of the turbine building structural steel beams at EL. 44-0.
05000266/FIN-2011009-052011Q3Point BeachLicensee-Identified ViolationA finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified by the licensee for the failure to ensure adequate instructions were adequately prescribed in procedures. Specifically, the licensee failed to ensure the receptacle 2PR-49 listed in Procedure AOP-30, Temporary Ventilation for Vital Areas, as one of the three potential power sources for transformer X-71 adequate for the transformer plug, was acceptable, in that the receptacle and transformer had difference phase connections. This transformer would be used to power temporary fans relied upon for design basis accident and the loss of the normal/fixed ventilations in the AFW and switchgear rooms. The performance deficiency 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. The SDP Phase I evaluation concluded the finding screened as of very low safety significance. This issue was entered into the licensees corrective action as AR01652555, as a corrective action, the licensee prepared an EC 271778 to modify the receptacle during the next Unit Refueling Outage. The inspectors also noticed procedure AOP-30 still showed 2PR-49 as one of the potential power sources. The inspectors were concerned there were no compensatory measures in place identifying that this power source could not be used and also identifying other receptacles in the area that could be utilized as an interim measure. The licensee entered the inspectors concern into their corrective action program as AR01682644.
05000237/FIN-2011003-022011Q2DresdenInadequate Instructions for the Inspection of Safety-Related Portions of the Intake StructureA finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to establish adequate instructions for inspecting bay 13 and portions of the intake structure surrounding the diesel generator cooling water pumps. Specifically, the procedure that provides guidance for inspecting these structures lacked specific instructions on how to detect and record degradation by erosion and corrosion. The licensee entered this issue into the corrective action program and initiated procedure revisions to provide further direction for capturing the degradation of these structures and related components. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead a more significant safety concern. The finding screened as of very low safety significance because it was a qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, a qualitative assessment of historic surveillance reports found the documented results acceptable. The inspectors determined the cause of this finding did not represent current licensee performance and no cross-cutting aspect was assigned.
05000237/FIN-2011003-032011Q2DresdenInadequate Acceptance Criteria for Testing Equipment Relied Upon to Mitigate the Consequences of a Lock and Dam FailureA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified by the inspectors for the licensees failure to establish adequate acceptance criteria for testing equipment relied upon to mitigate the consequences of a dam failure. Specifically, the acceptance criteria in DOS 0010-01, Dresden Dam Failure Equipment Test, did not consider additional steps required to demonstrate the ability of the screen refuse pumps to deliver water to the enclosure for the safety-related pumps to support operability of the isolation condensers. The licensee entered this issue into the corrective action program and initiated procedure revisions to include these additional steps in the procedures acceptance criteria. The performance deficiency was determined to be more than minor because it adversely affected the availability, reliability, and capability of mitigating systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding screened as of very low safety significance because it was a qualification deficiency confirmed not to result in loss of operability or functionality. Specifically, the licensee estimated that the time it would take to perform the additional steps not included in the procedures acceptance criteria was within the time required. The inspectors determined the cause of this finding did not represent current licensee performance and no cross-cutting aspect was assigned.
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