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	Q	Site	Significance	CCA	Identified by	Title	Description
05000237/FIN-2009005-01	2009Q4	Dresden	Green	H.11 H.12	Self-revealing	Operating Personnel Incorrectly Placed Clearance Tags	A finding of very low safety significance and associated NCV of Technical Specification 5.4.1 was self-revealed for the failure to meet the requirements of Clearance Order (CO) 69631 by removing shorting links instead of fuses as required by the CO on November 12, 2009. As a result, protective relaying was unintentially removed from the Unit 2 main power transformer TR-2, the unit auxiliary transformer TR-21, and the reserve auxiliary transformer TR-22. This issue was entered into the licensees CAP as Issue Report 992290. Corrective actions included: coaching of the individuals involved with the incorrect placing of the out-of-service and a placard on the device that was incorrectly repositioned was changed to include the specific equipment part number of the shorting links. The finding was determined to be more than minor because the finding could reasonably be viewed as a precursor to a significant event. The finding was evaluated using the SDP in accordance with IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, Checklist 6, dated May 25, 2004. This checklist stated that for a finding to require a Phase 2 or 3 determination, it would require an increase in the likelihood of a loss of offsite power or degrade the licensees ability to cope with a loss of offsite power. The ability of the licensee to cope with a loss of offsite power was not impacted because at least one emergency diesel generator was operable during the entire period. The inspectors determined that neither of these conditions were met so the finding screened as Green. This finding had a cross-cutting aspect in the area of Human Performance, Work Practices. H.4(a
05000237/FIN-2009005-02	2009Q4	Dresden	Green	H.12	NRC identified	NRC Inspector-Identified Control Room Alarm Isolation Valve Out-of-Position	The inspectors identified a finding of very low safety significance and associated NCV of Technical Specification 5.4.1 for the licensee failing to follow Dresden procedure DOP 2-1500-M1, LPCI System Mechanical Checklist, Revision 39. On September 24, 2009, the inspectors identified valve 2-1501-42A, U2 low pressure coolant injection (LPCI) A pump gland leak-off valve, was closed instead of open as required by DOP 2-1500-M1. With this valve closed instead of open, the control room alarm for LPCI pump seal leakage would not have been able to fulfill its function. The issue was entered into the licensees CAP as IR 969490. The licensees corrective actions included changing maintenance procedure DMP 1500-05, LPCI Pump Maintenance, step G.25.d to include the case drain valve equipment numbers and sign offs to position and verify the valves; and Operations Department Management addressed the operations department personnel about this issue. The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, the valve isolated an alarm in the control room. The inspectors concluded this finding was associated with the Mitigating Systems Cornerstone using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, dated January 10, 2008. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not have any documentation as to how or when the valve was placed into the position it was in. The design and location of the valve precluded that the valve was accidently placed into the position it was found in. Therefore, the inspectors concluded that either the failure to use human error prevention techniques or maintaining proper documentation of activities caused the mispositioning of valve 2-1501-42A. H.4.(a)
05000237/FIN-2009005-03	2009Q4	Dresden	Green	H.4 H.5	NRC identified	Preconditioning the Unit 2 Emergency Diesel Generator Prior to Performing TS Surveillance Requirements	The inspectors identified a finding of very low significance and associated NCV of 10 CFR 50 Appendix B, Criterion XI, Test Control, because the licensee unacceptably preconditioned the Unit 2 Emergency Diesel Generator (EDG) prior to performing Technical Specification (TS) Surveillance Requirements (SR) 3.8.1.19.c.4, 3.8.1.12.c.3, and 3.8.1.10. These TS SRs involved verifying that the EDG supplied steady state frequency would be acceptable following a loss of offsite power coincident with and without a loss of coolant accident, and following the loss of the largest post-accident load. Specifically, the inspectors identified that the licensee routinely performed governor oil change outage maintenance activities which involved a section that tuned the Unit 2 diesel governors response to a load change just prior to performing these TS SRs. This issue has been entered into the licensees CAP as IR 1000609. The licensee had not reached a conclusion on corrective actions by the end of the inspection period. This finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Unacceptable preconditioning the EDG could mask latent performance issues and affect the ability of the EDG to supply safety-related power to vital loads during an event. The inspectors performed a Phase 1 SDP evaluation and determined that this issue was Green because it did not result in an inoperable Unit 2 EDG. The failure to adequately coordinate the work activity of the preventive maintenance and post-maintenance testing with the TS SR activities was the principal contributor to this finding and was reflective of recent performance. This finding had a cross-cutting aspect in the area of Work Control. Specifically, the licensee did not appropriately coordinate work activities by incorporating actions to address the impact of the work as different job activities. The scheduling of the work activities resulted in the pre-conditioning of the EDG prior to performing the surveillance tests. H.3(b
05000237/FIN-2009005-04	2009Q4	Dresden			NRC identified	2/3 Emergency Diesel Generator (EDG) Overvoltage During Division I Undervoltage Surveillance	The inspectors identified an URI regarding the regulatory requirements associated with the circumstances surrounding the 2/3 EDG overvoltage event on November 16, 2009. On November 16, 2009, at 10:53 a.m., a nuclear station operator (NSO) was performing step I.11.c per DOS 6600-06, Bus Undervoltage and ECCS Integrated Functional Test for Unit 2/3 Diesel Generator to Unit 2, Revision 46. At this time, the operator was attempting to synchronize Bus 23-1 (powered from 2/3 EDG) to Bus 23 (powered from reserve auxiliary transformer 22). The operator stated that he was only monitoring running versus on-coming bus voltage meters, which are transformed down and are only relative to actual bus voltages. The operator stated that a loud pop noise was heard from the 902-3 panel. At this time, the operator noticed that the 23-1/24-1 digital volt meter read around 5600 volts (was previously around 4100 volts). The 2/3 EDG was then shutdown per DOS 6600-06 step I.12. On step I.12.c, the voltage regulator would not lower (remained upscale). The EDG stopped after the 6-minute cool down and DOS 6600-06 was stopped. The licensee generated EACE 994101-07, 2/3 Emergency Diesel Generator (EDG) Voltage Transient, to determine the cause, extent of condition and corrective actions for this event. The inspectors reviewed EACE 994101-07 and interviewed the NSO who had performed DOS 6600-06. The inspectors raised more questions regarding the capabilities of the control room simulator used for training, procedure adequacy and the corrective actions in place. The inspectors plan to review the licensees response to their questions to determine if there were any violations of NRC requirements and that appropriate corrective actions were applied. (URI 05000237/2009005-04; 05000249/2009005-04
05000237/FIN-2009005-05	2009Q4	Dresden	Green	H.11 H.12	Self-revealing	Failure to Follow the Master Refueling Procedure During Movement of Fuel Assembly JLU569	A finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, was self-revealed for the failure to properly move a fuel assembly to its specified location, in accordance with DFP 0800-01, Master Refueling Procedure. Specifically, on November 5, 2004, fuel assembly JLU569 was placed in position C4-E5, instead of C4-F5, as required by the procedure. The violation was placed into the licensees CAP in IR 990180. As corrective action, the licensee temporarily suspended all fuel handling activities, conducted a piece count of the spent fuel and stationed a second Senior Reactor Operator on the refueling bridge as additional oversight for follow-on fuel movements. Additionally the fuel handling crew associated with the event was suspended from future fuel moves, pending remedial training. Using the guidance contained in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, dated December 4, 2008, the inspectors determined that the finding was more than minor because the finding was associated with the configuration control and human performance attributes of the Barrier Integrity Cornerstone and impacted the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (i.e., fuel cladding) protect the public from radionuclide releases caused by an accident or event. Specifically, the shutdown margin and thermal management of the spent fuel pool(s) is affected by fuel assembly placement inside the pool(s). The inspectors determined the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b, question 6, which directed the inspectors to Appendix M, Significance Determination Process Using Qualitative Criteria. Because probabilistic risk assessment tools were not well suited for this finding, the criteria for using IMC 0609, Appendix M, were met. In determining the significance of this finding, regional management reviewed the licensees bounding analysis in the UFSAR, which demonstrated that regardless of the incorrect bundle position in the fuel pool, the design of the pool still maintained pool Keff less than .95. Based on the additional qualitative circumstances associated with this finding, regional management concluded the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Practices. Specifically, neither the Senior Reactor Operator (SRO), nor either of the two members of the fuel handling crew, adequately performed independent verification techniques that ensured the fuel assembly move was made in accordance with the Nuclear Component Transfer List, as required by DFP 0800-01. H.4(a
05000237/FIN-2009005-06	2009Q4	Dresden			Self-revealing	Mispositioning of a Unit 3 Control Rod at Power	Mispositioning of a Unit 3 Control Rod at Powe A finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the mispositioning of a Unit 3 control rod at power. Control rod G-11 was withdrawn one notch contrary to TS SR 3.1.3.3 requirements to insert each withdrawn control rod at least one notch. This was a performance deficiency. The violation was entered into the licensees CAP as IR 993634. Corrective actions included inserting control rod G-11 one notch back to the original position and suspending control rod movement while all rods were verified to be in their correct position. The operator was removed from shift duties and the oncoming shift was briefed of the event. The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attributes of human performance and configuration control of a control rod, and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the operator withdrew a control rod contrary to expected operation. This added positive reactivity and caused an unanticipated power increase. The inspectors evaluated the finding using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Fuel Barrier Cornerstone. Per Table 4a, any issue that involves the fuel barrier is screened as Green. This finding had no cross-cutting aspect.
05000237/FIN-2009005-07	2009Q4	Dresden			NRC identified	Changes to EAL HU6 Potentially Decreased the Effectiveness of the Plans without Prior NRC Approval	The inspectors reviewed changes implemented to the Dresden Station Radiological Emergency Plan Annex EALs and EAL Basis. In Revision 24, the licensee changed the basis of EAL HU6, Fire not extinguished within 15 minutes of detection within the protected area boundary, by adding two statements. The two changes added to the EAL basis stated that if the alarm could not be verified by redundant control room or nearby fire panel indications, notification from the field that a fire exists starts the 15-minute classification and fire extinguishment clocks. The second change stated the 15-minute period to extinguish the fire does not start until either the fire alarm is verified to be valid by additional control room or nearby fire panel instrumentation, or upon notification of a fire from the field. These statements conflict with the previous Dresden Station Annex, Revision 23, basis statements and potentially decrease the effectiveness of the Plans. Dresden Station Radiological Emergency Plan Annex, Revision 23, EAL HU6, initiating condition stated, Fire not extinguished within 15 minutes of detection, or explosion, within the protected area boundary. The threshold values for HU6 were, in part: 1) Fire in any Table H2 area not extinguished within 15 minutes of Control Room notification or verification of a Control Room alarm, or 2) Fire outside any Table H2 area with the potential to damage safety systems in any Table H2 area not extinguished within 15 minutes of Control Room notification or verification of a Control Room alarm. Table H2, Vital Areas, were identified as reactor building, auxiliary electric room, control room, diesel generator rooms, 4 kilovolt emergency core cooling system switchgear area, battery rooms, control rod drive and component cooling service water pump rooms, turbine building cable tunnel, turbine building safe shutdown areas, and crib house. The basis defined fire as combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is not required if large quantities of smoke and heat are observed. The basis for Revision 23, EAL HU6 thresholds 1 and 2 stated, in part, the purpose of this threshold is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. As used here, notification is visual observation and report by plant personnel or sensor alarm indication. The 15-minute period begins with a credible notification that a fire is occurring or indication of a valid fire detection system alarm. A verified alarm is assumed to be an indication of a fire unless personnel dispatched to the scene disprove the alarm within the 15-minute period. The report, however, shall not be required to verify the alarm. The intent of the 15-minute period is to size the fire and discriminate against small fires that are readily extinguished (e.g., smoldering waste paper basket, etc.). Revision 24 of the Dresden Station Radiological Emergency Plan Annex, changed the threshold basis for EAL HU6 by adding the following two statements: 1) If the alarm cannot be verified by redundant control room or nearby fire panel indications, notification from the field that a fire exists starts the 15-minute classification and fire extinguishment clocks, and 2) The 15-minute period to extinguish the fire does not start until either the fire alarm is verified to be valid by utilization of additional control room or nearby fire panel instrumentation, or upon notification of a fire from the field. The two statements added to the basis in Revision 24 conflict with the Revision 23 threshold basis and initiating condition. The changed threshold basis in Revision 24 could add an indeterminate amount of time to declaring an actual emergency until a person responded to the area of the fire and made a notification to the control room of a fire in the event that redundant control room or nearby fire panel indications were not available. Pending further review and verification by the NRC to determine if the changes to EAL HU6 threshold basis potentially decreased the effectiveness of the Plans, this issue was considered an Unresolved Item. (URI 05000237/2009005-07
05000237/FIN-2009005-08	2009Q4	Dresden	Green	H.7	Self-revealing	Procedural Deficiency Causing a Pressure Pulse Resulting in a Reactor Water Level Low-Low Group 1 Isolation Signal and Unit 3 Reactor Scram	A self-revealed finding involving a non-cited violation (NCV) of Technical Specification 5.4.1 was identified on October 3, 2009, due to the licensees failure to include essential information in DOP 1200-03, RWCU System Operation with the Reactor at Pressure, Revision 51, regarding startup of the reactor water cleanup system with the reactor at pressure. This procedural deficiency caused a pressure pulse that resulted in a reactor water level Low-Low Group 1 Isolation Signal and Unit 3 reactor scram. This event was entered into the licensees corrective action program (CAP) as Issue Report (IR) 974426. Corrective actions by the licensee included revising procedure DOP 1200-03. This finding was considered more than minor because it affected the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as at power operations. The finding was determined to be of very low safety significance because it did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigating equipment or functions will not be available. This finding has a cross-cutting aspect in the area of Human Performance (Resources) because the licensee did not provide complete, accurate and up-to-date procedures to plant personnel. H.2(c
05000237/FIN-2009005-09	2009Q4	Dresden	Green	H.8	Self-revealing	Failure to Ensure a Safety-Related Plug was Ordered and Installed in the 2/3 Emergency Diesel Generator Turbo Lube Oil Y Strainer	A finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion IV, Procurement Document Control, was self-revealed for the licensees failure to ensure a safety-related plug was ordered and installed where required in the 2/3 EDG turbo lube oil Y strainer. Instead, a non-conforming part was installed, which resulted in a one-half gallon per minute oil leak and removal of the diesel generator from service. The issue was entered into the licensees CAP as IR 926605. Corrective actions included inspection of all other diesel generators to ensure the nonconforming condition did not exist on another machine, revising the procurement documents to ensure that future parts include a pressure retaining pipe plug with approved material, and adding a requirement for a quality inspection to be performed to inspect the strainer for metallic pipe plug in blow down port. Individual procedure compliance issues were addressed through the stations performance improvement initiatives. The finding was determined to be more than minor because the finding was similar to IMC 0612, Appendix E, Example 5 c because an incorrect and inadequate part was installed and the system was returned to service. This performance deficiency impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Phase 3 SDP risk evaluation was performed by the regional Senior Risk Analyst who determined the risk significance of the finding to be less than 1.0E-6/yr delta core damage frequency (CDF) and less than 1.0E-7/yr delta LERF, which represents a finding of very low safety significance. Failure of plant personnel to question the plastic shipping plug before the equipment was installed and returned to service was not in compliance with MA-AA-716-008, Foreign Material Exclusion Program, and, therefore, inspectors determined that this event was cross-cutting in Human Performance, Work Practices, Procedural Compliance for failure of personnel to follow the procedure. H.4(b
05000237/FIN-2009005-10	2009Q4	Dresden			NRC identified	Electro-Hydraulic Control (EHC) Fluid Leaking From Stop Valve 3-5699-MSV4-FA Resulting in Forced Outage D3F49	The inspectors identified an unresolved item regarding the regulatory requirements associated with the circumstances surrounding the Unit 3 turbine trip on On November 5, 2009, at 8:53 p.m., Unit 3 Control Room received the following alarm: 903-7 B-6, EHC (electro-hydraulic control) RESERVOIR LVL HI/LO (reference IR 989641) indicating a rate of change in the EHC reservoir at 1.3 in 100 hrs or greater. A non-licensed operator (NLO) was dispatched to stage a barrel of EHC fluid for addition. Preparations were made for a heater bay entry to look for leaks. A Unit 3 heater bay entry was made and it was determined that the Unit 3 Main Turbine Stop Valve (MSV) # 4 had an EHC leak from the fast-acting solenoid valve (3-5699-MSV4-FA). The leak was determined to be approximately 4-5 gallons of fluid per hour. A report from the field was that reservoir level had dropped about 1.1 in the last 12 hours. Between 12:50 a.m. and 3:43 a.m. on November 6, 2009, the licensee added two barrels of EHC fluid to the EHC reservoir. On November 6, 2009, between 9:00 a.m. and 2:00 p.m., licensee management conducted meetings regarding the repair of the leak on MSV #4. The plan called for starting to down power Unit 3 to 650 Mwe for a planned 3:00 p.m. entry into the heater bay to repair the valve. The decision to go to 650 Mwe was to reduce the dose rate in the area and extend stay time for the repair. At approximately 3:00 p.m., while staging for entry to repair the leak, Operations personnel informed the NLO, staged to isolate the oil supply to the leaking valve, that level in the EHC reservoir was dropping quickly, and requested the NLO to enter the pipeway as soon as possible. At approximately 3:05 p.m., the NLO observed oil spraying profusely from the bottom area of #4 Main Stop Valve and the area of the solenoid that was going to be changed out. The NLO immediately contacted the control room to report what was observed and a decision was made to take the turbine offline. At 3:32 p.m., the Unit 3 Turbine was tripped. The licensee had not completed their root cause investigation by the end of the inspection period. The inspectors planned to review the root cause investigation to determine if there were any violations of NRC requirements and that appropriate corrective actions were applied. (URI 05000249/2009005-10
05000237/FIN-2009005-11	2009Q4	Dresden	Green		Licensee-identified	Licensee-Identified Violation	Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. A significant condition adverse to quality for both Unit 2 and Unit 3 containment cooling service water (CCSW) systems was identified by the licensee in RCR 776598-08, Dresden 3-1503-B, 3B Low Pressure Coolant Injection (LPCI) / Containment Cooling Heat Exchanger (HX) Failure to Meet Design Basis Heat Removal Capability Due to Inadequate Programmatic Control of Macrofoulants, Revision 1, on January 9, 2009. Procedure LS-AA-125, Corrective Action Program (CAP) Procedure, revision 13, defines significant condition adverse to quality (SCAQ), in part, as A condition which, if left uncorrected, could have a serious effect on safety or reliability. In addition, recurring deficiencies or errors that cannot be dispositioned or brought into conformance by established corrective action systems, are considered SCAQs. The inspectors determined that the conditions described in RCR 776598-08, met the licensees definition of a significant condition adverse to quality. Contrary to the above requirements, from January through September 2009, the licensee failed to take measures to assure that the cause of the condition (blockage of the LPCI heat exchangers) was determined and corrective action taken to preclude the repetition for a significant condition adverse to quality on both Unit 2 and Unit 3 CCSW systems. Specifically, the licensee failed to prevent the recurrence of Asiatic clam blockage in the 2A LPCI Hx tubes which resulted in the degraded thermal performance of the Hx. Licensee planned corrective actions included installation of a temporary modification to provide temporary keepfill that was expected to provide better chemical treatment of the CCSW piping upstream of the LPCI Hxs, and a permanent injection skid for biocide to provide for long term assurance of effective chemical treatment. This violation was determined to be of very low safety significance because even though the 2A LPCI Hx was degraded it was able to perform the required design safety function
05000237/FIN-2009006-01	2009Q2	Dresden	Green	H.4 H.5	NRC identified	Diesel-Driven Fire Pump Discharge Valve Found Out of Position	A finding of very low safety significance and associated non-cited violation of license conditions 2.E and 3.G for Units 2 and 3, respectively, was identified by the inspectors for the failure to restore the Unit 1 diesel-driven fire pump to an operable condition within seven days as required by Technical Requirements Manual (TRM) 3.7.i.A.1. Specifically, the Unit 1 fire pump discharge valve was found closed rendering the pump inoperable for greater than seven days. Upon discovery of the valve in the closed position the licensee repositioned the valve in the correct locked open position and initiated Action Requests (AR) 922581 and 922585. This finding is more than minor because the failure to provide the two required fire pumps could have resulted in a failure of the stations water based fire protection system should the Unit 2/3 fire pump have been out-of-service at the same time. The finding screened as very low safety significance because the performance of the system was not affected by the closed valve as the Unit 2/3 diesel-driven fire pump remained operable to provide water to the stations fire protection system, if required. This finding has a cross-cutting aspect in the area of human performance, work control because the licensee did not properly plan and coordinate activities consistent with nuclear safety. Specifically, the licensee failed to restore the Unit 1 diesel-driven fire pump to an operable condition within seven days as required by TRM 3.7.i.A.1 as a result of ineffective communications between licensee personnel to verify that valve 1-4199-109 was in its correct locked open position prior to declaring the pump operable (H.3(b))
05000237/FIN-2009007-01	2009Q4	Dresden	Green	H.2	NRC identified	Ventilation System One-Time Inspection Results	A finding of very low safety-significance (Green) was identified by the inspectors for the licensees failure to adequately evaluate and address an aging effect identified by the ventilation system one-time inspection program in accordance with the license renewal Program Basis Document B.1.23C. The licensee entered this issue into the corrective action program, and initiated periodic inspections to manage the aging effect. The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, failure to address the aging effect would not provide assurance that the intended function of in-scope ventilation systems would be maintained consistent with the current licensing basis through the period of extended operation. This finding is of very low safety-significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not ensure proper supervisory and management oversight of work activities, such that nuclear safety is supported. Specifically, supervisory expectations for follow-up were not adequately conveyed prior to the completion of the program. (H.4(c))
05000237/FIN-2009007-02	2009Q4	Dresden	Green		NRC identified	Unit 2 SBLC Tank Thickness Calculation Errors	A finding of very low safety-significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to accurately translate the design bases for the Standby Liquid Control (SBLC) tank into specifications, drawings, procedures, and instructions. Specifically, the SBLC tank wall thickness used in a design basis calculation was incorrect. The licensee initiated IR 983037 to address deficiencies in the calculation. The finding was determined to be more than minor because the finding 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 design basis calculations did not demonstrate that the tank will remain available following design basis seismic events. This finding is of very low safety-significance (Green) because it did not result in a loss of operability. The inspectors did not identify a cross-cutting aspect associated with this finding as it was not indicative of current performance.
05000237/FIN-2009007-03	2009Q4	Dresden	Green	H.7	NRC identified	Failure to Inspect the Non-EQ Electrical Connections Subject to Localized Adverse Environment	A finding of very low safety-significance was identified by the inspectors for the licensees failure to implement a program in accordance with the license renewal program basis Document B.1.33. Specifically, the licensee failed to develop and implement a program to perform visual inspections of the accessible non-environmentally-qualified electrical connections located in adverse localized environments. The licensee subsequently entered the issue into their corrective action program as AR00977284 to re-perform the inspection and revise documentations as required. The finding was not associated with violation of regulatory requirements. The finding was determined to be more than minor because, if left uncorrected, the finding would become a more safety-significant concern. The failure to perform a visual inspection of the subject connections did not assure that the intended functions of these connections would be maintained consistent with the current licensing basis through the extended period of operation. The finding was of very low safety-significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations. This finding has a cross-cutting aspect in the area of Human Performance for the resources component because implementing procedures did not include sufficient guidance defining the parameters of the program. H.2(c)
05000237/FIN-2009007-04	2009Q4	Dresden			NRC identified	Potential Non-Conservative Fatigue Analysis	As part of review of licensee Commitment Item 48, the inspectors identified an unresolved item (URI) related to a potential non-conservative analytical methodology concern described in NRC Regulatory Issue Summary (RIS) 2008-30, Fatigue Analysis of Nuclear Power Plant Components. Specifically, in the licensees fatigue analysis of the feedwater nozzle, only one value of stress was used for the evaluation of the actual plant transients. As part of the review of TLAA related to metal fatigue, the inspectors reviewed licensee calculations that assessed potential effects of reactor coolant on component fatigue life prior to the period of extended operation. During this review, the inspectors reviewed licensee corrective action documents related to their operating experience review of NRC RIS 2008-30 and noted the following: In AR 905504, Potential Issue with FatiguePro Monitoring of Feedwater Nozzle, originated April 9, 2009, the licensee identified that calculation EXLN-15Q-302, Dresden and Quad Cities Updated Analysis of Feedwater Nozzle, used a single stress component in the analysis of fatigue. This corrective action document concluded there was sufficient margin in the cumulative usage factor that included extended power uprate at 60 years of plant operation (end of period of extended operation), and a confirmatory analysis was not required as RIS 2008-30 was issued past the Dresden LRA approval. In IR 958725, LR B.1.34 AMP FatiguePro Software Issues, originated August 28, 2009, the licensee noted that RIS 2008-30 requested recent licensee renewal applicants (close to December 2008) to perform confirmatory analyses that demonstrate the simplified analyses provide acceptable results. This corrective action document indicated that additional calculations would be completed to demonstrate conservatism or equivalence compared to a full six stress component evaluation. The inspectors requested the licensees position regarding the need for a confirmatory analysis if a single stress component was used in their fatigue analyses. The detailed stress analysis requires consideration of six stress components, as discussed in ASME Code, Section III, Subsection NB, Subarticle NB-3200. RIS 2008-30 indicated that NRC staff has requested that recent license renewal applicants that have used this simplified methodology perform confirmatory analyses to demonstrate that the simplified analyses provide acceptable results. The licensee indicated that their position was no confirmatory analyses were required. The licensee documented further justification in their corrective action program: In AR 982528, Resolution of the FatiguePro Related Issue (RIS 2008-30), originated October 21, 2009, the licensee used the technical justification in Structural Integrity Associates (SIA) technical paper PVP2009-78136, Dispelling the Myths Behind Regulatory Issue Summary 2008-30 (Nuclear Plant Component Fatigue Sensitivity Analyses) that was written in response to RIS 2008-30 and published in the proceedings of the ASME 2009 Pressure Vessels and Piping Division Conference. This corrective action document concluded that the PVP2009-78136 paper provided a reasonable technical basis to support no further actions for plants that used the single stress component Greens Function approach employed by FatiguePro Software, and concluded that no confirmatory analyses were necessary. The inspectors reviewed an NRC supplemental Safety Evaluation Report related to the license renewal of the Vermont Yankee Nuclear Power Station (ADAMS ML091200162) dated May 21, 2009, that addressed the technical concern of using a single stress component in fatigue analysis determinations. The inspectors further discussed this design and licensing basis issue with NRC staff in the Office of Nuclear Reactor Regulation. After further discussion with the licensee, the licensee initiated IR 983556, NRC-Identified RIS 2008-30 Applies to Dresden Fatigue Calculation, dated October 23, 2009. The licensee indicated that confirmatory analyses would be performed that retained six stress components for all applicable calculations identified. Pending NRC review of licensee confirmatory analyses, this item will be tracked as an unresolved item (URI 5000237/2009007-04).
05000237/FIN-2009301-01	2009Q1	Dresden	Severity level IV	H.8	NRC identified	Failure to Provide Complete and Accurate Information to the NRC Associated with Verifying No Operating Test Item Duplication with the Audit Test	The inspectors identified a Severity Level IV Non-Cited Violation (NCV) of 10 CFR 55.40, Implementation, 10 CFR 50.9, Completeness and accuracy of information, and 10 CFR 55.49, Integrity of examinations and tests. For the Dresden Station March 2009 NRC Initial Operator License Examination, the inspectors identified that the examination author and the facility reviewer had initialed Step 2.b and Step 3.a.(3) of Form ES-201-2, Examination Outline Quality Checklist, on August 15, 2008, and August 19, 2008, respectively, and Step 1.c of Form ES-301-3 Operating Test Quality Checklist, on January 15, 2009, and January 20, 2009, respectively, which indicated that the operating test did not duplicate items from the applicants audit test, when, upon NRC review, it was determined that six of the 23 dynamic simulator scenario events, and one of the 15 Job Performance Measures (JPMs) for the Reactor Operator (RO) candidates were duplicated from the applicants audit test. The finding was determined to be more than minor, because the integrity of the NRC initial operator licensing examination could have been compromised if, but for detection by the NRC examiners, the NRC examination had been administered with the duplication of the operating test items from the applicants audit test. The finding was determined to be of very low safety significance because the duplication of operating test items was discovered by the NRC examiners prior to administration of the NRC examination, the duplicate test items were either removed from the audit test or the NRC exam changed to remove the duplication, and the facility implemented examination security requirements for the audit test similar to that which was required for the NRC examination. The inspectors concluded that this finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not define and effectively communicate expectations regarding procedural compliance and for personnel to follow procedures (i.e., in the development of the NRC initial operator license examination) (H.4(b)). (Section 4OA5.2)
05000237/FIN-2010002-01	2010Q1	Dresden	Severity level IV	H.7	NRC identified	Failure to Record the Identity of Personnel Performing Post-Maintenance Tests	A finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records, was identified by the inspectors for the licensees failure to record the identity of various personnel who performed seven post-maintenance tests (PMTs) related to Unit 3 EDG maintenance. Despite the PMTs being related to work on safety-related components, an activity affecting quality, neither the licensees procedure MA-AA-716-012, Post-Maintenance Testing, nor DAP 15-10, Post-Maintenance Testing Program, required the identity of the inspector or tester to be recorded. Completed corrective actions included adding PMT documentation requirements to DAP 15-10 and briefing individuals who perform PMTs. This finding was determined to be more than minor because the finding was similar to IMC 0612, Appendix E examples 1b since a portion of required records were irretrievably lost, and 2h since multiple examples were identified as failures to properly implement the same regulatory requirement. Following IMC 0612, Appendix B, it was apparent that this issue did not fall directly under a cornerstone and that incomplete information was recorded in the seven PMTs. Therefore, the Enforcement Policy was used to screen the severity in conjunction with the IMC 0612, Appendix E, Examples 1b and 2h. Since MA-AA-716-012, Post-Maintenance Testing, did not properly implement regulatory requirements, this finding has a cross-cutting aspect in the area of Human Performance, Resources because the licensee did not provide complete, accurate, and up-to-date procedures to plant personnel. H.2(c
05000237/FIN-2010002-02	2010Q1	Dresden	Green	H.14	NRC identified	Failure to Follow Technical Specification 5.5.4 Implementing Procedure	The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of Technical Specification 5.5.4 for the licensee failing to follow Step I.2.a and b of Procedure DOS 1500-08, Discharge of Containment Cooling Service Water (CCSW) From Low Pressure Coolant Injection (LPCI) Heat Exchanger (Hx) During CCSW Pump Operations, Revision 16. Specifically, the licensee failed to perform a tube leak test as required by DOS 1500-08 when activity exceeded 1.5E-6 microcuries/milliliter. The licensees corrective actions included a change to DOS 1500-08 to ensure personnel do not waive performance of the test procedure until tube leak checks are considered during non-routine samples of CCSW and revising the chemistry sampling procedure CY-DR-110-220, LPCI Service Water (CCSW) and Torus Water Sampling, to notify operations to evaluate performance of a tube leak check if activity exceeds 1.5E-6 microcuries/milliliter. The inspectors determined that the failure to perform a tube leak test or perform Calculated CCSW Sample Activity Limit and Canal Activity Calculations was contrary to DOS 1500-08, and was a performance deficiency. The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, had there been an actual LPCI Hx tube leak radioactivity could have been released. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Containment Barrier Cornerstone. All four questions on this table were answered no. There was no actual degradation of the containment barrier. Therefore, the issue screened as having very low safety significance. This finding had a cross-cutting aspect in the area of Human Performance, Decision Making because the licensee did not demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that it was unsafe in order to disapprove the action. Specifically, the licensee assumed the activity in the sample was coming from the floor drain system with no valid proof that was the case. H.1(b
05000237/FIN-2010002-03	2010Q1	Dresden	Green	H.1 H.2(d)	NRC identified	Failure to Meet Regulatory Commitment to Maintain Contingency Plans for Post-Accident Sampling	The inspectors identified a finding of very low safety significance for the failure to meet a regulatory commitment to maintain a contingency plan for obtaining highly radioactive samples of reactor coolant, the suppression pool, and drywell atmosphere for post-accident plant recovery planning. Specifically, the licensees contingency plan was not adequately maintained to ensure the High Radiation Sampling System (HRSS) functioned adequately or otherwise was demonstrated to be in a state of readiness to allow samples to be obtained within a two-week window. No violations of regulatory requirements were identified related to this finding. Corrective actions were being developed to ensure the licensees contingency plan commitments would be met. Those actions included a means to improve system ownership and establishment of an effective process for HRSS equipment maintenance and repair at a priority consistent with its intended use. The finding was more than minor because it impacted the facilities and equipment attribute of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective of ensuring capability to implement adequate measures to protect health and safety of the public in the event of a radiological emergency. Specifically, equipment intended to obtain highly radioactive samples that are used to assess reactor core condition as part of post-accident recovery activities was not demonstrated to be in a readiness condition consistent with the licensees contingency plan. The finding was determined to be of very low safety significance because it involved equipment, which supplements the licensees emergency plan for reentry and recovery activities as provided in the planning standard of 10 CFR 50.47(b)(8), and represented a planning standard problem associated with demonstrating functional readiness of that equipment. The finding was determined to be associated with a cross-cutting aspect in the area of human performance in the resources component, in that, the licensee failed to ensure that equipment to support its emergency plan was functional or otherwise was demonstrated to meet a defined status of operational readiness. H.2(d
05000237/FIN-2010002-04	2010Q1	Dresden	Green		NRC identified	Significance of Potentially Submerged Safety and Nonsafetyrelated Low Voltage Power and Control Power Cables	The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, licensee personnel failed to maintain safety-related cables in underground manholes from becoming repeatedly submerged, which resulted in subjecting the cables to an environment for which they were not qualified. As corrective action, the licensee generated work order (WO) 01271108 on September 24, 2009, to remove the seals on the conduit which contained the cables and which kept water from draining out of the conduit. This issue was entered into the licensees corrective action program as Issue Report (IR) 975308. The finding was determined to be more than minor because the finding 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 finding was of very low safety significance because it was a qualification deficiency that did not result in a loss of operability. The inspectors concluded that there was not a cross-cutting issue associated with this violation
05000237/FIN-2010002-05	2010Q1	Dresden	Green	P.5 P.2(b)	Self-revealing	Electro-Hydraulic Control (EHC) Fluid Leaking From Stop Valve 3-5699-MSV4-FA Resulting in Forced Outage D3F49	The failure of the Unit 3 Main Turbine Stop Valve (MSV) # 4 fast acting solenoid valve on November 6, 2009, resulted in a self-revealed finding of very low safety significance. The licensee failed to use the correct o-rings and bolts when replacing the Unit 3 MSV #4 fast acting solenoid valve during the Unit 3 refueling outage in 2008 which led to the failure. The equipment was not safety-related. Therefore, this finding did not result in a violation of regulatory requirements. The licensees corrective actions included revising maintenance procedure DEP 5600-01, Main Turbine Valve Solenoid and Servo Maintenance, to incorporate the actions described in GE Technical Information Letter 1594. The bolts on the U3 and U2 solenoid valves were replaced. The licensee did not determine that the o-rings were defective until after both this Unit 3 forced outage and the Unit 2 November 2009 refueling outage were complete. Therefore, one corrective action was to write a work order to change the o-rings on the solenoids for both units. In addition, corrective actions were put in place to address weaknesses in the evaluation of Operating Experience. The licensee addressed this issue in the corrective action program under Issue Reports 899829 and 989733. The inspectors determined that the use of o-rings, GE part number U472X000B906, in U3 turbine control valve solenoids, was contrary to Vendor Technical Information Program Binder D1180, General Electric Steam Turbine Generator (GEK5551), Tab 8, GE drawing 115D2402 (Revision 12), and GE Technical Information Letter (TIL) 1594, dated November 30, 2007, which required the use of o-rings, GE part number U472X000BS906, and was a performance deficiency. The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of procedure quality and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Initiating Events Cornerstone. The electro-hydraulic control leakage caused by one or more failed o-rings could have resulted in a turbine trip and reactor scram. However, the failure would not affect mitigating equipment or functions so the finding screened as having very low safety significance. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience because the licensee did not implement and institutionalize Operating Experience through changes to station processes, procedures, equipment, and training programs. P.2(b
05000237/FIN-2010002-06	2010Q1	Dresden	Green		Licensee-identified	Licensee-Identified Violation	As noted in Section 4OA3.2 of this report, in 2009, the licensee discovered that changes had been inappropriately made to the 2B reactor recirculation pump discharge valve above seat drain line. Title 10 CFR 50.59(d)(1) requires the licensee to maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to 50.59(c). Contrary to this requirement, on November 6, 2003, the licensee discovered that the configuration of the above seat drain line for the discharge valve of the 2B Reactor Recirculation pump, which comprises part of the reactor pressure boundary, had been previously modified so that it did not match the piping and instrumentation drawing (P&ID). The licensee could not locate any records that described the scope or date of the change, nor could they locate an evaluation that provided the basis for the determination that the change did not require a license amendment. This change resulted in the line not meeting the requirements of ASME NB3671.3 for Class 1 piping. The configuration of the above seat drain depicted on the P&ID showed a drain line piped to the equipment drain sump with two normally closed manual isolation valves. The configuration in which the above seat drain line was found during the 2003 refueling outage was a drain line with one closed valve and a threaded cap, which did not meet the requirements of ASME NB3671.3 for Class 1 piping. Issue Report 185073 was written, and a second isolation valve and a threaded cap were added to the drain line during the 2003 refueling outage to conform to ASME requirements. The P&ID was also updated to reflect the change. The finding associated with this violation is of very low safety significance (green) because it would not have likely resulted in exceeding the Technical Specification limit for Reactor Coolant System leakage and would not have affected other mitigating systems. Therefore, this would constitute a Severity Level IV violation per the NRC Enforcement Policy, Supplement I, Section D.5
05000237/FIN-2010002-07	2010Q1	Dresden	Green		Licensee-identified	Licensee-Identified Violation	The Code of Federal Regulations, Title 10, Part 50, Appendix B, Criterion III, Design Control, states, in part, Measures shall be established to assure that ... the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Electrical Standard IEEE-279-1968, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, required, in part, that the reactor protection system (RPS) was designed, such that, the system was not susceptible to a single point vulnerability. On October 30,2009, the licensee identified that the RPS pressure switches PS 2-0504-A, B, C, and D all shared a common sensing line with a single isolation valve, which created a single point vulnerability for the turbine stop valve closure (RPS function 8) and turbine control valve fast closure - trip oil pressure low (RPS function 9) reactor scram functions. This condition had existed since original construction. The inspectors reviewed the licensees corrective actions. The inspectors had no issues with the licensees corrective actions and determined that they were completed for Unit 2 and had an acceptable time table for completion of Unit 3. This incident was identified in the licensees corrective action program as Issue Reports 986676 and 996586, and documented in LER 237/2009-007-00, Reactor Protection System Nonconformance to a Design Standard. This violation was determined to be of very low safety significance because the safety function of the reactor protection system was supported by existing procedure guidance in the event the sensing line failed or was inadvertently isolated; therefore, this condition resulted in RPS functions 8 and 9 being operable, but nonconforming
05000237/FIN-2010002-08	2010Q1	Dresden	Green		Licensee-identified	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization. Contrary to this requirement, on November 3, 2009, it was identified that changes had been made, either intentionally or unintentionally, to the design of the plant by the removal of several Unit 2 core spray instrument piping supports, for an indeterminate period of time, without the application of appropriate design control measures. This finding was entered into the licensees corrective action program as IR 1002474. The licensees corrective actions included: restoration of the supports to the original design configuration; creation of a preventive maintenance activity to walk down piping supports prior to the drywell closeout; and an action to walkdown the Unit 3 drywell to verify all supports are connected and tight. The inspectors used Table 4b of IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings to evaluate the significance of the finding. The finding was determined to involve the degradation of equipment specifically designed to mitigate a seismic event. The impact of the finding was that a seismic event could have resulted in the failure of the core spray dp line which would cause a seismically-induced small loss of coolant accident (LOCA) event with the possible unavailability of the B core spray train. The finding screened as potentially risk significant using the screening criteria of Table 4b. The RIII Senior Risk Analyst (SRA) performed a phase 3 SDP evaluation of the finding using the Dresden Standardized Plant Analysis Risk (SPAR) Model and the Risk Assessment of Operational Events Handbook, Volume 2 External Events. The SRA used the seismic initiating event frequency for Dresden (4.58E-04/yr) from the handbook and assumed that any seismic event would cause the failure of the core spray dp line and result in a small LOCA. A conditional core damage probability was estimated using the SPAR model by assuming a small LOCA event occurred simultaneous with a loss of offsite power and the unavailability of the B core spray train. The seismic initiating event frequency was combined with the estimated conditional core damage probability to estimate the delta CDF. The delta CDF calculated was less than 1E-7/yr, which is a finding of very low safety significance (Green). The dominant core damage sequences involved a seismically-induced small LOCA followed by the failure of both high and low pressure injection
05000237/FIN-2010002-09	2010Q1	Dresden	Green		Licensee-identified	Licensee-Identified Violation	The licensee identified a finding of very low safety significance and associated Non-Cited Violation (NCV) of 10 CFR 50.54(t), Conditions of Licenses, for the failure to complete an independent review of all program elements of the emergency preparedness program. The independent assessment did not evaluate and document the adequacy of the interfaces with State and local governments at an interval not to exceed 12 months for all groups. Specifically, Quality Assurances assessment failed to evaluate the adequacy of interface with Will County in 2008. The licensee entered the issue in their corrective action program as AR-00889346 and all required audits were conducted in 2009 (Grundy, Kendal, Will, and the State of Illinois and Indiana). The deficiency was screened using the Emergency Preparedness SDP and determined to be more than minor because the finding adversely affected the EP Cornerstone objective. The failure to conduct the audit to evaluate the effectiveness of the EP program had the attribute associated with Offsite EP; specifically, the evaluation of the working relationship between the offsite and onsite emergency response organizations and programs. The inspector evaluated the finding using with IMC 0609, Appendix B, Sheet I, Failure to Comply flowchart. The audit program was noncompliant with a regulatory requirement not involving an EP planning standard or a risk significant planning standard; therefore, the finding was determined to be of very low safety significance (Green)
05000237/FIN-2010003-01	2010Q2	Dresden			NRC identified	High Pressure Coolant Injection (HPCI) Booster Pump Bearing Oil Levels	The inspectors identified an unresolved item (URI) regarding the impact of HPCI booster pump bearing oil levels being outside the acceptable band on the operability of the HPCI pump. During a walkdown of the Unit 2 HPCI system, the inspectors identified that the preferred level band indicators on the sightglasses for the HPCI booster pump inboard and outboard bearing oil reservoirs were not reliable since the indicators consisted of wires that could be slid up and down the sightglass. When informed by the inspectors, the licensee checked the indicators and identified that they do not conform to the type and position of the markings described in the HPCI booster pump maintenance procedure (MA-AB-734-448). The maintenance procedure defined the oil level band as 12 between the high and low levels. Since the sightglasses on all bearings in the field had a band of approximately 1, they were all non-conservative in at least one direction. Upon further questioning by the inspectors, the licensee measured the oil levels in each of the HPCI booster pump bearing oil level sightglasses for both Units 2 and 3. From these measurements, the inspectors identified that one oil level, the Unit 3 outboard bearing oil level, was below the low level described in the maintenance procedure. When the inspectors asked the licensee about the impact on operability of the pump with this oil level, the licensee indicated that they do not consider the oil band contained in the maintenance procedure as defining the operability of the pump, and they determined that the oil level was not low enough to consider the pump inoperable. When asked at what levels the licensee would consider the pump inoperable, licensee staff were unable to provide an answer. Further information is required from the licensee to determine whether the inaccurate oil level bands caused the oil for the bearings to be at levels such that the HPCI pump was inoperable. This is considered an URI pending further NRC review.
05000237/FIN-2010003-02	2010Q2	Dresden	Green		NRC identified	LPCI Room Heat-up Calculation Deficiencies	The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) related to whether the equipment in the low pressure coolant injection (LPCI) corner rooms would remain within environmental qualification limits during a design basis loss-of-coolant-accident. Specifically, the licensees heat-up calculation contained discrepancies and failed to evaluate the worst case effects post-extended power uprate when determining the heat-up in the LPCI corner rooms. This finding was entered into the licensees corrective action program as AR00763663, AR00742158, AR00883207, AR01055863, and AR01060243, and an operability call performed by the licensee concluded that there are sufficient conservatisms in the calculation that equipment in the corner rooms remained operable. The finding was more than minor because it was associated with the attribute of design control, which affected the Mitigating Systems Cornerstone objective of ensuring the availability and reliability of safety-related systems. This finding is of very low safety significance (Green) because the design deficiency was confirmed not to result in loss of operability or functionality. 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.
05000237/FIN-2010003-03	2010Q2	Dresden			NRC identified	Undocumented Technical Basis for change to EOP ATWS Mitigation Strategy	During the conduct of one of the dynamic simulator scenarios, the inspectors identified an unresolved item related to procedure implementation. 15 Enclosure During one of the dynamic simulator scenarios, conditions were simulated during an Anticipated Transient Without Scram (ATWS) that required the operating crew to lower reactor pressure vessel (RPV) water level in accordance with emergency operating procedure DEOP 400-5, Failure to SCRAM. The purpose of lowering RPV water level is to reduce core inlet sub-cooling and thus reduce the potential for power oscillations. DEOP 400-5, directs the operators to Terminate and Prevent all injection flow into the RPV except for flow from the CRD and Standby Liquid Control (Boron) systems. Contrary to the BWR Owners Group (BWROG) Emergency Procedure Guidelines (EPG) and Severe Accident Guidelines (Revision 2) which states that failure to completely stop RPV injection flow (with the exception of CRD, RCIC, and Standby Liquid Control) would delay the reduction in core inlet sub-cooling, thus increasing the potential for flux oscillations the crew was observed to implement this step in accordance with the licensees expectations, by decreasing the FWLC SETPOINT to -40 inches in incremental steps, such that the set point was always less than actual level. Using this method, feedwater flow was not actually stopped but level was dropped to -35 inches in approximately 1.5 minutes. When asked why the licensees procedural steps deviated from the BWROG EPG, the licensee stated that the deviation was necessary to prevent the loss of the Main Condenser heat sink (bypassing the Group 1 Isolation interlocks is performed in parallel and cannot be completed quick enough to prevent isolation of the Main Steam lines if flow is terminated completely). The BWROG EPG states that reducing reactor power and preventing power oscillations is of greater importance than preventing loss of the main condenser. Technical Specification 5.4.1 requires, in part, that written procedures/instructions be established, implemented, and maintained covering the emergency operating procedures required to implement the requirements of NUREG-0737, Clarification of TMI Action Plan Requirements, and NUREG-0737, Supplement 1, as stated in GenericLetter 82-33. NUREG-0737 and the associated Supplement 1 requires licensees to analyze transients and accidents, prepare emergency procedure technical guidelines, and develop symptom-based emergency operating procedures based on those technical guidelines. The BWROG EPG provides the technical basis for the development of the emergency operating procedures used by BWR licensees. Licensees are permitted to deviate from the BWROG guidelines provided they document the technical basis for the deviation. When asked to provide justification for the deviation from the BWROG EPG, the licensee was unable to do so. The licensee has initiated an engineering evaluation to provide the necessary basis for the deviation. This issue is an URI pending further NRC review and completion of the licensees actions to provide the necessary documentation to support the deviation: (URI 05000237/2010003-03; 05000249/2010003-03, Undocumented Technical Basis for Change to EOP ATWS Mitigation Strategy)
05000237/FIN-2010003-04	2010Q2	Dresden	Green		NRC identified	Failure to Monitor Unit 3 Drywell Temperature	A finding of very low safety significance was identified by the inspectors on May 3, 2010, for the failure to monitor Unit 3 (U3) drywell temperature per commitments made to the NRC within LER 88-22, Supplement 2 to ensure that equipment in the drywell was operating within environmental qualification limits. No violation of regulatory requirements occurred. The licensees corrective actions were to reinstate the temperature monitoring of both drywells, perform walkdowns of both drywells for correct placement and to verify functionality of drywell thermocouples, improve administrative requirements on plant engineering turnover from one engineer to another, and train engineers to focus on identifying and validating assumptions when performing or reviewing technical products. This finding was placed in the licensees corrective action program as IR 1064681. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to monitor drywell temperatures and evaluate those temperatures against environmental qualification limits would have resulted in the motor operator for valve 3-1301-1 to exceed its qualified life without the licensees knowledge after D3R21. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Mitigating Systems Cornerstone. The inspectors were able to answer the first question, is the finding a design or qualification deficiency confirmed not to result in loss of operability or functionality, yes and, therefore, the finding screened as Green. The inspectors did not identify a cross-cutting aspect associated with this finding because the age of the finding did not reflect current performance.
05000237/FIN-2010003-05	2010Q2	Dresden	Green	P.3	NRC identified	Failure to Provide An Adequate Procedure for Several Instrument Maintenance Surveillance Tests	The inspectors identified a NCV of Technical Specification (TS) 5.4.1 for the failure to provide an adequate procedure for the verification of correct installation and restoration of equipment during an instrument maintenance surveillance test in May 2010. The licensees corrective actions included a task to add requirements for an independent verification of the removal of volt-ohm meters (VOMs) in the ohm meter mode in 17 affected procedures. The licensee entered this finding into the corrective action program as issue report (IR) 1068559. Using IMC 0612, Appendix E, Examples of Minor Violations, issued on September 20, 2007, the inspectors determined that there were no similar examples to this finding in Appendix E. The inspectors referenced IMC 0612, Appendix B, Issue Screening, dated January 1, 2010. The inspectors determined that the finding was more than minor based on Block 9, Figure 2, paragraph 2.b, If left uncorrected, would the finding become a more significant safety concern. The inspectors determined that the failure to perform an independent verification that a testing configuration had been returned to normal could result in the inability of a system or component to perform its function which would be a more significant safety concern. No systems had been incorrectly returned to service as a result of the inadequate procedure and, therefore, this violation had very low safety significance. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution - Corrective Actions because the licensee did not address a previously identified safety issue in a timely manner.
05000237/FIN-2010003-06	2010Q2	Dresden	Green		Self-revealing	Failure To Perform An Adequate Inspection of Circulating Water Valve 3-4402-C	On January 22, 2010, a finding of very low safety significance was self-revealed for failure to perform an adequate inspection of the grease condition of the 3-4402-C valve actuator HBC gear box, which was contrary to the requirements of MA-AA-723-301,Periodic Inspection of Limitorque Model SMB/SB/SBD-000, Revision 3. No violation of regulatory requirements occurred because valve 3-4402-C 2 Enclosure was a nonsafety-related component. The licensee planned to drill inspection ports into and/or replace the HBC gear boxes for valves 2/3-34403-A(B)(C)(D) and 2/3-34402-A(B)(C)(D) and 2-34402-C and change the preventive maintenance requirement to perform a 12 year mechanical inspection of the HBC gear box. This finding was placed in the licensees corrective action program as IR 1034444, Failure of the 3-4402-C Condenser Inlet Valve. The finding was determined to be more than minor because the finding could be reasonably viewed as a precursor to a significant event. Specifically, valve 3-4402-C acted as an inlet in the circulating water system for the south water box. When the valve failed, it was almost completely closed. Had the valve failed open, circulating water would have been diverted from the condenser potentially causing a loss of vacuum that would have resulted in a reduction in power and/or a turbine trip and reactor trip. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, Initiating Events Cornerstone Column, Transient Initiators question 1, does the finding contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available, was answered no, and, therefore, screened as Green. This finding had no cross-cutting aspect due to the issues involved in this valve failure were not indicative of current performance.
05000237/FIN-2010003-07	2010Q2	Dresden	Green		NRC identified	2/3 Emergency Diesel Generator (EDG) Overvoltage during Division I Undervoltage	A finding of very low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XI, Test Control, was self-revealed on November 16, 2009, for the failure to perform adequate post-modification testing which allowed an improper wiring condition to exist for years before an inappropriate action by a nuclear station operator exposed the problem. The licensee generated issue report 994101 to address this issue. Corrective actions implemented by the licensee to address this issue included correcting the improper wiring configuration and creating a training request to have enhanced post-maintenance testing requirement training given to all plant engineers, design engineers and maintenance work planners. Using IMC 0612, Appendix B, Issue Screening, the inspectors determined that the finding was more than minor because it impacted the Mitigating Systems Cornerstone attribute of procedure quality to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors completed a Phase 1 significance determination of this issue using IMC 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The inspectors answered \"no\" to all questions in the Mitigating Systems Cornerstone column of Table 4a, Characterization Worksheet for IE, MS, and BI Cornerstones. Therefore, the finding screened as Green (very low safety significance). Because the behavior leading to this event occurred in 1986, the inspectors determined that this event was not indicative of current performance and, therefore, no cross-cutting area was affected.
05000237/FIN-2010003-08	2010Q2	Dresden			NRC identified	Potential Safety Significance of Free-Standing Stack Configuration during MPC Transfer from HI-TRAC Transfer Cask into HI-STORM Storage Cask	As part of review of the licensees control of heavy loads associated with the design and installation of a physical restraint to the stack configuration during transfer activities of the MPC from the HI-TRAC transfer cask into the HI-STORM storage cask, the inspectors identified an unresolved item (URI) related to the potential safety significance of a free-standing stack configuration during a postulated design basis seismic event. Specifically, the licensee had utilized a free-standing stack configuration during prior dry fuel storage campaigns. During a weekly teleconference with the licensees Byron Station and the NRC on February 9, 2010, the NRC inspectors noted concerns with the seismic analysis of the proposed free-standing stack configuration during transfer activities of the MPC from HI-TRAC transfer cask into HI-STORM storage cask. Noting similarities in the analysis performed for the Dresden free-standing stack configuration, the licensee entered the concern into its corrective action program, AR 1028862, Questions Raised by NRC Regarding Cask Stackup, dated February 10, 2010. In addition, Exelon initiated AR 1031363, NRC Questions on Byron Specific Calculation, dated February 12, 2010. Corrective actions are being processed through Exelon AR 1031363 to ensure future safe handling of the cask. Based on final corrective actions, the safety significance of 50 Enclosure the historical use of a free-standing stack configuration may need to be addressed, as noted in AR 1028862. Pending NRC review of the licensees associated corrective actions, this concern will be tracked as an unresolved item (URI 05000237/2010003-08; 05000249/2010003-08).
05000237/FIN-2010003-09	2010Q2	Dresden	Green		Licensee-identified	Licensee-Identified Violation	On April 16, 2010, a plant engineer performed an ultrasonic examination (UT) of the 3B core spray pump suction piping and identified that an air bubble existed in the piping. The purpose of the UT was to determine post-maintenance operability. The test procedure required an examination within one foot of the vent valve on the suction piping. The engineer found no air in that location and called the test satisfactory. However, the engineer performed additional UT examination of the piping farther away from the vent valve and found air in that location. The licensee declared the 3B core spray subsystem operable after the completion of the UT examination. The engineer neither evaluated the size of the air bubble for operability nor contacted operations personnel or wrote an issue report so that operability could be evaluated until April 19, 2010. The air bubble was eventually documented in IR 1058558, Air Ribbon Discovered in 3B Core Spray Suction, and was determined to not impact the operability of the 3B Core Spray subsystem. The fact that the air bubble was identified and that no IR was written until three days later was documented in IR 1058966, UT Inspection Process Improvement. Exelon Procedure LS-AA-125, Corrective Action Program (CAP) Procedure, Revision 14, Section 4.1.2, states, in part, that at any time any question of current or past operability arises, then initiate an Issue Report. Air in the suction piping of emergency core cooling system piping raised a question of operability. This violation was not greater than Green because the air bubble did not result in the inoperability of the 3B core spray subsystem. The adequacy of the PMT procedure as addressed in IR 1057049, Air Void U2 HPCI Discharge Piping Above Acceptance Criteria.
05000237/FIN-2010004-01	2010Q3	Dresden			NRC identified	Failure to Address NRC Concerns Regarding a Reactor Building Closed Cooling Water (RBCCW) Line Break in the Unit 3 Reactor Building	During a walkdown of the Unit 2 and Unit 3 reactor buildings the inspectors identified: That there were RBCCW pipes directly above the bermed areas surrounding the safety-related busses 23-1, 24-1, 33-1, and 34-1. If those pipes were to fail the bermed area around the busses would hold water in potentially resulting in the failure of power to all the low pressure ECCS pumps The inspectors identified that on Unit 3 there was only one hole (not several) in the floor inside the bermed area around busses 33-1 and 34-1 and that hole had a one and onehalf inch lip around it which was not notched. This configuration did not appear to be evaluated at the time the SER was written. The licensee performed an evaluation after the inspectors brought this condition to their attention. The review of the licensees evaluation is an unresolved item. (URI 05000237/2010004-01; 05000249/2010004-01)
05000237/FIN-2010004-02	2010Q3	Dresden			NRC identified	Failure to Seal Holes in the Floor Above the Emergency Core Cooling System (ECCS) Corner Rooms	During a walkdown of the Unit 2 and Unit 3 reactor buildings the inspectors identified: That there were holes in the floor on both units, which would allow flood water to bypass the berms around the stairways to the ECCS corner rooms. The holes in the floor could also potentially result in a loss of all ECCS pumps. In regard to the first observation, the inspectors reviewed a letter from the NRC to Commonwealth Edison (the licensee) dated August 20, 1982. The subject was, SEP (Systematic Evaluation Program) Topic III-5.B, Pipe Break Outside Containment - Dresden Nuclear Power Station Unit 2. The enclosure to the letter was the NRCs Safety Evaluation Report (SER) for SEP Topic III-5.B. In the safety evaluation, the NRC reviewed the licensees response to a previous NRC concern about the failure of the RBCCW piping above the 23-1 and 24-1 switchgear on Unit 2. The licensee responded that there were several holes in the floor inside the bermed area around busses 23-1 and 24-1. The largest hole had a one and one-half inch lip around it. The licensee stated that the lip would be notched and that the holes would be sufficient to let the water drain before it could get high enough to impact the safety-related busses. There was no mention of Unit 3 in the safety evaluation In regard to the second observation, the inspectors reviewed DR PSA-012, Internal Flood Evaluation Summary and Notebook, dated May 2009. This document supported the licensees probabilistic risk assessment, but was not part of the licensing basis. This document stated that the berms around the ECCS corner room stairs were credited in the internal flooding analysis. A review of the licensing basis to determine the design requirements of the ECCS corner room stairway berms was an unresolved item. (URI 05000237/2010004-02; 05000249/2010004-02)
05000237/FIN-2010004-03	2010Q3	Dresden	Green	H.13	NRC identified	Installation of Nonconforming Material Into a Safety-Related System	The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components for the installation of a commercially dedicated part for use in a safety-related system which failed testing acceptance criteria on October 6, 2008. The licensees corrective actions included replacing the nonconforming material on November 11, 2009. The licensee made procedure changes to clarify the requirements for documentation of the technical justification of accepting discrepancies. The licensee entered this finding into the corrective action program as issue report (IR) 1068559. 2 Enclosure The finding was determined to be more than minor because the finding was similar to IMC 0612, Appendix E, Example 5c (dated August 11, 2009). The inspectors determined the finding could be evaluated using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Barrier Integrity Cornerstone. The inspectors answered all four questions in Table 4a, No, therefore, the inspection finding screened as having very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance - Decision Making. Specifically, there was a systematic process to ensure that discrepancies identified in the commercial grade dedication process were properly resolved, which was not followed.
05000237/FIN-2010004-04	2010Q3	Dresden	Green		NRC identified	Failure to Identify and Correct Test Procedures to Assess the As-Found Trip Setpoint for Pressure Switches that Satisfy Technical Specification Functions	The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure that conditions adverse to quality associated with preconditioning were promptly identified and corrected. The licensees corrective actions included actions for Engineering to evaluate all the Technical Specification functions that do not have test valves installed on their pressure switches and justify the potential unacceptable preconditioning as acceptable or take other actions as appropriate. The licensee entered this finding into the corrective action program as issue report (IR) 1120159. The finding was determined to be more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors did not identify any cross-cutting aspect associated with this finding. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors answered No to all questions in the Mitigation System Cornerstone column of Table 4a, Characterization Worksheet for IE, MS, and BI Cornerstones, therefore, the finding screened as Green (very low safety significance).
05000237/FIN-2010005-01	2010Q4	Dresden	Green	P.3	NRC identified	Failure to Provide Robust Indication of Adequate Oil Level in High Pressure Coolant Injection Booster Pump Sight Glasses	The inspectors identified a non-cited violation of Technical Specification 5.4.1 for the licensees failure to implement the section of the high pressure coolant injection (HPCI) booster pump maintenance procedure that prescribes how to robustly mark the HPCI booster pump sight glasses to indicate the acceptable oil levels for the bearings of the Unit 2 and 3 HPCI booster pumps. Upon being informed of the condition, the licensee verified, using measurements, that the oil level for the bearings was adequate and entered the condition in the corrective action program to provide a more robust indication of the acceptable oil level. This finding is of greater than minor safety significance because, if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, if a more robust minimum level indication is not used, the wires could slide down the sight glass to the point that they do not prevent operators from allowing the oil level to drop below the minimum acceptable level for the pump to perform its safety function. The finding impacted the Mitigating Systems Cornerstone because it involved degradation of HPCI. It is not greater than Green because it did not result in the loss of operability of the HPCI system. The inspectors determined that this finding has a cross-cutting aspect in the area of Problem Identification and Resolution under the component Corrective Action Program because the licensee did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity.
05000237/FIN-2010005-02	2010Q4	Dresden	Green	H.12	NRC identified	Failure to Perform Magnetic Particle Examination on the 3/2/1501-20/20-10 U3 LPCI Support in Accordance with Procedures	A finding of very low safety significance (Green) 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 failure to accomplish activities affecting quality in accordance with procedures. Specifically, the licensees vendor non-destructive examination (NDE) examiner failed to perform a magnetic particle (MT) examination in accordance with procedures on the 3/2/1501-20/20-10 Unit 3 low pressure coolant injection support. The licensee initiated corrective action document Issue Report (IR) 01135770 to address the issue. The finding 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 failure to perform an adequate MT examination could have allowed undetected flaws to remain in service. This finding is of very low safety significance (Green) because the inspectors answered No to all of the characterizations worksheet questions in Table 4a of MC 0609.04. Specifically, no indications were identified when the examination was re-performed. This finding has a cross-cutting aspect in the area of Human Performance for the Work Practices component because the licensee proceeded in the face of uncertainty and unexpected circumstances by continuing to perform the examination after the equipment became damaged. The licensees examiner also elected to continue after identifying there was material present on the pipe in a location that could interfere with the exam. In addition, due to different circumstances surrounding the exam such as: component location, equipment weight, and environmental conditions, the examiner became tired. Nonetheless, the examiner elected to continue to perform the examination in this condition.
05000237/FIN-2010005-03	2010Q4	Dresden	Green		NRC identified	Drywell Equipment Drain Sump Discharge Valves 2-2001-5 and 2-2001-6 Body to Diaphragm Leakage	The inspectors identified an unresolved item regarding the operability of the Unit 2 drywell floor drain sump and drywell equipment drain sump inboard and outboard discharge air operated valves and their ability to perform their function as primary containment isolation valves. On October 19, 2010, the Unit 2 drywell floor drain sump (DWFDS) outboard discharge air operated valve (AOV), 2-2001-106, was declared inoperable due to local observation of a pinhole water leak between the body of the valve and the diaphragm while the valve was stroking closed after pumping the Unit 2 DWFDS. The licensee generated issue report (IR) 1127948, PCIV 2-2001-106 DWFDS DISCH VLV INOP, to address this issue. The licensee determined that the cause of the leak was relaxation of the valve bolting torque. Air operated valve 2-2001-106 was repaired by torquing the bolts back to the original torque value. In addition, the licensee generated operability evaluation (OpEval) number 10-007, Drywell Primary Containment Isolation System Equipment and Floor Drain Valves. The licensee determined that the leakage noted was not indicative of a diaphragm leak, which would originate from between the bonnet and diaphragm or the stem and bonnet of the valve. On October 27, 2010, the Unit 2 drywell equipment drain sump (DWEDS) pump inboard discharge air operated valve, 2-2001-5, and the Unit 2 drywell equipment drain sump pump outboard discharge air operated valve, 2-2001-6, exhibited pinhole leaks, 20 drops and 1 drop respectively, between the body of the valve and the lower side of the diaphragm while the valves were stroking closed after pumping the Unit 2 DWEDS. The licensee generated IR 1131662, DWEDS 2-2001-5 and 2-2001-6 BODY-DIAPHRAGM LEAKAGE, to address this issue. The licensee determined that based on OpEval 10-007, the leakage was not indicative of a diaphragm leak; therefore, valves 2-2001-5 and 2-2001-6 remained operable. As a compensatory action, the licensee performs weekly inspection of the valves during the pumping operation. By the end of the inspection period, both valves continued leaking and multiple IRs had been generated as a result of the weekly inspections performed. The inspectors questioned the licensee regarding the ability of these valves to meet the local leak rate test (LLRT) acceptance criteria and Technical Specification 3.6.1.3, Primary Containment Isolation Valves (PCIVs). The ability of the Unit 2 drywell equipment drain sump inboard and outboard discharge air operated valves to perform their function as primary containment isolation valves is considered an unresolved item pending further review of the licensees response.
05000237/FIN-2010005-04	2010Q4	Dresden	Green		NRC identified	Adequacy of High Pressure Core Injection System Low Pressure Testing	The inspectors identified an unresolved item regarding the low pressure testing of the HPCI system. The Technical Specifications require the testing of the HPCI system during reactor startup from a refueling outage prior to exceeding 150 psig. On November 26, 2010, during the post Unit 3 refueling outage low pressure testing of the HPCI system, the system experienced abnormal parameters. System flow cycled between about 4000 gpm to about 7000 gallons per minute (gpm). System discharge pressure cycled between 300 psig to about 500 psig. These oscillations resulted in a severe shaking of the system discharge piping. The operator in the control room did not observe the severity of these oscillations. The inspectors interviewed the control room operator performing the surveillance test. The control room operator stated that HPCI flow was observed in the control room to oscillate between 5000 and 5600 gpm. The more severe oscillations were observed by the system manager monitoring a computer point outside the control room. The licensee documented the flow oscillations in IR 1145149, Abnormal U3 HPCI Pump Parameters During Low Pressure Run. In this IR the licensee stated that the oscillations appeared to be the result of pump runout. Centrifugal pump runout is a condition where the pump flow is beyond the point where net positive suction head (NPSH) required exceeds NPSH available and the pump starts to cavitate. Prolonged operation in this condition can cause damage to the pump and the piping. The inspectors observed that IR 1145149 required no actions to attempt to determine why or if the pump ran out. The licensee accepted the test results because in the control room a flow rate of 5000 gpm was observed. The inspectors interviewed the control room operator who performed the test. The test procedure stated that HPCI test return flow to the condensate storage tank shall be throttled such that discharge pressure shall be maintained greater than 98 psig above reactor pressure. The operator stated that he thought this meant to get discharge as close to 98 psig above reactor pressure as possible. This would have made pump discharge pressure close to 250 psig. That was why he fully opened the test return valve. The inspectors determined through interviews that in previous performances of this surveillance test operators did not normally open the test control valve fully and maintained pump discharge pressure greater than 600 psig. The inspectors questioned licensee engineering management that if a HPCI injection occurred at a reactor pressure of 150 psig and then the HPCI injection valve fully opened without throttling, would the HPCI pump also runout? Licensee engineering management did not know the answer to this question. The licensee felt the problem was with the flow characteristics of the test valve but did not know if the injection valve would have significantly different flow characteristics. The licensee felt that the valve vendor would have to be consulted to get an assessment of the flow characteristics of the valves. If the HPCI pump would runout during injection at a discharge pressure of 150 psig, then the testing performed would be inadequate to determine HPCI operability. The licensee evaluation of system performance at low discharge pressure is an unresolved item pending inspector review for adequacy.
05000237/FIN-2010007-01	2010Q2	Dresden	Green	H.14	NRC identified	Failure to Establish Heat Exchanger Inspection Procedures Appropriate for the Circumstances	The inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, having very low safety significance for the failure to establish inspection procedures that were appropriate for the circumstances for the NRC Generic Letter (GL) 89-13 program heat exchangers. Specifically, procedures did not contain adequate guidance for partially blocked heat exchanger tubes found to be completely blocked. As a result, the licensee did not evaluate the as-found conditions of the 2/3 Emergency Diesel Generator (EDG) jacket water heat exchangers; therefore, did not determine the heat exchangers were not bounded by applicable design documents. The licensee entered this issue into its corrective action program. The performance deficiency was determined to be more than minor because it was associated with the mitigating system cornerstone attribute of procedure quality and affected the cornerstone objective. This finding was of very low safety significance because it was a qualification deficiency confirmed not to result in loss of operability or functionality. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not use a conservative assumption in decision making. Specifically, the licensee did not use a conservative assumption when establishing the acceptance criteria for the inspection of heat exchangers.
05000237/FIN-2010007-02	2010Q2	Dresden	Green	H.14	NRC identified	Improper Reclassification of LPCI Pump Mechanical Seals	The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for improperly classifying Low Pressure Coolant Injection (LPCI) pump mechanical seals as non-safety-related. The licensee subsequently concluded that the seals should be classified as safety-related based upon Exelon procedure SM-AA-300 and parts classification Guide M-1994-300, and reclassified them as safety-related. The performance deficiency was determined to be more than minor because if left uncorrected, it would become a more significant safety concern. This finding was of very low safety significance because it was a qualification deficiency confirmed not to result in loss of operability or functionality. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action. H.1(b)
05000237/FIN-2010007-03	2010Q2	Dresden	Green		NRC identified	Failure to Perform Adequate Testing to Confirm Acceptable Fast Bus Transfer Time	The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to have an adequate post-maintenance test for circuit breakers to confirm fast bus transfer capability. Specifically, the licensee failed to ensure that either vendor overhaul procedures or the station procedure for receipt inspection confirmed that breaker timing tests were performed after the circuit breakers were overhauled at a vendor facility. The licensee entered this issue into its corrective action program. The finding was more than minor because if left uncorrected, the finding could have the potential to lead to a more significant safety concern. This finding was of very low safety significance (Green) because the test deficiency was confirmed not to result in loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not reflective of current performance.
05000237/FIN-2010007-04	2010Q2	Dresden	Green		Licensee-identified	Licensee-Identified Violation	A finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the licensee for the failure to assure that 125 VDC and 250 VDC safety-related batteries were installed in accordance with their seismic qualification. During a walkdown of the 125 VDC and 250 VDC batteries, the licensee noted that the ethafoam spacers between individual battery cells and against battery racks were not tight in spots due to 1/4 inch gaps and should have thicker or additional ethafoam. The finding was more than minor due to impacting the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The licensee entered this nonconformance into its corrective action program as AR01032718 and AR01054241 and initiated work orders to evaluate and replace the ethafoam spacers with properly sized material. To establish a reasonable assurance of operability, the licensee reviewed seismic experience database reports from the Seismic Qualification Utility Group (SQUG). Experience from actual seismic events has shown that batteries without spacers or with thinner spacers have survived earthquakes stronger than the Dresden design basis earthquake.
05000237/FIN-2010402-01	2010Q1	Dresden			Licensee-identified	Security
05000237/FIN-2010502-01	2011Q2	Dresden	Green		NRC identified	(Traditional Enforcement) Changes to EAL Basis Decreased the Effectiveness of the Plan without Prior NRC Approval	The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54(q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan. Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 24, which indefinitely extended the start of the 15-minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner. The violation affected the NRCs ability to perform its regulatory function because it involved implementing a change that decreased the effectiveness of the emergency plan without NRC approval. Therefore, this issue was evaluated using Traditional Enforcement. The NRC determined that a Severity Level IV violation was appropriate due to the reduction of the capability to perform a risk significant planning standard function in a timely manner. The licensee entered this issue into its corrective action program and revised the EAL basis to restore compliance. The finding was more than minor using IMC 0612, because it is associated with the emergency preparedness cornerstone attribute of procedure quality for EAL and emergency plan changes, and it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Therefore, the performance deficiency was a finding. Using IMC 0609, Appendix B, the inspector determined that the finding had a very low safety significance because the finding is a failure to comply with 10 CFR 50.54(q) involving the risk significant planning standard 50.47(b)(4), which, in this case, met the example of a Green finding because it involved one Unusual Event classification (EAL HU6).
05000237/FIN-2010502-02	2011Q2	Dresden	Green		NRC identified	Changes to EAL Basis Decreased the Effectiveness of the Plan without Prior NRC Approval	The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54(q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan. Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 24, which indefinitely extended the start of the 15-minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner. The violation affected the NRCs ability to perform its regulatory function because it involved implementing a change that decreased the effectiveness of the emergency plan without NRC approval. Therefore, this issue was evaluated using Traditional Enforcement. The NRC determined that a Severity Level IV violation was appropriate due to the reduction of the capability to perform a risk significant planning standard function in a timely manner. The licensee entered this issue into its corrective action program and revised the EAL basis to restore compliance. The finding was more than minor using IMC 0612, because it is associated with the emergency preparedness cornerstone attribute of procedure quality for EAL and emergency plan changes, and it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Therefore, the performance deficiency was a finding. Using IMC 0609, Appendix B, the inspector determined that the finding had a very low safety significance because the finding is a failure to comply with 10 CFR 50.54(q) involving the risk significant planning standard 50.47(b)(4), which, in this case, met the example of a Green finding because it involved one Unusual Event classification (EAL HU6).

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