IR 05000237/2013007

Download: ML13179A220

Text

June 27, 2013

Mr. Michael Senior Vice President, Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO), Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT: DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 COMPONENT DESIGN BASES INSPECTION (CDBI) INSPECTION REPORT 05000237/2013007; 05000249/2013007

Dear Mr. Pacilio:

On May 30, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection (CDBI) at your Dresden Nuclear Power Station, Units 2 and 3. The enclosed report documents the results of this inspection, which were discussed on May 30, 2013, and on March 1, 2013, with Mr. D. Czufin, and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of this inspection, four NRC-identified findings of very low safety significance were identified. Three of the findings involved violations of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy. No cross-cutting aspects were assigned to these findings. If you contest the subject or severity of the NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident inspectors Office at the Dresden Nuclear Power Station. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/ Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR-25

Enclosure:

Inspection Report 05000237/2013007; 05000249/2013007

w/Attachment:

Supplemental Information cc w/encl: Distribution via ListServŽ Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket Nos: 05000237; 05000249 License Nos: DPR-19; DPR-25 Report No: 05000237/2013007; 05000249/2013007 Licensee: Exelon Generation Company, LLC Facility: Dresden Nuclear Power Station, Units 2 and 3 Location: Morris, IL Dates: January 28 through May 30, 2013 Inspectors: A. Dunlop, Senior Engineering Inspector, Lead M. Munir, Engineering Inspector, Electrical J. Corujo Sandin, Engineering Inspector, Mechanical B. Palagi, Senior Operations Inspector G. Morris, Electrical Contractor B. Sherbin, Mechanical Contractor Approved by: Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety 1 Enclosure

SUMMARY

IR 05000237/2013007; 05000249/2013007; 01/28/2013 - 05/30/2013; Dresden Nuclear Power Station, Units 2 and 3; Component Design Bases Inspection (CDBI). The inspection was a 3-week onsite baseline inspection that focused on the design of components. The inspection was conducted by regional engineering inspectors and two consultants. Four Green findings were identified by the inspectors. Three of the findings were considered Non-Cited Violations (NCVs) of NRC regulations. The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using IMC 0609, "Significance Determination Process" dated June 2, 2011. Cross-cutting aspect are determined using IMC 0310, "Components Within the Cross Cutting Areas" dated October 28, 2011. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy dated January 28, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process" Revision 4, dated December 2006.

A. NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to assure and verify adequate voltage was available at the air start solenoid associated with Unit 2 and Unit 2/3 emergency diesel generators. Specifically, the licensee failed to assure the minimum available voltage at the air start solenoid met the minimum rated voltage value for the solenoid. The licensee entered this finding into their Corrective Action Program and provided test results and calculations to reasonably conclude the currently installed air start solenoid valves would energize at the minimum calculated available voltage. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. (Section 1R21.3.b.(1))

Green.

The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to correctly calculate the minimum air volume and pressure required to actuate the Target Rock Safety/Relief Valve air accumulators. Specifically, when calculating the minimum required air volume in the accumulator, the licensee failed to include the volume of air needed to stroke the air operator from closed to open. The licensee entered this finding into their Corrective Action Program and verified through a preliminary calculation there would be sufficient air in the accumulators for the valves to perform their safety function.

The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. (Section 1R21.3.b.(2)) Green: The inspectors identified a finding of very low safety significance concerning motor-operated valve differential pressure calculation with respect to Dresden's anticipated transient without scram (ATWS) analysis. Specifically, the inspectors identified the design differential pressure used in calculation for the high pressure coolant injection (HPCI) steam supply valve did not address the significantly higher differential pressure that would be applied across the motor-operated valve during an ATWS event. The licensee entered this finding into their Corrective Action Program and verified through a preliminary calculation the HPCI steam supply valve would have sufficient thrust to open against the higher differential pressure to allow HPCI to function during ATWS event. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. (Section 1R21.3.b.(3))

Green.

The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to ensure the isolation condenser would be capable of performing its safety function under design conditions. Specifically, the licensee was unable to justify the assumption the heat transfer rate would remain the same once the isolation condenser tubes began to become exposed. The licensee entered this finding into their Corrective Action Program and instituted a standing order to maintain the shellside water level and temperature in a more restrictive band. In addition, the licensee contracted a vendor to develop a calculation and additional bases for the design assumptions. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors had reasonable doubt the system would have been able to perform its safety function during the initial 20 minutes of operation if called upon under design conditions. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. (Section 1R21.3.b.(4))

B. Licensee-Identified Violations

No violations were identified.

REPORT DETAILS

REACTOR SAFETY

Cornerstone:

Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Component Design Bases Inspection

.1 Introduction

The objective of the component design bases inspection is to verify the design bases have been correctly implemented for the selected risk significant components and the operating procedures and operator actions are consistent with design and licensing bases. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The Probabilistic Risk Assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for which there are no indicators to measure performance. Specific documents reviewed during the inspection are listed in the Attachment to the report.

.2 Inspection Sample Selection Process The inspectors used information contained in the licensee's Probabilistic Risk- Assessment and the Dresden Standardized Plant Analysis Risk Model to identify one scenario to use as the basis for component selection.

The scenario selected was a loss of main feedwater event. Based on this scenario, a number of risk significant components were selected for the inspection. The inspectors also used additional component information such as a margin assessment in the selection process. This design margin assessment considered original design reductions caused by design modification, power uprates, or reductions due to degraded material condition. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as performance test results, significant corrective actions, repeated maintenance activities, Maintenance Rule (a)(1) status, components requiring an operability evaluation, NRC resident inspectors input of problem areas/equipment, and system health reports. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. A summary of the reviews performed and the specific inspection findings identified are included in the following sections of the report.

The inspectors also identified procedures and modifications for review that were associated with the selected components. In addition, the inspectors selected operating experience issues associated with the selected components. This inspection constituted 22 samples as defined in Inspection Procedure 71111.21-05.

4 Enclosure

.3 Component Design

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), Technical Specifications (TS), design basis documents, drawings, calculations and other available design basis information, to determine the performance requirements of the selected components. The inspectors used applicable industry standards, such as the American Society of Mechanical Engineers (ASME) Code, Institute of Electrical and Electronics Engineers (IEEE) Standards and the National Electric Code, to evaluate acceptability of the systems' design. The NRC also evaluated licensee actions, if any, taken in response to NRC issued operating experience, such as Bulletins, Generic Letters (GLs), Regulatory Issue Summaries (RISs), and Information Notices (INs). The review was to verify that the selected components would function as designed when required and support proper operation of the associated systems. The attributes that were needed for a component to perform its required function included process medium, energy sources, control systems, operator actions, and heat removal. The attributes to verify that the component condition and tested capability was consistent with the design bases and was appropriate may include installed configuration, system operation, detailed design, system testing, equipment and environmental qualification, equipment protection, component inputs and outputs, operating experience, and component degradation. For each of the components selected, the inspectors reviewed the maintenance history, preventive maintenance activities, system health reports, operating experience-related information, vendor manuals, electrical and mechanical drawings, and licensee Corrective Action Program documents. Field walkdowns were conducted for all accessible components to assess material condition and to verify that the as-built condition was consistent with the design. Other attributes reviewed are included as part of the scope for each individual component. The following 17 components (samples) were reviewed: High Pressure Coolant Injection (HPCI) Pump/Turbine (3-2302): Hydraulic calculations were reviewed to ensure design requirements for flow and pressure were translated as acceptance criteria for pump in-service testing (IST). The inspectors reviewed IST results to assess potential component degradation and impact on design margins. Surveillance procedures for the HPCI pump were reviewed to ensure TS surveillance requirements were met. Maintenance and calibration procedures were reviewed to ensure instrument setpoints were consistent with design basis assumptions. In addition, the licensee actions to NRC Bulletin 88-04, "Potential Safety-Related Pump Loss," were reviewed to ensure pump minimum flow requirements were addressed. The inspectors also reviewed licensing and design basis requirements for the HPCI pump related to commitments made as a result of Extended Power Uprate (EPU). HPCI Auxiliary Oil Pump (3-2303-AOP): Surveillance testing of the HPCI pump was reviewed to ensure the auxiliary oil pump operated when required to support the HPCI pump. Recent oil filter change-out and oil sample documentation were reviewed to ensure oil quality was maintained. The inspectors also reviewed the following electrical attributes associated with 250 Vdc pump motor: minimum allowable motor dc voltage, schematic drawing control interlocks, power supply, 5 Enclosure overload selection criteria, short circuit protective device selection, power circuit voltage drop, correct power division assigned, and control and power cable routing. HPCI Turbine Steam Supply Line Valve (3-MOV 2301-3): The inspectors reviewed the motor-operated valve (MOV) calculations, including required thrust, weak link, and maximum differential pressure, to ensure the valve was capable of functioning under design and licensing bases conditions. Diagnostic and IST results were reviewed to verify acceptance criteria were met and performance degradation would be identified. In addition, the following electrical attributes were reviewed: valve motor size, power circuit voltage drop through both armature and field reversing winding, control circuit voltage drop at minimum volt, control and power cable routing, valve develops adequate torque at low voltage, motor starter size, thermal overload criteria/selection, short circuit protective device selection, short time duty category, thermal overload protection during testing, and protective device testing. HPCI Injection Check Valve (3-2301-7): The inspectors reviewed the design, seat leakage and exercise testing of the check valve. The inspectors reviewed the IST results to verify acceptance criteria were met and performance degradation would be identified. Documentation related to recent disassembly and inspection of the check valve was reviewed to ensure valve integrity was maintained.

HPCI Discharge Isolation Valve to Feedwater (3-2301-8): The inspectors reviewed MOV calculations and analysis to ensure the valve was capable of functioning under design conditions. These included calculations for required thrust, maximum differential pressure, degraded voltage, and valve weak link analysis. Diagnostic testing and IST surveillance results, including stroke time and available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified. In addition, the following electrical attributes were reviewed: valve motor size, power circuit voltage drop through both armature and field reversing winding, control circuit voltage drop at minimum volt, control and power cable routing, valve develops adequate torque at low voltage, motor starter size, thermal overload criteria/selection, short circuit protective device selection, short time duty category, thermal overload protection during testing, and protective device testing. HPCI Cooling Water Return to Booster Pump Valve (3-2301-48): The inspectors reviewed MOV calculations and analysis to ensure the valve was capable of functioning under design conditions. These included calculations for required thrust, maximum differential pressure, degraded voltage, and valve weak link analysis. Diagnostic testing and IST surveillance results, including stroke time and available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified. In addition, the following electrical attributes were reviewed: valve motor size, power circuit voltage drop through both armature and field reversing winding, control circuit voltage drop at minimum volt, control and power cable routing, valve develops adequate torque at low voltage, motor starter size, thermal overload criteria/selection, short circuit protective device selection, short time duty category, thermal overload protection during testing, and protective device testing.

Isolation Condenser (IC) (3-1302): The inspectors reviewed calculations, including the design bases heat removal capability, thermal performance tests heat removal capabilities, tube plugging limit, and seismic mounting calculations.

The recent surveillance test results were reviewed and a component walkdown was performed. A review of the makeup sources of water was included in conjunction with the review of the IC makeup pumps and system. The inspectors also reviewed the impact on the IC caused by the EPU. IC Makeup Pumps (2/3-43122A(B)): The inspectors reviewed pump calculations, including vortex prevention, water requirements, fuel consumption of the associated diesel engine, net positive suction head (NPSH) requirements, and the system hydraulic performance calculations. Test results for the pumps and associated diesel engine were reviewed to ensure they were capable to perform their indented function. In addition, the related component's procedures were reviewed.

The inspectors performed walkdowns of the components and reviewed for potential ambient effects that could impact the operation as described in the site's design and licensing basis documents. The inspectors also reviewed the components' seismic classification to ensure the pumps would be available to perform their function as required. The inspectors reviewed the control schematics of the IC make-up pumps.

IC Condensate Outlet Valve (3-1301-3): The inspectors reviewed MOV calculations, including the thrust, differential pressure, and weak link calculations. As part of the review the inspectors included the recent test results for IST and diagnostics. A significant effort was devoted to reviewing the historic and current health of the valve. The review included but was not limited to: operational difficulties, conditions which resulted in the valve being inoperable, operability determinations of the valve, maintenance history, and the leak tightness of the valve. In addition, the following electrical attributes were reviewed: valve motor size, power circuit voltage drop through both armature and field reversing winding, control circuit voltage drop at minimum volt, control and power cable routing, valve developing adequate torque at low voltage, motor starter size, thermal overload criteria/selection, short circuit protective device selection, short time duty category, thermal overload protection during testing, and protective device testing.

Isolation Condenser Makeup Valve (3-4399-74): The inspectors reviewed MOV calculations, including the thrust, differential pressure, and weak link calculations. The inspectors performed walkdowns of the component and reviewed for potential ambient effects which could impact operation. The inspectors also reviewed the valve's seismic classification. In addition, the following electrical attributes were reviewed: valve motor size, power circuit voltage drop through both armature and field reversing winding, control circuit voltage drop at minimum volt, control and power cable routing, valve develops adequate torque at low voltage, motor starter size, thermal overload criteria/selection, short circuit protective device selection, short time duty category, thermal overload protection during testing, and protective device testing. Automatic Depressurization System (ADS) Valves (3-203-3A/B/C/D/E): There are five ADS valves, four of which are electrically operated electromatic valves, and one pneumatically operated Target Rock valve, which is a dual function ADS 7 Enclosure valve and steam safety relief valve (SRV). For the Target Rock valve, the inspectors reviewed calculations used for sizing of the air accumulator to ensure the valves were capable of functioning under loss of normal air conditions. The inspectors also reviewed recently completed leak rate testing performed for the air system connected to the accumulators to verify the acceptance criteria were appropriate. The inspectors also reviewed licensing and design basis requirements for the ADS valves related to commitments made as a result of EPU. In addition, results of recent valve actuators stroke testing were reviewed to ensure pressure relief functions are operable. For the electromatic valves, the inspectors reviewed electro-pneumatic functionality of the ADS valves. The inspectors also verified the voltage adequacy at the solenoid coils.

250 Vdc Batteries (2 and 3) and Battery Chargers (3 and 2/3): The inspectors reviewed the following attributes associated with the batteries: electrolyte temperature range, hydrogen monitoring, duty cycle/mission time, minimum electrolyte temperature, replacement criteria, sizing, and short circuit capacity.

The inspectors reviewed the following attributes associated with the battery chargers: design requirements, charger alarms, and electrical protection. The inspectors also reviewed room temperature requirements and hydrogen generation calculations. 250 Vdc Motor Control Centers (MCC) (3A and 3B): The inspectors reviewed the following attributes: Bus/MCC engineered safety feature separation division, battery electrical protection, main bus 2 short circuit rating, short circuit rating, control of swing charger, input power sources, connected loads, short circuit capability, voltage drop to the buses, bucket replacements, protective device selection, coordination, and component ratings. Emergency Diesel Generator (EDG) (3-6601) and EDG Output Breaker: The inspectors reviewed EDG loading calculation, EDG output breaker interrupting rating, short circuit calculation for bus 34-1, voltage drop calculation to verify voltage available at EDG air start solenoid valve and closing coil of EDG output breaker, electrical schematic diagram for EDG output breaker, and protective relaying for the EDG. The inspectors also reviewed the adequacy of the EDG grounding resistor with respect to the system charging current. Voltage and frequency requirements for the EDG as identified in the TS were reviewed. The inspectors also reviewed completed TS surveillances including the endurance test to ensure that the loading test was conducted per the power factor requirements in the TS.

4160 Vac Bus (34/34-1): The inspectors reviewed the short circuit calculation, degraded voltage calculation including the degraded voltage relay setpoint calculation, protective relaying and coordination calculation, auxiliary power system analysis calculation, voltage drop calculation to verify adequacy of voltage at the breakers' closing coils, and electrical schematic diagrams of feeder breakers to buses 34 and 34-1. The inspectors also reviewed work orders related to breaker testing. 480 Vac Switchgear Bus (39/MCC 39-7): The inspectors reviewed the short circuit calculation, degraded voltage calculation, protective relaying and coordination calculation, auxiliary power system analysis calculation, voltage 8 Enclosure drop calculation for both power and control circuit, and electrical schematic diagrams of feeder breakers to buses 39 and 39-7. The inspectors also reviewed the low pressure coolant injection swing bus MCC's 39-7/38-7 undervoltage, overvoltage, frequency relays settings, including the electrical schematic diagrams of feeder breakers to 38-7.

Unit 3 Reactor Protection System (RPS) Trip Relays: The inspectors reviewed overvoltage, undervoltage, and underfrequency relay setpoint calculation for the electrical protection assembly associated with the RPS electric power monitoring including the time delay setpoint error analysis. The inspectors also reviewed electrical schematic diagrams and the TS required surveillance including calibration of the relays.

b. Findings

(1) Adequate Voltage not Assured for EDG Air Start Solenoid Valve

Introduction:

The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to assure and verify adequate voltage was available at the air start solenoid associated with Unit (U) 2 and 2/3 EDGs. Specifically, the licensee failed to assure the minimum available voltage at the air start solenoid met the minimum rated voltage value for the solenoid.

Description:

During a walkdown of the U3 EDG, the inspectors noted the air start solenoid valve nameplate indicated a voltage operating range of 90 to 140 Vdc. The inspectors questioned what the worst case voltage would be available to operate the solenoid. The licensee indicated that no 125 Vdc calculation existed for any of the EDG air start solenoid valves. However, based on calculation DRE03-0025, "Baseline Calculation for 125 Vdc ELMS-DC Conversion to DCSDM", the licensee determined that the minimum available voltages at the 125 Vdc buses that supply the EDG control panels were 93.747 Vdc, 98.431 Vdc, 95.968 Vdc for U2 EDG, U2/3 EDG, and U3 EDG, respectively. The licensee then calculated the voltage drops from the control panels to the individual solenoid valves and determined the minimum voltages available at the air start solenoids were 86.417 Vdc, 76.661 Vdc, and 88.638 Vdc, respectively. Since all these voltages were below the vendor specified minimum voltage rating of 90 Vdc, the licensee initiated Action Request (AR) 01472605. The licensee justified the continued functionality/operability of these valves based on a quality receipt inspection test conducted in 2010. The test for U2 and U3 EDG air start solenoids indicated the pick-up voltages were less than or equal to 85 Vdc and the pick-up voltage for U2/3 EDG solenoid was 68.8 Vdc. The inspectors questioned the validity of these values because the testing was performed at an ambient temperature of 16.9 degrees Celsius (°C), which was less than the assumed EDG room design temperature of 50°C. During the inspection, the licensee tested an air start solenoid from the warehouse at various input port air pressures and determined the voltages required for the solenoid pick-up. However, all these tests were conducted at an ambient temperature of 16.9°C and not the higher room design temperature. The licensee followed up this test with a design basis calculation to determine the voltages available at the air start solenoid valves at an ambient temperature of 50°C. Although this was a minor revision to existing calculation DRE03-0025, the licensee used a different methodology. The inspectors had 9 Enclosure some concerns with the new methodology that were documented in AR01475007. The licensee was able to demonstrate, however, the two methodologies calculated similar results. The calculation determined the voltages available at both the U2 and U2/3 EDGs air start solenoid valves were below the vendor minimum voltage rating of 90 Vdc, while the U3 EDG voltage was above 90 Vdc. All of the solenoids installed were tested to pick-up at a voltage below their calculated available voltage to ensure the solenoids would function as required. Corrective actions for this issue included revising the quality receipt inspection test to require the solenoid valve to pick-up at or less than 70 Vdc and revising Model Work Order, "4 Y PMs to replace the EDG solenoid valves," to require solenoid testing to pick-up at or less than 75 Vdc.

Analysis:

The inspectors determined the failure to assure and verify that adequate voltage was available to energize the air start solenoid was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to assure and verify adequate voltage was available at the air start solenoid could have affected the capability of EDGs and other safety-related equipment powered by the EDGs to respond to initiating events. Although the licensee provided test results that indicated the currently installed air start solenoid valves would energize at the less than 90 Vdc, at the time of discovery there was reasonable doubt as to the operability of the solenoids. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase I-Initial Screening and Characterization of Findings," Table 4a for the Mitigating System cornerstone. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. Specifically, the solenoid valves were verified to be able to operate at the lower available voltage. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. Specifically, the finding was related to original plant design, which had not been reviewed as part of recent licensee activities.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program. Contrary to the above, as of March 1, 2013, the licensee's design control measures failed to verify the adequacy of control voltage for safety-related air start solenoid valves. Specifically, the licensee failed to assure or verify by testing that the air start solenoid valves associated with the U2 and U2/3 EDGs would energize at the available minimum voltage under worst case design basis conditions. Specifically, the licensee did not have a design basis calculation and/or a periodic testing program, but relied upon a receipt inspection test to justify continued operability of the solenoid valves.

10 Enclosure Because this violation was of very low safety significance and it was entered into the licensee's Corrective Action Program as AR01472605 and AR01475007, which confirmed the valves would function under the available minimum voltage, required revising the testing associated with the receipt inspection process and preventive maintenance task, and reconciling the differences with the associated calculation, this violation is being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 5000237/2013006-01; 05000249/2013006-01, Adequate Voltage not Assured for EDG Air Start Solenoid Valve). (2) Non-conservative Sizing Calculation for Target Rock SRV Air Accumulators

Introduction:

The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to correctly calculate the minimum air volume and pressure required to actuate the SRVs. Specifically, when calculating the required air volume needed in the accumulator, the inspectors identified that the licensee failed to include the volume of air needed to stroke the air operator from closed to open.

Description:

The accumulator on the Target Rock SRV (2/3-0203-3A) provided pneumatic pressure and volume for valve actuation on a loss of normal pneumatic supply pressure. The pneumatic system was only required in the relief valve actuation mode, when the valve functions as an ADS valve. When acting as a safety valve, the valve actuates due to steam pressure acting against spring pressure at a setpoint of 1135 psig. The Target Rock SRV function to prevent reactor over-pressurization was not impacted by this issue.

The inspectors reviewed calculation NUC-60, "Air Accumulator System Analysis to Ensure Operability in a Loss of Coolant Accident [LOCA] for 1(2)-0203-3A at Quad Cities and 2(3)-0203-3(A) for Dresden." The calculation purpose was to determine the air volume and pressure required in the accumulator for the Target Rock SRV air actuators. The calculation provided documentation that the SRV would be capable of actuating for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the loss of normal instrument air. The actual design basis of the accumulator required sufficient capacity for the valve to perform five actuations at atmospheric drywell pressure, and two actuations at 70 percent of drywell design pressure. A critical parameter in this calculation was the volume above the diaphragm in the pilot actuator of the valves. The calculation used 15 cubic inches for the volume based on information in the Target Rock SRV Tech Manual. The inspectors questioned the basis for 15 cubic inches as to whether it included the volume of air expanded in the piston when the valve stroked. The licensee contacted the vendor, who provided additional information that the volume above the diaphragm expands an additional 25 cubic inches upon actuation. As such, this additional expansion reduced the pressure available for subsequent actuations.

The licensee initiated AR01479044 and performed an operability evaluation for the Target Rock SRVs. The licensee accommodated the additional 25 cubic inches of expansion by crediting the temperature and pressure rise in the accumulators following a LOCA, since the accumulators would heat up and internal pressure would increase. The results determined there was sufficient air volume/pressure for the SRVs to perform their function; however, there was only a minimal amount of design margin in the evaluation.

11 Enclosure

Analysis:

The inspectors determined the failure to correctly calculate the minimum air volume and pressure required to actuate the SRVs was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the errors identified in the sizing calculation resulted in over-predicting the air accumulators' performance that could potentially render the SRVs incapable of performing the required number of strokes for their safety function. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase I-Initial Screening and Characterization of Findings," Table 4a for the Mitigating System cornerstone. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. Specifically, a preliminary calculation concluded that despite the loss of design margin in the air accumulators' capacity, the ADS system would have performed its safety function as required. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. Specifically, the finding was related to original plant design, which had not been reviewed as part of recent licensee activities.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, as of March 1, 2013, the licensee failed to assure that the original plant design for the Target Rock SRV pneumatic accumulator minimum operability limits were correctly translated into specifications, drawings, procedures, and instructions. Specifically, when calculating the required air volume for the accumulators, the licensee failed to include the volume of air needed to stroke the air operator from closed to open.

Because this violation was of very low safety significance and it was entered into the licensee's Corrective Action Program as AR01479044, which preliminarily concluded the accumulator sizing was adequate and required revising the affected calculation, this violation is being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000237/2013007-02; 05000249/2013007-02, Non-conservative Sizing Calculation for Target Rock SRV Air Accumulator). (3) Failure to Ensure Functionality of HPCI Steam Supply Valve during an ATWS

Introduction:

The inspectors identified a finding of very low safety significance (Green) concerning MOV differential pressure (d/p) calculation with respect to Dresden's anticipated transient without scram (ATWS) analysis. Specifically, the inspectors identified the d/p used in calculation for the HPCI steam supply valve (2/3-2301-3) did not address the significantly higher d/p that would be applied across the MOV during an ATWS event.

Description:

The inspectors reviewed design calculation DRE03-0015, "High Pressure Coolant Injection Motor Operated (MOV) Design Basis Document and Differential 12 Enclosure Pressure Calculation," which formed the bases for design d/p under which the HPCI system MOVs need to operate against. The HPCI steam supply valve was currently analyzed to open against 1120 psid, which was based on the worst case pressures associated with a design basis LOCA. The inspectors noted in TR051DR, "ATWS Analysis for the Introduction of SVEA-96 Optima 2 Fuel at Dresden Units 2 and 3," that the MOV could get an opening signal at a pressure as a high as 1450 psig for ATWS scenario Pressure Regulator Fails Open [PRFO]. This could occur in an ATWS event early in the scenario when the main steam safety valves lift and pressurize the drywell above the initiation setpoint for initiating HPCI due to high drywell pressure. The inspectors were concerned whether the HPCI steam supply valve would be able to open under the higher d/p. The licensee initiated AR01486532 to address the inspectors' concern, and performed a new thrust profile of the MOV using a more refined Electric Power Research Institute methodology. A preliminary MIDACALC analysis using the new thrust profile indicated the MOV would have sufficient thrust to open against the higher d/p. Therefore, the inspectors concluded the HPCI steam supply valve would be functional to mitigate the consequences of an ATWS event.

Analysis:

The inspectors determined the failure to evaluate whether the HPCI steam supply valve would function during an ATWS was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not account for the higher d/p the HPCI steam supply valve would have to operate against during an ATWS event.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase I-Initial Screening and Characterization of Findings," Table 4a for the Mitigating System cornerstone. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. Specifically, the HPCI steam supply valve was reevaluated to confirm the valve would have sufficient thrust to open against the higher d/p to allow HPCI to function during ATWS event.

Enforcement:

The inspectors determined this Finding (FIN) does not involve enforcement action because no violation of a regulatory requirement was identified. The licensee entered this issue in the Corrective Action Program as AR01486532 (FIN 05000237/2013007-03; 05000249/2013007-03, Failure to Ensure Functionality of HPCI Steam Supply Valve during an ATWS). (4) Failure to Ensure Isolation Condenser (IC) Would Perform Its Safety-Related Function Under Design Conditions.

Introduction:

The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to demonstrate via test or calculation the IC would be capable of performing its safety function under design bases condition.

13 Enclosure

Description:

The safety-related IC system functions as a heat sink for decay heat removal from the reactor vessel following a reactor scram and isolation from the main condenser. Each IC (one per unit) consists of two tube bundles immersed in a large water storage tank. The tubes are U-tubes style. There are 121 tubes per bundle (242 tubes per IC). Currently, U3 has 34 tubes plugged: 24 plugged in the U3 West bundle and 10 plugged in the U3 East bundle. Unit 2 has no plugged tubes. The IC system operates by natural circulation. During operation, clean demineralized water on the shell side condenses the reactor coolant steam/water inside the IC tubes. Make-up water to the shell side is required to be manually initiated within 20 minutes using one of the various IC water make-up options. As stated in the UFSAR section 5.4.6 and TS Basis 3.5.3, the design bases for the IC are: (1) remove 252.5 MBTU/hr, which is equivalent to the decay heat rate 8.8 minutes after an automatic reactor shutdown, and (2) provide sufficient heat removal capability for 20 minutes of operation without makeup water. The TS require a number of surveillance requirements (SR) be performed to ensure these bases are met, including the following: SR 3.5.3.1: SR 3.5.3.4: Verify the IC system heat removal capability to remove design heat load [i.e., 252.5 MBTU/hr]. The TS bases for the values in SR 3.5.3.1 were documented in calculation BSA-D-95-07, "Dresden Isolation Condenser Performance." This calculation, which was based on a starting water level and temperature in the shell of 6 feet and 210 degrees Fahrenheit (°F), concluded the IC would be able to meet its required heat removal capability. However, the calculation noted in order to accomplish the required heat removal capability, the liquid water level in the shell would drop to approximately the last quarter of the tube bundle (26.3 percent tube bundle height) after 20 minutes of operation, exposing 3/4 of the tube bundles. Preliminary calculations by the inspectors determined the shell side water level would reach the top of the tube bundles approximately 13 minutes after the IC initiation if starting at a level of 6 feet. Although a large portion of the tube bundles would be exposed, the licensee assumed the heat removal capability remained constant, that is, the heat removal capability of a two-phase froth created as a result of the violent boiling and water carryover on the shellside of the IC was equivalent to the heat transferred accomplished by complete submergence. The licensee reasoned the froth created would be high enough to cover the entire tube bundles and the entire bundle would be covered by either liquid water or a two-phase mixture (froth). The licensee assumed if a tube was covered by froth, it could be considered wetted and would result in an equivalent heat removal capability as the tubes submerged under liquid water.

After reviewing calculation BSA-D-95-07 and other applicable documents, the inspectors had the following concerns regarding the IC capacity to meet its required heat removal capability under design bases conditions: The licensee did not provide a basis in BSA-D-95-07 to support the assumption the heat removal capability would remain the same for tubes covered by froth as for the tubes submerged under liquid water. During the inspection, the licensee provided 14 Enclosure some qualitative correlations to justify the height of the froth and its expected heat removal capability. The inspectors, with technical support from the Office of Nuclear Reactor Regulation, agreed the information provided was not sufficient to reasonably justify the heat removal capability of the tubes covered by the froth would be the same as those covered by liquid water. The information provided did not correlate to the specific condition in the IC (e.g., no forced flow, horizontal versus vertical tubes).

The calculation did not account for the plugged tubes in the U3 IC. The majority of the tubes plugged were located in the top and bottom of the bundles (because of the u-tube type configuration). This meant a significant portion of the tubes would not provide any heat transfer to maintain the boiling/frothing for the U3 IC.

Calculation BSA-D-95-07 was developed in 1996 using parameters effective at that time. When Dresden implemented the EPU, the rated thermal power was increased from 2527 megawatt thermal (MWt) to 2957 MWt; effectively a 17 percent power increase. The EPU was achieved without an increase in the maximum reactor pressure during a transient (hence no increase in the maximum saturation temperature). However, it did impact the time the reactor remained at a higher pressure following a transient. For the IC system, this resulted in additional time for the system to match the decay heat generated in the reactor. Until the IC system matched the decay heat rate, the reactor would lose inventory via blowdown through the SRVs. The heat removal capability of the IC system matched the decay heat rate 5 minutes after an automatic shutdown prior to the EPU, while after the EPU it would take the system 8.8 minutes. This resulted in the delta temperature between the tube side fluid and the shellside fluid would remain at its maximum value for longer time period, which could result in a lower level in the IC shell at 20 minutes. The licensee agreed the decay heat after the EPU was greater; however, concluded there would be no impact to the final level in the IC. The licensee indicated the level in the reactor vessel would be lower due to the relief valves being opened for the additional 3.8 minutes.

As part of EPU, the licensee provided General Electric (GE) a constant heat removal capability of 252.5MBTU/hr as the capacity of the ICs for use in the EPU evaluations. However, if heat removal capability was not constant, the anticipated reactor water level would be impacted. Specifically, the safety analysis for the IC calculated that after 8.8 minutes, the heat removal capability was capable of matching the reactor generated decay heat rate. As a result, the safety analysis assumed after this point no additional reactor water inventory would be lost through the SRVs. However, if the heat removal capability of the IC resulting from exposed tubes degrades below the decay heat generated in the core, pressure in the reactor could increase and potentially cause the relief valves to cycle open. This would result in additional inventory lost through the SRVs and a lower reactor vessel water level than calculated in the current safety analysis. As part of SR 3.5.3.4, the licensee conducted actual performance tests on the capacity of the IC. The inspectors noted the tests started with a shell water level of at least 7 feet and was typically stopped before the level in the shell reached the tubes bundles. The licensee used the data acquired to extrapolate to design conditions and calculate the heat removal capability. Because the liquid water level was not allowed to drop significantly below the tops of the tubes, the inspectors 15 Enclosure concluded these tests, by themselves, did not demonstrate a heat removal capability of 252.5MBTU/hr would be maintained for the first 20 minutes of IC operation.

The original IC design assumed after 20 minutes of operation the tube bundles would remain covered with water. The licensee later determined the assumed moisture carryover (water escaping through the vent on the shell side) was not conservative. The licensee developed BSA-D-95-07 in 1996 to determine actual water levels in the shellside (at 20 minutes) and to provide the bases for the TS required shellside water levels and temperature. When the calculation was completed and the licensee determined the tubes would be exposed during the 20 minutes of operation without makeup, the licensee failed to update the UFSAR description to state the tubes would become exposed. This error resulted in incorrect information being provided to the NRC when the licensee submitted Improved TS in 2000. This issue is further discussed in Section 1R21.3.b.(5) of this report. These concerns could potentially impact the ability of the IC to perform its safety function under design conditions. However, although TS required a minimum shell water level of at least 6 feet and a shell water temperature of no more than 210°F, the licensee had administratively maintained the level above 7 feet and the temperature at no more than 153°F. Based on multiple discussions with the licensee, review of applicable calculations and preliminary analysis developed by the licensee staff; the inspectors were satisfied that under these more restrictive limits, the licensee was able to demonstrate reasonable assurance of operability of the IC. These measures, previously administratively controlled by procedure, were now being additionally controlled by a standing order until the licensee completed permanent corrective actions. The licensee has contracted a vendor to develop a calculation and additional bases to support the previously unverified assumption. The licensee initiated AR01509103 to address this issue.

Analysis:

The inspectors determined the failure to ensure the IC would have been capable of performing its safety function under design conditions was a performance deficiency the licensee could have foreseen and correct. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors had reasonable doubt the IC system would have been able to perform its safety function during the initial 20 minutes of operation if called upon under design conditions. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase I-Initial Screening and Characterization of Findings," Table 4a for the Mitigating System cornerstone. The finding screened as very low safety significance (Green) because the finding was a design deficiency that did not result in a loss of operability or functionality. Specifically, the licensee administratively controlled shellside water level of at least 7 feet and shell water temperature at or below 153°F; such that the IC tubes remained covered during a significant portion of the 20 minutes prior to the initiation of makeup. This provided the inspectors with reasonable assurance of operability. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. Specifically, the 16 Enclosure finding was related to an old design issue, which had not been reviewed as part of recent licensee activities.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, as of May 30, 2013, the licensee failed to ensure via calculation or test the IC would be capable of performing its safety function under design conditions. Specifically, the licensee was unable to justify the assumption the heat removal capability would remain constant once the IC tubes began to become exposed. As part of their corrective actions, the licensee instituted a standing order to ensure shellside water level and temperature are maintained in a more restrictive band. In addition the licensee contracted a vendor to develop a calculation and additional bases for their design assumptions. Because this violation was of very low safety significance, and it was entered into the licensee's Corrective Action Program as AR1509103, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000237/2013007-04; NCV 05000249/2013007-04, Failure to Ensure IC Would Perform Its Safety-Related Function Under Design Conditions.) (5) Inaccurate Information Provided During Improved TS Submittal

Introduction:

The inspectors identified an Unresolved Item (URI) associated with the information provided to the NRC in a license amendment application. Specifically, in a letter dated March 3, 2000, the licensee inaccurately stated the TS limits established for the IC provided sufficient capability for 20 minutes of operation without makeup water, before beginning to uncover the tube bundles. However, based on the licensee's 1996 design calculation, the IC tube bundles would become significantly uncovered.

Description:

As part of the review for the NCV documented in Section 1R21.3.b.(4) of this report, the inspectors noted the licensee had provided the NRC inaccurate information in a previous correspondence. In a letter dated March 3, 2000, the licensee requested the NRC to approve a license amendment to upgrade to Improved TS. The proposed change included adding explicit values for the IC minimum starting shell side water level and maximum starting water temperature. The previous version of the TS had no explicit values listed, so the limits were controlled by the licensee administratively. As a result, the proposed change was intended to be more restrictive on plant operation. The inaccurate information was included as part of the above mentioned letter under Enclosure A, "Dresden ITS [Improved Technical Specification] Conversion Document," Volume 5. Specifically, the basis for TS SR 3.5.3.1 stated, in part: "Based on a scram from 2552.3 MWt (101% RTP), a minimum water level of 6 removal capability for 20 minutes of operation without makeup water, before beginning to uncover the tube bundles [emphasis added]. The volume and temperature allow sufficient time for the operator to provide makeup to the condenser."

17 Enclosure On March 26, 1996, the licensee developed calculation BSA-D-95-07, "Dresden Isolation Condenser Performance," in order to determine whether the IC would meet its safety function. The calculation determined the water in the shell side would boil down below the top of the tube bundles before makeup was established based on the proposed TS. Specifically, the licensee determined the liquid water level could reach as low as 26.3 percent tube bundle height and result in 75 percent of the tube bundles being uncovered. Therefore, in 1996 the licensee determined the tube bundles would actually become uncovered, before make-up could be established. This was contrary to both statements in the UFSAR and TS Bases that stated the tubes would remain covered for the first 20 minutes.

As part of process for implementing BSA-D-95-07 as a design calculation, the licensee should have updated the UFSAR and TS Bases statements be in agreement with the new calculation results. However, the licensee did not update the documents as required. This discrepancy was identified during a subsequent NRC inspection on December 18, 2003. The licensee documented the discrepancy under AR 00191696, "Uncovering of the Isolation Condenser Tubes." As part of the corrective actions, the licensee initiated change request 04001 and 04003 to correct the TS Bases and UFSAR respectively. Although the discrepancy was identified, it was not recognized that the inaccurate information was previously submitted to the NRC as part of the Improved TS process. Based on the licensee's 1996 design calculation, a portion of the tube bundles would be exposed before the makeup could be established at the IC proposed limits contrary to information provided to the NRC in the March 3, 2000, submittal. The NRC was unaware of this discrepancy and approved the requested amendment by letter dated March 30, 2001. Based on the new information, it was unclear as to whether the NRC would have approved the IC values included in TS. As stated in Section 1R21.3.b.(4), the licensee implemented administrative limits on the IC level and water temperature; therefore, this unresolved item does not involve an immediate safety concern. This issue is considered Unresolved Item (URI) pending additional information regarding the adequacy of the existing TS values for the IC. (URI 5000237/2013007-05; 05000249/2013007-05, Inaccurate Information Provided During Improved TS Submittal)

.4 Operating Experience

a. Inspection Scope

The inspectors reviewed five operating experience issues (samples) to ensure that NRC generic concerns had been adequately evaluated and addressed by the licensee. The operating experience issues listed below were reviewed as part of this inspection: GL 1987-06, "Periodic Verification Of Leak Tight Integrity Of Pressure Isolation Valves"; GL 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems"; IN 2010-25, "Inadequate Electrical Connections"; IN 2012-06, "Ineffective Use of Vendor Technical Recommendations"; and IN 2012-11, "Age Related Capacitor Degradation."

b. Findings

No findings of significance were identified.

.5 Modifications

a. Inspection Scope

The inspectors reviewed three permanent plant modifications related to selected risk significant components to verify that the design bases, licensing bases, and performance capability of the components had not been degraded through modifications. The modifications listed below were reviewed as part of this inspection effort: EC 338243, Unit 3 HPCI Piping Steam Void EC; EC 366256, Add Interposing Relay for Unit 3 Diesel Generator Output Breaker; and EC 366676, Add Interposing Relay for Diesel Generator Output Breaker Unit 2.

b. Findings

No findings of significance were identified.

.6 Operating Procedure Accident Scenarios

a. Inspection Scope

The inspectors performed a detailed reviewed of the procedures listed below associated with the selected scenario, the Loss of Main Feed Water. For the procedures listed time critical operator actions were reviewed for reasonableness, in plant action were walked down with a qualified operator, and any interfaces with other departments were evaluated. The procedures were compared to UFSAR, design assumptions, and training materials to assure for constancy.

The following operating procedures were reviewed in detail: DEOP 0100-00, "RPV Control"; DEOP 0400-05, "Failure To Scram"; DOA 2300-03, "High Pressure Coolant Injection System Local Manual Operation"; DOA 6600-01, "Diesel Generator Failure"; DOP 1300-03, "Manual Operation Of The Isolation Condenser"; DOP 1300-08, "Diesel Fuel Oil Transfer System Unavailable"; DOP 1300-09, "Isolation Condenser Makeup Pump Local Operation"; and DOS 1300-03, "Isolation Condenser Makeup Pump Quarterly Operability."

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Review of Items Entered Into the Corrective Action Program

a. Inspection Scope

The inspectors reviewed a sample of the selected component problems that were identified by the licensee and entered into the Corrective Action Program. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions related to design issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the Corrective Action Program. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report. The inspectors also selected three issues that were identified during previous CDBIs to verify the concern was adequately evaluated and corrective actions were identified and implemented to resolve the concern, as necessary. The following issues were reviewed: NCV 05000237/249/2007-006-01; Inadequate Acceptance Criteria in 125 Vdc Station Battery Service Test Procedures; NCV 05000237/249/2007006-02; Adequate Control Voltage for 4160 Breaker's Closing Coil was not Assured; and NCV 05000237/249/2010007-03; Failure to Perform Adequate Testing to Confirm Acceptable Fast Bus Transfer Time.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

.1 (Closed) Unresolved Item 05000237/2012002-01; 05000249/2012002-01:

Failure to Report an Unanalyzed Condition that Could Significantly Degrade Plant Safety As documented in Inspection Report 05000237/2012002; 05000249/2012002, the licensee initiated an immediate operability determination to assess the impact of a void in the common discharge piping between the shutdown cooling (SDC) and LCPI systems. Specifically, while on SDC in Mode 3 and a LOCA were to occur, the SDC and LPCI common discharge piping would void resulting in flashing and pipe damage during the 40 seconds it takes for the SDC isolation valves to close on a Group 3 isolation signal. The licensee concluded the LPCI system would remain operable; however, 20 Enclosure additional analysis was required to determine if the event was plausible and if the plant was bounded by previous analysis.

The licensee completed a prompt operability evaluation which confirmed LPCI was operable and the condition was not unanalyzed. The licensee justified prompt operability after 60 days based on 1) a GE BWR Owners Group Technical Report "Effects of Voiding in emergency core cooling system (ECCS) Drywell Injection Piping," 2) the probability of occurrence of the aforementioned scenario being on the order of 1E-08 and was therefore not a credible event and did not require further analysis, and 3) the fact that preliminary results done at Quad Cities and Duane Arnold Energy Center for similar issues indicated that all operability requirements for piping were met. The licensee justified reportability as no loss of safety function occurred and therefore this issue was not reportable. This issue was unresolved because the inspectors had several concerns with the licensee's justification and the licensee indicated an ongoing technical evaluation would demonstrate this condition would not challenge the ability of LPCI to inject. During this inspection, the inspectors reviewed the licensee's justification and identified the following concerns: The BWR Owners Group Technical Report 0000-0088-8669-R0, "Effects of Voiding in ECCS Drywell Injection", referenced in AR01341563 was not applicable for the circumstances at Dresden. Specifically, a. The report was a qualitative assessment developed to address specific concerns of GL 2008-01. The focus of the report was on potential voids downstream of the first normally closed motor operated isolation valve. However, the report's emphasis was on air voids, mainly introduced due to inappropriate fill and vents. It specifically stated the scope of review would be limited to voids existing prior to an accident or transient. Under the postulated scenario steam voids (not air) would be created as a result of an accident. b. The report did discuss some flashing in the discharge segment of piping would be expected and was designed accordingly. However, it was not clear to what extent the original design of the plant accounted for flashing. In particular it could not be determined if flashing in the section of concern was expected and accounted for in the original design. EC 389067, "Evaluate LPCI Injection Piping Post Accident while Operating in SDC Mode", Revision 0, concluded, based on the specific conditions postulated by the EC, the pressure in the discharge sections of piping would not drop below the saturation pressure. As a result, no flashing would have occurred. The inspectors identified a number of discrepancies with the EC.

a. The EC did not account for the plant's design bases conditions when the subject LOCA could occur. Based on discussions with the licensee, the parameters used were intended to determine past operability. The water temperature in the discharge piping was based on a starting reactor water temperature (controlled administratively) of 330F instead of 350F (the maximum water temperature which could be present at the start of SDC).

21 Enclosure b. The expected temperature drop through the SDC heat exchangers was assumed to be 70F. This was based on historically observed SDC heat exchanger performance, not on the performance expected during a worst-case scenario (i.e. a hot day when river water temperature was at its design maximum). c. There was a numerical error, which resulted in a 10F temperature drop, which was not accounted. The inspectors determined the above discrepancies impacted the conclusions of the EC and would have required the licensee to either analyze for potential dynamic transient effects or change the parameters of the evaluation to remove conservatism.

While reviewing the ARs associated with the issue, the inspectors noted the licensee also used EC 389067 for a design bases assessment and to justify no significant changes to existing evaluations and no procedure revisions. The inspectors were concerned the licensee based a number of conclusions (affecting future operation of the plant) on an EC, which was not intended to bound all potential design conditions. Similarly, because the EC was intended to review for past operability, it was not subjected to the typical rigor required for a design calculation/evaluation As a result, during the CDBI inspection, the licensee contracted an outside vendor to develop a design basis RELAP5 evaluation. The RELAP5 model showed that under the scenario of concern, the system would not have been subjected to loads that would have challenged operability of the LPCI system.

Lastly, the inspectors were concerned with the licensee's determination on AR1353772, Assignment #4 and AR1312222, Assignment #2. These concluded no further actions, in addition to the ones already performed, were needed for Dresden. These conclusions were based on the BWROG-TP-12-025, "Position Paper-Guidance for Alignment and Venting of RHR Loop for Post-LOCA LPCI Mode Operation." The licensee stated: "Oyster Creek and Dresden have separate SDC systems from LPCI. They are not affected by these BWROG products and therefore no additional actions to Oyster Creek and Dresden are appropriate" The inspectors noted the licensee did not interpret the scope sections of the BWROG document appropriately. Although the scope section stated the developed product did not contain recommendations for BWR/3 designs (Dresden), the reasoning provided was: "-because SDC system components are separate from the RHR subsystems and are therefore not impacted by the transient of concern described in Objective Section I of this document." The objectives section provided the following description: "To provide participating stations with guidance to successfully perform alignment and venting of [RHR] system/subsystems/components prior to [LPCI] mode operation following a [LOCA] occurring at or below the SDC mode permissive pressure. Recommendations contained within this product are for procedural enhancements for RHR system realignment to LPCI from SDC while in Mode 3" 22 Enclosure As noted above, the "transient of concern" discussed in the BWR Owners Group Technical Report was the RHR realignment - which is not the transient of concern for Dresden (initiation of RHR when the separate SDC system is in operation). As a result the Technical Report did not apply to Dresden, but it did not exclude Dresden from the effects of the initiating RHR while in SDC is operating.

In an effort to resolve the inspectors' concern, the licensee contacted GE and verified the piping segments in question were actually designed and supported to the same standards and requirements as the sections of pipe expected to undergo flashing during a full power LOCA. With this information the inspectors had reasonable assurance the event in question would not have any significant adverse effects on the operability of the LPCI system and had been incorporated into the original design of the plant. The inspectors reviewed the examples provided in NUREG-1022, "Event Reporting Guidelines," Revision 2, and determined the safety function of maintaining the core covered and cooled during a LOCA event would not have been prevented in this scenario.

Based on the inspectors review, the licensee was able to verify the scenario in question was part of the plant's original design. The inspectors concluded that no violation of regulatory requirements occurred with regards to the reportability of this issue. This unresolved item is closed.

4OA6 Meeting(s)

.1 Exit Meeting Summary On May 30, 2013, the inspectors presented the inspection results to Mr. D. Czufin, and other members of the licensee staff.

The licensee acknowledged the issues presented.

.2 Interim Exit Meeting Summary On March 1, 2013, the inspectors presented the inspection results to Mr. D. Czufin, and other members of the licensee staff.

The licensee acknowledged the issues presented. Several documents reviewed by the inspectors were considered proprietary information and were either returned to the licensee or handled in accordance with NRC policy on proprietary information. ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

D. Czufin, Site Vice President

S. Marik, Station Plant Manager

H. Dodd, Regulatory Assurance Manager

J. Fox, Design Engineer

G. Graff, Nuclear Oversight Manager

B. Kapellas, Operations Director

J. Knight, Director, Site Engineering

B. Madderon, Electrical Design Engineer

M. Mitidiero, Electrical Design Engineer

M. McDonald, Maintenance Director

M. McDonald, Design Engineering - Rapid Response Team

P. O'Brien, Regulatory Assurance - NRC Coordinator

D. Phalen, Programs Engineer - Maintenance Rule

R. Ruffin, Licensing Engineer

J. Sipek, Work Management Director

R. Urbanek, System Engineer - Electrical

P. Wojkiewicz, Design Engineering Manager Nuclear Regulatory Commission A.Stone, Chief, Engineering Branch 2

G. Roach, Senior Resident Inspector

D. Meléndez-Colón, Resident Inspector

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000237/2013007-01;

05000249/2013007-01 NCV Adequate Voltage not Assured for EDG Air Start Solenoid Valve (Section 1R21.3.b.(1))

05000237/2013007-02;

05000249/2013007-02 NCV Non-conservative Sizing Calculation for Target Rock SRV Air Accumulators (Section 1R21.3.b.(2))

05000237/2013007-03;

05000249/2013007-03 FIN Failure to Ensure Functionality of HPCI Steam Supply Valve during an ATWS (Section 1R21.3.b.(3))

05000237/2013007-04;

05000249/2013007-04 NCV Failure to Ensure Isolation Condenser Would Perform Its Safety-Related Function Under Design Conditions (Section 1R21.3.b.(4))

05000237/2013007-05;

05000249/2013007-05 URI Failure to Provide Complete and Accurate Information on the Isolation Condenser when ITS was Submitted (Section 1R21.3.b.(5))

Closed

05000237/2012002-01;

05000249/FIN-2012002-01 URI Failure to Report an Unanalyzed Condition that Could

Significantly Degrade Plant Safety (Section 4OA5.1)

05000237/2013007-01;

05000249/FIN-2013007-01 NCV Adequate Voltage not Assured for EDG Air Start Solenoid Valve (Section 1R21.3.b.(1))

05000237/2013007-02;

05000249/FIN-2013007-02 NCV Non-conservative Sizing Calculation for Target Rock SRV Air Accumulators (Section 1R21.3.b.(2))

05000237/2013007-03;

05000249/FIN-2013007-03 FIN Failure to Ensure Functionality of HPCI Steam Supply Valve during an ATWS (Section 1R21.3.b.(3))

05000237/2013007-04;

05000249/FIN-2013007-04 NCV Failure to Ensure Isolation Condenser Would Perform Its Safety-Related Function Under Design Conditions (Section 1R21.3.b.(4))

Attachment

LIST OF DOCUMENTS REVIEWED

The following is a list of documents reviewed during the inspection.

Inclusion on this list does not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that selected sections of portions of the documents were evaluated as part of the overall inspection effort.

Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it, unless this is stated in the body of the inspection report. CALCULATIONS Number Description or Title Revision 70213-06 Dresden Unit 3 Battery Room Hydrogen Generation 0 8231-03-19-1 LPCI Swing Bus MCC's 39-7/38-7) Relay Settings 002 88-17-19-2 Second Level Undervoltage Relay Setpoint - Unit 3 004 8900-15-E0-S Re-evaluate Anchorage Assemblies for Isolation Condenser

6 8956-56-19-2 250

VDC System Short Circuit Current 3 8982-19-19-3 Nonsize 2 Motor Control Center (MCC) Control Voltage Contactor Circuit Length Fed from Switchgear 39 002 9198-18-19-1 Replacement of 250 V BC 2 and 125 V BC 2A 3A 9198-18-19-2 Replacement of 250 V BC 2/3 and 125 V BC 2 3A 9198-18-19-3 Replacement of 250 V BC 3 and 125 V BC 3A 3A 9198-18-19-4 Replacement of 250 V BC 2/3 and 125 V BC 3 3A 93839-46-19-1 Diesel Generator 3 Loading Under Design Bases Accident Condition 003

BSA-D-95-07 Dresden Isolation Condenser Performance 0

BSA-D-99-04 Reconstitution of Isolation Condenser Design Bases with Respect to Decay Heat Loads and Long Term Makeup Requirements 1A

CE-DR-020 MOV Weak Link Evaluation (M02-3-4399-74) 0

DR-019-E002 4kV Bus 23-1/33-1 and 24-1/34-1 Coordination Study 003A

DR-150-M-002 Iso-Condenser Make-up Pump, Minimum Flow Recirc Line 0

DR-265-M-001 MOV Differential Pressure and Thrust Window Calculation for 2(3)-4399-74 1

DRE 13-0006 EDG Neutral Grounding Transformer/Resistor Calculation 000 DRE00-0038 Reactor Protection System (RPS) Electric Power Monitoring - Overvoltage, Undervoltage, and Underfrequency Time Delay Setpoint Error Analysis 0 DRE00-0041 Overvoltage, Undervoltage, and Underfrequency Relay Setpoint Calculation for the Electrical Protection Assembly (EPA) Associated with the RPS 0 DRE00-0054 HPCI Condensate Storage Tank Level Error Analysis 0 DRE02-0034 Motor Operated Valve AC Motor Terminal Voltage Calculation for Dresden System 1501, Unit 3 00H DRE02-0045 Motor Operated Valve AC Motor Terminal Voltage Calculation for Dresden System 0202, Unit 3 00B DRE02-0049 HPCI Valve Open Permissive Pressure Switches 1 DRE03-0011 Isolation Condenser System Combined DBD and DP Calculation 1 DRE03-0015 HPCI MOV Design Basis Document and Differential Pressure Calculation 0B DRE03-0025 Baseline Calculation for 125

VDC

ELMS-DC Conversion to DCSDM 001D DRE04-0003 Baseline Calculation for 250 VDD

ELMS-DC Conversion to DCSDM

F DRE04-0019 Auxiliary Power Analysis for Dresden Unit 3 005

DRE-3-1301-3 MidaCalc Results for DC MOV 3-1301-3 4

DRE-3-2301-3 MidaCalc Results for DC MOV 3-2301-3 3

DRE-3-2301-48 MidaCalc Results for DC MOV 3-2301-48 2

DRE-3-2301-8 MidaCalc Results for DC MOV 3-2301-8 4

Attachment CALCULATIONS Number Description or Title Revision DRE96-0124 Dresden HPCI NPSH Temperature Limits 1 DRE96-0126 Motor Terminal Voltage for Dresden 250

VDC MOVs 001C DRE96-0149 Breaker Settings for Bus 28 and 29 003C DRE96-0206 HPCI Pump Discharge Pressure For 5000 gpm Flow To Reactor Vessel 1 DRE96-0215 Pressure Drop Analysis For HPCI Exhaust Steam Piping 1 DRE96-0267 480 V MCC Circuit Protection for 125/250V Battery Chargers 0A DRE97-0068 Minimum Flow Through The LPCI And HPCI Pump Minimum Flow Lines 2C DRE97-0135 HPCI Pump Discharge Pressure Indication Accuracy For Normal Conditions 1 DRE97-0252 Sizing of CCSW Pipe and Flow Orifices for ECCS Room Coolers 3 DRE98-0030 Determination Of Setpoint Of CST Low-Low Level Switches To Prevent Potential Air Entrainment From Vortexing During HPCI Operation

0B DRE98-0077 HPCI Room Thermal Response with Reduced Room Cooler Capability 1 DRE99-0011 Isolation Condenser Makeup Pumps Hydraulic Performance 1A DRE99-0012 Frictional Pressure Losses In HPCI Turbine Steam Supply Piping

0 DRE99-0013 Hydraulic Performance Of The HPCI System

2D DRE99-0075 Seismic Analysis of Isolation. Condenser Pump House 0

EC 363289 D3 250 V Battery Modified Performance Test Profile 0

EC 366730 Evaluation of the Dresden 2/3 Isolation Condenser System Restoration Following a GR V Isolation During Operation 0

ERC-025115-1 Equivalency Evaluation for Cutler Hammer DC MCC Cubicle 0

NUC-60 Air Accumulator System Analysis for 1(2)-0203-3AB at Quad and 2(3)-0203-3AB for Dresden 2

OTC-267 Thrust and Torque Calcs for 3-2301-8 4

OTC-296 Thrust and Torque Calcs for 3-2301-48 2

OTC-309 Thrust and Torque Calculations 3-1301-3 2

CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION Number Description or Title Date

01348180-04 Additional Information Regarding Voiding in SDC Mode 03/01/13

01468025 Incorrect Reference in Calculation DRE96-0126, Rev 01C 01/28/13

01468442 Leak in Roof of IC Makeup Pump Blockhouse 01/29/13

01468979 House Keeping Issue During Walkdown 01/30/13

01469090 NRC Identified Issues Petro-Guard Pump 01/30/13

01469502 IC Diesel Makeup Pump Building Foundation Buoyancy 01/31/13

01469666 Typo Found on

DR-0190E-002 01/31/13

01469906 Unsecured Eyewash Stations in IC Makeup Pump Room 01/31/13

01470323 EDGs 3250 KVA Mislabeled as 3125 KVA on Various Drawings 02/01/13

01470362 EDG Neutral Grounding Transformer 02/01/13

01470397 Typo Found in Calc 9389-46-19-1 02/01/13

01472101 Battery Charger Ripple Not Measured per PCM Template 02/06/13

01472146 Incorrect Buses Identified in Calculation of Fault Current 02/06/13

01472605 EDG Air Start Solenoid Voltage Calculation 02/07/13

01472634 Unit 3 250 V Modified Performance Test Acceptance Criteria 02/07/13

01473651 Calc. DRE96-0149 & DRE96-0267 Not Revised per Mod. 02/11/13

01473934 U2/3 IC Makeup Pumps Seismic Classification 02/11/13

Attachment CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION Number Description or Title Date

01474442 Calc DRE03-0015 Needs Clarification 02/12/13

01474859 Wrong Motor HP Shown on Key Diagram for MOV 3-2301-3 02/14/13

01474952 Issued Identified in CDBI Walkdown With NRC in

DOA 2300-03 02/13/13

01475007 DRE03-0025 Calculation Technique Differences 02/13/13

01475450 250

VDC Syst. Op Proc. Fails to Identify TRM Requirements 02/14/13

01475559 Calculation 9389-46-19-1 Typo & Admin Error

02/14/13

01477214 Battery Test Surveillance Procedures to Include Required Acceptance Criteria 02/19/13

01477345 Incorrect Statement in Calculation DRE99-0011 02/19/13

01477844 Outdated Information in Calc DRE03-0011 02/20/13

01477922 Incorporate Testing Criteria for Single TOL Testing 02/21/13

01478006 Incorrect Reference Used in Calculation 02/20/13

01478014 Calculation Typo and Admin Error 02/20/13

01478184

DOA 2300-03 Enhancements Identified 02/21/13

01479044 Calc

NUC-60 Requires Revision From Critical Param 02/22/13

01479066 Calculation Formula Administrative Discrepancy 02/22/13

01479627 JOG Classification of MOVs 02/25/13

01479930 Safety Concern, Extension Cord Plug Exposed to Elements 02/25/13

01480302 Enhancement to Engine Radiator Overpressure Line 02/26/13

01480460 Calculation 01849.00 E(B) Should have been Superseded when

MOV 4399-74 Motor was Changed to a DC Motor

02/26/13

01480842 Basis for Short Circuit Interrupting Ratings at MCCs 2 & 3 02/27/13

01480843 Calculation Incorrectly Identifies the CB Duty as the CB Rating 02/17/13

01481035 Incorrect Reference to TS in Proc. 8300-15 02/27/13

01481053 Inconsistencies in MOV HP on Key Diagrams 02/27/13

01481422 Wrong Test Current used for Maintenance Testing of TOLs 02/28/13

01481926 HP Ratings Listed on Drawing 12E-2321 & 12E-3321 03/01/13

01481936 Reference in Proc.MA-AA-723-325 not Applicable to Dresden 03/01/13

01482030 Petroguard Pump Demonstration 03/01/13

01486532 Increase Required Opening d/p for

MOV 2301-3 03/12/13

01490144 Revision to Calculation

BSA-D-95-07 Required 03/20/13

CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION Number Description or Title Date

00210558 Proof Needed to Conclude that Cold Makeup Splash 03/24/04

00630890 Battery Service Test Criterion Requires Revision 05/17/07

00723893 3C Main Steam Line Electromatic RV Temperature Indication 01/17/08

00748689 HPCI Oil Pressure Inadequate 03/12/08

00754896 MOV Testing Development Of COF Acceptance Criteria 03/26/08

00773115 U3 HPCI Room Sump Has Oil In It, Needs Clean Out 05/08/08

00867982 Increase Trend In Unit 3 E ERV Tailpipe Temperature 01/16/09

01032550 HPCI Aux. Oil Pump Not Torqued to Seismic Requirement 10/23/02

01069334 Lack of Documented 4KV Auto Bus Transfer Analysis 05/14/10

01071620 NRC Identified Missing Procedural Requirement 05/20/10

01071691 4 KV Fast Bus Transfer Timing Tests with AMHG GCBS 05/20/10

Attachment CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION Number Description or Title Date

01109362 U3 IC Tube Side Temp High Following

DOS 1600-05 09/03/10

01110939 Review of

IR 1109362 3-1301-3 Valve Seat Leakage 09/08/10

01138662 Discrepancies Noted During 3C ERV Surv. 11/10/10

01142554 Unit 3 Isolation Condenser Operability Review Results 11/19/10

01144497 IC Thermal Testing - Root Cause CA's Not Fully Incorporated 11/24/10

01144605 Isolation Condenser Eddy Current Testing 11/24/10

01145149 Abnormal HPCI Unit 3 Pump Parameters During Low Pressure

11/26/10

01184914

2-1301-1 Didn't Indicate Full Open When Given Open Signal 03/08/11

01312222 Potential Re-Evaluation of SDC Piping for

GL 2008-01 01/11/12

01341563 NRC Questions Timeliness of Response to

IR 1312222 03/15/12

01348180 Additional Information Regarding Voiding In SDC Mode 03/30/12

01353772 Recommend BWRs Revise RHR Procedures To Address

IN 2010-11 04/13/12

01386885 Group V Isolation During

DIS 1300-07 07/10/12

01415535 No Valid As-Found Testing For

MOV 09/19/12

01417005 3A Cond/Cond BST PP Did Not Initially Start on Start Signal 09/22/12

01453610 Installed Contactors Do Not Meet All Acceptance Criteria 12/18/12

01467200 Reclassify CCF in

IR 1417005 as Consequential

CCF 01/25/13

01467668 EOC Delivery Date Not Met for Fast Bus Transfer Analyses 01/28/13

DRAWINGS Number Description or Title Revision 12E-2311 Key Diagram 480 V

MCC 29-2

AU 12E-2312 Key Diagram 480 V

MCC 28-3

AA 12E-2321 Key Diagram Unit 2 250

VDC MCC No. 3

AP 12E-2321A Key Diagram 250

VDC Non-Essential MCC No.

2-8350-1A B 12E-2321B Key Diagram 250

VDC Non-Essential MCC No.

2-8350-1B B 12E-2328 Single Line Diagram Emergency Power System O 12E-2389K Schematic Diagram 250V DC Battery Charger 2 A 12E-2389M Schematic Diagram 250V DC Battery Charger 2/3 A 12E-3301, sh. 1 Single Line Diagram

AP 12E-3301, sh. 2 Single Line Diagram

AO 12E-3301, sh. 3 Single Line Diagram

AN 12E-3302A Station Key Diagram 4160V and 480V Switchgears 480V MCCs V 12E-3303 Key Diagram 4160V Switchgears 31, 32, 33 & 34 S 12E-3303, sh.1 Key Diagram 4160V Switchgear 33 and 34 D 12E-3304 Key Diagram 4160V Switchgears 33-1 &

34-1 V 12E-3306 Key Diagram Reactor Building 480V Switchgears 38 & 39 Y 12E-3311 Key Diagram 480 V

MCC 38-2 and 39-2

AU 12E-3320 Key Diagram Reactor Building 480V

MCC 38-4, 38-7 & 39-7

AI 12E-3321 Key Diagram Unit 3 250

VDC MCC No. 3

AG 12E-3321A 250V DC Non-Essential MCC 3-8350-1A B 12E-3325 120 and120/240V AC Distribution Essential Service Bus,Sh.2

AH 12E-3325, sh. 1 Key Diagram 120 VAC & 120/240 VAC Distribution Essential, Instrument & Reactor Protection Buses AD

Attachment DRAWINGS Number Description or Title Revision 12E-3325, sh. 2 Key Diagram 120 VAC & 120/240 VAC Distribution Essential, Instrument & Reactor Protection Buses

AH 12E-3325, sh. 3 Key Diagram 120 VAC & 120/240 VAC Distribution Essential, Instrument & Reactor Protection Buses

AC 12E-3336 Relay, Metering and Excitation Diagram Standby Diesel Generator 3 U 12E-3343 Schematic Diagram 4160V Bus 34 Main & Reserve Feed G.C.B.'S

AA 12E-3344, sh. 3 Schematic Diagram 4160V Bus 34 Bus 34-1 Main Feed Breaker V 12E-3344, sh. 4 Schematic Diagram 4160V Bus 34-1 Main Feed Breaker W 12E-3346, sh. 2 Schematic Diagram 4160V Bus 34-1 Standby Diesel 3 Feed & 24-1 Tie Breaker

AR 12E-3346, sh.1 Schematic Diagram 4160V Bus 34-1 Standby Diesel 3 Feed & 24-1 Tie Breaker

AT 12E-3350A, sh. 1 Schematic Diagram Standby Diesel Generator 3 Engine Control and Generator Excitation

AN 12E-3350A, sh. 2 Schematic Diagram Standby Diesel Generator 3 Engine Control and Generator Excitation

AJ 12E-3374 Schematic Diagram 480V Miscellaneous Auxiliaries Part 2

AB 12E-3384 Schematic Diagram Isolation Condenser Valve 4399-74 A 12E-3389D Schematic Diagram 250V DC Battery Charger 3 B 12E-3389F Schematic Diagram 125V DC Battery Charger 3 A 12E-3461 Schematic Diagram Auto Blowdown Target Rock Valve 203-3A

AY 12E-3462 Schematic Diagram Auto Blowdown Electromatic Relief Valves 203-3C, 203-3D & 203-3E

AG 12E-3462, sh. 2 Schematic Diagram Auto Blowdown Part 2

AE 12E-3462A Schematic Diagram Auto Blowdown Electromatic Relief Valve G 12E-3484 Schematic Diagram Isolation Condenser Valve 4399-74 S 12E-3507 Schematic Diagram Isolation Condenser Valve 1301-3

AE 12E-3527, sh. 3 HPCI Sensors and Auxiliary Relays

AP 12E-3527A HPCI Valve and Turbine Auxiliary Relays H 12E-3529 Schematic Diagram HPCI Valves 2301-03, -08

AK 12E-3530 Schematic Diagram HPCI Valve 2301-48 X 12E-3644 Wiring Diagram Standby DG 3 Engine Equipment Control Panel

V 12E-3662D Wiring and Schematic Diagram 480V AC Reactor Building

MCC 39-7

AI 12E-3662E Wiring and Schematic Diagram 480V AC Reactor Building

MCC 39-7 F 12E-3684A Reactor Building 250

VDC MCC 3A, Part 1 N 12E-3684D Reactor Building 250

VDC MCC 3B, Part 1 N 12E-3684H Reactor Building 250

VDC MCC 3A, Part 4 H 12E-3904D Cable Tabulation Cables 33750 to 33799 W 12E-3904F Cable Tabulation Cables 33850 to 33899 W 12E-3904G Cable Tabulation Cables 33900 to 33949 Y 12E-3904L Cable Tabulation Cables 34100 to 34149 V 12E-3906G Cable Tabulation Cables 36200 to 36249

AB 12E-6320 Wiring and Schematic Iso-Condenser Makeup Water System B 12E-6321 Schematic & Wiring Diagram ISO Condenser Makeup Water System ISCO Building

E 12E-6321A Schematic Diagram Diesel Driven Make-up Pumps 2/3-43122A and 2/3-43122B Engine Control Panels 2223-126A & B D

Attachment DRAWINGS Number Description or Title Revision 12E-6322 Wiring and Schematic Iso-Condenser Makeup Water System F 12E-6324 Wiring and Schematic Iso-Condenser Makeup Water System B 12E-6762A Key Diagram Makeup Demin Sys. 480 V MCC 1 D 66-2-5636C1 Setting Plan Isolation Condenser 5 D-6760-4 15000 Gal Diesel Fuel Oil Storage Tanks 4

ISI-551 System Pressure Test Walkdown Iso. Drywell Second Floor EL. 537'-11/4"

G M-29, sh. 1 Diagram of L.P. Coolant Injection Piping CH M-32 Diagram of Shutdown Reactor Cooling Piping BC M-35 Diagram Of Demineralized Water System Piping DZ M-345, sh. 1 Diagram of Main Steam Piping BB M-345, sh. 2 Diagram of Main Steam Piping QH M-347 Control and Instrumentation Diagram of Reactor A M-354

Diagram of HPCI

CT M-355 Diagram of Service Water Piping SC M-357, sh. 1 Diagram of Nuclear Bolier

BV M-357, sh. 2 Diagram of Nuclear Bolier

BQ M-357, sh. 3 Diagram of Nuclear Bolier

E M-359 Diagram of Isolation Condenser Piping BM M-3639, sh.2 Battery Room Ventilation C M-4203 Flow Diagram Isolation Condenser Make Up System E M-4204 Flow Diagram Miscellaneous System

E

MISCELLANEOUS

Number Description or Title Date or Revision

Correspondence NRC to Commonwealth Edison Re. DC TSs 09/18/95

Correspondence between EATON and Exelon on DC Rating 02/18/13

GNB Industrial Power to Exelon Re. Battery Maintenance 02/08/13

Maintenance Rule Expert Panel Minute Notes 08/22/12

Station Response to IE Bulleting No. 76-01 03/17/76

Isolation Condenser Makeup Pumps' Engine Oil Analysis 09/07/12

Pump Manual, "Installation, Operation, and Maintenance Instructions for Goulds Pumps Model 3410" 1987

Technical Brochure, "2AM32-P Gasoline Engine Driven Self-Priming Petroleum Pump" 2007

ComEd Response Letters to NRC Regarding IE Bulletin 88-04, Safety-related Pump Loss 07/11/88, 02/07/89, 01/08/90

Lube Oil Oil Analysis Report, Unit ID 3-2302 01/14/13 0000-0088-8669-R0 BWR Owners' Group Technical Report Effects of Voiding in ECCS Drywell Injection Piping 09/08 2004-0036 UFSAR Change 04003 and Tech Spec Bases Change for SR 3.5.3.1 0 257HA353AB GE Design Spec , HPCI System 3 8300c-f-250

VDC System Health Report 250

VDC 3Q-2012 ACMP

1287893 Unit 3 HPCI Discharge Piping Void Monitoring 0

Attachment MISCELLANEOUS

Number Description or Title Date or Revision

BWROG-TP-12-025

LPCI-Shutdown Cooling Alignment Committee (377); Position Paper-Guidance for Alignment and Venting of RHR Loop for Post-LOCA LPCI Mode Operation 0

EC 338243 Modification of Opening Logic for HPCI Valve 3-2301-8 by Adding Pressure Switch 0

EC 363437 Unit 3 Isolation Condenser Heat Removal Capability with Tube Plugging 1

EC 366367 Iso Condenser Makeup Pump Capacity Test Evaluation 06/28/07

EC 382128 Isolation Condenser Eddy Current Testing Tube Plug 0

EC 386623 Evaluation of HPCI

2-2301-8 Seat Leakage Test Results 0

EC 387177 Acceptance Criteria for HPCI Discharge Piping Due to Back Leakage at 2301-7 and 2301-8 Valves 0

EC 389067 Evaluate LPCI Piping Post Accident While Operating in SDC Mode 0 ER2002-9984 Unit 3 Isolation Condenser Elevated Temperatures 02/12/02

GEK 786 Equipment Manual Ch. 22 Isolation Condenser

05/72

GEK-15545 HPCI Pump/Turbine Manual 0

GE-NE-A22-00103-10-01 Dresden and QC Extended Power Uprate (Task T0900): Transient Analysis 0

GE-NE-A22-00103-74-01 Dresden and QC Extended Power Uprate (Task T0317): Isolation Condenser

0

NX-7831-437-1 Vendor Manual, Target Rock Safety Relief Valve, Model 67F 10/15/76

OPL-3 Parameters for Dresden Unit 3 Cycle 19 Transient Analysis 06/14/04

RS-03-140 Correspondence between Exelon and NRC on DC System Technical Specification Changes 07/29/03

SEC-DR-99-077 250

VDC System Load Flow, Volts Drop & Cable Ampacity 04/22/99 T0300 GE EPU Task Report: Nuclear Boiler 0 T0400 GE EPU Task Report: Containment System Response 1 T0404 GE EPU Task Report: HPCI 0 T0611 GE EPU Task Report: Appendix R Fire Protection 1 T0900 GE EPU Task Report: Transient Analysis 0 T0902 GE EPU Task Report: ATWS 0 T0903 GE EPU Task Report: Station Blackout 0

TR 1001257 EPRI Capacitor Performance Monitoring Project 12/2000 TR021DR LOCA Analysis for

SVEA-96 Optima 2 Fuel at Dresden Units 2 and 3

TR051DR ATWS Analysis for the Introduction of

SVEA-96 Optima 2 Fuel at Dresden Units 2 and 3 0

VP-025115-4 Eaton/C-H TOL Selection Verification Plan 03/14/03 VTIP D2021 Isolation Condenser Makeup Pumps Vendor Manual 06/17/92 VTM D1315 GNB Classic Batteries Installation and Operating Instructions 02/2004 VTM D1622, V.1 125

VDC Battery Charger 01/17/13 VTM D1622, V.2 250

VDC Battery Charger 12/05/12

Attachment MODIFICATIONS

Number Description or Title Date or Revision E12-3-97-201 Gland Seal Leak-Off Upgrade 03/11/97 E12-3-98-206 Upgrade HPCI Exhaust Drain Pot and Gland Seal Lines from Non-Safety to Safety-Related 01/08/99

EC 366256 Add Interposing Relay for Unit 3 Diesel Generator Output Breaker

000

EC 366676 Add Interposing Relay for Diesel Generator Output Breaker Unit 2 000

OPERABILITY EVALUATIONS

Number Description or Title Date 00-052 Isolation Condenser Primary Containment Isolation Valve 3-1301-3 1 07-005 EDG 4KV Breaker Closing Coil Minimum Pick Up Voltage Acceptance 11/20/07 13-004 HPCI MOV's 2(3)- 2301-3 May Not Be Able to Open Against ATWS High Steam Line Pressure 03/15/13 13-003 2(3)-0203-3A Target Rock Safety Relief Valves 02/27/13

PROCEDURES

Number Description or Title Revision

CC-AA-304 Component Classification 5 Dan 902(3)-3, D-4 Isolation Condenser Level Hi/Lo Annunciator Response 11 DEOP 0100-00 RPV Control 10 DEOP 0400-05 Failure To Scram 16

DES 8300-01 Maintenance of Cutler Hammer DC Contactors 10

DES 8300-03 Maintenance of Cutler Hammer DC Motor Starters 10

DES 8300-07 Unit 2(3) Weekly Station Battery Inspection 10

DES 8300-15 Unit 3 250 V Battery Service Test 17, 20

DES 8300-16 Unit 2(3) Battery Monthly Surveillance 16

DES 8300-17 Unit 2(3) Battery Quarterly Surveillance 16

DES 8300-19 Unit 3 250 V Battery Modified Performance Test 16

DES 8300-20 Unit 3 125 V Battery Modified Performance Test 11

DES 8300-24 Unit 3 125 V Battery Service Test 15

DGA-12 Partial or Complete Loss of AC Power 70

DOA 5750-01 Ventilation System Failure 58

DOA 6600-01 Diesel Generator Failure 16

DOP 1200-13 RWCU System Operation With Reactor at Power 61

DOP 1300-01 Standby Operation of the Isolation Condenser System 50

DOP 1300-02 Automatic Operation of Isolation Condenser 24

DOP 1300-03 Manual Operation Of The Isolation Condenser 33

DOP 1300-07 Filling Isolation Condenser Makeup Pumps Fuel Oil Day Tanks Using Diesel Fuel Oil Transfer System 7

DOP 1300-08 Filling Isolation Condenser Makeup Pumps Fuel Oil Day Tanks With Diesel Fuel Oil Transfer System Unavailable 7

DOP 1300-09 Isolation Condenser Makeup Pump Local Operation 05

DOP 2300-03 HPCI System Manual Startup and Operation 39

Attachment PROCEDURES

Number Description or Title Revision

DOP 6600-10 Filling Isolation Diesel Generator Day Tanks with Diesel Fuel Oil Storage Tank Transfer Pumps Unavailable 09

DOP 6900-01 250

VDC Electrical System 35

DOP 8300-01 Station 125/250

VDC Battery Equalizing Charge 06

DOP 8300-02 Station 250

VDC Battery Equalizing Charge 02

DOS 0010-34 Petro-Guard Pump Testing 02

DOS 1300-01 Isolation Condenser Five Year Heat Removal Capability Test 39

DOS 1300-03 Isolation Condenser Makeup Pump Quarterly Operability 18

DOS 1300-05 2/3A(B) Isolation Condenser Makeup Pump Capacity Test 7

DOS 6600-01 Diesel Generator Surveillance Tests 122

DOS 6600-12 Diesel Generator Tests Endurance and Margin/Full Load Rejection /ECCS/Hot Start 57

DOS 7000-13 LLRT Limit Bases for 3-1301-3 4

DOS 8300-07 Weekly Station Battery Inspection

11

DTS 6900-01 Annual Change of Pilot Cells for Station Batteries 09

ER-AA-302-1009 Final JOG MOV Periodic Verification Program Implementation 1

MM-AA-723-325 Molded Case Circuit Breaker Testing 11

MM-DR-8300-1001 Battery Systems Supplemental Information 01

SURVEILLANCES (COMPLETED) Number Description or Title Date

DOS 7000-08 LLRT Performed on 3-1301-3 11/12/06 11/03/10

DOS 2300-10 HPCI High/Low Pressure Operability Test and HPCI Quarterly IST Comprehensive IST

03/25/11 12/22/12

DOS 7100-02 Leak Rate Test of Target Rock Pneumatic System 11/18/12

DOS 1600-05 Quarterly Valve Stroke Timing (IST) 09/11/12

WORK DOCUMENTS

Number Description or Title Date

SR 00054036 Remove TOL Trip Checks from PMID in Hard Wired Buckets 02/14/13

SR 00054037 Generate New PM for Outage Trip Check of Hard Wired

TOL 02/14/13

WO 00506290 D3 3RFL PM Surveillance Limitorque Operator 3-1301-3 11/24/07

WO 00514350 MOV Diagnostic Testing 3-2301-8 11/16/08

WO 00563285 6Y

PM 250 VDC Bkr 3-1301-3 05/09/11

WO 00748439 6Y

PM 250 VDC Bkr 3-2301-8 03/21/11

WO 00748441 6Y

PM 250 VDC Bkr 3-2301-48 03/21/11

WO 00748443 6Y

PM 250 VDC Bkr 3-2301-3 03/21/11

WO 00801264 6Y

PM 250 VDC Bkr 3-4399-74 05/09/11

WO 00880955 Disassemble and Inspect Check Valve 3-2301-7 11/11/08

WO 00916413 Inspect/Calibrate HPCI Room Temperature Controller 12/07/09

WO 00977704 D3 250 V Battery Modified Performance Test 10/21/10

WO 00997320 OP D3 5Y TS Isolation Condenser Heat Removal Test 11/28/10

Attachment WORK DOCUMENTS

Number Description or Title Date

WO 01078509 D2 250 V Battery Modified Performance Test 10/29/11

WO 01096688 IST Replace Unit 3 HPCI Turbine Exhaust Rupture Disks 03/19/12

WO 01198950 PMID 5051-01, Inspect HPCI Oil Filters/Clean Elements 03/23/11

WO 01199963 Open/Clean/Inspect/Eddy Current Test HPCI Room Cooler 03/21/11

WO 01213514 HPCI 3 Steam Line Hi Flow Isolation Channel Calibration 07/29/11

WO 01221226 HPCI 3 Turbine Pressure Switch Calibration 03/24/11

WO 01246496 D 2/3 2Y Com 'A' Iso Cond Make-Up Pump Capacity Test 06/26/11

WO 01246499 D 2/3 2Y Com 'B' Iso Cond Make-Up Pump Capacity Test 06/26/11

WO 01252748

Battery Jumper 02/22/11

WO 01370227 RFL PM TS Cal Testing of RPS Bus/3A-1 and 3A-2 EPA's 11/18/12

WO 01377151 10 YR PM on 250 V Battery Charger 3 01/10/13

WO 01388735 D3 Refueling

TS 250

VDC Surveillance (Service Test) 11/08/12

WO 01424681 D2/3 AN PM Petro-Guard Pump Functional Test 02/14/12

WO 01506646 4 YR

PM Inspect 4 KV Breaker 10/26/12

WO 01546365 D3 QTR TS Valve Timing (IST) 07/10/12

WO 01551437 QTR TS HPCI MOV Operability Surveillance 09/19/12

WO 01555419 D 2/3 Qtr Com 'B' Iso Cond Make-Up Pump Operability 09/29/12

WO 01555420 D 2/3 Qtr Com 'A' Iso Cond Make-Up Pump Operability

09/29/12

WO 01559257 Quarterly D3

TS 250

VDC Surveillance 10/11/12

WO 01577945 Quarterly D2

TS 250

VDC Surveillance 12/26/12

WO 01582913 Quarterly D3

TS 250

VDC Surveillance 01/10/13

WO 01589982 Monthly D3

TS 250

VDC Battery Surveillance 12/06/12

WO 01598017 Monthly D3

TS 250

VDC Battery Surveillance 01/10/13

WO 01603293 Weekly D3

TS 250

VDC Surveillance 01/03/13

WO 01606590 Weekly D2

TS 250

VDC Surveillance 01/16/13

WO 01608860 Weekly D3

TS 250

VDC Surveillance 01/23/13

WO 01616331 Weekly D2

TS 125

VDC Battery Surveillance 02/18/13

WO 01616332 Weekly D3

TS 250

VDC Battery Surveillance 02/18/13

WO 01616334 Weekly D3

TS 250

VDC Battery Surveillance 02/18/13

WO 01616337 Weekly D3

TS 250

VDC Battery Surveillance 02/18/13

WO 01616338 Weekly D3

TS 125

VDC Battery Surveillance 02/18/13

WO 01616339 Weekly D3

TS 125

VDC Battery Surveillance 02/18/13

WO 01617845 Operational Test of the Petro-Guard Portable Fuel Oil Transfer Pump 04/02/13

WO 99065371 EM D3 6Y PM Surv Limitorque VLV Oper 3-4399-74 05/05/03

WO 99065808 MM D2/3 5Y PM Clean B Isol Cond M/U 02/18/02

WO 99156455 MOV Diagnostic Testing 3-2301-48 12/19/07

WO 99269739 6 Year Lube/Inspection of Limitorque Valve Operator 3-2301-48 12/19/07

Attachment

LIST OF ACRONYMS

USED °C Celsius Degrees °F Fahrenheit Degrees

ADAMS Agencywide Document Access Management System

ADS Automatic Depressurization System

AR Action Request

ASME American Society of Mechanical Engineers

ATWS Anticipated Transient without Scram

BWR Boiling Water Reactor

BWROG Boiling Water Reactor Owner's Group

CDBI Component Design Bases Inspection

CNO Chief Nuclear Officer

CFR Code of Federal Regulations d/p Differential Pressure

DRS Division of Reactor Safety

EC Engineering Change

ECCS Emergency Core Cooling System

EDG Emergency Diesel Generator

EPU Extended Power Uprate

FIN Finding

GE General Electric

GL Generic Letter

HPCI High Pressure Coolant Injection

IC Isolation Condenser

IEEE Institute of Electrical & Electronic Engineers

IMC Inspection Manual Chapter

IN Information Notice

IR Inspection Report

IST Inservice Testing

LLC Limited Liability Company

LOCA Loss of Coolant Accident

LPCI Low Pressure Coolant Injection

MBTU /hr Million British Thermal Units/hour

MCC Motor Control Center

MOV Motor-Operated Valve

MW t Megawatt Thermal

NCV Non-Cited Violation

NPSH Net Positive Suction Head

NRC [[]]

U.S. Nuclear Regulatory Commission

PARS Publicly Available Records System

PM Preventative Maintenance

RHR Residual Heat Removal

RIS Regulatory Issue Summary

RPS Reactor Protection System

RPV Reactor Pressure Vessel

RTP Reactor Thermal Power SDC Shutdown Cooling

Attachment

SDP Significance Determination Process

SR Surveillance Requirement

SRV Safety Relief Valve

TS Technical Specification U Unit

UFSAR Updated Final Safety Analysis Report

URI Unresolved Item Vac Volts Alternating Current Vdc Volts Direct Current WO Work Order

M. Pacilio -2- In accordance with 10

CFR 2.390 of the

NRC 's "Rules of Practice," a copy of this letter and its enclosure, and your response (if any) will be available electronically for public inspection in the

NRC Public Document Room or from the Publicly Available Records System (PARS) component of

NRC 's Agencywide Documents Access and Management System (

ADAMS),

accessible from the

NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). Sincerely, /

RA / Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-237; 50-249 License Nos.

DPR -19;

DPR -25 Enclosure: Inspection Report 05000237/2013007; 05000249/2013007 w/Attachment: Supplemental Information cc w/encl: Distribution via ListServŽ