Text

IR 05000454-12-007, 05000455-12-007(DRS), on 05/14/2012 - 6/15/2012, Byron Station, Units 1 and 2, Component Design Bases Inspection (CDBI)
	ML12212A309
Person / Time
Site:	Byron
Issue date:	07/27/2012
From:	Ann Marie Stone; NRC/RGN-III/DRS/EB2
To:	Pacilio M; Exelon Generation Co, Exelon Nuclear
References
	IR-12-007
	Download: ML12212A309 (37)
	v • d • e

July 27, 2012

SUBJECT:

BYRON STATION, UNITS 1 AND 2 COMPONENT DESIGN BASES INSPECTION 05000454/2012007; 05000455/2012007(DRS)

Dear Mr. Pacilio:

On June 15, 2012, the U.S. Nuclear Regulatory Commission, (NRC) completed a Component Design Bases Inspection, (CDBI) at your Byron Station, Units 1 and 2. The enclosed report documents the results of this inspection, which were discussed on June 15, 2012, with Mr. T. Tulon, and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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, five NRC-identified findings of very low safety significance were identified. The findings involved a violation 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 If you contest the subject or severity of this NCV, 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 Inspector Office at the Byron Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at Byron Station. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, 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 document system (ADAMS). ADAMS is accessible from the NRC Website 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. 50-454; 50-455 License Nos. NPF-37; NPF-66

Enclosure:

Inspection Report 05000454/2012007; 05000455/2012007(DRS)

w/Attachment: Supplemental Information

REGION III==

Docket Nos: 05000454; 05000455 License Nos: NPF-37; NPF-66 Report No: 05000454/2012007; 05000455/2012007(DRS)

Licensee: Exelon Generation Company, LLC Facility: Byron Station, Units 1 and 2 Location: Byron, IL Dates: May 14, 2012, through June 15, 2012 Inspectors: A. Dunlop, Senior Engineering Inspector, Lead C. Brown, Engineering Inspector, Electrical E. Sanchez Santiago, Engineering Inspector, Mechanical R. Langstaff, Senior Engineering Inspector, Operations J. Chiloyan, Electrical Contractor C. Baron, Mechanical Contractor Approved by: Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS IR 05000454/2012007; 05000455/2012007(DRS); 05/14/2012 - 06/15/2012; Byron Station, Units 1 and 2; 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. Five Green findings were identified by the inspectors. The findings were considered Non-Cited Violations (NCV) of NRC regulations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC)

0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs 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 Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure qualified components were installed in the plant. Specifically, purchase orders did not specify the minimum pickup voltage for NEMA Size 1 through Size 4 safety-related motor-control contactors such that the installed contactors were not rated to function at the design basis minimum voltage. The licensee entered the issue into their corrective action program and based on a sample testing of contactors demonstrated there was adequate margin between the highest found minimum-pickup voltage and the design basis pickup voltage.

The performance deficiency was determined to be more than minor because it was associated with the Mitigating System Cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, having installed contactors that may not function under degraded voltage conditions could affect the operability of multiple safety-related structures, systems and components during an event. The finding screened as of very low safety significance (Green) because the finding involved a design or qualification deficiency that did not result in a loss of operability. 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 Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure the component cooling water (CCW) system was capable of withstanding a reactor coolant pump thermal barrier break. Specifically, when assuming a single failure of the automatic isolation function, the licensee failed to evaluate the break effect on the CCW system during the 3 minutes postulated to isolate the leak. The licensee entered the issue into their corrective action program; verified the CCW system would be able to withstand the postulated event, and planned to perform a detailed evaluation of the effect of a thermal barrier break on the CCW system.

1 Enclosure

The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Design Control and objective of ensuring the capability of the system to respond to an initiating event to prevent undesirable consequences. Specifically, the failure to evaluate the effect of the thermal barrier rupture on the CCW system created reasonable doubt whether the system would be capable of withstanding the applied forces of this event. The finding screened as very low safety significance (Green) because the design deficiency 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 and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to specify in a design calculation the allowable relay setpoint calibration tolerances. Specifically, the acceptance criteria used in relay setting calibration procedures was not bounded by the relay setting design calculations. The licensee entered this finding into their corrective action program and verified the calibrated relay settings would still provide adequate electrical protection coordination capability.

The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to adequately evaluate the design requirements of the relay settings could have resulted in a loss-of-relay coordination and could allow a fault on one piece of equipment to propagate to other safety-related equipment outside the designed isolation boundary. The finding screened as very low safety significance (Green) because the finding was design deficiency confirmed not to 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 Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to adequately analyze potential design basis internal flooding events in the auxiliary building. Specifically, the licensees analysis did not account for the possible single failure of an essential service water motor-operated isolation valve or its associated power supply, which would have prevented break isolation within 30 minutes.

The licensee entered the issue into their corrective action program; verified essential service water piping in the auxiliary building would meet the crack exclusion pipe stress criteria, and planned to the revise the flooding analysis.

The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Design Control and objective of ensuring the capability of the system to respond to an initiating event to prevent undesirable consequences. Specifically, the failure to adequately analyze potential design basis internal flooding events in the auxiliary building would affect the capability of safety-related equipment to withstand the postulated event. The finding screened as very low safety significance (Green) because the design deficiency did not result in a loss of operability or functionality. The inspectors did not identify a 2 Enclosure

cross-cutting aspect associated with this finding because the finding was not representative of current performance. (Section 1R21.4.b.(1))

Cornerstone: Barrier Integrity

Green: The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to provide a means to detect and isolate a leak in the emergency core cooling flow path within 30 minutes, which was contrary to the Updated Final Safety Analysis Report. Specifically, the licensee failed to provide a means to detect and isolate a leak within 30 minutes in that neither sump alarms nor radiation monitors were provided for the safety injection pump rooms. The licensee entered the issue into their corrective action program and planned to evaluate options for modifications to address detection of emergency core cooling system leakage.

The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity cornerstone attribute of Design Control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.

Specifically, the failure to provide a means to detect and isolate a leak in the emergency core cooling flow path within 30 minutes could result in a delayed isolation of such a leak after an accident and result in a greater radionuclide release to the auxiliary building and the environment. The finding screened as very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. (Section 1R21.6.b.(1))

B. Licensee-Identified Violations No violations were identified.

3 Enclosure

1. REACTOR SAFETY Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity 1R21 Component Design Bases Inspection (71111.21)

.1 Introduction The objective of the component design bases inspection is to verify that design bases have been correctly implemented for the selected risk significant components and that 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 (PRA) 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 licensees PRA and the Byron Station Standardized Plant Analysis Risk-Model to identify a scenario to use as the basis for component selection. The scenario selected was a reactor coolant pump loss of seal cooling 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 inspector 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 24 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 18 components were reviewed:

Residual Heat Removal (RHR) Pump (1RH01PA): The inspectors reviewed design analyses associated with RHR pump capacity, net positive suction head (NPSH), and minimum flow to verify the equipments capacity to perform its required functions. The pump test procedures and recent results were reviewed to verify the actual capability of the installed equipment. The inspectors reviewed industry experience issues associated with the pump thrust bearing loading to verify appropriate maintenance intervals were being maintained. The inspectors reviewed industry experience associated with operation of the RHR pump under recirculation conditions for extended periods of time to verify this operation would not result in unacceptable RHR system pressures and temperatures. The potential susceptibility of the pump to internal flooding events was reviewed to verify the capability of the pump to perform its required function. The inspectors reviewed a sample of operating procedures associated with the pump under normal and accident conditions.

5 Enclosure

RHR Pump Recirculation Valve (1RH611): The inspectors reviewed motor-operated valve (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 inservice testing (IST) surveillance results, including stroke time and available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified. The setpoint calibration was reviewed to ensure the valve would open to provide a minimum flow path for the RHR pump and close to ensure design flow requirements were met. Short-circuit calculations, breaker interrupting ratings, thermal overload (TOL) sizing calculations, TOL rating, voltage drop calculations, MOV motor terminal voltage requirements, MOV control circuit requirements and logic were reviewed to ensure the valve would function as required.

Essential Service Water (SX) Pump (1SX01PA): The inspectors reviewed design analyses associated with SX pump capacity and NPSH to verify the equipments capacity to perform its required functions. The inspectors also reviewed test procedures and recent results to verify the actual capability of the installed pump.

The inspectors reviewed the performance changes associated with replacement of pump rotating elements, including the potential effects on the pump motors and electrical distribution system. The potential susceptibility of the SX pump to internal flooding events was reviewed to verify the capability of the pump to perform its required function after any design basis event. The inspectors reviewed a sample of operating procedures associated with the pump under normal and accident conditions. The inspectors also reviewed the motor feeder circuit sizing, to ensure adequacy of ampacity, short circuit current capability, and voltage requirements at the motor terminals under the most limiting conditions.

Electrical calculations were also reviewed to ensure the adequacy of the motor feeder circuit phase and ground protective device trip settings.

SX Strainer (1SX01FA): The inspectors reviewed the capability of the SX strainer to perform its required functions. The inspectors reviewed inspection procedures and recent results to verify the actual capability of the installed strainer. Operating procedures associated with the strainer under normal and accident conditions were reviewed, including the capability of the operators to clean the strainer without normal electrical power available.

SX Unit Cross-tie Valve (1SX005): The inspectors reviewed calculations, operations history, and design requirements to verify the equipments capacity to perform its required functions. The inspectors reviewed the design differential pressure for this valve to verify its capability to operate under the most limiting conditions. The inspectors also reviewed MOV test procedures and recent results to verify the actual capability of the installed equipment. Short-circuit calculations, breaker interrupting ratings, TOL sizing calculations, TOL rating, voltage drop calculations, MOV motor terminal voltage requirements, MOV control circuit requirements and logic were reviewed to ensure the valve would function as required.

6 Enclosure

Centrifugal Charging (CV) Pump (1CV01PA): The inspectors reviewed design analyses associated with CV pump capacity, NPSH, and minimum flow to verify the equipments capacity to perform its required functions. The inspectors also reviewed test procedures and recent results to verify the actual capability of the installed pump. The weak pump/strong pump analysis was reviewed to verify that all operating pumps would have adequate minimum flow under all conditions. The inspectors reviewed the performance changes associated with replacement of pump rotating elements. The inspectors reviewed operating procedures associated with the pump under normal and accident conditions.

The inspectors also reviewed the motor feeder circuit sizing to ensure adequacy of ampacity, short circuit current capability, and voltage requirements at the motor terminals under the most limiting conditions. Electrical calculations were reviewed to ensure the adequacy of the motor feeder circuit phase and ground protective device trip settings.

CV Reactor Coolant Pump Seal Injection Valve (1CV8355A): 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. Breaker interrupting ratings, short-circuit calculations, TOL sizing calculations, voltage drop calculations, TOL rating, MOV motor terminal voltage requirements, MOV control circuit requirements and logic were reviewed to ensure the valve would function as required.

CV Pump Suction Check Valve (1CV8546): The inspectors reviewed maintenance history, operations history, and design requirements to verify the check valves capacity to perform its required functions. The inspectors also reviewed test procedures and recent results to verify the actual capability of the installed check valve. The inspectors reviewed an analyses associated with a material change of a valve component.

Component Cooling Water (CCW) Pump (1CC01PA): The inspectors reviewed design analyses associated with the CCW pump capacity, acceptance criteria, and seismic evaluations to verify the equipments capacity to perform its required functions. The inspectors reviewed operating procedures associated with the pump under normal and accident conditions. The inspectors reviewed surveillance results including IST quarterly pump testing, flow verification, and performance testing to verify acceptance criteria were met and performance degradation could be identified. The inspectors also reviewed the motor feeder circuit sizing, to ensure adequacy of ampacity, short-circuit current capability, and voltage requirements at the motor terminals under the most limiting conditions.

Electrical calculations were also reviewed to ensure the adequacy of the motor feeder circuit phase and ground protective device trip settings.

7 Enclosure

Reactor Coolant Pump Cooling from CCW Valve (1CC9413A): 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. Breaker interrupting ratings, short-circuit calculations, TOL sizing calculations, voltage drop calculations, TOL rating, MOV motor terminal voltage requirements, MOV control circuit requirements and logic were reviewed to ensure the valve would function as required.

CCW Heat Exchanger (1CC01A): The inspectors reviewed the design and licensing basis of the heat exchanger to ascertain compliance with system operation requirements. The inspectors reviewed calculations to determine fouling criteria to ensure the heat exchanger was meeting its design requirements and the results of the calculations were properly translated into the procedure acceptance criteria. The inspectors also reviewed test procedure for appropriate acceptance criteria; including the testing and inspection results to verify compliance with heat exchanger program requirements.

CCW Surge Tank (1CC01T): The inspectors reviewed design analyses associated with the surge tank capability to perform its required functions. The inspectors also reviewed the capacity and level set points for the tank to ensure compliance with design requirements. Internal/external inspection results were reviewed to ensure the integrity of the tank was being monitored and maintained.

The inspectors reviewed minimum and maximum temperature limits as well as refill capability and overpressure calculations to ensure the tank was capable of performing its intended safety function under different operating scenarios.

CCW Heat Exchanger Outlet Header Crosstie Isolation Valve (1CC9467B): The inspectors reviewed the design and licensing basis of the valve to ascertain compliance with system design function. The inspectors reviewed testing results to ensure compliance with the IST program. Operating procedures were reviewed to verify capability to operate the valve under various scenarios.

Unit Auxiliary Transformer (UAT) 141-1: The inspectors reviewed transformer windings configuration, circuit breaker control schematics, protective relay settings, and surveillance test records to assess the status and maintenance condition of the transformer. The inspectors reviewed the electrical distribution system calculations and performed independent calculations using the transformer nameplate data to determine the adequacy of the transformer to supply required power to the associated 4160 Vac switchgear and to verify short circuit current interrupting duty requirements were within the switchgear breaker ratings. To determine whether the 6.9kV and 4.16kV neutral grounding resistors were adequately maintained, the inspectors reviewed the completed preventive maintenance test procedures to verify that the measured resistance values of the grounding resistors remained within manufacturers tolerances.

8 Enclosure

Emergency Diesel Generator (EDG) (1DG01KA): The inspectors reviewed the load voltage drop calculation, maximum and minimum voltage profile, and DC field flashing circuit design. The inspectors also verified that the EDGs start properly under degraded voltage conditions. Surveillance test results were reviewed to ensure TS requirements were met. The inspectors reviewed the adequacy and appropriateness of design assumptions and calculations related to EDG protection and relay coordination during test mode and during emergency operation. The EDG output breaker control logic diagrams were reviewed to verify the breaker tripping and closing logic was consistent with design basis description and interlocking requirements. The inspectors reviewed the adequacy of the EDGs high resistance neutral grounding equipment and whether appropriate periodic maintenance and measurements were performed to ensure the design basis of the ground fault detection was maintained.

The inspectors performed independent calculations of available fault current contributions from the EDG, UAT, and SAT (system auxiliary transformer) for postulated phase and ground faults on safety Bus 141 and compared them with the relay setting calculations to verify the appropriateness of the applied over current and voltage relay settings.

Switchgear Bus 141: The inspectors reviewed circuit breaker control schematics, protective relay settings, loss of voltage and degraded voltage relay settings, and electrical distribution system calculations to assess the status and maintenance condition of the equipment and to verify the adequacy of bus and circuit breaker load capacity and short circuit interrupting ratings for full loading and emergency loading. Operating and maintenance test procedures were reviewed to assess whether component operation and alignments were consistent with design and licensing bases assumptions. The switchgear protective relay testing procedures and recently completed calibration test results were reviewed to verify the acceptance criteria for tested parameters were supported by calculations, or other engineering documents. The inspectors performed independent calculations of available fault current contributions from the EDG, UAT, and SAT for postulated phase and ground faults and compared them with the phase and ground over-current relay setting calculations to verify the appropriateness of the applied over-current relay settings. The inspectors also reviewed the 4160 Vac Bus 141 loss of voltage and bus over-current relay settings to ensure adequate coordination was maintained between the bus over-current and the bus undervoltage relay settings to ensure the over-current relays function as designed during postulated electrical bus faults.

480 Vac ESF Bus 131X: The inspectors reviewed bus design capability, bus and load breaker protection settings, control circuits for adequate power, electrical separation/isolation, seismic qualifications, and degraded voltage effects on 480/120 Vac motor control relays to ensure conformance with applicable design standards.

9 Enclosure

120 Vac Instrument Bus 111: The inspectors reviewed bus design capability related to loading and short circuit protection, breaker ratings and settings to prevent spurious tripping and protection against low magnitude faults, breaker coordination, and electrical separation/isolation to ensure conformance with applicable design standards. Test procedures and associated results were reviewed to verify bus components were adequately tested and degradation would be identified. The inspectors also reviewed the design compliance with the single failure criterion.

b. Findings (1) Non-Conforming 480/120 Vac Motor Control Contactors Introduction: The inspectors identified a finding of very low safety significance (Green)

and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to ensure qualified components were installed in the plant. Specifically, purchase orders did not specify the minimum pickup voltage for NEMA Size 1 through size 4 safety-related motor-control contactors such that the installed contactors were not rated to function at the design basis minimum voltage.

Description: During the review of 480 Vac engineered safety features (ESF) Bus 131x, the inspectors requested information on periodic testing of the minimum pickup voltage for NEMA Size 1 and 2 Contactors used in the control circuits for safety-related motors.

The licensee responded that no periodic testing was performed as all of the size 1 through 4 contactors were purchased under a standard specification to have minimum pickup voltages of 70 percent of the rated voltage (84 Vac for 120 Vac rated size 1 and 2 contactors and 322 Vac for 460 Vac rated size 3 and 4 contactors). The 70 percent minimum pickup voltage had been specified in the original specification (F-2755, Proposed Technical Data for 480 Volt Motor Control Centers). During the inspection, the licensee discovered that the contactor supplier (Westinghouse) was not currently testing at the 70 percent pick-up voltage value and initiated Action Request (AR) 1368220. The AR documented that contactors purchased since 1999 had been tested at 85 percent of the rated voltage and that the population of affected contactors could include all of the 120 Vac contactors in the plant. Operability of the installed contactors was supported by the results of 2011 testing in which eight contactors tested had a highest pickup voltage of 78 Vac; 6 volts below the minimum of 84 Vac. This AR also stated that the 70 percent minimum pickup voltage was applicable to the size 3 and 4 contactors; however, no operability concern existed as they were powered from 480 Vac power with minimum voltage drops due to the physical location and short cable runs.

The licensee stated that all of the contactors had been purchased certified to the Westinghouse Environmental and Seismic Qualification reports 11210-CCR-1 and 11210SCR-1 associated with Westinghouse General Order INU-11210; however, the inspectors review revealed that the purchase orders did not include any minimum pickup voltage requirements. The licensee had been relying only on the certification to the qualification reports to maintain the original technical design and qualification basis for the equipment. As a result, the licensee had not verified the minimum pickup voltage for the contactors actually installed in the safety-related applications in the plant. The inspectors verified that Calculation 19-AQ-24, Voltage Drop on 480V-120V AC Control Transformer Circuits, assumed the 70 percent minimum pickup voltage and that adequate margin existed between the calculated voltages and the 84 volt minimum 10 Enclosure

pickup value. For the size 3 and 4 contactors, Appendix F to Calculation 19-AQ-63, 480V Switchgear and MCC Voltages, demonstrated that the minimum expected voltage at the most limiting 480V ESF bus was 88.4 percent or 424 Vac. Since the size 3 and 4 contactors have a published minimum pickup rating of 85 percent of 460 Vac (391 Vac), there was no operability concern; however, Specification F-2755 requirements did not match the actual purchased components.

On June 4, 2012, the licensee received information from Westinghouse that confirmed that none of the contactors had been tested to the 70 percent standard. In addition, the licensee tested an additional 35 NEMA Size 1 Contactors previously removed from the plant. The worst case minimum pickup voltage was 76 volts, which was less than the 70 percent design standard of 84 volts. However, on June 11, 2012, the licensee initiated AR1376793 to document that the 120 Vac contactors purchased before 1999 had only been tested to a 75 percent of rated voltage standard. As a result, all of the installed size 1 or 2 contactors were non-conforming to the design basis. The licensee generated this AR due to the inspectors concerns with the operability determination process. Specifically the engineering staff had received information that the installed contactors were non-conforming to the design, but had decided that it did not need to go back to shift management for another operability call since the information did not contradict the information provided in the original AR. The inspectors questioned the applicability of Procedure OP-AA-108-115, Operability Determinations (CM), in this situation. Specifically, Step 4.1.14, stated RE-EVALUATE, as necessary, SSC operability following a change in conditions or as additional information about the cause of the degradation/non-conformance becomes known. Shift management had not been informed that all of the installed contactors were non-conforming to design or of the results of the additional contactor testing. The on-duty shift manager agreed and the licensee initiated AR1368220. The shift manager concluded that there was reasonable assurance of operability, but also requested a full operability evaluation. The inspectors reviewed EC 0000389469, OP EVAL 12-006 - Westinghouse NEMA Size 1 and Size 2 Contactor Pick-Up Voltage Concerns, and had no concerns with the determination that the contactors were operable but non-conforming.

Analysis: The inspectors determined that the failure to specify the minimum pickup voltage in the purchase orders for NEMA Size 1 through 4 safety-related contactors was a performance deficiency. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the capability of the systems to respond to an initiating event to prevent undesirable consequences. Specifically, having installed contactors that may not function under minimum voltage conditions could affect the operability of multiple safety-related structures, systems and components during an 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 contactors tested were shown to operate below the 70 percent minimum pickup rating in the design calculation.

11 Enclosure

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 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, from initial construction until May 18, 2012, the design basis minimum pickup voltage was not specified in purchase order specifications and no testing had been performed to verify the minimum pickup voltage for the installed safety-related motor-control contactors. Because this violation was of very low safety significance and was entered into the licensees corrective action program as AR01368220 and AR01376793, this violation is being treated as a Non-Cited Violation (NCV), consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 5000254/2011009-01; 5000265/2011009-01, Non-Conforming 480/120 Vac Motor Control Contactors).

(2) Failure to Verify the CCW System Capability to Withstand a Thermal Barrier Break:

Introduction: The inspectors identified a finding of very low safety significance (Green)

and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the CCW system was capable of withstanding a reactor coolant pump (RCP) thermal barrier break. Specifically, when assuming a single failure of the automatic isolation function, the licensee failed to evaluate the break effect on the CCW system during the 3 minutes postulated to isolate the leak.

Description: Section 9.2.2.4.4, Shared Function, of the UFSAR stated that the RCP thermal barrier outlet header had a flow indicating switch to close MOV CC685 in the event of high flow, which would be an indication of a tube rupture in a RCP thermal barrier heat exchanger. If the MOV or switch did not operate properly, MOV CC9438 could be used to manually isolate the CCW return line from the RCP thermal barrier.

The licensee analyzed the thermal barrier break event in FAI/02-75 Byron/Braidwood Units 1 and 2 TREMOLO Analysis for MOV 1/2CC9438, to ensure the MOV could be closed. The evaluation determined the maximum differential pressure across CC9438 following the rupture of a single tube in an RCP thermal barrier heat exchanger during full power conditions. The results were then included in the MOV thrust calculations, which verified the valve would function as required under these conditions. For the postulated scenario, it was estimated that it would take 3 minutes to manually close CC9438 if CC685 fails to automatically close. During these 3 minutes, water at reactor coolant system (RCS) pressure (2250 psi) and temperature (550°F) would be entering the CCW system, which would be normally at 40 psi pressure and 105°F. The calculation determined 195 gallons per minute (gpm) of RCS water would be entering the CCW system and flashing to steam causing the pressure and temperature in the line to rapidly increase.

12 Enclosure

Based on this scenario, the inspectors were concerned the pressure in the CCW system would exceed the design pressure limits for the piping. Although the break would occur in piping rated to withstand the event, the high pressures would also impact downstream piping with a lower rated design pressure. The inspectors were also concerned with the potential effects of water hammer on the piping and piping supports, after the valve was closed and the voided portions of the piping quickly collapsed. The licensee did not have an evaluation addressing these potential effects on the CCW system, creating reasonable doubt the system would be able to withstand the conditions resulting from an RCP thermal barrier break. This issue was documented in AR01377834.

The licensee performed a preliminary evaluation and determined the CCW system would be able to withstand the postulated event. As part of the corrective actions, the licensee was planning to perform a more thorough evaluation of this unanalyzed condition to also demonstrate this event will not cause them to be outside their design basis and the bounds of the ASME Code.

Analysis: The inspectors determined that the failure to verify the CCW system was capable of withstanding a RCP thermal barrier break was a performance deficiency.

Specifically, when assuming a single failure of the automatic isolation function, the licensee failed to evaluate the break effect on the CCW system during the 3 minutes postulated to isolate the leak. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the capability of the systems to respond to an initiating event to prevent undesirable consequences.

Specifically, by failing to evaluate the effect of the thermal barrier rupture on the CC system there was reasonable doubt whether the system was capable of withstanding the forces applied during the course of this 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 of very low safety significance (Green) because the finding involved a design or qualification 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. Specifically, the finding was related to a calculation performed in 2004, 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 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 calculations methods or by the performance of a suitable testing program.

Contrary to the above, as of June 15, 2012, the licensee failed to verify the adequacy of design; specifically they failed to verify the CCW system was able to withstand the loads resulting from a thermal barrier rupture scenario in conjunction with a single failure.

Because this violation was of very low safety significance, it was entered into the licensees corrective action program as AR01377834, and the licensee planned to 13 Enclosure

perform an evaluation of the CCW system in response to a thermal barrier break, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000454/2012007-02; 05000455/2012007-02, Failure to Verify the CCW System Capability to Withstand a Thermal Barrier Break).

(3) Non-Conservative Calibration Tolerance Limits for Electrical Relay Settings Introduction: The inspectors identified a finding of very low safety significance (Green)

and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to specify in a design calculation the allowable relay setpoint calibration tolerances. Specifically, the acceptance criteria used in relay setting calibration procedures was not bounded by the relay setting design calculations.

Description: During the review of licensees protective relay trip setpoint calibration procedures and relay setting calculations to verify whether the applied relay settings were consistent with the designed basis calculations, the inspectors noted that the stated allowable relay setpoint setting tolerances in the relay setting calibration procedure MA-MW-772-701, Calibration of Over-current Protective Relays, were neither specified nor analyzed in the design basis relay setting calculation 19-AN-3, Protective Relay Settings For 4.16kV ESF Switchgear. The acceptance criteria in relay setting calibration procedure were not bounded by the relay setting calculations to ensure the relay settings achieved selective tripping under postulated electrical fault or overload conditions. Following discovery, the licensee performed a preliminary evaluation for affected components using the worst case scenario of relay setpoint tolerances stated in the relay setting calibration procedure and concluded that at the limits of the setting tolerances, the relay setpoints would not always meet the acceptance criteria, and selective tripping was no longer ensured. The licensee entered this finding into their corrective action program as AR01377764. Based upon the actual as-left setpoints of the affected components, the licensee determined they would still perform their required safety basis functions.

Analysis: The inspectors determined that the failure to establish adequate relay setpoint tolerances in relay setpoint setting calibration procedures and verify the effects on relay coordination margin in relay setting calculations for relays used on 4.16kV emergency safety feature switchgears 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 Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the allowed relay setpoint calibration limits would not ensure the calibration activities implement the design basis established by the relay setting calculations. At the limits of the allowable relay setting tolerances, selective tripping, during electrical faults was no longer ensured, and, the loss of relay coordination could allow a fault on one piece of equipment to propagate to other safety-related equipment outside the designed isolation boundary.

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 of very low safety significance (Green) because the finding involved a design deficiency that did not result in a loss of operability or functionality.

14 Enclosure

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 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 June 15, 2012, the acceptance criteria established in relay setting calculations were not translated into relay setpoint calibration procedures.

Because this violation was of very low safety significance, it was entered into the licensees corrective action program as AR01377764, and the licensee planned to revise the calibration procedures, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 5000454/2012007-03; 05000455/2012007-03), Non-Conservative Calibration Tolerance Limits for Electrical Relay Settings).

.4 Operating Experience a. Inspection Scope The inspectors reviewed six operating experience issues 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:

Industry Issue (AT 00371721), Water Solid RH During SBLOCA;

NER QC-12-013, Diesel Generator Technical Specification Frequency and Voltage Variation not Considered in Loading Calculations;

IN 1991-18, Switchyard Problems that Contribute to Loss of Offsite Power;

IN 1993-92, Plant Improvements to Mitigate Common Dependencies in Component Cooling Water Systems;

IN 2005-30, Safe Shutdown Potentially Challenged by Unanalyzed Internal Flooding Events and Inadequate Design; and

IN 2011-14, Component Cooling Water System Gas Accumulation and Other Performance Issues.

b. Findings (1) Design Analyses Did Not Adequately Address Potential Flooding of the Auxiliary Building Introduction: The inspectors identified a finding of very low safety significance (Green)

and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to adequately analyze potential design basis internal flooding events in the auxiliary building. Specifically, the licensees analysis did not account for the possible single failure of an SX isolation valve or its associated power supply, which would have prevented break isolation within 30 minutes.

15 Enclosure

Description: The inspectors reviewed calculations associated with the consequences of postulated moderate energy piping cracks within the auxiliary building. Specifically, design calculation 3C8-1281-001, Auxiliary Building Flood Level Calculations, assumed that the most limiting pipe cracks could be isolated by operator action within 30 minutes.

Calculation 3C8-0787-001, Confirmation of Safe Shutdown Capability after Auxiliary Building Flooding, evaluated the consequences of the postulated floods, including consideration of an active single failure. In addition, the UFSAR discussed the capability of the plant to withstand flooding resulting from SX pipe cracks in the auxiliary building.

The inspectors reviewed these calculations and identified portions of SX piping that could not be isolated within 30 minutes in the event of a single failure of an SX motor-operated isolation valve (1/2SX001A/B) or its associated power supply. In some cases, other isolation valves would not be accessible to the operators and failure to isolate this piping could result in a leakage path from the safety-related SX cooling tower basin to the lower levels of the auxiliary building.

In response to this issue, the licensee reviewed existing pipe stress analyses and verified that this SX piping in the auxiliary building would meet the crack exclusion pipe stress criteria and that these flood sources may be removed from the design basis. The licensee also indicated that there were existing corrective actions to revise these flooding calculations, but these specific vulnerabilities had not been identified. This single failure issue was documented in AR01378533. The licensee also documented the need to revise the UFSAR in AR01377546.

Analysis: The inspectors determined that the failure to adequately analyze potential design basis internal flooding events in the auxiliary building was a performance deficiency. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the capability of the systems to respond to an initiating event to prevent undesirable consequences. Specifically, the failure to adequately analyze potential design basis internal flooding events in the auxiliary building would affect the capability of safety-related equipment to withstand the postulated 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 of very low safety significance (Green) because the finding involved a design or qualification deficiency that did not result in a loss of operability or functionality.

Specifically, based on the licensees review of existing pipe stress analyses, these specific SX pipe cracks would not need to be postulated in accordance with the design basis.

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.

16 Enclosure

Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires in part, design control 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 June 13, 2012, design control measures failed to translate the design basis into design calculations and into plant procedures. Specifically, the licensee did not adequately analyze potential design basis internal flooding events in the auxiliary building based on a postulated worse case single-failure. Because this violation was of very low safety significance, it was entered into the licensees corrective action program as AR01378533 and AR01377546, and the licensee was revising the flooding analysis, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000454/2012007-04; 05000455/2012007-04, Design Analyses Did Not Adequately Address Potential Flooding of the Auxiliary Building).

.5 Modifications a. Inspection Scope The inspectors reviewed four 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:

DCP 9400205, Emergency Diesel Generator Governor Upgrade;

EC 382815, Replacement 1A SX Pp Requires IEE;

EC 379433, Calculate BHP Values for SX Pump Based on Field Test Data Gathered During B2R15; and

WO 00343479, Replacement of Charging Pump Rotating Elements.

b. Findings No findings of significance were identified.

.6 Operating Procedure Accident Scenarios a. Inspection Scope The inspectors performed a detailed review of the procedures listed below associated with the selected scenario, the reactor coolant pump loss of seal cooling event and internal flooding scenarios in the auxiliary building. For the procedures listed, time critical operator actions were reviewed for reasonableness, in plant action were walked down with a licensed operator, and any interfaces with other departments were evaluated. The procedures were compared to UFSAR, design assumptions, and training materials to assure for constancy.

17 Enclosure

The following operating procedures were reviewed in detail:

1BEP ES-1.3; Transfer to Cold Leg Recirculation, Unit 1;

BOP CC-14; Post LOCA Alignment of the CC System;

BOP CC-10; Alignment of the U-0 CC Pump and U-0 HX to a Unit;

1BOA PRI-7; Essential Service Water Malfunction, Unit 1;

0BOA PRI-8; Auxiliary Building Flooding, Unit 0;

1BOA PRI-6; Component Cooling Malfunction, Unit 1;

1BOA RCP-1; Reactor Coolant Pump Seal Failure, Unit 1; and

1BOA RCP-2; Loss of Seal Cooling, Unit 1.

b. Findings (1) Failure to Provide Means to Detect Leak in Emergency Core Cooling Flow Path:

Introduction: The inspectors identified a finding of very low safety significance (Green)

and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to provide a means to detect and isolate a leak in the emergency core cooling system (ECCS) flow path within 30 minutes, which was contrary to the UFSAR. Specifically, the licensee failed to provide a means to detect and isolate a leak within 30 minutes in that neither sump alarms nor radiation monitors were provided for the safety injection (SI) pump rooms.

Description: Section 6.3.2.5, System Reliability, of the UFSAR stated that the ECCS had been designed and proven by analysis to withstand any single credible active or passive failure during the recirculation phase of an accident. Additionally, Section 6.3.2.5 of the UFSAR stated that the design of the auxiliary building and related equipment was based upon handling of leaks up to a maximum of 50 gpm. The UFSAR also stated that means were provided to detect and isolate such leaks in the ECCS flow path within 30 minutes. As part of the review of the licensees application for a license, the NRC stated that the applicant has provided a system of water-level monitors and radiation detectors located in each compartment that contains ECCS components, (Section 6.3.2, Evaluation of Single Failures, of NUREG-0876, Safety Evaluation Report related to the operation of Byron Station, Units 1 and 2, Docket Nos. STN 50-454 and STN 50-455, Commonwealth Edison Company, February 1982). In addition, by letter dated April 20, 1982, the licensee informed the NRC that the auxiliary building is equipped with leak detection sumps which will detect any leakage above normal rates.

The inspectors noted that the RHR and containment spray (CS) pump rooms were equipped with sump alarms. However, such alarms were powered from non-safety-related power sources and would not initially be available after a loss of offsite power (LOOP). The alarms would be repowered from safety-related buses after the transfer from the injection phase to the recirculation phase of an accident was completed. Leaks collected by the floor drain system would eventually be detected, although it would take over an hour for a 50 gpm leak to activate a collection tank high 18 Enclosure

level alarm. The CV pump rooms, RHR pump rooms, and CS pump rooms were also equipped with air sampling radiation monitors for detection of leaks. However, similar to sump alarms, it was not clear that these monitors would be available initially after a LOOP. Like the sump alarms, they would be repowered from a safety-related power supply after the transfer to the recirculation phase was completed. Additionally, the auxiliary building was equipped with radiation monitors for the ventilation stacks.

However, it was not clear that increased radioactivity due to a leak would be readily detected due to dilution and, even if detected, the location would not be readily identifiable. Operating procedures directed operators to check auxiliary building radiation trends after the transfer to the recirculation phase and, if required, after power was restored to equipment. Operators performed rounds of auxiliary building areas on an 8-hour frequency.

Based on the above information, the inspectors concluded that leaks from ECCS components and piping may not be detected and isolated for certain areas such as the SI pumps rooms and portions of the auxiliary building not having sump alarms or local radiation detectors. In response to the inspectors concerns, the licensee initiated AR01378257. The licensee planned to evaluate options for modifications to either the plant or the plant design bases to address detection of ECCS leakage.

The licensee determined that the dose consequences associated with an ECCS component leak that was not readily detected would be bounded by a single active failure of an electrical train, which had been previously analyzed and found to be acceptable. Additionally, the licensee calculated that a 50 gpm leak would have to be present for over 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> before the NPSH for RHR pumps would be affected due to the loss of inventory in the containment sump.

Analysis: The inspectors determined that the failure to provide a means to detect and isolate a leak in the ECCS flow path within 30 minutes was contrary to the UFSAR and was a performance deficiency. Specifically, the licensee failed to provide a means to detect and isolate a leak within 30 minutes in that neither sump alarms nor radiation monitors were provided for the SI pumps rooms. The finding was determined to be more than minor because the finding was associated with the Barrier Integrity cornerstone attribute of Design Control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to provide a means to detect and isolate a leak in the ECCS flow path within 30 minutes could result in a delayed isolation of such a leak after an accident and result in a greater radionuclide release to the auxiliary building and the environment.

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 Barrier Integrity Cornerstone. The finding screened as of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment.

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.

19 Enclosure

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. The UFSAR stated that means were provided to detect and isolate a leak in the ECCS flow path within 30 minutes.

Contrary to the above, from original construction through June 15, 2012, the licensee failed to provide a means to detect and isolate a leak in the ECCS flow path within 30 minutes. Specifically, the licensee failed to provide a means to detect and isolate a leak within 30 minutes in that neither sump alarms no radiation monitors were provided for the SI pumps rooms. Because this violation was of very low safety significance, it was entered into the licensees corrective action program as AR 01378257, and the licensee planned to evaluate options for modifications to address detection of ECCS leakage, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000454/2012007-05; 05000455/2012007-05, Failure to Provide Means to Detect Leak in ECCS Flow Path).

4. OTHER ACTIVITIES 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 five issues that were identified during previous CDBIs to verify that the concern was adequately evaluated and corrective actions were identified and implemented to resolve the concern, as necessary. The following issues were reviewed:

AR00897537, AC Power Feed to River Screen House;

AR00898000, Auxiliary Feedwater Suction Pressure Calculation Enhancement;

AR00892033, Perform Monitoring Ultrasonic Testing of 0SX10AB-8;

NCV 05000454/455/2009007-01, Failure to Maintain/Extend the Qualification Basis for Molded-Case Circuit Breakers (MCCBs) Used in Safety-Related Applications Greater than 20 Years; and

NCV 05000454/455/2009007-02, Inadequate Analysis of Molded-Case Circuit Breaker Test Data.

20 Enclosure

b. Findings No findings of significance were identified.

4OA6 Meeting(s)

.1 Exit Meeting Summary On June 15, 2012, the inspectors presented the inspection results to Mr. T. Tulon, 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 21 Enclosure

SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee T. Tulon, Site Vice President B. Youman, Plant Manager D. Basina, Electrical Engineer E. Blondin, Mechanical/Structural Design Manager S. Briggs, Operations Manager A. Corrigan, Mechanical, Structural Design Manager J. Feimster, Design Engineering Manager D. Gudger, Regulatory Assurance Manager N. Halsey, Mechanical Engineer E. Hernandez, Engineering Director T. Hulbert, Reg Assurance NRC Coordinator B. James, Instrument Maintenance Superintendent C. Keller, Electrical I&C S. Kerr, WM Director M. Krawczyk, System Engineer B. Runde, System Engineer E. Stender, Mechanical Engineer D. Sargent, Structural Engineer H. Welt, Operations SRO L. Zurawiski, Nuclear Oversight Nuclear Regulatory Commission A. M. Stone, Chief, Engineering Branch 2, DRS B. Bartlett, Senior Resident Inspector J. Robbins, Resident Inspector J. Hafeez, Reactor Inspector, DRS LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened/Closed 05000454/2012007-01; NCV Non-Conforming 480/120 Vac Motor Control Contactors 05000455/2012007-01 (Section 1R21.3.b.(1))05000454/2012007-02; NCV Failure to Verify the CCW System Capability to Withstand 05000455/2012007-02 a Thermal Barrier Break (Section 1R21.3.b.(2))05000454/2012007-03; NCV Non-Conservative Calibration Tolerance Limits for 05000455/2012007-03 Electrical Relay Settings (Section 1R21.3.b.(3))05000454/2012007-04; NCV Design Analyses Did Not Adequately Address Potential 05000455/2012007-04 Flooding of the Auxiliary Building (Section 1R21.4.b.(1))05000454/2012007-05; NCV Failure to Provide Means to Detect Leak in Emergency 05000455/2012007-05 Core Cooling Flow Path (Section 1R21.6.b.(1))

1 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 002-M-014 RHR MOV Differential Pressure Calculation 4 002-M-063 Byron Unit 1 and Unit 2 CCW System MOV Differential Pressure Calculation 000A 002-M-064 CVCS Differential Pressure Calculation 1B 19-AN-1 Relay settings for Generator, MPT, UAT and SAT 5C 19-AN-3 Protective Relay Settings For 4.16KV ESF Switchgear 16C 19-AN-4 Protective Relay Settings for 4160V Non-Safety-Related Switchgear 15B 19-AN-5 Diesel Generator Protective Relay Settings 3 19-G-1 Cable Ampacity 1 3C8-0787-001 Confirmation of Safe Shutdown Capability After Auxiliary Building Flooding 2 3C8-1281-001 Auxiliary Building Flood Level Calculations 12 95-044 Thermal Endurance Evaluation of CV and SX Pumps 0 ATD-0250 Determination of Hydraulic Characteristics for SX System MOVs 1H ATO-0021 Heat Load to the Ultimate heat Sink During Station Blackout 1 ATO-0024 SX System Alignment Variations For A Single-Unit LOCA 1 BYR 98-211 RHR/ECCS Pump Flow &Pressure Accuracy Evaluation 0F BYR 98-212 RHR Pump ASME Surveillance Instrument Accuracy 2 BYR01-089 Motor Operated Valves (MOV) Actuator Motor Terminal Voltage and Thermal 1 Overload Sizing Calculation - CV System BYR03-096 Chemical and Volume Control Strong Pump I Weak Pump Interaction on 0 Recirculation Flow BYR04-016 RHR, SI, CV, and CS Pump NPSH During ECCS Injection Mode 2 BYR05-098 NPSH for RHR Pumps During RC System Mid-Loop Operation 0 BYR06-029 SI/RHR/CS/CV System Hydraulic Analysis in Support of GSI-191 3 BYR06-058 NPSHA for RHR & CS Pumps During Post-LOCA Recirculation 1 BYR07-058 Component Cooling Water NPSH Adequacy 0 BYR07-059 RH System Heat-up During RH Pump Recirculation Without CC Flow Through 0 the RH Heat Exchanger BYR07-089 Review of Aerofin's Evaluation Report for Handhole Cover of SX Strainers 0 BYR07-091 Assessment of Short-Circuit Withstand Capability of a 5KV Power Cable 0 BYR-1 CV8355A AC Motor Operated Globe Valve Calculation 2 BYR-1 RH611 Thrust/Torque MidaCalc for 1RH611 3 BYR11-031 Evaluate 1SX005 Flange Joint for Reduced Number of Studs 0 BYR-1CC9413A Midas Data Sheet for 1CC9413A 3 BYR-1CC9438 Midas Data Sheet for 1CC9438 3 BYR-1CV8355A Midas Data Sheet for 1CV8355A 2 BYR-1SX005 AC Motor Operated Butterfly Valve Calculation 5 BYR95-022 HELB/MELB Evaluation for DCP 9500108 0 BYR96-259 SX System FLO-SERIES Analysis 2E BYR97-158 SX Water Temperature Rise Due to Pump Heat 0 BYR-97-387 CVCS MOV Differential Pressure Calculation 0 BYR97-467 Component Cooling Heat Exchanger Tube Plugging Evaluation 3 2 Attachment

CALCULATIONS Number Description or Title Revision CN-SEE-04-14 Addenda to Design Report EM-4868, Rev.1 for Non-Stellite Bearing Block 0 COD-013935 Seismic Qualification Review of Westinghouse (NSSS) Class 0 CQD-038779 Modification of Shaft Mechanical Seals for 1,2SX01PA, B 0 CWS-CAE-472C Flow Switch Setpoints 0 DD-RH-030382 RHR Orifice Plate Sizing for Flow Limiting on RHR Pump Recirculation 1 EC376794 CC System Evaluation 1 EC378827 Technical Evaluation of Potential Gas Voids in CC System 0 FAI/02-75 Byron/Braidwood Units 1 &2 TREMOLO 3 Analysis for MOV 1/2CC9438 0 FIA/02-75 Byron/Braidwood Units 1&2 - TREMOLO 3 Analysis for MOV 1/2CC9438 0 MAD 90-0060 Essential Service Water System Hydraulic Analysis 1 MAD 90-0094 Essential Service Water System Station Blackout Analysis 0K MAD 91-0080 Service Water Model Calibration 2 MSC-BB-001 MOV Seismic Qualification Re-evaluation at Byron and Braidwood Due to 0 Revised Operating Loads Design Input for Westinghouse Supplied Valves NED-M-MSD-38 Seismic Qualification Reevaluation of the MOVs Listed Below 2 PSA-B-98-08 Byron/Braidwood ECCS Flow Calculations for Safety Analysis 3D SX2-76 SX Pump Head Check 4A T-1 Calculations to verify suitability of the EDG neutral grounding resistor 1 19-AQ-24 Voltage Drop on 480-120Vac Control Transformer Circuits 7 19-AQ-63 Division Specific Degraded Voltage Analysis 6 19-AQ-65 Overvoltage Evaluation 4160-480 V Unit Substation Transformer Tap Changes 0 BYR01-087 Motor Operated Valves (MOV) Actuator Motor Terminal Voltage & Thermal 001A Overload Sizing Calculation - Component Cooling (CC) System BYR01-095 Motor Operated Valves (MOV) Actuator Motor Terminal Voltage and Thermal 000 Overload Sizing Calculation - Essential Service Water (SX) System CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION Number Description or Title Date 01367734 Seismic Concern on the SX/CC Make-Up Mod Install 05/17/12 01367989 WCAP-17308 Requires Detailed Review 05/13/12 01368220 ESF MCC Contactors Not Tested At Assumed Pickup Voltages 05/18/12 01369698 Assumption in Calc PSA-B-98-08 05/23/12 01372521 NRC CDBI - Regarding Baffle Plate Welds in CC Tanks 05/30/12 01373143 OPEX Evaluation Computation Incorrect 06/01/12 01373652 NRC CDBI -No 50.59 for Operator Time Reduction 06/02/12 01374953 NRC CDBI Identified a Discrepancy with MOV DB and DP Calc 06/06/12 01376426 NRC CDBI - Misleading Opening Description of NER 04/10/12 01376793 CDBI Followup on MCC Contactors (IR 1368220) 06/11/12 01377546 Change Required to Eliminate UFSAR Deficiency 06/13/12 01377764 Protective Relay Setting Tolerances 06/14/12 01377770 NRC CDBI - Questions on CC Surge Tank Baffle Plate Weld 06/14/12 01377834 NRC CDBI - Lack of Formal Analysis 06/13/12 01377869 NRC-CDBI - Preliminary Testing Results For C-H Contactors 06/14/12 01377901 CDBI - Followup on MCC Contactors (IR 1376793) 06/14/12 01377946 NRC CDBI - NRC Concern With Operability Determination Procedure 06/14/12 01378257 CDBI, Question about ECCS Leakage 06/15/12 01380744 Action Tracking Needed For Size 3 And 4 Contactors 06/22/12 3 Attachment

CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION Number Description or Title Date 00253533 Floor Plug Design Requirements 09/16/04 00544821 Water Solid RH During SBLOCA 10/16/06 00629356 CDBI FASA - Westinghouse DC Contractor Pick-Up Voltage Issue 05/11/07 00665855 1B D/G Output BKR Manually Opened Due to Abnormal Indication 08/29/07 00670260 Relay House South Battery Capacity Degradation 09/11/07 00741054 DG Frequency Not Addressed on Calcs 02/26/08 00763421 NER NC-08-019 Yellow NRC CDBI findings - IN 2008-02 04/15/08 00771016 Very Light Surface Corrosion, Spotty Corrosion on 1CC01A 05/02/08 00810972 Review MCCB Breaker 07/22/09 00853952 IST Trend Observation for 1CC01PA Differential Pressure 12/08/08 00892033 Perform Monitoring UT on SX10AB-8 03/12/09 00897537 2009 CDBI Issue - AC Power Feed to the River Screen House 03/25/09 00897630 02 Review Exelon Fleet Practices Concerning Testing & Replacement of MCCBs 07/15/09 00898000 2009 CDBI Issue - AF Suction Pressure Calculation Enhancement 03/26/09 00898543 Westinghouse TB 06-02 Review Issue -2009 CDBI 03/27/09 00907731 OOT Safety-Related HFB Breakers Installed Since 09/2003 04/15/09 00916063 1CV8355A - ASSY - MOV 1A RC PP Seal WTR INJ Inlet ISOL VLV 05/05/09 00920470 U1 CC Surge Tank Level Dropping 05/15/09 00930284 NRC Finding Documented in Inspection Report (MCCBs) 06/11/09 00930301 NRC Finding Documented In Inspection Report (MCCB Test Data) 06/11/09 00968169 Indication of Leak BY From SX Isolations to OCC01A 09/08/09 00971233 1CV8355A - ASSY - MOV 1A RC PP Seal WTR INJ Inlet ISOL VLV 09/27/09 01018676 Review Needed of Braidwood IR 1018119 - CDBI FASA Actions 01/20/10 01046109 Byron Review of BWD CDBI IR 1043396 03/22/10 01056849 Minor Dry Chem Buildup at Stem/Packing Area 1CC9413A 04/13/10 01056849 1CC9413A - MOV U-1 RC PPS SUP UPST ISOL VLV 04/13/10 01076087 2B D/G Voltage Pegged High 06/02/10 01147743 RH Miniflow Closure Accident Analysis Concern 12/02/10 01149502 1CV8355A - MOV 1A RC PP Seal WTR INJ Inlet ISOL VLV 12/06/10 01192924 Piping Configuration on 1SX005 Prevents Stud Removal 03/27/11 01212704 Actuator Indicator Needs To Be Installed 05/06/11 01220072 UAT 141-1 Oil Sample Port Plugged 05/21-11 01225159 Replace Lockout Relay For UAT 141-1 06/06/11 01247004 03 NRC IN 2011-14 Component Cooling Water Sys Gas Accumulation 09/20/11 01250432 NRC Questions Regarding Byron/Braidwood Exceptions to R.G. 1.9 08/11/11 01258339 Byron Station DG 24-Hour Run Needs To Be Revised 09/01/11 01258343 Byron DG Full Load Reject SURV May Need To Be Enhanced 09/01/11 01258653 Byron IR for DG 24 Hour Run Applicable to Braidwood 09/02/11 01258666 Byron IR for DG Full Load Reject Applicable to Braidwood 09/02/11 01342772 Perform DG Full Load Reject at Rated Power Factor in A1R16 03/19/12 01346061 Perform DG Full Load Reject At Rated Power Factor 03/27/12 01348279 NOS ID: Errors Impacting SBO Calculation 03/30/12 01351084 Incorrect Implementation of DG TS License Amendment 04/06/12 01354220 Need to Replace Primary Rosettes on S.O. #01Y017B4-7 04/16/12 01359198 DG Full Load Reject Testing 04/26/12 01359466 Commitment Discrepancy in 2000 TS Change for DG AOT 04/27/12 01362200 0BOSR WF-SA1 Acceptance Criteria Needs Engineering Calc 05/03/12 4 Attachment

DRAWINGS Number Description or Title Revision 118E02 Tank Component Cooling Surge Vol. 2,000 Gal 6E-0-4030VA13 Schematic Diagram Auxiliary Building Charcoal Booster Fan 0A (0VA03CA) O 6E-0-4030VC01 Schematic Diagram Control Room HVAC System Supply Fan 0A - T OVC01CA 6E-1-4001A Station One Line Diagram O 6E-1-4002A Single Line Diagram Generator, Main Power & Unit Auxiliary Transformer S Unit 1 6E-1-4002B Single Line Diagram System Auxiliary Transformer & 6.9KV Switchgear K 6E-1-4002C Single Line Diagram 4.16KV SWGR Bus 141 & 143 Diesel Generator 1A & T 480V SWGR 6E-1-4007A Byron - U1- Key Diagram 480V ESF Substation Bus 131X (1AP10E) M 6E-1-4008C Tabulation of Trip Settings 480V Auxiliary Building ESF MCC 131X1 AN (1AP21E) Part 1 and 131X1A (1AP21EA)

6E-1-4016A Relaying & Metering Diagram Unit Auxiliary Transformer 141-1 I 6E-1-4018A Relaying & Metering Diagram 4160V ESF SWGR Bus 141 U 6E-1-4018C Relaying & Metering Diagram 4160V SWGR Bus 143 L 6E-1-4020A Relaying & Metering Diagram Diesel Generator 1A-1DG01KA Generator T Control Part 1 6E-1-4020B Relaying & Metering Diagram DG 1A-1DG01KA Generator Control & Engine V Governor Control System Part - 2 6E-1-4030AP23 Schematic Diagram System Auxiliary Transformer 142-1 Feed to 4.16KV X ESF SWGR. BUS 141-ACB #1412 6E-1-4030AP25 Schematic Diagram Reserve Feed From 4.16KV ESF SWGR BUS 241 TO AE 4.16KV ESF SWGR Bus 141 & ACB 1414 6E-1-4030AP26 Schematic Diagram Bus Tie Breaker - ACB # 1411 K 6E-1-4030AP41 Schematic Diagram Unit Auxiliary Transformer 141-1 Feed to 4.16KV J SWGR Bus 143-ACB#1431 6E-1-4030AP42 Schematic Diagram Unit Auxiliary Transformer 142-1 Feed to G 4.16KV SWGR Bus 143-ACB#1432 6E-1-4030AP45 Schematic Diagram 4160V SWGR Bus 143 Undervoltage Relays I 6E-1-4030AP60 Byron -U1 - Schematic Diagram 480V ESF SWGR. 131X (1AP10E) F Manually Operated Breakers 6E-1-4030DC01 Schematic Diagram 125V DC Battery Charger III (1DC03E) N 6E-1-4030DG01 Schematic Diagram DG 1A Feed to 4.16KV ESF SWGR BUS 141 ACB Z

1413 6E-1-4030DG31 Schematic Diagram DG 1A Starting Sequence Control 1DG01KA, Part-1 AM 6E-1-4030DG32 Schematic Diagram DG 1A Starting Sequence Control 1DG01KA Part-2 AF 6E-1-4030DG35 Schematic Diagram DG 1A Generator Control 1DG01KA 6E-1-4030DG36 Schematic Diagram DG 1A Generator & Eng. Governor Control 1DG01KA N 6E-1-4030VD01 Schematic Diagram Diesel Generator Room 1A HVAC System Ventilation K Fan 1A - 1VDo1CA 6E-1-4030VP01 Schematic Diagram Reactor Containment Fan Cooler 1A - Low Speed S 1VP01CA 6E-1-4030VP02 Schematic Diagram Reactor Containment Fan Cooler 1A - High Speed T 1VP01CA 6E-1-4030VP05 Schematic Diagram Reactor Containment Fan Cooler 1C - Low Speed U 1VP01CC 6E-1-4030VP06 Schematic Diagram Reactor Containment Fan Cooler 1C - High Speed T 1VP01CC 6E-1-4611M Unit 1 Internal/External Wiring Diagram 4160V ESF SWGR Bus 141 Cub 12 N A-206 Auxiliary Building Pump Floor Plan Area 6 AB 5 Attachment

DRAWINGS Number Description or Title Revision A-207 Auxiliary Building Basement Floor Plan Area 2 BC A-208 Auxiliary Building Basement Floor Plan Area 3 CC A-209 Auxiliary Building Basement Floor Plan Area 5 AB A-210 Auxiliary Building Basement Floor Plan Area 6 BO A-211 Auxiliary Building Basement Floor Plan Area 7 Y A-212 Auxiliary Building Basement Floor Plan Areas 5 & 7 AG AC-3 AC One Line Diagram 0 CC-1 CC System 25 M-126, Sh. 1 Diagram of Essential Service Water BA M-126, Sh. 2 Diagram of Essential Service Water AD M-243, Sh. 3 Auxiliary Building Piping C M-243, Sh. 4 Auxiliary Building Piping E M-257, Sh. 1 Auxiliary Building Piping P M-42, Sh. 1A Diagram of Essential Service Water AP M-42, Sh. 1B Diagram of Essential Service Water AP M-42, Sh. 2A Diagram of Essential Service Water BA M-42, Sh. 2B Diagram of Essential Service Water BB M-42, Sh. 3 Diagram of Essential Service Water AZ M-42, Sh. 4 Diagram of Essential Service Water AN M-42, Sh. 5A Diagram of Essential Service Water AF M-42, Sh. 5B Diagram of Essential Service Water AF M-42, Sh. 6 Diagram of Essential Service Water BB M-42, Sh. 7 Diagram of Essential Service Water AE M-62 Diagram of Residual Heat Removal BD M-64 Diagram of Chemical & Volume Control & Boron Thermal Regen AE M-64, Sh. 1 Diagram of Chemical & Volume Control & Boron Thermal Regen AE M-64, Sh. 2 Diagram of Chemical & Volume Control & Boron Thermal Regen AG M-64, Sh. 3A Diagram of Chemical & Volume Control & Boron Thermal Regen AY M-64, Sh. 3B Diagram of Chemical & Volume Control & Boron Thermal Regen AS M-64, Sh. 4A Diagram of Chemical & Volume Control & Boron Thermal Regen K M-64, Sh. 4B Diagram of Chemical & Volume Control & Boron Thermal Regen L M-64, Sh. 5 Diagram of Chemical & Volume Control & Boron Thermal Regen AV M-64, Sh. 6 Diagram of Chemical & Volume Control & Boron Thermal Regen AL M-64, Sh. 7 Diagram of Chemical & Volume Control & Boron Thermal Regen AM M-64, Sh. 8 Diagram of Chemical & Volume Control AD M-64A Diagram of Chemical & Volume Control & Boron Thermal Regen C M-66, Sh. 1A Diagram of Component Cooling AW M-66, Sh. 1B Diagram of Component Cooling AJ M-66, Sh. 3A Diagram of Component Cooling AT M-66, Sh. 3B Diagram of Component Cooling AN M-66, Sh. 4D Diagram of Component Cooling AQ S-671 Auxiliary Building Essential Service Water Pump Room EL. 330-0 W 6 Attachment

MISCELLANEOUS Number Description or Title Date or Revision Letter: Commonwealth Edison to NRC - Auxiliary Building Flooding 04/12/82 ComEd and PECO Transmission Planning Criteria 02/27/12 2012-119 Ultrasonic Thickness Calibration Sheet 05/01/12 20897-DB-BYR-CC MOV Design Basis Document; Component Cooling Water 1 B1A4141.M97 Report ELMS DATA: Running Voltage Summary 05/15/12 SL 101 B1A4141.M97 Report MCC 131X1 Bus Short Circuit Information 05/17/12 SL 112 B1A4141.M97 Report ELMS DATA: Short Circuit Summary for Low Voltage Buses 05/15/12 SL 103 BB-PRA-012 Internal Flood Evaluation Summary Notebook 6 BRW-SE-1997-676 10CFR50.59 - PDP Out of Service for Extended Period of Time 09/24/98 EC 366877 Min. Wall Thickness Evaluation for Lines 0SX10AB-8 & 0SXB3AB-2 Eval 0 EC 379179 Review of Byron Station Bus 141 Molded Case Circuit Breaker (MCCB) 000 Testing Results (B1R16)

ECR 393171 Proceduralize TCCP to Install Gasoline-Powered Generators to Provide 0 AC Power to the SX Make-up Pump Battery Chargers in the Event that the RSH Losses all AC Power for an Extended Period ER-AA-321, IST Pump Evaluation Form 11/28/06, Attachment 4 11/17/08, 03/25/11 ER-AA-321-1005 Condition Monitoring Plan 4 ESC-284 Electrical Engineering Reference For Relays and Current Transformers 01/26/79 For Medium Voltage Switchgear F/L-2708 Unit Auxiliary Transformers Specifications 08/20/75 F/L-2737-01 4160 AND 6900 Volt Switchgear Specification 05/11/76 F-2755 L-2755 Proposal Technical Data for 480 Volt Motor Control Centers, Byron Station 09/08/77

- Units 1 and 2, Braidwood Station - Units 1 and 2 FASA 780286 Readiness Review for 2009 NRC Component Design Basis Inspection 0 FASA 01288088 Readiness Review for 2012 NRC Component Design Basis Inspection 02/14/12 IST-BYR-BDOC-V-03 Inservice Testing Bases Document N/A Memo 131813 Premature Degradation of the RHR Pump Thrust Bearing 11/03/89 MOV-DB-BYR-CV MOV Design Basis Document; Chemical & Volume Control 3 MPM- Maintenance Program for Westinghouse Type DB-50 Reactor Trip Circuit 0 WOGRTSDB50-01 Breakers and Associated Switchgear Manual OP Eval 12-006 Westinghouse NEMA Size 1 & 2 Contactor Pick-up Voltage Concerns 000 PE Eval 51481/51482 Item Equivalency Evaluation for the replacement Neutral Grounding 07/20/06 Resistor for the UAT 6.9KV winding PO 00000358 SR Procurement Specification 1 MODIFICATIONS Number Description or Title Date DCP 9400205 Emergency Diesel Generator Governor Upgrade 08/16/96 EC 382815 Replacement 1A SX Pp Requires IEE 04/27/10 EC 379433 Calculate BHP Values for SX Pump Based on Field Test Data Gathered During 04/27/10 B2R15 7 Attachment

PROCEDURES Number Description or Title Revision 0B0A ELEC-1 Degraded SWYD Voltage Unit 0 9 0BOA PRI-8 Auxiliary Building Flooding, Unit 0 4 0BOSR WF-SA1 Auxiliary Building Floor Drain Semi-Annual Surveillance 6 0BVSR DC-3b Unit Common 125V DC Relay House Battery Bank System 2 (South) 2 Performance Discharge Test 1/2BHSR DG-1 Diesel Generator 18 month Electrical Inspection 12 1BEP ES-1.3 Transfer to Cold Leg Recirculation, Unit 1 201 1BEP-1 Loss of Reactor or Secondary Coolant, Unit 1 202 1BOA ELEC-2 Table A (Loss of INST Bus 111 Effects) 102 1BOA PRI-1 Excessive Primary Plant Leakage 106 1BOA PRI-6 Component Cooling Malfunction, Unit 1 107 1BOA PRI-7 Essential Service Water Malfunction, Unit 1 105 1BOA RCP-2 Loss of Seal Cooling 105 1BOSR 5.5.38.CC.5-1c Comprehensive IST Surveillance Requirements for Component Cooling 0 Pump 1CC01PA 1BOSR 5.5.8.CV.S-1c Comprehensive IST Requirements for Centrifugal Charging Pump 2 1CV01PA 1BOSR 5.5.8.RH.5-1c Group A IST Requirements for Residual Heat Removal Pump 1RH01PA 1 1BOSR 5.5.8.SX.5-1c Comprehensive IST Requirements for the Essential Service Water (SX) 2 Pump 1SX01PA and Unit 1 SX Pumps Discharge Check Valves 1BOSR 5.c.2-1 Charging/Safety Injection System Flow Balance 1 1BOSR 6.6.2-1 Reactor Containment Fan Cooler Monthly Surveillance 27 BAR 1-20-B1 UAT 141-1 Unit Trouble 52 BAR 1-20-D1 UAT 141-1 Oil Flow Low Temp High 1 BAR 1-2-A4 CC Pump Trip 6 BAR 1-2-A5 CC Surge Tank Level High Low 7 BAR 1-2-E4 CC Surge Tank Auto-M/U On 8 BAR 1-7-E3 RCP Therm Barr CC Wtr Temp High 51 BOP CC-10 Alignment of the U-0 CC Pump and U-0 CC HX to a Unit 26 BOP CC-14 Post LOCA Alignment of the CC System 9 BOP RH-5 RH System Startup for Recirculation 24 BOP RH-6 Operation of the RH System in Shutdown Cooling 40 BOP RH-7 Boration of the RH System 11 BOP RH-8 Filling the Refueling Cavity for Refueling 20 BOP RH-9 Pump Down of the Refueling Cavity to the RWST 25 BOP SX-22 Essential Service Water Leak Isolation 5 BOP SX-4 Essential Service Water Strainer Manual Operation 12 BVP 600-10 Auxiliary Power System Breaker Program 0 BVP 800-30 Essential Service Water Fouling Monitoring Program 14 CC-AA-109 Equipment Abandoned via Operational Configuration Change 6 LS-AA-1110 Reportable Event SAF 1.24: Notification of Failure to Comply or 16 Existence of a Defect in a Procured Component LS-AA-120 Issue Identification and Screening Process 14 LS-AA-125 Corrective Action (CAP) Program 16 MA-AA-716-210-1001 Performance Centered Maintenance (PCM) Templates 9 MA-AA-725-102 Preventive Maintenance on Westinghouse Type DHP 4KV ,6.9 and 13.8KV Circuit Breakers 6 MA-AA-725-103 Preventive Maintenance of Westinghouse 4KV, 6.9KV, and 13.8KV 2 Switchgear Cubicles 8 Attachment

PROCEDURES Number Description or Title Revision MA-AP-723-450 Molded Case Circuit Breaker ODEN Testing 2 MA-AP-725-101 Preventive Maintenance on Westinghouse 480V Switchgear Cubicles 5 MA-AP-725-562 Preventive Maintenance on Westinghouse Type DS 480V Circuit 6 Breakers MA-BY-723-053 Station Battery Charger 18 Month Surveillance 17 MA-BY-773-503 Unit 1 - 6.9KV UAT and SAT Breakers Relay Routine 2 MA-BY-773-511 Unit 1 - 480 V Unit Substation Feed Breaker Relay Routine 1 MA-MW-772-701 Calibration of Over-current Protective Relays 3 OO-AA-102-102 General Area Checks and Operator Field Rounds 11 OP-AA-108-107 Switchyard Control 3 OP-AA-108-115 Operability Determinations (CM-1) 11 WC-AA-8003 Interface Procedure Between COMED/PECO and EXELON Generation 3 for Design Engineering and Transmission Planning Activities WORK ORDERS Number Description or Title Date 00149305 Internal/External CC Surge Tank PCM Inspection 03/17/08 00343479 Document Acceptability of Replacement CV Pump Rotating Element 11/04/04 00362498 Revise Drawings to Support Installation of Replacement Filnor Neutral 10/10/06 Grounding Resistors on UATs 00847710-01 Manually Backwash SX Strainer Due to Loss of Power N/A 00974466 125V DC SRH BATT SYS 2 South 09/12/07 01058018 1A DG Cubicle Prevent Maint Bus 141 Cub 6 01/16/09 01116821 Replace FF and VR Relays in 1PL07J-9OFF 01/13/09 01119377 DC111 BAT CHGR Breaker Inspection SUB 131X COMPT 4B 01/22/10 01119876-01 Perform MOV Operator Inspection - 1CV8355A 09/18/09 01123163 Perform Preventive Maintenance Inspection and Testing UAT 141-1 10/12/09 01155284 DG 1A Crosstie To Bus241 Sa 242-1 and Crosstie to Bus141 01/20/10 01158660 1RH611Position Indication Test 01/26/10 01170764 1A D/G ESF Actuation and Non-Emer Trip and Gen Trip Surv. 03/18/10 01181874 Unit CC Crosstie IST Valve Strokes - Required During Cold Shutdown 04/02/10 01228567 UAT 141-1 Feed Bus 143 ACB 1431 Relay Routine Cal 10/18/10 01237087 Diagnostic Testing of MOV 1RH611 04/27/12 01252181 UAT 141-1 Feed To Bus 159 05/16/11 01252526 Diesel Generator Electrical Inspection 01/13/11 01252527 1A DG Relay Routine Cal Bus 141 Cub 6 01/14/11 01261677-01 "A" Train Essential Service Water Valve Indication Test 02/10/11 01266811 UAT 141-1 Feed to Bus 157 Relay Routine 04/14/11 01266838 Bus 143 Undervoltage Relay Routine 03/28/11 01266933-01 UT Inspection 04/02/11 01267580 Bus 157 Bus Undervoltage Relay Routine Cal 03/23/11 01273485 Bus 141 Tech Spec Undervoltage Relays Cub 5 03/20/11 01273789-01 Containment Isolation Valve Stroke Test 03/17/11 01275004 1A Diesel Generator Sequencer Test 03/25/11 01275830 1A Diesel Generator Safe S/D Sequence and Single Load Reject 03/26/11 01278245 Unit 1 Generator Relay Routine Calibration 04/20/11 01288621 Battery Charger Operability Test 08/15/11 9 Attachment

WORK ORDERS Number Description or Title Date 01290054 125V Battery Charger Operability Test 09/02/11 01305163 DG 1A Crosstie To Bus241 Sat242-1 and Crosstie to Bus141 07/14/11 01322212 1A D/G ESF Actuation and Non-Emer Trip and Gen Trip Surv 11/09/11 01325230 Unit CC Crosstie IST Valve Strokes - Required During Cold Shutdown 07/29/11 01331609 1A DG Breaker Prevent Maint Bus 141 Cub 6 01/14/11 01353253 1A DG Vent Fan Breaker Inspection SUB 131X COMPT 3D 01/12/11 01409331 Perform Calibration of 1FIS-0611 04/23/12 01462990-01 STT for lSX005 11/10/11 01466625 1CC01PA Comprehensive IST Requirements for Component Cooling Water 11/11/11 01484344 ASME Surveillance Requirements for RHR Mini Flow Valve 1RH611 01/24/12 01509398-01 Essential Service Water System Surveillance 02/21/12 06029654 SC-Inspect/Operate/Lubricate Disconnect 11/04/09 20030407 Diagnostic Testing of MOV 1RH611 04/08/03 10 Attachment

LIST OF ACRONYMS USED

°F Fahrenheit Degrees AC Alternating Current ADAMS Agencywide Document Access Management System AR Action Request ASME American Society of Mechanical Engineers CDBI Component Design Bases Inspection CFR Code of Federal Regulations CCW Component Cooling Water CS Containment Spray CV Chemical and Volume Control DC Direct Current DRP Division of Reactor Projects DRS Division of Reactor Safety ECCS Emergency Core Cooling System ECP Engineering Change Package EDG Emergency Diesel Generator ESF Engineered Safety Feature GL Generic Letter gpm Gallons per Minute IMC Inspection Manual Chapter IN Information Notice IP Inspection Procedure IR Inspection Report IST Inservice Testing kV Kilovolt LOOP Loss of Off-site Power MOV Motor-Operated Valve NCV Non-Cited Violation NPSH Net Positive Suction Head NRC U.S. Nuclear Regulatory Commission PARS Publicly Available Records System PRA Probabilistic Risk assessment psi Pounds Per Square Inch RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal SAT System Auxiliary Transformer SDP Significance Determination Process SI Safety Injection SSC Systems, Structures, and Components SX Emergency Service Water TOL Thermal Overload TS Technical Specification UAT Unit Auxiliary Transformer UFSAR Updated Final Safety Analysis Report Vac Volts Alternating Current WO Work Order 11 Attachment

IR 05000454/2012007

Text

SUBJECT:

Dear Mr. Pacilio:

Enclosure:

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