IR 05000498/2013007

IR 05000498-13-007; 05000499-13-007; on 01/13/2014 - 02/13/2014; South Texas Project; Baseline Inspection, NRC Inspection Procedure 71111.21, Component Design Bases Inspection

ML14087A141
Person / Time
Site:	South Texas
Issue date:	03/28/2014
From:	Thomas Farnholtz Region 4 Engineering Branch 1
To:	Koehl D South Texas
Farnholtz T
References
IR-13-007
Download: ML14087A141 (64)

Text

March 28, 2014

SUBJECT:

SOUTH TEXAS PROJECT - THE NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000498/2013007; 05000499/2013007

Dear Mr. Koehl:

On February 13, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the South Texas Project Units 1 and 2. On March 20, 2014, the inspectors discussed the results of this inspection with T. Powell, Site Vice President and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented seven findings of very low safety significance (Green) in this report. All of these findings involved violations of NRC requirements. Further, inspectors documented a licensee-identified violation which was determined to be of very low safety significance. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance of the NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the South Texas Project. The information you provide will be considered in accordance with the NRC Inspection Manual Chapter 0305.

If you disagree with the characterization of 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 IV, and the NRC Resident Inspector at South Texas Project. In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, 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 NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Thomas R. Farnholtz, Chief Engineering Branch 1 Division of Reactor Safety

Dockets No.: 50-498; 50-499 Licenses No.: NPF-76; NPF-80

Enclosure: Inspection Report 05000498/2013007; 05000499/2013007 w/ Attachment 1: Supplemental Information Attachment 2: Detailed Risk Evaluations for the South Texas Project Component Design Bases Inspection

Electronic Distribution for the South Texas Project

SUMMARY

IR 05000498/2013007; 05000499/2013007; 01/13/2014 - 02/13/2014; South Texas Project; baseline inspection, NRC Inspection Procedure 71111.21, Component Design Bases Inspection.

The report covers an announced inspection by a team of five regional inspectors and two contractors. Seven Green violations were identified. The final significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process. Findings for which the significance determination process does not apply may be Green or be assigned a severity level after the NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to February 11, 2014, the licensee failed to adequately verify by analysis that safety-related nuclear steam supply system instrumentation loads would be capable of operating at the minimum inverter output voltage, when the inverter is fed from the station battery, and when considering the actual voltage drop to the load. In response to this issue, the licensee performed a preliminary voltage drop analysis that supported an immediate operability determination. This finding was entered into the licensees corrective action program as Condition Report 14-2017.

The team determined that failure to maintain design control of the nuclear steam supply system instrumentation power supply load was a performance deficiency. This finding was more than minor because if left uncorrected, it would lead to a more significant safety concern. Specifically, the incorrect analysis resulted in a reasonable question of operability of nuclear steam supply system instrumentation to operate at the minimum inverter output voltage, when the inverter is fed from the station battery, and when the actual voltage drop to the load for that condition was considered. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green)because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.1).

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to January 13, 2014, the licensees preventive maintenance Procedures OPMPO5-NA-002, 4160V Gould Breaker Test, and OPMP05-NA-0018 4160 Volt Gould HK Breaker Overhaul/Lubrication, failed to assure that the 4160 VAC Gould circuit breakers would perform satisfactorily in service when the licensee performed maintenance prior to completing as-found tests to verify the circuit breakers would function properly. In response to this issue, the licensee validated that the components had passed their required surveillance tests and remained operable. This finding was entered into the licensees corrective action program as Condition Reports14-738 and 14-1633.

The team determined that failure to establish a test and maintenance program which ensures that safety-related 4160 VAC Gould circuit breakers would perform satisfactorily in service was a performance deficiency. This finding was more than minor because if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to perform as-found tests prior to performing maintenance in preventive maintenance procedures was a significant programmatic deficiency which could cause unacceptable conditions to go undetected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather.

This finding had a crosscutting aspect in the area of human performance, documentation component because the licensee failed to create and maintain complete, accurate, and up-to-date documentation. [H.7] (Section 1R21.2.3).

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to January 13, 2014, the licensees preventative and post-maintenance procedures for safety-related 480 VAC Westinghouse DS circuit breakers failed to include manufacturers recommended testing of breaker control circuits at the minimum expected control voltage levels postulated to exist at the device terminals during design basis events. In response to this issue, the licensee validated that the components had passed their required surveillance tests and remained operable. This finding was entered into the licensees corrective action program as Condition Reports 11-4895 and 14-738.

The team determined that the failure to include manufacturers recommended testing of safety-related circuit breaker control circuits at the voltages postulated to exist at the device terminals during design basis events or to provide justification for not performing the tests was a performance deficiency. This finding was more than minor because if left uncorrected, it would lead to a more significant concern. Specifically, the failure to perform the breaker testing at reduced voltage using minimum expected control voltage levels could cause unacceptable conditions to go undetected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green)because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of problem identification and resolution, evaluation component because the licensee failed to thoroughly evaluate the issue to ensure that resolution addressed causes and extent of condition commensurate with their safety significance. [P.2] (Section 1R21.2.5).

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, which states, in part, Activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures.

Specifically, on November 3, 2013, maintenance personnel performing a maintenance activity change and performing the second party technical review did not initial and date the change that was performed for reactor containment fan cooler 12C backdraft damper as required by Procedure MG-0006, Work Execution and Closeout Guideline,

Revision 11, step 6.2.3. In response to this issue, the licensee initiated revisions to the associated work order instructions and established as-found trend data for backdraft damper 12C. This finding was entered into the licensees corrective action program as Condition Reports 14-1820 and 14-1836.

The team determined that failure to follow Procedure MG-0006 to complete the preventative maintenance work order on reactor containment fan cooler 12C as instructed was a performance deficiency. This finding was more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, not performing a proper procedure change does not ensure a proper technical review of the change and had the potential to challenge the availability and capability of the reactor containment fan cooler. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather.

This finding had a cross-cutting aspect in the area of human performance, resources component because procedures were not available to ensure successful work performance. [H.1] (Section 1R21.2.9).

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to February 13, 2014, documented requirements in purchase specification 3V259VS0005 were not correctly translated into specifications, drawings, and instructions evaluated in calculations MC-06482 and MC-06482A for the safety injection pump room coolers. In response to this issue, the licensee revised the associated calculations and established that the room coolers remained operable. This finding was entered into the licensees corrective action program as Condition Report 14-2673.

The team determined that the failure to maintain design control of the safety injection pump room cooler was a performance deficiency. This finding was more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of Design Control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining design control and performing a proper heat transfer calculation had the potential to challenge the availability, reliability, and capability of the safety injection pump room cooler and in turn the safety function of safety injection pumps. In accordance with Inspection Manual Chapter 0609,

Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.16).

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to January 28, 2014, the licensee failed to adequately verify by analysis that the AF-19 valve motor had adequate voltage available to close the valve when required during postulated high energy line break conditions. In response to this issue, the licensee performed a preliminary battery sizing and voltage analysis and verified that the valve motor had sufficient voltage to close when required by the failure modes and effects analysis. This finding was entered into the licensees corrective action program as Condition Report 14-1374.

The team determined that the failure to evaluate and translate the requirements for adequate voltage available at the AF-19 valve motor to close the valve during postulated high energy line break conditions was a performance deficiency. This finding was more than minor because if left uncorrected, it would lead to a more significant safety concern.

Specifically, the failure to analyze and translate the relevant requirements resulted in a condition where there was a reasonable question on the capability of the valve to close when required during postulated high energy line break conditions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green)because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. (Section 1R21.2.17).

Cornerstone: Initiating Events

Green.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate qualitative and quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, prior to January 29, 2014, the licensee failed to include appropriate qualitative and quantitative criteria in emergency operating procedures, off-normal operating procedures, and annunciator response procedures that are used during a loss of all seal cooling to a reactor coolant pump to prevent increased risk of a reactor coolant pump seal loss of coolant accident. In response to this issue, the licensee implemented changes to the affected procedures and communicated the changes to the operating staff. This finding was entered into the licensees corrective action program as Condition Report 14-1635.

The team determined that the failure to include appropriate qualitative and quantitative criteria in emergency operating procedures, off-normal operating procedures, and annunciator response procedures for a loss of all seal cooling to a reactor cooling pump was a performance deficiency. This finding was more than minor because it adversely affected the Initiating Events Cornerstone attribute of Procedure Quality and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.

Specifically, operating procedures did not contain appropriate attributes to ensure timely action to prevent an increased likelihood of a reactor coolant pump seal loss of coolant accident. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012,

Exhibit 1, Initiating Events Screening Questions, the team determined a detailed risk evaluation was necessary because, after a reasonable assessment of degradation, the finding could result in exceeding the reactor coolant system leak rate for a small loss of coolant accident. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the change to the core damage frequency would be less than 1E-7 per year (Green). This finding had a cross-cutting aspect in the area of human performance, training component because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. [H.9] (Section 1R21.4).

B. Licensee-Identified Findings

• A violation of very low safety significance that was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and associated corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

REACTOR SAFETY

This inspection of the component design bases verifies that plant components are maintained within their design and licensing bases. Additionally, this inspection provides monitoring of the capability of the selected components and operator actions to perform their design bases functions. As plants age, modifications may alter or disable important design features making the design bases difficult to determine or obsolete. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for which there are no indicators to measure performance.

1R21 Component Design Bases Inspection

To assess the ability of the South Texas Project equipment and operators to perform their required safety functions, the team inspected risk significant components and the licensees responses to industry operating experience. The team selected risk significant components for review using information contained in the South Texas Project probabilistic risk assessments and the U.S. Nuclear Regulatory Commissions (NRC) standardized plant analysis risk model. In general, the selection process focused on components that had a risk achievement worth factor greater than 1.3 or a risk reduction worth factor greater than 1.005. The items selected included components in both safety-related and nonsafety-related systems. The team selected the risk significant operating experience to be inspected based on its collective past experience.

.1 Inspection Scope

To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed calculations to independently verify the licensee's conclusions. The team also verified that the conditions of the components were consistent with the design basis and that the tested capabilities met the required criteria.

The team reviewed maintenance work records, corrective action documents, and industry operating experience records to verify that licensee personnel considered degraded conditions and their impact on the components. For the review of operator actions, the team observed operators during simulator scenarios, as well as during simulated actions in the plant.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions because of modifications, and margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results; significant corrective actions; repeated maintenance; Title 10 CFR 50.65(a)1 status; operable, but degraded conditions; the resident inspector input of problem equipment; system health reports; industry operating experience; and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in-depth margins.

The inspection procedure requires a review of 15 to 25 total samples that include risk-significant and low design margin components, containment related components, and operating experience issues. The sample selection for this inspection was 18 components, 4 containment related components, 5 operating experience items, and 4 event based activities associated with the components. The selected inspection and associated operating experience items supported risk significant functions including the following:

a. Electrical power to mitigation systems: The team selected several components in the electrical power distribution systems to verify operability to supply alternating current (AC) and direct current (DC) power to risk significant and safety-related loads in support of safety system operation in response to initiating events such as loss of offsite power, station blackout, and a loss-of-coolant accident with offsite power available. The team selected the following components:

• Safety-Related Nuclear Steam Supply System Inverter/Rectifier B IV-1203

• Reactor Coolant Pump C Underfrequency Relay

• 4160 VAC Class 1E Switchgear, Bus A

• Emergency Diesel Generator Output Circuit Breaker B

• 480 VAC Class 1E, Bus B

• Steam Generator Power Operated Relief Valve Control Circuit b. Components necessary to mitigate radiation releases: The team reviewed components required to perform isolation functions to prevent an unmonitored release of radiation.

The team selected the following components:

• Normal and Supplementary Containment Purge Valves

• Post-Accident Sampling System Containment Isolation Valves

• Reactor Containment Fan Coolers

• Containment Electrical Penetrations

c. Mitigating systems needed to attain safe shutdown: The team reviewed components and support systems required to perform the safe shutdown of the plant. The team selected the following components:

• Steam Generator Power Operated Relief Valves FV-7411, FV-7421, FV-7431, and FV-7441

• Technical Support Center Diesel Generator

• Positive Displacement Pump

• Electrical Auxiliary Building Heating, Ventilation, and Air Conditioning

• Auxiliary Feedwater Cross Connect Air Operated Valves

• Safety Injection Pump Room Coolers

• Auxiliary Feedwater System Steam Generator Isolation Motor Operated Valves

• Essential Cooling Water Screen Wash System

.2.1 Safety-Related Nuclear Steam Supply System Inverter/Rectifier B IV-1203

a. Inspection Scope

The team reviewed the updated safety analysis report, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with safety-related inverter IV-1230 to ensure design basis requirements were met. The team also performed walkdowns and conducted interviews with system and design engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Short circuit calculations, inverter sizing calculations, coordination studies, and voltage drop calculations.

• One-line diagrams and design basis documents for the inverter electrical distribution system to identify requirements and interfaces.

• Preventive maintenance activities to verify the inverter system maintained according to manufacturer recommendations.

• Periodic load testing to demonstrate system capability.

• Vendor documentation to verify distribution panel branch circuit load and load voltage requirements properly translated into inverter sizing and voltage drop calculations.

• Alarm response procedures for monitored conditions and operator response.

• Past modifications associated with the inverter for design basis considerations.

b. Findings

Failure to Properly Evaluate Safety-Related Equipment Electrical Load Requirements when Verifying the Adequacy of Voltage from the Nuclear Steam Supply System Inverter/Rectifier

Introduction.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that the applicable design basis requirements associated with the safety-related nuclear steam supply system instrumentation electrical load requirements were correctly translated into the plant design.

Description.

The team requested the inverter sizing and voltage drop studies to review and verify that the Class 1E inverters and distribution system were capable of providing sufficient voltage and current to the critical loads. The team found that the licensee failed to perform an adequate voltage analysis and design verification to demonstrate that the Class 1E inverters would be capable of providing sufficient voltage to safety-related nuclear steam supply system instrumentation loads fed from inverter backed power distribution panel DP-1203.

The team found that the voltage analysis performed for the nuclear steam supply system instrumentation in power cable sizing verification calculation, EC-5038, did not include the correct instrumentation power supply load information from the nuclear steam supply system vendor. For example, for DP-1203, circuit number 3, nuclear steam supply system process cabinet, the vendor load was 1619 watts at 0.85 power factor and 118 volts, or approximately 1905 volt-amperes. However, the team found that the licensee evaluated only 1190 volt-amperes in calculation EC-5038 and did not evaluate the voltage drop to the load based on the expected load current that the nuclear steam supply system instrumentation cabinet power supplies would require at the minimum inverter output voltage conditions. The team also found that nuclear steam supply system instrumentation loads were incorrectly depicted on the inverter one-line diagram.

The one-line diagram incorrectly showed the nuclear steam supply system process cabinet load as 1619 volt-amperes. The value of load volt-amperes on the one-line diagram for DP-1203, circuit number 3, was found by the team to be understated by approximately 286 volt-amperes (1905 volt-amperes minus 1619 volt-amperes) in the example discussed above. Nonetheless, the licensee confirmed that the correct value of load was considered in the inverter sizing analysis and the station battery sizing was performed correctly. The licensee found during their review on this issue that there were other similar errors made on the one-line diagram for DP-1203 where load watts were incorrectly represented as load volt-amperes.

The team determined that the use of incorrect load data and voltage drop methodology contributed both to understating the load current required by the instrumentation power supplies and overestimating the acceptance value for the maximum circuit length for the conductor size that was utilized for the load. The licensee performed a preliminary voltage drop analysis that supported an immediate operability determination that provided assurance the nuclear steam supply system instrumentation power supplies would operate within the manufacturers voltage design limit for the minimum conditions of inverter output voltage.

Analysis.

The team determined that failure to maintain design control of the nuclear steam supply system instrumentation power supply load was a performance deficiency.

This finding was more than minor because if left uncorrected, it would lead to a more significant safety concern. Specifically, the incorrect analysis resulted in a reasonable question of operability of nuclear steam supply system instrumentation to operate at the minimum inverter output voltage, when the inverter is fed from the station battery, and when the actual voltage drop to the load for that condition was considered. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, prior to February 11, 2014, the licensee did not assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures and instructions. Specifically, the licensee failed to adequately verify by analysis that safety-related nuclear steam supply system instrumentation loads would be capable of operating at the minimum inverter output voltage, when the inverter is fed from the station battery, and when considering the actual voltage drop to the load.

In response to this issue, the licensee performed a preliminary voltage drop analysis that supported an immediate operability determination. This finding was entered into the licensees corrective action program as Condition Report 14-2017. Because this finding is of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000498/2013007-01, and 05000499/2013007-01, Failure to Properly Evaluate Safety-Related Equipment Electrical Load Requirements when Verifying the Adequacy of Voltage from the Nuclear Steam Supply System Inverter/Rectifier.

.2.2 Reactor Coolant Pump C Under Frequency Relay

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the under frequency relay. The team also performed walkdowns, and conducted interviews with system and design engineering personnel to ensure the capability of this component to perform its desired design basis function.

Specifically the team reviewed:

• Schematics for the under frequency relay.

• Vendor setpoint analysis and acceptance requirements.

• Calculations for determining relay setting and safety analysis limit.

• Surveillance testing to demonstrate relay setting and performance in accordance with vendor requirements.

b. Findings

No findings were identified.

.2.3 4160 VAC Class 1E Switchgear, Bus A

a. Inspection Scope

The team reviewed the updated safety analysis report, design basis documents, the current system health report, calculations, maintenance and test procedures, and condition reports associated with the A Train of 4160 VAC bus E1A. The team also performed walkdowns and conducted interviews with engineering personnel to ensure the capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

• Circuit one-line diagrams.

• Bus loading study during normal plant operation and design basis accident load conditions.

• Vendor data on available short circuit current.

• Calculated short-circuit current at loads for the bus.

• Breaker coordination study for the bus.

• Vendor data for the bus and associated circuit breakers.

• Cable sizing requirements and analyses.

• Preventive maintenance and surveillance test procedures.

• Completed surveillance and maintenance documentation.

• Modifications performed.

b. Findings

Improper Sequencing of Maintenance of 4160 VAC Circuit Breakers Prior to As-Found Tests

Introduction.

The team identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, involving the licensees failure to establish a test program which demonstrates that components will perform satisfactorily in service. Specifically, the licensee failed to record as-found test values prior to performing maintenance of cycling, cleaning, and lubricating for 4160 VAC circuit breakers.

Description.

The team reviewed five-year preventive maintenance procedures and the overhaul/lubrication procedure for 4160 VAC circuit breakers. During the review, the team identified that Procedure OPMPO5-NA-002, 4160V Gould Breaker Test, and Procedure OPMP05-NA-0018, 4160 Volt Gould HK Breaker Overhaul/Lubrication, directed maintenance personnel to clean, adjust, and manipulate the physical condition of 4160 VAC circuit breaker contacts, insulators, and other critical circuit breaker components before performing an as-found test, to determine if the circuit breakers would have performed their intended design function.

For example, Procedure OPMPO5-NA-002, 4160V Gould Breaker Test, Revision 26, Section 5.16.2 and 5.16.4 directs maintenance personnel to record as-found and as-left readings for breaker closing and opening time. Prior to performing these tests, Sections 5.6 and 5.7 direct the maintenance personnel to perform contact and insulation cleaning, and lubrication, respectively. Step 5.12.1 directs the breaker to be cycled.

Steps 5.6, 5.7, and 5.12.17 are completed before any as-found tests are performed to verify the functionality of the critical components of the circuit breaker, such as coil operations (electrical operation).

The team reviewed the data sheet resulting from the November 30, 2010, Inspect/Periodic Lube/Test in the preventative maintenance performed on 4160 VAC Standby Diesel Generator 12 output circuit breaker using Procedure OPMPO5-NA-0002, Revision 22. Those results show that maintenance personnel documented the same results for as-found and as-left for the coil opening and closing times tested parameters; therefore, the team determined that the preventive maintenance, as performed, could mask existing conditions such as unacceptable contact resistance, setpoint drift, and mechanical binding. Additionally, the procedure resulted in the inability to verify past functionality of 4160 VAC Gould circuit breakers such as the 4160 VAC Diesel Generator 22 output circuit breaker B2PKSGOE1B14.

Analysis.

The team determined that failure to establish a test and maintenance program which ensures that safety-related 4160 VAC Gould circuit breakers would perform satisfactorily in service was a performance deficiency. This finding was more than minor because if left uncorrected, it would lead to a more significant safety concern.

Specifically, the failure to perform as-found tests prior to performing maintenance in preventive maintenance procedures was a significant programmatic deficiency which could cause unacceptable conditions to go undetected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather.

This finding had a crosscutting aspect in the area of human performance, documentation component because the licensee failed to create and maintain complete, accurate, and up-to-date documentation [H.7].

Enforcement.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to the above, prior to January 13, 2014, the licensee failed to establish a test program that assured that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service was identified and performed in accordance with written test procedures which incorporated the requirements and acceptance limits contained in applicable design documents. Specifically, prior to January 13, 2014, the licensees preventive maintenance Procedures OPMPO5-NA-002, 4160V Gould Breaker Test, and OPMP05-NA-0018 4160 Volt Gould HK Breaker Overhaul/Lubrication, failed to assure that the 4160 VAC Gould circuit breakers would perform satisfactorily in service when the licensee performed maintenance prior to completing as-found tests to verify the circuit breakers would function properly. In response to this issue, the licensee validated that the components had passed their required surveillance tests and remained operable. This finding was entered into the licensees corrective action program as Condition Reports14-738 and 14-1633. Because this finding is of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000498/2013007-02, 05000499/2013007-02, Improper Sequencing of Maintenance of 4160 VAC Circuit Breakers Prior to As-Found Tests.

.2.4 Emergency Diesel Generator Output Circuit Breaker B

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the B Train 4160 VAC Emergency Diesel Generator output breaker. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Schematics and control wiring diagrams of record.

• Functional logic diagram of circuit breaker and breaker coordination.

• Preventive maintenance procedures.

• Vendor manual and specifications.

• Load calculations of record and supporting documentation.

• Calculations of record for protection settings and alarms.

• Load Coordination studies.

• Completed preventive maintenance work orders.

b. Findings

No findings were identified.

.2.5 480 VAC Class 1E Bus B

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the B Train 480 VAC Bus E1B1/E1B2. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

• Circuit one-line diagrams.

• Bus loading study during normal plant operation and design basis accidents.

• Vendor Data on available short circuit current.

• Vendor installation and maintenance manuals.

• Electrical distribution system load flow/voltage drop, short circuit, and electrical protection and coordination calculations.

• Protective device settings and circuit breaker ratings to confirm operation during worst-case short circuit conditions.

• Circuit breaker preventive maintenance inspection and testing procedures.

• Completed preventive maintenance work orders.

b. Findings

Failure to Establish an Adequate Test Program for Safety-Related 480 VAC Circuit Breakers

Introduction.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, involving the licensee's failure to establish an adequate test program for 480 VAC circuit breakers. Specifically, the licensees preventive maintenance program did not include manufacturers recommended breaker operability testing at reduced voltage using minimum expected control voltage levels.

Description.

During a review of documents related to 480 VAC circuit breaker preventive maintenance Procedure OPMP05-NA-0008, Westinghouse 480 Volt Breaker Test, the team determined that manufactures recommended reduced voltage testing was not included in the procedure and therefore not performed on Westinghouse DS 480 VAC circuit breakers. The licensee utilizes Westinghouse DS 480 VAC type circuit breakers in safety-related 480 VAC electrical systems.

On March 16, 2011, the licensee documented the failure of a 480 VAC 1T-4C reduced voltage test in Condition Report 11-4895. The breaker shunt trip coil failed to operate at 70 VDC. In reviewing the corrective action to this failure, the licensee recommended the revision of safety-related 480 VAC breaker maintenance Procedure OPMP05-NA-0008, Westinghouse 480 Volt Breaker Test, to include the performance of reduced voltage testing as part of the preventive maintenance program. As of January 12, 2014, this procedure had not been revised to perform this test. The procedure has been used to perform preventive maintenance on some safety-related 480 VAC breakers without performing the reduced voltage test since the failure was identified.

In Condition Report 11-4895, the licensee recognized the importance of performing the reduced voltage test by stating that reduced voltage testing provides information on the operation of the open and close coils and their interaction with their respective Trip and Close components (Latches, Bushings, Rollers, linkages). The licensee further stated that Review of equipment history has found occasions where breakers have failed an as-found low voltage test and have then passed it after the breaker has been overhauled. The licensee has since initiated Condition Report 14-738 and a plan of action to revise the procedure to include the reduced voltage tests.

Analysis.

The team determined that the failure to include manufacturers recommended testing of safety-related circuit breaker control circuits at the voltages postulated to exist at the device terminals during design basis events or to provide justification for not performing the tests was a performance deficiency. This finding was more than minor because if left uncorrected, it would lead to a more significant concern. Specifically, the failure to perform the breaker testing at reduced voltage using minimum expected control voltage levels could cause unacceptable conditions to go undetected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of problem identification and resolution, evaluation component because the licensee failed to thoroughly evaluate the issue to ensure that resolution addressed causes and extent of condition commensurate with their safety significance [P.2].

Enforcement.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to the above, prior to January 13, 2014, the licensee failed to establish a test program that assured that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, the licensees preventative and post-maintenance procedures for safety-related 480 VAC Westinghouse DS circuit breakers failed to include manufacturers recommended testing of breaker control circuits at the minimum expected control voltage levels postulated to exist at the device terminals during design basis events. In response to this issue, the licensee validated that the components had passed their required surveillance tests and remained operable. This finding was entered into the licensees corrective action program as Condition Reports 11-4895 and 14-738. Because this finding is of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000498/2013007-03, 05000499/2013007-03, Failure to Establish an Adequate Test Program for Safety-Related 480 VAC Circuit Breakers.

.2.6 Steam Generator Power Operated Relief Valve FV-7441 Control Circuit

a. Inspection Scope

The team reviewed the updated safety analysis report, design basis documents, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the steam generator power operated relief valve FV-7441 control circuit. The team also performed walkdowns and conducted interviews with engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Schematic and wiring diagrams.

• One-line and control diagrams for block valve motor starter.

• Calculation for voltage available at valve motor terminal during degraded voltage conditions.

• Cable routing for the power operated relief valve and associated hydraulic pumps.

• Modifications performed on the motor operator and control and starter circuit.

b. Findings

No findings were identified.

.2.7 Normal and Supplementary Containment Purge Valve Operation and Qualification.

Compliance with Containment Systems Branch BTP 6-4

a. Inspection Scope

The team reviewed the updated final safety analysis report, the current system health report, selected drawings, motor-operated and air-operated valve qualification calculations, maintenance and test procedures, condition reports and design change packages associated with the normal and supplementary containment purge isolation valves. The team conducted a walk down of the supplementary containment purge valves outside of containment. The team also conducted interviews with engineering personnel to ensure the capability of these components to perform their desired design basis functions. Specifically, the team reviewed:

• Health reports associated with normal and supplementary purge valves for the last several years.

• Maintenance work order history and corrective action program reports from 1989 to 2013 for any common problems or issues.

• Weak link calculations for the normal and supplementary motor operated valves.

• Weak link analyses for qualifying the air/spring operated valves in the supplementary containment purge system.

• Calculation NC-7121, critical mass flow rate through the supplementary purge valve following a loss of coolant accident.

• Calculation 34753/2-48, operability analysis and test report on the supplementary containment purge isolation valves.

• Documentation and drawings provided regarding ducting and piping downstream of supplementary containment purge valves and the potential for downstream piping and supports to become missiles that could damage the purge isolation valves.

b. Findings

No findings were identified.

.2.8 Post Accident Sampling System Containment Isolation Valves

a. Inspection Scope

The team reviewed the updated safety analysis report, design basis documents, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the post-accident sampling system containment isolation valves to ensure design basis requirement specifications were met. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• License amendment request and related safety evaluation report for removal of the post-accident sampling system from the technical specifications.

• License amendment request and related safety evaluation report for the approval of the risk informed inservice testing program.

• License amendment request and related safety evaluation report for the graded quality assurance risk exemption.

• Implementation of license commitments from the license amendment requests.

• Implementation of alternate post-accident containment and reactor coolant sampling and the ability to determine associated emergency action levels.

b. Findings

No findings were identified.

.2.9 Reactor Containment Fan Coolers

a. Inspection Scope

The team reviewed the updated safety analysis report, design basis documents, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the reactor containment fan coolers to ensure design basis requirement specifications were met. The team also performed walkdowns and conducted interviews with operations and system engineering personnel to ensure the capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• License amendment request and related safety evaluation report for the approval of the risk informed inservice testing program.

• Purchase specification requirements and comparison with the design basis documents and the heat transfer calculations.

• The original startup flow balance calculation, including annubar flow element constants.

b. Findings

Failure to Follow Preventative Maintenance Procedure for Reactor Containment Fan Cooler

Introduction.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to perform activities affecting quality prescribed by documented procedures of a type appropriate to the circumstances and accomplished in accordance with these procedures. Specifically, the licensee deviated from Procedure MG-0006, Work Execution and Closeout Guideline, Revision 11 without performing a procedure change including justification and adequate review.

Description.

On October 30, 2013, Unit 1 control room operators noticed that reactor containment fan cooler 12C inlet temperature was less than outlet temperature by 6 degrees F. Control room logs require a condition report be generated when the temperature difference is greater than 5 degrees F. Consequently, the licensee generated Condition Report 13-12589 to document and evaluate the issue.

On November 1, 2013, the licensee inspected reactor containment fan cooler using preventative maintenance work order 452023. It was determined that the backdraft damper would not return to the closed position and was stiff to operate. Per the work instructions, maintenance personnel were to inspect the damper and operating linkage for issues. Maintenance personnel initiated a one-time change per MG-0006 to change the spring assembly because it did not appear to have the proper tension. Later in the work instructions it requires maintenance to full stroke the backdraft damper. The instructions included a clarifying statement that full stroke is interpreted as against a mechanical stop. Maintenance personnel attempted to perform this step multiple times but the backdraft damper would not go full closed.

At this point maintenance personnel returned to the shop and discussed it with other personnel, and the supervisor, and was informed that the damper only opens about 20 percent during normal operation. On November 3, 2013, maintenance personnel returned to the damper and verified that it did close from the 20 percent position and subsequently closed out the work order. The 20 percent position was collective institutional knowledge that did not have any documented basis for this damper or any of the other reactor containment fan cooler backdraft dampers. Due to the orientation of the dampers in the system they will see different flows and different open positions. No technical data or basis is documented to support the 20 percent statement.

In accordance with Procedure MG-0006, step 6.2.3, Work Instruction Alteration, a pen and ink change which does not alter the scope may be made by the work supervisor or planner. This change SHALL be initialed and dated by the person performing the change and SHALL receive a second party technical review which is documented by initial and date. It is clear that the maintenance department understood the requirements for changes because a one-time change was performed for replacement of the spring. But because the 20 percent position was collective institutional knowledge it was not recognized as a change to the work order and potentially a change that could impact the operability of the reactor containment fan cooler. The licensees corrective actions included generating several feedback forms to revise work order instructions to allow personnel to complete the maintenance instructions as written, documenting the as-found condition of backdraft damper 12C to allow for trending, and evaluating human performance concerns for using institutional knowledge without basis.

Analysis.

The team determined that failure to follow Procedure MG-0006 to complete the preventative maintenance work order on reactor containment fan cooler 12C as instructed was a performance deficiency. This finding was more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, not performing a proper procedure change does not ensure a proper technical review of the change and had the potential to challenge the availability and capability of the reactor containment fan cooler. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather.

This finding had a cross-cutting aspect in the area of human performance, resources component because procedures were not available to ensure successful work performance [H.1].

Enforcement.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to the above, on November 3, 2013, the licensee failed to ensure that activities affecting quality was prescribed by documented procedures of a type appropriate to the circumstances and failed to accomplish these activities in accordance with these procedures. Specifically, maintenance personnel performing a maintenance activity change and performing the second party technical review did not initial and date the change that was performed for reactor containment fan cooler 12C backdraft damper as required by Procedure MG-0006, Work Execution and Closeout Guideline, Revision 11, step 6.2.3. In response to this issue, the licensee initiated revisions to the associated work order instructions and established as-found trend data for backdraft damper 12C. This finding was entered into the licensees corrective action program as Condition Reports 14-1820 and 14-1836. Because this finding is of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000498/2013007-04, Failure to Follow Preventative Maintenance Procedure for Reactor Containment Fan Cooler.

.2.10 Containment Electrical Penetrations

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the Electrical Containment Penetrations. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

• One-line diagrams for containment electrical penetrations.

• Cable schedule and routing for the containment electrical penetration cables.

• Cable sizing and material requirements for the penetrations.

• Design documentation and analyses for the seals used for the penetrations.

• Cable protection evaluation for the penetration cables.

• Leak tests and surveillance tests performed on the penetrations and associated cables.

• Preventive maintenance activities performed on the penetrations and associated cables.

b. Findings

No findings were identified.

.2.11 Steam Generator Power Operated Relief Valves FV-7411, FV-7421, FV-7431, and FV-

7441

a. Inspection Scope

The team reviewed the updated safety analysis report, the current system health report, selected drawings, maintenance and test procedures, condition reports and design change packages associated with the steam generator power operated relief valves for both units. The team also conducted walkdowns of both units and conducted interviews with engineering personnel to ensure the capability of these components to perform their desired design basis functions. Specifically, the team reviewed:

• Health reports associated with the power operated relief valves for the last several years.

• Maintenance work order history and corrective action program reports from 1989 to 2013 for any common problems or issues.

• Design change packages, DCP 08-9595-10 and DCP 08-9595-11, steam generator power operated relief valve fail close modification and associated 10 CFR 50.59 screenings, evaluations, and updated final safety analysis report changes.

• Operability determinations associated with failure of power operated relief valves to close and the effects of leakage on valve operability (Condition Reports 10-18770-1, 10-18770-16, and 12-21808-4).

• Surveillance test procedures and test results.

b. Findings

No findings were identified.

.2.12 Technical Support Center Diesel Generator

b. Inspection Scope

The team reviewed the updated safety analysis report, the current system health report, selected drawings, calculations, maintenance and test procedures, condition reports and design change packages associated with the Technical Support Center diesel generators. The team also conducted walk downs of the Technical Support Center diesel generator in Unit 2. In addition, the team conducted interviews with engineering personnel to ensure the capability of these components to perform their desired design basis functions. Specifically, the team reviewed:

• Schematics for the emergency diesel generator start and trip circuits and for generator breaker close and trip circuits.

• Maintenance work order history and corrective action program reports from 1989 to 2013 for any common problems or issues.

• One-line diagrams for the electrical distribution system to identify requirements and interfaces.

• Vendor nameplate and equipment specification data.

• Calculations for determining diesel generator load under design basis conditions.

• Calculations and supporting documentation for determining horsepower and other power system loads on the diesel generator.

• Preventive maintenance procedures to verify the diesel generator is maintained according to manufacturer recommendations.

• Analysis of flooding potential for the Technical Support Center diesel generator building in both units and the effects on plant requirements for Technical Support Center diesel generators.

• Periodic load testing to demonstrate system capability for design basis conditions.

b. Findings

No findings were identified.

.2.13 Positive Displacement Pump

a. Inspection Scope

The team reviewed the updated safety analysis report, the current system health report, selected drawings, calculations, maintenance and test procedures, condition reports and design change packages associated with the positive displacement pumps. The team also conducted walk downs of the positive displacement pump in Unit 2. The team also conducted interviews with engineering personnel to ensure the capability of these components to perform their desired design basis functions. Specifically, the team reviewed:

• Maintenance work order history and corrective action program reports from the past five years.

• Surveillance test results for the positive displacement pumps. Emphasis was placed on the capability of the positive displacement pump to be powered from the Technical Support Center diesel generator.

• Procedures utilized to power the positive displacement pump from the Technical Support Center diesel generator.

b. Findings

No findings were identified.

.2.14 Electrical Auxiliary Building Heating Ventilation and Air Conditioning

b. Inspection Scope

The team reviewed the updated safety analysis report, the current system health report, selected drawings, calculations, technical specifications, maintenance and test procedures, condition reports, operating procedures, and design change packages associated with the electrical auxiliary building heating, ventilation, and air conditioning system. The team also reviewed operating procedures that direct operations personnel to take certain actions upon loss of all heating, ventilation, and air conditioning in the electrical auxiliary building. The team conducted walk downs of the electrical auxiliary building heating, ventilation, and air conditioning systems. The team also conducted interviews with engineering, probabilistic risk assessment, and operations personnel to ensure the capability of these components to perform their desired design basis functions. Specifically, the team reviewed:

• System health reports associated with the electrical auxiliary building heating, ventilation, and air conditioning systems and components.

• Maintenance work order history and corrective action program reports from the past five years.

• Procedures 0POP04-HE-0001, Loss of EAB or Control Room HVAC and 0POP10-HE-0001, Loss of EAB HVAC for procedure logistical planning, validation, and implementation.

• Separation criteria for the electrical auxiliary building heating, ventilation, and air conditioning chilled water piping, as described in the South Texas Project updated final safety analysis report.

b. Findings

No findings were identified.

.2.15 Auxiliary Feedwater Cross Connect Air Operated Valves

a. Inspection Scope

The team reviewed the updated safety analysis report, design basis documents, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the auxiliary feedwater cross connect air operated valves. The team also performed walkdowns and conducted interviews with system engineering and operations personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• Operations procedures and training for positioning or determining that the valves are in the locked neutral position.

b. Findings

No findings were identified.

.2.16 Safety Injection Pump Room Coolers

a. Inspection Scope

The team reviewed the updated safety analysis report, design basis documents, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the safety injection pump room coolers to ensure design basis requirement specifications were met. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the monitoring of potential degradation.

• License amendment request and related safety evaluation report for the approval of the risk informed inservice testing program.

• Purchase specifications requirements and comparison with the design basis documents and the heat transfer calculations.

• The original startup flow balance calculation, including annubar flow element constants.

b. Findings

Failure to Maintain Design Control of Safety Injection Pump Room Cooler

Introduction.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to assure that the applicable design basis requirements, associated with the safety injection pump room coolers, were correctly translated into the plant design.

Description.

While conducting a review of the safety injection pump room cooler heat transfer properties and assumption used in Calculations MC-06482, Essential Chilled Water / EAB HVAC Design Basis Loads with Capacity of 300 Tons Per Train, Revision 3, and MC-06482A, Essential Chilled Water Minimum Flow Requirements for EAB, CRE, FHB, and MAB Coolers, Revision 0, the team noted a discrepancy in one of the design parameters. The team determined that the purchase specification for the room cooler, 3V259VS0005, Specification for Safety Class Air Handling Units, Revision 2, identifies the numbers of fins per inch as a maximum of 8. However, calculations MC-06482 and MC-06482A both use 11 fins per inch. The assumption in the calculation specifies that all coolers have 8 fins per inch except the safety injection pump room coolers which have 11. The licensee initiated Condition Report 14-2673 and determined that the purchase specification for the spare room cooler appeared to have been used incorrectly as the source data for the heat transfer calculations.

The original design was to have a separate cooler for each safety injection pump, but was changed during construction to only use one cooler. The change was not verified to be correctly translated to the design documentation that performed the heat transfer calculations to determine the maximum room air temperature to ensure the operability of the safety injection pumps. Ultimately, in addition to the number of fins per inch being incorrect, it was identified that the number of tubes per row, the face area, the height of the face area, and the tube length were also incorrect. These are the original calculations of record and, as such, have been incorrect prior to February 2014. The licensees corrective actions included correcting the errors in calculations MC-06482 and MC-06482A and determining that the room coolers remained operable as design margin existed between the calculated maximum room air temperature, 95 degrees F, and the design room air temperature, 120 degrees F.

Analysis.

The team determined that the failure to maintain design control of the safety injection pump room cooler was a performance deficiency. This finding was more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of Design Control and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not maintaining design control and performing a proper heat transfer calculation had the potential to challenge the availability, reliability, and capability of the safety injection pump room cooler and in turn the safety function of safety injection pumps. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, 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, prior to February 13, 2014, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions.

Specifically, documented requirements in purchase specification 3V259VS0005 were not correctly translated into specifications, drawings, and instructions evaluated in calculations MC-06482 and MC-06482A for the safety injection pump room coolers. In response to this issue, the licensee revised the associated calculations and established that the room coolers remained operable. This finding was entered into the licensees corrective action program as Condition Report 14-2673. Because this finding is of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000498/2013007-05, 05000499/2013007-05, Failure to Maintain Design Control of Safety Injection Pump Room Cooler.

.2.17 Auxiliary Feedwater Steam Generator Isolation Motor Operated Valves

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the Auxiliary Feedwater Steam Generator isolation motor operated valves. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• Maximum expected differential pressure, required stem thrust, and stroke time calculations.

• Calculations and design basis documents to ensure acceptance criteria for tested parameters were valid to support operation under accident conditions.

• Component maintenance history and corrective action program reports to verify that degraded conditions were being appropriately addressed.

• Procedures for preventive maintenance, inspection, and testing.

• Calculations for determining minimum motor terminal voltage under design and licensing basis conditions.

• Calculations for determining minimum contactor terminal voltage under design and licensing basis conditions.

• Environmental design requirements under design and licensing basis conditions.

b. Findings

Failure to Evaluate the Adequacy of Voltage Available at AF-19 Valve Motor to Close the Valve During Postulated High Energy Line Break Conditions

Introduction.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to assure that applicable regulatory requirements and the design basis, are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to evaluate the adequacy of voltage available at the AF-19 valve motor operator to close the valve during postulated high energy line break conditions.

Description.

The team found that valve AF-19 was required to close to terminate auxiliary feedwater flow under certain accident scenarios (for example, for high energy line break conditions, such as main steam and steam generator blowdown system line breaks) as stated in the Updated Final Safety Analysis Report, Table 10.4-3a, HELB Failure Modes and Effects Analysis of Auxiliary Feedwater System Electrical Equipment in IVC. However, on review of the Class 1E DC system battery sizing and system voltage calculation, the team found that closing valve AF-19, when required, was not modeled in the analysis and the adequacy of voltage at the valve motor for closing the valve was not determined. The licensee performed a preliminary battery sizing and voltage analysis during the inspection to address this error and verified that the valve motor had sufficient voltage to close when required by the failure modes and effects analysis.

Analysis.

The team determined that the failure to evaluate and translate the requirements for adequate voltage available at the AF-19 valve motor to close the valve during postulated high energy line break conditions was a performance deficiency. This finding was more than minor because if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to analyze and translate the relevant requirements resulted in a condition where there was a reasonable question on the capability of the valve to close when required during postulated high energy line break conditions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, prior to January 28, 2014, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to adequately verify by analysis that the AF-19 valve motor had adequate voltage available to close the valve when required during postulated high energy line break conditions. In response to this issue, the licensee performed a preliminary battery sizing and voltage analysis and verified that the valve motor had sufficient voltage to close when required by the failure modes and effects analysis. This finding was entered into the licensees corrective action program as Condition Report 14-1374. Because this finding is of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement Policy:

NCV 05000498/2013007-06, 05000499/2013007-06, Failure to Evaluate the Adequacy of Voltage Available at AF-19 Valve Motor to Close the Valve During Postulated High Energy Line Break Conditions.

.2.18 Essential Cooling Water Screen Wash System

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current system health report, selected drawings, maintenance and test procedures, and condition reports associated with the essential cooling water screen wash system. The team also performed walkdowns and conducted interviews with system engineering personnel to ensure the capability of this component to perform its desired design basis function. Specifically, the team reviewed:

• License Amendments to remove the screen wash system from technical specification surveillances.

• System differential pressure start setpoint verification calculation and procedure.

• Management of bio-fouling in the essential cooling water pond.

• Procedures for preventive maintenance and inspection.

• Component maintenance history and corrective action program reports to verify that degraded conditions were being appropriately addressed.

b. Findings

No findings were identified.

.3.1 Inspection of NRC Information Notice 1989-54 Potential Overpressurization of the

Component Cooling Water System

b. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 1989-54 Potential Overpressurization of the Component Cooling Water System to verify the licensee performed an applicability review and took appropriate corrective actions, if appropriate, to address concerns that could result from the failure of the component cooling water tubing within the thermal barrier heat exchanger of the reactor coolant pump. The team verified that the licensees review adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.3.2 Inspection of NRC Information Notice 1990-26 Inadequate Flow of Essential Service

Water to Room Coolers and Heat Exchangers for Engineered Safety-Feature Systems

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 1990-26 Inadequate Flow of Essential Service Water to Room Coolers and Heat Exchangers for Engineered Safety-Feature Systems to verify that the licensee performed an applicability review and took appropriate corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses potential problems resulting from using the wrong flow and pressure drop relationship in establishing adequate flow of essential service water to room coolers for engineered safety-feature systems and from failing to establish or maintain balanced flow in essential service water systems.

The team verified that the licensees review adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.3.3 Inspection of NRC Information Notice 2010-23, Malfunctions of Emergency Diesel

Generator Speed Switch Contacts

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 2010-23, Malfunctions of Emergency Diesel Generator Speed Switch Contacts, to verify that the licensee performed an applicability review and took appropriate corrective actions, if appropriate, to address the concerns described in the information notice. The team verified that the licensees review in Condition Report 10-24261 adequately addressed the issues in the information notice. Additionally, the team reviewed actions completed in Condition Report 11-11508 to verify that corrective actions were being implemented.

b. Findings

No findings were identified.

.3.4 Inspection of NRC Information Notice 2012-11, Age Related Capacitor Degradation

a. Inspection Scope

The team reviewed the licensees evaluation of Information Notice 2012-11, Age Related Capacitor Degradation to verify that the licensee performed an applicability review and took appropriate corrective actions, if appropriate, to address the concerns described in the information notice. This information notice discusses potential problems resulting from adversely affected capacitors due to age causing the epoxy insulation to harden and crack over time. This degrades the capacitor, allowing a high flow of current and excessive heating. The excessive heat can then ignite the epoxy. The team verified that the licensees review adequately addressed the issues in the information notice.

b. Findings

No findings were identified.

.4 Results of Reviews for Operator Actions:

The team selected risk-significant components and operator actions for review using information contained in the licensees probabilistic risk assessment. This included components and operator actions that had a risk achievement worth factor greater than two or a Birnbaum value greater than 1E-6.

a. Inspection Scope

The team observed operators during simulator scenarios associated with the selected components, as well as observing simulated actions in the plant.

The selected operator actions were:

• Scenario 1, Part 1: The scenario was designed to place the crew in a situation where they will need to trip a Reactor Coolant Pump that had lost all seal cooling and place the positive displacement pump in service to restore seal cooling.

• Scenario 1, Part 2: The scenario used time compression to move the crew later in the event timeline. Issues with the positive displacement pump resulted in a loss of seal cooling to reactor coolant pump 1B for a period of time and resulted in a small loss of coolant accident. During the implementation of procedure 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant, Step 20, the crew determines that cold leg recirculation capability cannot be verified and transitions to procedure POP05-EO-EC11, Loss of Emergency Coolant Recirculation. The crew must then perform a blend to refill the refueling water storage tank.

• Scenario 2: The scenario was designed to place the crew in a situation where reactor containment fan coolers are needed and do not operate automatically such that operator action is needed to place the reactor containment fan coolers in service. To accomplish this, the scenario initiates a loss of coolant accident in containment followed by reduced containment spray availability. Containment pressure rises and the reactor containment fan coolers are needed to control containment pressure.

• In-plant job performance measure: This job performance measure was designed for a plant operator to demonstrate the correct field actions required to refill the refueling water storage tank due to a loss of emergency coolant recirculation following a loss of coolant accident.

b. Findings

Failure to Develop Adequate Procedures for Loss of All Seal Cooling to a Reactor Coolant Pump

Introduction.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to have procedures with appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee failed to develop adequate procedures for responding to a loss of all seal cooling to a reactor coolant pump.

Description.

On January 29, 2014, two operating crews were observed during simulator scenarios that required operators to respond to a loss of all seal cooling to a reactor coolant pump. As evidenced by the performance of the crews, Procedures 0POP05-EO-EO00, Reactor Trip or Safety Injection, Revision 22, 0POP04-RC-0002, Reactor Coolant Pump Off Normal, Revision 32, and 0POP09-AN-04M8, Annunciator Lampbox 04M8 Response Instructions, Revision 39, were inadequate for an accident sequence that involved the loss of component cooling water to a reactor coolant pump thermal barrier followed by a loss of the only available centrifugal charging pump. This simulated condition resulted in a complete loss of seal cooling to one reactor coolant pump. Specifically, both operating crews failed to restore seal injection to the affected reactor coolant pump using the positive displacement pump within the six minute timeframe outlined by the licensees probabilistic risk assessment to prevent the increased risk of a reactor coolant pump seal loss of coolant accident. Further, the operating crew who performed the scenario validation in the simulator on January 13, 2014, also failed to restore seal injection using the positive displacement pump.

Additionally, one crew took actions that would have further degraded the potential seal failure by failing to stop the affected reactor coolant pump within one minute and then initiating seal injection with seal inlet temperature above 230 degrees F, which is contrary to the direction provided by Procedure 0POP04-RC-0002, Reactor Coolant Pump Off Normal, Revision 32. To restore seal injection, the operating crew utilized Procedure 0POP09-AN-04M8, Annunciator Lampbox 04M8 Response Instructions, Revision 39, which did not contain the same caution to avoid restoring seal injection once seal inlet temperature exceeded 230 degrees F.

The associated procedures did not include sufficient direction to ensure that reactor coolant pump seal cooling was restored within the risk-significant timeframe. Further, Procedure 0POP09-AN-04M8, Annunciator Lampbox 04M8 Response Instructions, Revision 39, did not contain adequate direction to prevent further degradation of the reactor coolant pump seal.

Analysis.

The team determined that the failure to include appropriate qualitative and quantitative criteria in emergency operating procedures, off-normal operating procedures, and annunciator response procedures for a loss of all seal cooling to a reactor cooling pump was a performance deficiency. This finding was more than minor because it adversely affected the Initiating Events Cornerstone attribute of Procedure Quality and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, operating procedures did not contain appropriate attributes to ensure timely action to prevent an increased likelihood of a reactor coolant pump seal loss of coolant accident. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the team determined a detailed risk evaluation was necessary because, after a reasonable assessment of degradation, the finding could result in exceeding the reactor coolant system leak rate for a small loss of coolant accident. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the change to the core damage frequency would be 1E-7 per year (Green). This finding had a cross-cutting aspect in the area of human performance, training component because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values [H.9].

Enforcement.

The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate qualitative and quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, prior to January 29, 2014, the licensee failed to include appropriate qualitative and quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee failed to include appropriate qualitative and quantitative criteria in emergency operating procedures, off-normal operating procedures, and annunciator response procedures that are used during a loss of all seal cooling to a reactor coolant pump to prevent increased risk of a reactor coolant pump seal loss of coolant accident.

In response to this issue, the licensee implemented changes to the affected procedures and communicated the changes to the operating staff. This finding was entered into the licensees corrective action program as Condition Report 14-1635. Because this finding is of very low safety significance and has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: 05000498/2013007-07, 05000499/2013007-07, Failure to Develop Adequate Procedures for Loss of All Seal Cooling to a Reactor Coolant Pump.

4OA2 Identification and Resolution of Problems

The team reviewed actions requests associated with the selected components, operator actions, and operating experience notifications. Any related findings are documented in prior sections of this report.

4OA6 Meetings, Including Exit

On February 14,2014, the team leader presented the preliminary inspection results to T. Powell, Site Vice President, and other members of the licensees staff. On March 20, 2014, the inspectors discussed the final results of this inspection with T. Powell, Site Vice President and other members of the licensees staff. The licensee acknowledged the findings during the meeting. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.

• Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, Activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to the above, on January 22, 2014, an activity affecting quality was not accomplished in accordance with procedures. Specifically, the licensee failed to follow Procedure 0PGP03-ZX-0002A, Condition Report Process Implementation, Revision 1, step 4.4 to ensure that a prompt operability determination was completed on two reactor containment fan coolers with backdraft dampers found approximately 50 percent open. The finding was determined to be of very low safety significance because the safety function was never lost and was determined to be operable but degraded. The licensee entered this issue in their corrective action program as Condition Reports 14-1102, 14-1106, and 14-2726.

ATTACHMENT 1 -

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

C. Albury, Supervisor, Nuclear Fuel and Analysis

M. Berg, Manager, Design Engineering

M. Berg, Engineer, Maintenance Engineering

C. Bowman, General Manager, Engineering

B. Brown, Supervisor, Maintenance Engineering

J. Cook, Supervisor, Design Engineering

R. Dunn, Manager, Nuclear Fuel and Analysis

K. Frazier, Supervisor, Systems Engineering

C. Georgeson, Supervisor, Design Engineering

D. Gore, Supervisor, Nuclear Fuel and Analysis

W. Harris, Engineer, Systems Engineering

A. Hasan, Engineer, Systems Engineering

E. Heacock, Electrical Engineer, DP Engineering

G. Hildebrandt, Manager, Operations

Q. Huynh, Mechanical Engineer, Design Engineering

R. Kersey, Supervisor, Design Engineering

R. Lacey, Electrical Engineer, Design Engineering

H. Leon, Electrical Engineer, Design Engineering

M. Meier, Vice President, Corporate Services

J. Milliff, Manager, Operations Support

J. Morris, Engineer, Regulatory Affairs

L. Peter, Plant General Manager

T. Powell, Site Vice President

R. Savage, Engineer, Regulatory Affairs

NRC personnel

F. Sanchez, Senior Resident Inspector

N. Hernandez, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000498/2013007-01,

05000499/2013007-01 NCV Failure to Properly Evaluate Safety-Related Equipment Electrical Load Requirements when Verifying the Adequacy of Voltage from the Nuclear Steam Supply System Inverter/Rectifier

05000498/2013007-02,

05000499/2013007-02 NCV Improper Sequencing of Maintenance of 4160 VAC Circuit Breakers Prior to As-Found Tests

05000498/2013007-03,

05000499/2013007-03 NCV Failure to Establish an Adequate Test Program for Safety-

Related 480 VAC Circuit Breakers

Opened and Closed

05000498/2013007-04 NCV Failure to Follow Preventative Maintenance Procedure for Reactor Containment Fan Cooler

05000498/2013007-05,

05000499/2013007-05 NCV Failure to Maintain Design Control of Safety Injection Pump Room Cooler

05000498/2013007-06,

05000499/2013007-06 NCV Failure to Evaluate the Adequacy of Voltage Available at AF-19 Valve Motor to Close the Valve During Postulated High Energy Line Break Conditions

05000498/2013007-07,

05000499/2013007-07 NCV Failure to Develop Adequate Procedures for Loss of All Seal Cooling to a Reactor Coolant Pump