IR 05000266/2015008

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IR 05000266/2015008, 05000301/2015008; on 02/23/2015 03/27/2015; Point Beach Nuclear Plant (Pbnp), Units 1 and 2; Component Design Bases Inspection
ML15126A155
Person / Time
Site: Point Beach  NextEra Energy icon.png
Issue date: 05/05/2015
From: Christine Lipa
NRC/RGN-III/DRS/EB2
To: Mccartney E
Point Beach
References
IR 2015008
Download: ML15126A155 (34)
v • d • e


Text

UNITED STATES May 5, 2015

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2, COMPONENT DESIGN BASES INSPECTION REPORT 05000266/2015008; 05000301/2015008

Dear Mr. McCartney:

On March 27, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection at your Point Beach Nuclear Plant, Units 1 and 2. The enclosed report documents the results of this inspection, which were discussed on March 27, 2015, with you and other members of your staff.

Based on the results of this inspection, one NRC-identified finding of very-low safety significance was identified. The finding involved a violation of NRC requirements. However, because of its very-low safety significance, and because the issue was entered into your Corrective Action Program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy If you contest the subject or severity of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Point Beach Nuclear Plant. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at Point Beach Nuclear Plant. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 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/

Christine A. Lipa, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27

Enclosure:

Inspection Report 05000266/2015008; 05000301/2015008 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 50-266; 50-301 License Nos: DPR-24, DPR-27 Report No: 05000266/2015008; 05000301/2015008 Licensee: NextEra Energy Point Beach, LLC Facility: Point Beach Nuclear Plant, Units 1 and 2 Location: Two Rivers, WI Dates: February 23-27, 2015; March 9-13, 2015; and March 23-27, 2015 Inspectors: B. Jose, Senior Engineering Inspector, Lead C. Phillips, Project Engineer, Operations V. Meghani, Engineering Inspector, Mechanical J. Gilliam, Engineering Inspector, Electrical H. Leake, Electrical Contractor C. Edwards, Mechanical Contractor Approved by: Christine A. Lipa, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

Inspection Report 05000266/2015008, 05000301/2015008; 02/23/2015-03/27/2015; Point

Beach Nuclear Plant (PBNP), Units 1 and 2; Component Design Bases Inspection.

The inspection was a 3-week onsite baseline inspection that focused on the design of components. The inspection was conducted by four regional engineering inspectors, and two consultants. One Green finding was identified by the inspectors. The finding was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green,

White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas effective date December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated July 9, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process Revision 5, dated February 201

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation of Title 10, Code of Federal Regulations (CFR) Part 50,

Appendix B, Criterion XVI, Corrective Action, for the licensees failure to implement timely corrective actions to address the longstanding issue of electrical power cables that have not been verified to be sized or protected in accordance with their design bases, as described in PBNPs Final Safety Analysis Report Section 8.0.1. Specifically, the licensee failed to correct known deficiencies regarding: (1) power cables with operating currents in excess of their current-carrying capacities; (2) power cables that are not protected against overload in accordance with the National Electrical Code; and (3) power cables for which their current-carrying capacities are undetermined. Although various corrective action documents have been initiated since these issues first came to light in the 1990 to 1991 time period, the licensee has not taken appropriate actions to correct the conditions adverse to quality to this date. The licensee entered this finding into their Corrective Action Program as Condition Report (CR) 02035020 and CR 02035680, with recommended actions to perform ampacity analysis for applicable cables, verify cables are protected against overload in accordance with the National Electrical Code, verify cable ampacities are higher than their respective load currents, and perform an evaluation to determine why this issue has not been resolved and address the safety culture aspect.

The inspectors determined the licensees failure to promptly correct the conditions adverse to quality regarding electrical power cables was a performance deficiency warranting a significance determination. The performance deficiency was determined to be more than minor, and a finding in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Design Control attribute of the Reactor Safety, Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding in accordance with IMC 0609.04,

Phase 1, Initial Screening and Characterization of Findings. The finding 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 on 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 inspectors identified a crosscutting aspect associated with this finding in the area of Human Performance, associated with the Design Margin component, because the licensee failed to ensure equipment is operated within design margins, and margins are carefully guarded and changed only through a systematic and rigorous process. [H.6] (Section 1R21.3.b (1))

Licensee-Identified Violations

No violations were identified.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R21 Component Design Bases Inspection

Main article: IP 71111.21

.1 Introduction

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

Specific documents reviewed during the inspection are listed in the Attachment to the report.

.2 Inspection Sample Selection Process

The inspectors used information contained in the licensees PRA, and the Point Beach Nuclear Plant (PBNP), Units 1 and 2, Standardized Plant Analysis Risk Model as the basis for component selection. Based on this, a number of risk-significant components, including those with Large Early Release Frequency (LERF) implications, were selected for the inspection.

The inspectors also used additional component information such as a margin assessment in the selection process. This design margin assessment considered original design reductions caused by design modification, power uprates, or reductions due to degraded material condition. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as performance test results, significant corrective actions, repeated maintenance activities, Maintenance Rule (a)(1) status, components requiring an operability evaluation, U.S.

Nuclear Regulatory Commission (NRC) resident inspector input of problem areas/equipment, and system health reports. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. A summary of the reviews performed and the specific inspection findings identified are included in the following sections of the report.

The inspectors also identified procedures for review that were associated with the selected components. In addition, the inspectors selected operating experience issues associated with the selected components.

This inspection constituted 20 samples as defined in Inspection Procedure 71111.21-05.

(13 Non-LERF components, 2 LERF components, and 5 operating experience)

.3 Component Design

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, Technical Specifications (TSs), design basis documents, drawings, calculations and other available design basis information, to determine the performance requirements of the selected components. The inspectors used applicable industry standards, such as the American Society of Mechanical Engineers Code, Institute of Electrical and Electronics Engineers (IEEE) Standards, and the National Electric Code, to evaluate acceptability of the components design. The NRC also evaluated licensee actions, if any, taken in response to NRC issued operating experience, such as Bulletins, Generic Letters (GLs),

Regulatory Issue Summaries, and Information Notices (INs). The review was to verify the selected components would function as designed when required, and support proper operation of the associated systems. The attributes that were needed for a component to perform its required function included process medium, energy sources, control systems, operator actions, and heat removal. The attributes to verify the component condition and tested capability was consistent with the design bases, and was appropriate may include installed configuration, system operation, detailed design, system testing, equipment and environmental qualification, equipment protection, component inputs and outputs, operating experience, and component degradation.

For each of the components selected, the inspectors reviewed the maintenance history, preventive maintenance activities, system health reports, operating experience-related information, vendor manuals, electrical and mechanical drawings, and licensee Corrective Action Program documents. Field walkdowns were conducted for all accessible components to assess material condition, including age-related degradation, and to verify that the as-built condition was consistent with the design. Other attributes reviewed are included as part of the scope for each individual component.

The following 15 components, including 2 with LERF implications, were reviewed:

  • 125 Vdc Battery D-105: The inspectors reviewed various electrical documents for the 125 Volts Direct Current (Vdc) D-105 Battery, including battery sizing and short circuit current calculations, TS Surveillance requirements (Service and Performance tests) to confirm the 125 Vdc system health and sufficient capacity existed for the battery to perform its safety function. The inspectors also reviewed a sample of surveillance, service, and performance test results and procedures to ensure batteries are being tested in accordance to TSs and IEEE standards. A review of various discharge tests was performed to verify the battery capacity was adequate to support the design basis duty cycle requirements, and to verify the battery capacity meets the requirements of the TS. In addition, maintenance procedures were reviewed to ensure maintenance activities (i.e. electrical termination/connection, torque requirements, no-oxide grease, etc.) were being performed according to IEEE standards and vendor manuals. The inspectors also completed a system walk down to see material conditions of the batteries and if there was any indication of degradation. The inspectors reviewed corrective action documents, trend data, and the System Health Report
  • 125 Vdc Battery Charger D-107: The inspectors reviewed electrical calculations associated with the safety-related D-107 Battery Charger. These included sizing, voltage drop, and capacity calculations. The review verified methodology, design inputs, assumptions, and results. Battery Charger surveillances, corrective actions, and performance history were reviewed to ensure acceptance criteria were met, and performance degradation would be identified. In addition, the test procedures were reviewed to determine whether maintenance and testing activities for the battery charger were in accordance with vendor/ industry recommendations. The review also verified the battery charger met the TS requirements. The electrolytic capacitors of the battery charger were also reviewed to verify they are being replaced within the recommended frequency.

In addition, the physical and material condition of the charger was visually inspected and corrective action documents were reviewed.

  • Unit 1 White Instrument Bus Inverter: The inspectors reviewed the circuit diagrams, the short circuit current calculation, and the coordination calculation to confirm the short circuit duty and the proper coordination between the panel fuses and branch circuit cabling with the upstream protective device. The inspectors reviewed the physical and material condition by visual inspection and review of corrective action documents in order to verify identification of adverse trends. The inspectors also reviewed seismic qualification, voltage drop and minimum voltage calculations. The calculation review verified methodology, design inputs, assumptions, and results. The inspectors also reviewed recent Condition Reports (CRs), operability evaluations, and operating procedures.
  • Unit 1 4160 Vac Division 1 Bus A-05: In addition to the generic list of attributes listed above, the inspectors reviewed electrical diagrams, calculations, and procedures, including system short circuit and load flow calculations. Incoming breaker protective relay trip setpoints were reviewed to evaluate the adequacy of the switchgear bus and breakers to carry anticipated loads under limiting conditions, and to withstand and interrupt maximum available faults. The inspectors also reviewed the voltage profile of the offsite system, voltage drop calculations, and the undervoltage relay settings to assess adequacy of voltage at the terminals of the safety-related loads, and ability to remain connected to offsite power under worst operating and accident conditions. Sizing of the incoming feeder cable was reviewed to determine its capability under worst case accident conditions.
  • 1A Residual Heat Removal Pump (1P-10A): In addition to the generic list of attributes listed above, the inspectors reviewed the piping and instrumentation diagram, system flow path, pump capacities, and in-service testing procedures and trending results. Also, the inspectors reviewed calculations related to pump head, flow, and net positive suction head (NPSH) to ensure the pumps are capable of performing their accident mitigation function. An overview of the post-accident containment pressure/temperature analysis was performed to verify assumptions regarding residual heat removal (RHR) flowrate inputs to this analysis were consistent with the RHR system hydraulic network analysis. The inspectors reviewed system operating procedures to ensure they were consistent with design requirements. A walkdown was performed to assess material condition of the pump and supporting components. The inspectors reviewed elementary diagrams to confirm the pump operation conformed to the design requirements. The inspectors reviewed the one-line diagram and the motor overload protection calculation to confirm proper selection of the motor circuit and motor overload protection. The voltage drop calculations were reviewed to determine whether the motor had adequate voltage for starting and running under degraded voltage conditions, and the motor circuit cabling had adequate ampacity. The inspectors also reviewed available control voltage to verify it was adequate for operation of the motor starter contactor. The maximum power demand of the pump was reviewed to verify it was properly reflected in alternating current (AC) distribution system and diesel generator loading analyses.
  • 1A Residual Heat Removal Heat Exchanger (HX) (1HX-11A): In addition to the generic list of attributes listed above, the inspectors reviewed related Westinghouse Electric design basis support documentation including the procurement specification and data sheet. The inspectors reviewed the licensees heat exchanger inspection procedures, recent inspection/test results and trending data to assess the licensees efforts to maintain the performance capability of this equipment. The licensees tube plugging analysis was reviewed to confirm adequate margin on heat transfer capability has been maintained after recent maintenance activities to plug a number of tubes.
  • Component Cooling Water Motor Operated Valve Inlet to RHR HX-11A (CCW-MOV-CC738A): The inspectors reviewed the design basis documents, and the Motor Operated Valve (MOV) Program documents. The inspectors reviewed calculations, including required thrust, weak link, and maximum differential pressure, to ensure the valve was capable of functioning under design and licensing bases conditions. The inspectors reviewed recent surveillance and inservice test (IST) results as well as the system health report and preventive maintenance records. The inspectors also reviewed the seismic qualification documentation for the MOV. The inspectors reviewed the MOV actuator terminal voltage, and thermal overload sizing and setpoint calculations. The review verified methodology, design inputs, assumptions, and results. The inspectors also verified separation from other trains and divisions by reviewing electrical drawings and cable routing information.
  • Service Water Ring Header Isolation Valves (SW-2869/2870): The inspectors reviewed the design basis documents, and the MOV Program documents. The inspectors noted the licensee had performed an evaluation in 1995, and de-classified the valve motors from safety-related to non-safety related. Since the valves did not have any active safety-related function, they were removed from the GL 89-10 MOV Program. The inspectors reviewed the AC power supply to the valve motors to confirm they received adequate voltage to operate when called upon. This review determined the valves are manually actuated, and that the motors and associated circuits have been re-classified as non-safety related.

The inspectors also noted the NRC staff had reviewed the licensee evaluation, and determined the de-classification and the removal from the MOV Program were acceptable as documented in a Task Interface Agreement memorandum, TIA-95-0264. Considering the valves did not have any active safety-related function, and were not included in the MOV Program, the inspectors did not perform additional reviews.

  • Unit 1 Component Cooling Water Surge Tank and Instrumentation: The inspectors reviewed the design basis documents for the Unit 1 component cooling water (CCW) surge tank as well as calculations determining the tank level set points, and the available NPSH for the CCW pumps under bounding conditions. The inspectors reviewed the seismic qualification documentation including the anchorage evaluations, and the associated modification. The inspectors reviewed the preventive maintenance work orders associated with tank periodic inspections and non-destructive examinations. The inspectors also reviewed the operator actions required to isolate the in-leakage prior to a tank overflow or relief condition. The inspectors reviewed level setpoint analyses, and the modification package for the instrumentation of the CCW Surge Tanks.
  • Unit 1 Turbine Driven Auxiliary Feed Water Pump/Turbine (1P-29): In addition to the generic list of attributes listed above, the inspectors reviewed the piping and instrumentation diagram, system flow path, pump capacities, and in-service testing procedures/trending results. Also, the inspectors reviewed calculations related to pump head, flow, and NPSH to ensure the pumps are capable of performing their accident mitigation function. The inspectors reviewed system operating procedures to ensure they were consistent with design requirements.

A walkdown was performed to assess material condition of the pump and supporting components.

  • Turbine Driven Auxiliary Feed Water Pump Discharge Check Valve to Steam Generator 1A (1AF-106): In addition to the generic list of attributes listed above, the inspectors reviewed the forward flow and back flow surveillance test procedures, and most recent test results for this valve. The valves procurement specification and bill of materials were also reviewed.
  • 1A Safety Injection Pump (1P-15A): The inspectors reviewed safety injection pump design bases including the flow requirements, design capacity and head, NPSH, and seismic requirements as well as design calculations for determining the minimum allowable IST acceptance criteria, and the minimum water level requirements for pump protections considering NPSH and vortexing. Seismic qualification documentation for the pump including anchorage evaluation was also reviewed. Inspectors reviewed the system health reports, preventive maintenance records, and the surveillance/IST records. The inspectors reviewed elementary diagrams to confirm the pump operation conformed to the design requirements. The inspectors reviewed the one-line diagram and the motor overload protection calculation to confirm proper selection of the motor circuit and motor overload protection. The voltage drop calculations were reviewed to determine that the motor had adequate voltage for starting and running under degraded voltage conditions, and that the motor circuit cabling had adequate ampacity. The inspectors also reviewed control voltage to verify it was adequate for operation of the motor breaker. The maximum power demand of the pump was reviewed to verify it was properly reflected in AC distribution system, and diesel generator loading analyses.
  • Safety Injection Pump Minimum Flow Common Air Operated Valve (1SI-897A):

The inspectors reviewed the design basis of the air-operated valve including requirements for the normal and failure position. Inspectors reviewed the thrust and weak link calculations as well as the seismic qualification documents for consistency with the Seismic Qualifications Utility Group methodology.

Inspectors reviewed the IST requirements identified in the PBNP IST Program, and also reviewed recent test documentation to verify completion of tests per the program requirements. Inspectors also reviewed the recent system health report, and the preventive maintenance documents.

  • Main Steam Atmospheric Dump Valve (1MS-AOV-2015) (LERF Related): In addition to the generic list of attributes listed above, the inspectors reviewed the valves procurement specification, assembly drawing and bill of materials. The IST acceptance criteria, trend data, procedures and completed work orders were also reviewed. The environmental qualification for limiting temperature, and radiation conditions for the valve were reviewed as well.
  • Main Steam Isolation Valve (1MS-2017) (LERF Related): The inspectors reviewed the design basis of the main steam isolation valve (MSIV) including the basis for its closure time requirements. The inspectors reviewed the IST Program requirements, and surveillance tests to verify their performance as required with acceptable results. The inspectors also reviewed the system health report, and the preventive maintenance records. The inspectors reviewed the Direct Current voltage drop calculation for the MSIV solenoid valves to confirm they received adequate voltage to operate during the most limiting conditions.

b. Findings

(1) Failure to Promptly Correct Conditions Adverse to Quality Regarding Electrical Power Cable Sizing and Protection
Introduction:

The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Title 10, Code of Federal Regulations (CFR)

Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to implement timely corrective actions to address the longstanding issue of electrical power cables that have not been verified to be sized or protected in accordance with their design bases, as described in PBNPs Final Safety Analysis Report (FSAR)

Section 8.0.1. Specifically, the licensee failed to correct known deficiencies regarding:

(1) power cables with operating currents in excess of their current-carrying capacities;
(2) power cables that are not protected against overload in accordance with the National Electrical Code; and
(3) power cables for which their current-carrying capacities are not determined. Although various corrective action documents have been initiated since these issues first came to light in the 1990 to 1991 time period, the licensee has not taken appropriate actions to correct the conditions adverse to quality to this date.
Description:

PBNPs FSAR Section 8.0.1, states, All cables are designed using conservative margins with respect to their current-carrying capacities. Section 8.0.1.2, states, Supports for cable trays are designed in accordance with the tray manufacturers recommendation based on 100 percent tray load. In general, cable trays are loaded such that power and control trays are filled less than 30 percent. Cables in trays are derated by factors recommended by Insulated Cable Engineers Association.

Section 8.0.1.10, states, All cables are protected against overload in accordance with the National Electrical Code. In 1990, the NRC conducted an Electrical Distribution System Functional Inspection (EDSFI). The Inspection Report Number 50-266/90-201 and 50-301/90-201, dated June 1, 1990, identified 210 cable tray sections for power cables did not conform to the FSAR and the original design criteria with respect to tray fill and ampacity. The inspectors documented this condition as Deficiency 90-201-05, Non-Conformance to Design Basis Criteria for Electrical Cable Tray Fill and Cable Ampacity Derating. On April 5, 1990, the licensee initiated Non-Conformance Report (NCR) N-90-092 to address this condition. On August 3, 1990, the licensee responded to the EDSFI concerns, in a letter to the NRC entitled "Response to Inspection Reports 50-266/90-201 and 50-301/90-201 Electrical Distribution System Functional Inspection, which included an action to develop a tool to provide the ability to evaluate the steady state electrical loading on plant equipment (including cables) for various combinations of plant operating conditions. In this letter, the licensee further stated: We have begun a process to formally evaluate the ampacity of cables running through these trays and evaluate the adequacy of the ampacity to perform the required functions of the cables.

On December 11, 1991, the licensee issued Calculation N-90-047, The Derated Ampacity of Cables in Power and Control Cable Trays with Greater than 30 percent Fill, Revision 0. This calculation discusses, on pages 10-11, analytical results that concluded 86 cables were found to have operating current in excess of their current-carrying capacities. It states, of the 86 cables that have a calculated ampacity less than the operating current, 16 supply safety related loads. The calculation further discusses, on pages 10-11, analytical results that concluded 488 cables were found to have overload protective device amperage settings that are in excess of the maximum settings allowed by the National Electrical Code. On December 11, 1991, the licensee issued Condition Report (CR) 01223215 entitled Cables have Calculated Ampacity Less than Operating Current. However, this CR was closed without correcting the issues identified in N-90-047. No CR from that time period was found that addressed the 488 cables that did not meet the FSAR requirement for overload protection.

On August 12, 1992, the licensee issued Action 5 to NCR N-90-092, to Modify the Cable and Raceway Data System (CARDS) to automatically determine the ampacity ratings of electrical cables. On December 18, 1992, this Action was closed because the same issue was being tracked in Commitment Tracking item 90232. Commitment 06068, stated, the CARDS data base contains all the data necessary to calculate the allowable ampacity of electrical cables. A function will be added to the CARDS system which will calculate the allowable ampacity of each cable based on an algorithm provided by Wisconsin Electric. The value determined will be retained in the database as an attribute of the cable. On January 22, 2003, the licensee issued corrective action document CR 01226467 entitled Non-Compliance with FSAR for Cable Overload Protection, which stated, There are several occurrences where the breaker does not protect the cable in accordance to article 240 of the national electrical code. A number of actions were initiated under this CR, none of which, corrected the breaker settings to protect the cables. The CR is still open, but the only open action is No. 05, which has a due date of May 15, 2015, to Resolve Ampacity Related Calculations on quarantine.

On October 30, 2007, the licensee issued corrective action document CR 01331133, which identified the current EDISON cable, and raceway data system (replacement for CARDS) is not being used to automatically generate cable ampacity values, contrary to Commitment 06068 discussed above. This CR was closed on June 4, 2013 without resolution of the issue of power cables for which no determination of current-carrying capacity has been made. Resolution of this issue was characterized as a long term corrective action. On September 30, 2009, the licensee issued Calculation 2009-0026, which discusses, on page 24, that safety-related cable ZGD0316A is not protected by its protective devices (fuses). On September 30, 2009, the licensee issued corrective action document CR 01373101, which identified that cable ZGD03016A is not protected against overloads in accordance with the National Electrical Code. This issue has not been corrected to this date, and it is being tracked as a long term corrective action.

In summary, since 1990, the licensee generated numerous corrective action documents to address concerns at PBNP associated with cable sizing and protection. However, timely and adequate corrective actions to address the non-conformances have not been implemented. Some of the corrective action documents deferred to long term corrective actions that are unclear regarding the specific conditions adverse to quality that have yet to be corrected.

The licensee entered this finding into their Corrective Action Program as CR 02035020 and CR 02035680 with recommended actions to perform ampacity analysis for applicable cables, verify cables are protected against overload in accordance with the National Electrical Code, verify cable ampacities are higher than their respective load currents, and perform an evaluation to determine why this issue has not been resolved and address the safety culture aspects.

Analysis:

The inspectors determined that the licensees failure to promptly correct the conditions adverse to quality regarding electrical power cables was a performance deficiency warranting a significance determination. The performance deficiency was determined to be more than minor, and a finding in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Design Control attribute of the Reactor Safety, Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding in accordance with IMC 0609.04, Phase 1-Initial Screening and Characterization of Findings. The finding 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 on 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 inspectors identified a cross-cutting aspect associated with this finding in the area of Human Performance, associated with the Design Margin component, because the licensee failed to ensure equipment is operated within design margins, and margins are carefully guarded and changed only through a systematic and rigorous process. [H.6]

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, Measures shall be established to assure conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, as early as the 1990 to 1991 time period, the licensee was aware of non-conformances with FSAR Section 8.0.1 requirements regarding power cables with operating currents in excess of their analyzed current-carrying capacities, power cables not protected against overload in accordance with the National Electrical Code, and power cables for which their current-carrying capacities are undetermined. These non-conformances have not yet been corrected. Because this violation was of very-low safety significance, and it was entered into the licensees Corrective Action Program as CR 02035020 and CR 02035680, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000266/2015008-01; 05000301/2015008-01, Failure to Promptly Correct Conditions Adverse to Quality Regarding Electrical Power Cable Sizing and Protection)

.4 Operating Experience

a. Inspection Scope

The inspectors reviewed five operating experience issues to ensure that NRC generic concerns had been adequately evaluated and addressed by the licensee. The operating experience issues listed below were reviewed as part of this inspection:

  • IN 88-45: Problems in Protective Relay and Circuit Breaker Coordination;
  • IN 2013-14; Potential Design Deficiency in Motor-Operated Valve Control Circuitry;
  • IN 2012-17; Inappropriate Use Of Certified Material Test Report Yield Stress And Age-Hardened Concrete Compressive Strength In Design Calculations;
  • Power Conversion Products Technical Bulletin TB-143001-01 PCP Edge Card Connector and Terminals.

b. Findings

No findings were identified.

.5 Modifications

a. Inspection Scope

The inspectors reviewed six permanent plant modifications related to selected risk-significant components to verify the design bases, licensing bases, and performance capability of the components had not been degraded through modifications. The modifications listed below were reviewed as part of this inspection effort:

  • Engineering Change (EC) 99-49341, Replacement of LT-618 T-12 CC Surge Tank Level Transmitter;
  • EC 94-091B, Auxiliary Building Design Enhancements;
  • EC 278265, Unit 1 & 2 P-15A, B Safety Injection Pump Anchor Bolt Drawing Discrepancies;
  • EC 94-653, Raised Covers for Overfilled Cable Trays.

b. Findings

No findings were identified.

.6 Time Critical Operator Action Review

a. Inspection Scope

The inspectors observed licensed operators perform a Loss of Coolant Accident (LOCA)with Loss of 4160 V Electrical Bus scenario that required the performance of the following time critical operator actions:

  • Safety-Related Battery Charger Restoration (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
  • Switchover from Refueling Water Storage Tank to Containment Sump Recirculation (34 to 10 percent)
  • Restore Auxiliary Building Ventilation within 30 Minutes after Establishing Containment Sump Recirculation The inspectors also observed licensed operators perform a LOCA with Loss of 4160 V Electrical Bus scenario that required the performance of the following time dependent Operator Action:
  • Secure Containment Spray Injection on a Large Break LOCA no Earlier than 40 Minutes The Inspectors also observed licensed operators perform the following:
  • Reactor Coolant Pump Thermal Barrier Failure with the Failure of 1CC 761A to Close Scenario

b. Findings

No findings were identified.

OTHER ACTIVITIES

4OA2 Identification and Resolution of Problems

.1 Review of Items Entered Into the Corrective Action Program

a. Inspection Scope

The inspectors reviewed a sample of the selected component problems identified by the licensee and entered into the Corrective Action Program. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues, and to evaluate the effectiveness of corrective actions related to design issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification, and incorporation of the problem into the Corrective Action Program. The specific corrective action documents sampled and reviewed by the inspectors are listed in the Attachment to this report.

The inspectors also selected five issues identified during previous Component Design Bases Inspections to verify that the concern was adequately evaluated, and corrective actions were identified and implemented to resolve the concern, as necessary. The following issues were reviewed:

  • FIN 5000266/301/2011009-04; Turbine Building Structural Steel Floor beams did not Meet AISC Requirements;
  • NCV 05000266/301/2008009-01; Equalizing Charge Voltage not Bounded by Battery Room H2 Generation Calculation;
  • NCV 05000266/301/2008009-02; Non-Conservative Design Basis for Primary Auxiliary Buildings Heat-up; and
  • NCV 05000266/301/2008009-04; RHR Pump Suction Pressure Gages Repeatedly found to be out of Tolerance.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On March 27, 2015, the inspectors presented the inspection results to Mr. E. McCartney, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. Several documents reviewed by the inspectors were considered proprietary information, and were either returned to the licensee or handled in accordance with NRC policy on proprietary information.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

E. McCartney, Site Vice President
J. Jensen, Vice President Fleet Support
D. Deboer, Plant General Manager
M. Millen, Licensing Manager
L. Chistensen, Licensing Project Manager
S. Aerts, Performance Improvement Manager
R. Webber, Operations Director
R. Seizert, EP Manager
R. Parker, Chemistry Manager
J. Pierce, Training Manager
L. Hawki, Engineering Manager
G. Strharsky, NOS Manager
D. Forter, Project Site Manager
B. Woyak, Engineering Manager
P. Wild, Engineering Design Manager
M. Rosseau, Engineering Design Supervisor
A. Gustafson, Operations Training Supervisor
B. Gerbers, Design Engineering
C. Gerbers, Design Engineering

J. Hudson; Design Engineering

T. Lensmire, Design Engineering
K. Locke, Licensing

U.S Nuclear Regulatory Commission

M. Shuaibi, Deputy Director, DRS
D. Oliver, Senior Resident Inspector
K. Barclay, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000266/2015008-01; NCV Failure to Promptly Correct Conditions Adverse to
05000301/2015008-01 Quality Regarding Electrical Power Cable Sizing and Protection (Section 1R21.3.b.(1))

Discussed

None.

LIST OF DOCUMENTS REVIEWED

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