IR 05000331/2017008

NRC Design Bases Assurance Inspection (Team): Inspection Report 05000331/2017008 (Jen)

ML17181A472
Person / Time
Site:	Duane Arnold
Issue date:	06/30/2017
From:	Jeffers M NRC/RGN-III/DRS/EB2
To:	Dean Curtland NextEra Energy Duane Arnold
References
IR 2017008
Preceding documents:	ML17047A665 (16 February 2017)
Download: ML17181A472 (27)

Text

June 30, 2017

SUBJECT:

DUANE ARNOLD ENERGY CENTER - NRC DESIGN BASES ASSURANCE INSPECTION (TEAM): INSPECTION REPORT 05000331/2017008

Dear Mr. Curtland:

On May 19, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed a triennial baseline Design Bases Assurance Inspection (Team) at your Duane Arnold Energy Center.

The enclosed report documents the results of this inspection, which were discussed on May 19, 2017, with yourself, and other members of your staff.

Based on the results of this inspection, two NRC-identified findings of very-low safety significance were identified. The findings involved a violation of NRC requirements. However, because of their very-low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as Non-Cited Violations in accordance with Section 2.3.2 of the NRC Enforcement Policy If you contest the violation(s) or significance of these Non-Cited Violations, 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; and the NRC resident inspector at the Duane Arnold Energy Center.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC resident inspector at the Duane Arnold Energy Center. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Mark T. Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-331 License No. DPR-49 Enclosure:

IR 05000331/2017008 cc: Distribution via LISTSERV

SUMMARY

Inspection Report 05000331/2017008, 05/01/2017 - 05/19/2017; Duane Arnold Energy Center;

Design Bases Assurance Inspection (Team).

The inspection was a 2-week onsite baseline inspection that focused on the design of components and modifications to mitigating systems. The inspection was conducted by regional engineering inspectors and two consultants. Two Green findings were identified by the inspectors. The findings were considered Non-Cited Violations (NCVs) 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 0609, Significance Determination Process, dated April 29, 2015. Cross-cutting aspects are determined using Inspection Manual Chapter 0310,

Aspects Within the Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated November 1, 2016. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6, dated July 2016.

Cornerstone: Mitigating Systems

Green.

A finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to designate the function of the residual heat removal service water (RHRSW) system strainer bypass valves as safety-related and establish the proper maintenance activities and testing associated with safety-related components. Specifically, the licensee modified the safety function of these bypass valves to provide cooling river water to the residual heat removal system heat exchangers. The licensee entered the issue into the Corrective Action Program as Condition Report (CR) 02205409. Corrective actions include classifying the open function for the RHRSW strainer bypass valves as safety-related and to re-evaluate the in-service testing requirements for the bypass valves based on the revised classification.

The inspectors determined that the failure to designate the function of the RHRSW bypass valves as safety-related and establish proper maintenance activities and testing associated with safety related components was contrary to 10 CFR Part 50, Appendix B,

Criterion III, Design Control, requirements and was a performance deficiency. This finding was greater-than-minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. The finding was of very-low safety significance because it was a design or qualification deficiency that did not represent a loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of the licensees current performance. (Section 1R21.3.b(1))

Green.

A finding of very-low safety significance and associated NCV of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to evaluate gas transport through the modified piping configuration.

Specifically, the licensee failed to perform an evaluation to ensure that sufficient vent flow velocity could be achieved for a sufficient time in the modified vent piping configuration to adequately remove any accumulated gas from the top of the A Core Spray Pump discharge piping and sweep it down through approximately 11 feet of added downward vertical vent piping. The licensee entered the issue into the Corrective Action Program as CR 02204664, CR 02205642, and CR 02205957. Corrective actions include to evaluate current venting methods, to determine enhancements, and to determine acceptance criteria for venting such as minimum flow rate and minimum venting time, as necessary to ensure detection and removal of any potential void.

The inspectors determined that the failure to evaluate the modified vent line piping configuration to ensure that any gas in the top of the A Core Spray Pump discharge piping would be adequately vented down through the vertical and horizontal sections of added piping was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very-low safety significance because it was a design or qualification deficiency that did not represent a loss of operability or functionality. The finding had an associated cross-cutting aspect in the Human Performance area of Teamwork, where Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the licensee failed to coordinate the modification activities between engineering disciplines for structural piping and fluid dynamics. (Section 1R21.4.b(1)) [H.4]

REPORT DETAILS

1R21 Design Bases Assurance Inspection (Team)

.1 Introduction

The objective of the design bases assurance inspection is to verify that design bases have been correctly implemented for the selected risk-significant components, modifications, 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 inspection also monitors the implementation of modifications to structures, systems, and components as modifications to one system may also affect the design bases and functioning of interfacing systems as well as introduce the potential for common cause failures. The Probabilistic Risk Assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for which there are no indicators to measure performance.

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

.2 Inspection Sample Selection Process

The inspectors selected risk significant components and operator actions for review using information contained in the licensees Probabilistic Risk Assessment and the Duane Arnold Energy Center Standardized Plant Analysis Risk Model. In general, the selection was based upon the components and operator actions having a risk achievement worth of greater than 1.3 and/or a risk reduction worth greater than 1.005.

Based on this process, a number of risk-significant components, including those with Large Early Release Frequency implications, were selected for the inspection. The operator actions or operating procedures selected for review included actions taken by operators both inside and outside of the control room during postulated accident scenarios associated with the selected components.

The inspectors 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 reductions caused by design modification, or 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 action, repeated maintenance activities, Maintenance Rule (a)(1) status, components requiring an operability evaluation, system health reports, and U.S. Nuclear Regulatory Commission (NRC) resident inspector input of problem areas/equipment. 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 modifications to mitigating systems for review. In addition, the inspectors selected procedures and operating experience issues associated with the selected components.

This inspection constituted 16 samples (8 components [1 component with large early release frequency implications, valve MO-2400], 6 modifications, and 2 operating experiences) as defined in Inspection Procedure 71111.21M-02.01.

.3 Component Design

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), 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 Standards, and the National Electric Code, to evaluate acceptability of the systems design. The NRC also evaluated licensee actions, if any, taken in response to NRC issued operating experience, such as Bulletins, Generic Letters, Regulatory Issue Summaries, and Information Notices. The review was to verify that the selected components would function as designed when required and support proper operation of the associated systems. The attributes that were needed for a component to perform its required function included process medium, energy sources, control systems, operator actions, and heat removal. The attributes to verify that the component condition and tested capability was consistent with the design bases and was appropriate may include installed configuration, system operation, detailed design, system testing, equipment and environmental qualification, equipment protection, component inputs and outputs, operating experience, and component degradation.

For each of the components selected, the inspectors reviewed the maintenance history, preventive maintenance activities, system health reports, operating experience-related information, vendor manuals, electrical and mechanical drawings, and licensee Corrective Action Program (CAP) 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 eight components (samples) were reviewed:

480 Volts Alternating Current (VAC) Essential Motor Control Center (MCC)

(B32): The inspectors reviewed calculations that determined the design basis loading, and reviewed starter and breaker sizes and breaker overcurrent settings to verify the setting would ensure reliable operation during design basis conditions. The inspectors also reviewed vendor documentation and the calculated available short circuit current at the MCC to assess the capability of the selected bus and breakers to withstand and interrupt short circuit conditions.

The inspectors also reviewed the load flow and voltage analysis for the MCC to confirm the adequacy of voltage at selected components during design basis conditions. The inspectors reviewed an Engineering Change (EC) that installed new motor starters to confirm the starters capability to operate reliably during degraded grid voltage and design basis seismic conditions. The inspectors performed a walkdown of the MCC to assess the observable material condition and the operating environment for adverse impact on the equipment.

4160VAC - 480VAC Supply Transformer to Essential Load Center (1X31):

The inspectors reviewed load flow calculations and vendor data to verify the transformer had sufficient capacity to support the required loads under worst case accident loading and that voltage supplied to the load center was adequate under degraded grid voltage conditions. Short circuit calculations were reviewed to ensure load center breakers were adequately sized for the transformers contribution to the available fault current. The overcurrent protective relaying for the feeder breaker to the transformer was reviewed to determine whether it provided adequate protection to the transformer, coordination with the load center bus breakers, and whether there would be any adverse interactions within the protection scheme that would reduce system reliability. The inspectors reviewed calibration procedures and records for the transformer feeder breaker protective relays to confirm that the relays were maintained as required and whether there were any adverse performance trends. The inspectors performed a walkdown of the transformer to assess the observable material condition and the operating environment for any adverse impact on the equipment.

125 Volts Direct Current (VDC) Battery (1D2): The inspectors reviewed applicable sections of the UFSAR and TSs to determine the battery design requirements and licensing commitments. The inspectors also reviewed the battery sizing calculation to verify the capability of the battery to support momentary and continuous loading for the duration of the duty cycle during accident and station blackout conditions. The voltage drop calculation was also reviewed to confirm the capability of the battery to supply adequate voltage to the loads under limiting conditions for the duration of the duty cycle. The inspectors reviewed the battery testing procedures and the results of recent tests to verify that periodic tests conformed to the TS requirements and industry standards.

The inspectors also reviewed a sampling of completed surveillance tests, service duty discharge tests, and performance tests. The review of various discharge tests was to verify the battery capacity was adequate to support the design basis duty cycle requirements and to verify that the battery capacity meets TS requirements.

125VDC Battery Charger (1D12): The inspectors reviewed applicable sections of the UFSAR and TSs to determine the battery chargers sizing requirements and licensing commitments. The inspectors also reviewed the battery charger sizing calculation to confirm its capability to maintain the battery in a charged state and to recharge the battery in a timely manner following a loss of offsite power event.

The battery charger testing procedures and recent test results were reviewed to confirm that testing conformed to the TS requirements and that test results supported design requirements. The inspectors reviewed a sample of recent incident reports to confirm the capability of the battery charger to support system demands.

125VDC Bus (1D10) & 250VDC Bus (1D40): The inspectors reviewed short circuit calculations and verified the interrupting ratings of the fuses and the molded case circuit breakers were above the calculated short circuit currents.

The voltage calculations were reviewed to determine if adequate voltage would be available for the medium voltage and low voltage switchgear circuit breaker open and close coils and spring charging motors. The inspectors also reviewed the short circuit and coordination calculations to assure coordination between the motor feed breaker open and close control circuit fuses, and supply breakers and to verify the interrupting ratings of the control circuit fuses and the control power feed breaker.

Residual Heat Removal (RHR) Heat Exchanger B (1E201B): The inspectors reviewed the heat removal capacity of the heat exchanger to ascertain that it is capable of removing the plants decay heat following shutdown under the most limiting conditions. This included a review of the heat exchanger design specification with the data sheets associated with each mode of operation, parameters of performance test results, and the heat exchangers analysis with the calculated fouling factor and heat transfer coefficient. The inspectors also reviewed the source and flow rate of the cooling water flowing through the heat exchangers tubes to validate the availability of cooling water with respect to the engineering analyses of the heat exchanger. The inspectors also reviewed the licensees modification that instituted and proceduralized the bypass of the residual heat removal service water (RHRSW) river water strainer as the safety-related flowpath to ascertain its impact to components affected by the change. The inspectors also reviewed the licensees response to NRC Generic Letter 89-13, Service Water Problems Affecting Safety-Related Equipment, dated July 18, 1989, the schedule of inspection and cleaning of the heat exchanger and the number of plugged tubes in the heat exchanger and compared the review results to engineering analyses.

High Pressure Coolant Injection (HPCI) Turbine Control Valve (HV-2200): The inspectors reviewed the HPCI start signals that initiated the HPCI turbine booster pump and main pump, the design of the HPCI turbine control valve including the valve size, its capability to operate in the steam environment, the valves functioning with the turbine governor and the feedback signal that maintained the desired pump revolutions per minute, in order to ascertain the capability of the HPCI system to deliver its flow rate assuming the most limiting accident conditions. The inspectors also reviewed the protection of the pump from turbine overspeed to verify that the HPCI system is protected from an unexpected failure due to overspeed. The inspectors also reviewed the emergency operating procedures that directed the operation of the HPCI system under specific plant accident conditions in order to validate that the system is capable to perform its design function under all emergency operating procedures accident conditions.

Reactor Core Isolation Cooling (RCIC) Steam Supply Inboard Isolation Valve (MO-2400): The inspectors reviewed motor-operated valve calculations and analyses to ensure the valve was capable of functioning under design conditions.

These included calculations for required thrust, maximum differential pressure, and valve weak link analysis. Diagnostic testing and insevice testing surveillance results, including stroke time and available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified.

The inspectors reviewed control logic and schematic diagrams to confirm that the operation of the valve conformed to design requirements and operating procedures. The inspectors also reviewed the circuit protection and the thermal overload application of the Limitorque motor operator to confirm that the circuit was adequately protected and that the valve was capable of performing its intended safety function during a design basis accident. Voltage drop calculations were reviewed to verify the motor and its associated control circuits had adequate voltage under degraded voltage conditions.

b. Findings

(1) Failure to Evaluate Effect of Crediting Bypass Line As Safety-Related Flowpath

Introduction:

The inspectors identified a finding of very-low safety significance (Green),and an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to designate the function of the RHRSW strainer bypass valves as safety-related and establish the proper maintenance activities and testing associated with safety-related components. Specifically, the licensee modified the safety function of these bypass valves to provide cooling river water to the RHR heat exchangers when the pressure differential across the RHRSW strainer reaches 6 pounds per square inch differential (psid) to assure the continued operation of the RHR heat exchanger with sufficient supply of cooling water to remove the plants decay heat load.

Description:

Following two occasions where the pressure differential across the RHRSW strainer pegged high (greater than 13 psid), the licensee performed a design modification in 2006, that provided a safety-related flow path for RHRSW cooling water to the RHR heat exchanger, bypassing the RHRSW strainer when the pressure differential across the strainer reaches 6 psid. However, the inspectors identified that the licensee failed to evaluate all the effects of this modification, such as the safety designation of the bypass path, the effects of opening the bypass path on the RHR heat exchanger with respect to debris reaching the heat exchanger, and the proper setpoint at which the bypass line should be opened. This bypass valve is located at the bottom of a pipe riser, and as such, it is subjected to river water containing silt and debris which accumulate over time and settle at the valve. The inspectors determined that the last time water flowed through this valve was 11 years ago, but the valve has not been subjected to subsequent routine maintenance activities or testing.

Analysis:

The inspectors determined that the failure to designate the function of the RHRSW bypass valves as safety-related and establish proper maintenance activities and testing associated with safety related components was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, requirements and was a performance deficiency. This finding was greater-than-minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern.

Specifically, the bypass valves could degrade with time and if opened when the strainer pressure differential reaches 6 psid, the bypass line could contain debris and silt which may impede the operation of the valve or may be dislodged with the RHRSW flow into the RHR heat exchanger.

The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on October 17, 2016. In accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, 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. Specifically, the inspectors determined the strainer bypass valves have been cycled several times in the past 15 years for different reasons. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of the licensees current performance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, states in part that, Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Criterion III also states that, the design control measures shall provide for verifying or checking the adequacy of design.

Contrary to the above, in 2006, the licensee modified the RHRSW strainer bypass line to be a safety-related flow path but did not subject the RHRSW strainer bypass valves to design control measures commensurate with those applied to the original design of the safety-related RHRSW system. Specifically, the licensee failed to designate the function of the RHRSW strainer bypass valves as safety-related and establish the proper maintenance activities and testing associated with safety-related components.

The licensee entered the issue into the CAP as CR 02205409. Corrective actions include classifying the open function for the RHRSW strainer bypass valves as safety-related and to re-evaluate the inservice testing requirements for the bypass valves based on the revised classification. Because this finding was of very-low safety significance (Green) and was entered into the licensees CAP, this violation is being treated as an NCV, consistent with Section 2.3.2a of the NRC Enforcement Policy.

(NCV 05000331/2017008-01; Failure to Evaluate Effect of Crediting Bypass Line As Safety-Related Flowpath)

.4 Mitigating System Modifications

a. Inspection Scope

The inspectors reviewed 6 permanent plant modifications to mitigating systems that had been installed in the plant during the last 3 years. This review included in-plant walkdowns for portions of the modified 125VDC, 480VAC MCC, Core Spray (CS), HPCI, RHR, and RHRSW systems. The inspectors reviewed the modifications to verify that the design bases, licensing bases, and performance capability of the components had not been degraded through modifications. The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:

the supporting design and licensing basis documentation was updated;

the changes were in accordance with the specified design requirements;

the procedures and training plans affected by the modification have been adequately updated;

the test documentation as required by the applicable test programs has been updated; and

post-modification testing adequately verified system operability and/or functionality.

The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The modifications listed below were reviewed as part of this inspection effort:

EC 156061 (Engineering Change Package1871), 480V MCC Bucket Replacement;

EC 156099 (Engineering Change Package 1906), Support Modification of HPCI Suction Piping;

EC 272555, Add Battery Cells to 1D1 and 1D2 125VDC Batteries;

EC 280492, RHRSW Pump Motor Cooler Piping Reroute;

EC 278843, A CS High Point Vent Reroute, EBB017; and

EC 283914, Wetted Cable Replacement for RHR Pumps.

b. Findings

(1) Failure to Evaluate Gas Transport through Modified Piping Configuration

Introduction:

The inspectors identified a finding of very-low safety significance (Green),and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate gas transport through the modified piping configuration. Specifically, the licensee failed to perform an evaluation to ensure that sufficient vent flow velocity could be achieved for a sufficient time in the modified vent piping configuration to adequately remove any accumulated gas from the top of the A CS Pump discharge piping and sweep it down through approximately eleven feet of added vertical vent piping.

Description:

Technical Specification Surveillance Requirement 3.5.1.1 required that, for each emergency core cooling system injection/spray subsystem, locations susceptible to gas accumulation be verified monthly to be sufficiently filled with water to ensure emergency core cooling system operability. The previous high point vent for the A CS Pump discharge piping was located in the Reactor Water Cleanup heat exchanger room which was a locked high radiation area. To reduce dose and risk to venting personnel, the licensee installed EC 278843 to reroute the high point vent piping release point to a lower dose area outside of the high radiation area using 3/4-inch Schedule 160 pipe. The modification installed approximately eleven feet of horizontal piping from the high point of the CS piping to penetrate through the wall of the Reactor Water Cleanup heat exchanger room. The new vent piping was also routed down approximately 11 feet which placed the new release point approximately 11 feet lower than the previous release point. The modification installation was completed January 30, 2017. During the inspection, the inspectors noted that modification EC 278843 failed to contain an evaluation to demonstrate that the water flow through the vent piping would be of sufficient velocity to entrain any gas accumulated near the high point of the CS piping and overcome the gas buoyancy forces and sweep the gas down the approximately 11 feet of vertical piping and out of the new lower release point. The licensee captured the concerns in its CAP as CR 02205957. The inspectors noted that Surveillance Test Procedure (STP) 3.5.1.14A, A Core Spray System Water Fill Test, was used to vent the A CS system discharge piping through the new piping and satisfy Technical Specification Surveillance Requirement 3.5.1.1. Due to lack of specificity, the inspectors questioned if the procedure was performed in a manner that resulted in a sufficient velocity for a sufficient time to sweep any accumulated gas down the piping and out the vent. The licensee captured the concern as CR 02205142, which as an immediate corrective action had personnel vent the B CS discharge piping which had a vent piping configuration similar to the A CS vent piping and determine flow rate.

During the inspection, the inspectors observed the licensee perform the venting procedure. The inspectors observed that the flowrate during the venting was approximately 2.5 gallons per minute. The inspectors requested the licensee to evaluate the flowrate and piping configuration in accordance with the methodology in the NRCs Final Safety Evaluation for National Energy Institute Topical Report NEI-09-10, Revision 1a, Guidelines for Effective Prevention and Management of System Gas Accumulation, issued March 2013. The licensee calculated the Froude Number as 2.13 and concluded that any accumulated gas would be swept out of the new vent piping.

The inspectors performed alternate calculations which resulted in similar calculated values and conclusions as the licensee. Although not specified in the procedure, the inspectors concluded venting practices observed provided adequate developed flow to transport gas and the typical time of venting was sufficient to identify and remove any gas accumulation. The licensee captured the inspectors concern that the venting procedure was not specific enough in CR 02204664 and CR 02205642.

Analysis:

The inspectors determined that the licensee s failure to evaluate the modified vent line piping configuration to ensure that any gas in the top of the A CS Pump discharge piping would be adequately vented down through the vertical and horizontal sections of added piping was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to have an evaluation that ensured that the new piping configuration will detect and remove any gas accumulation commensurate with the original design could result in the CS pump discharge piping voiding sufficient to cause hydraulic transients or otherwise degrade the CS systems ability to perform its safety function.

The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on October 17, 2016. 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 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.

Specifically, inspectors observed licensee personnel vent the CS system using the current procedure. Although not specified in procedure STP 3.5.1.14A, inspectors concluded venting practices observed provided adequate developed flow to transport gas and the typical time of venting (as indicated by craft that would perform the surveillance) would be sufficient to identify any gas accumulation.

The inspectors determined this finding had an associated cross-cutting aspect in the Human Performance area of Teamwork, where Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the licensee failed to coordinate the modification activities between engineering disciplines for structural piping and fluid dynamics. [H.4]

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, states in part that, Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Criterion III also states that, the design control measures shall provide for verifying or checking the adequacy of design.

Contrary to the above, on January 30, 2017, the licensee completed the modification to the A CS system vent line but failed to assure accumulated gas voids would be vented commensurate with the original design for venting of the CS system. Specifically, the licensee failed to perform an evaluation to ensure that sufficient vent flow velocity could be achieved for a sufficient time in the modified vent piping configuration to adequately remove any accumulated gas from the top of the A CS pump discharge piping.

The licensee entered the issue into the CAP as CR 02204664, CR 02205642, and CR 02205957. Corrective actions include to evaluate current venting methods, to determine enhancements, and to determine acceptance criteria for venting such as minimum flow rate and minimum venting time, as necessary to ensure detection and removal of any potential void. Because this finding was of very-low safety significance (Green) and was entered into the licensees CAP, this violation is being treated as an NCV, consistent with Section 2.3.2a of the NRC Enforcement Policy.

(NCV 05000331/2017008-02; Failure to Evaluate Gas Transport through Modified Piping Configuration)

.5 Operating Experience

a. Inspection Scope

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

Information Notice 86-14, Supplement 2: Overspeed Trips of AFW, HPCI and RCIC Turbines; and

Part 21 No. 2013-09-00, Wedge Pin Failure of an Anchor/Darling Double-Disc Gate Valve at Browns Ferry Nuclear Plant Unit 1.

b. Findings

No findings were identified.

.6 Operating Procedure Accident Scenarios

a. Inspection Scope

The inspectors performed a detailed reviewed of the procedures listed below associated with some of the selected inspection samples. For the procedures listed, time operator actions were reviewed for reasonableness and any interfaces with other departments were evaluated. The procedures were compared to UFSAR descriptions, design assumptions, and training materials to assure for consistency. The following operating procedures were reviewed in detail:

SEP 301.1, Torus Vent via SBGT, Revision 10;

OI 149 QRC2, Torus Cooling initiation, Revision 9;

OI 149 QRC3, Manual LPCI initiation, Revision 5;

OI 416, RHR Service Water System, Revision 67;

OI 416 QRC1, RHRSW Rapid Start, Revision 7; and

STP 3.5.1-14A, A CS fill test, Revision 10.

In addition, operator actions were observed during the performance of a small break loss of coolant accident (LOCA) concurrent with a loss of off-site power on the station simulator. For the selected operator actions, the inspectors performed a margin assessment and detailed review of the operator actions listed below. Where possible, margins were determined by the review of the assumed design basis and UFSAR response times and performance times noted during the small break LOCA scenario observed on the station simulator. The following operator actions were reviewed:

Rapid reactor pressure vessel depressurization following a small break LOCA;

RHR system alignment following a small break LOCA;

RHRSW system alignment following a small break LOCA;

Drywell spray initiation following a small break LOCA; and

RHRSW strainer hi-differential pressure.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Review of Items Entered Into the Corrective Action Program

a. Inspection Scope

The inspectors reviewed one sample of the selected component problems identified by the licensee and entered into the CAP. The inspectors reviewed the 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 CAP. The specific corrective action documents sampled and reviewed by the inspectors are listed in the attachment to this report.

The inspectors also selected one issue identified during previous NRC 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 issue was reviewed:

NCV 05000331/2014008-03, Failure to Include Minimum Required System Voltage as an Acceptance Criterion in the 125 VDC Station Battery Surveillances Test Procedures.

b. Findings

No findings were identified.

4OA6 Management Meeting(s)

Exit Meeting Summary

On May 19, 2017, the inspectors presented the inspection results to Mr. D. Curtland, and other members of the licensee staff. The licensee acknowledged the issues presented.

The inspectors confirmed that none of the potential report input discussed was considered proprietary. Several documents reviewed by the inspectors were considered proprietary information and were either returned to the licensee or controlled in accordance with NRC policy on proprietary information.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

D. Curtland, Site Director

P. Hanson, Site Engineering Director

M. Davis, Licensing Manager

T. Erger, Operations Shift Manager

T. Below, Electrical Design Engineer

C. Catino, Fleet Corporate MOV Program Engineer

Z. Cloe, Mechanical Design Engineer

P. Collingsworth, System Engineer

B. Hendrickson, System Engineer

S. Huebsch, Design Engineering Supervisor

J. Kuehl, MOV Program Engineer

G. Migliuolo, System Engineer

M. Mills, Fleet Corporate MOV Program Engineer

B. Murrell, Licensing Senior Engineer

T. Weaver, Senior Licensing Engineer

U.S. Nuclear Regulatory Commission

C. Norton, Senior Resident Inspector

J. Steffes, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000331/2017008-01 NCV Failure to Evaluate Effect of Crediting Bypass Line As Safety Related Flowpath (Section 1R21.3.b(1))

05000331/2017008-02 NCV Failure to Evaluate Gas Transport through Modified Piping Configuration (Section 1R21.4.b(1))

Discussed

None.