IR 05000266/2017007

NRC Design Bases Assurance Inspection (Teams): Inspection Report 05000266/2017007; 05000301/2017007

ML17222A556
Person / Time
Site:	Point Beach
Issue date:	08/09/2017
From:	Jeffers M NRC/RGN-III/DRS/EB2
To:	Coffey R Point Beach
References
IR 2017007
Preceding documents:	ML17088A724 (29 March 2017)
Download: ML17222A556 (29)

Text

UNITED STATES ust 9, 2017

SUBJECT:

POINT BEACH NUCLEAR PLANTNRC DESIGN BASES ASSURANCE INSPECTION (TEAMS): INSPECTION REPORT 05000266/2017007; 05000301/2017007

Dear Mr. Coffey:

On July 13, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed a Triennial Baseline Design Bases Assurance Inspection (Teams) at your Point Beach Nuclear Plant.

The enclosed report documents the results of this inspection, which were discussed on July 13, 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 violations 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 Point Beach Nuclear Plant. 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 Nos: 50-266; 50-301 License Nos: DPR-24; DPR-27 Enclosure:

IR 05000266/2017007; 05000301/2017007 cc: Distribution via LISTSERV

SUMMARY

Inspection Report 05000266/2017007; 05000301/2017007, 06/26/2017 - 07/13/2017; Point

Beach Nuclear Plant; Design Bases Assurance Inspection (Teams).

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 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 April 29, 2015. Cross-cutting aspects are determined using IMC 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6, dated July 201

NRC-Identified

and Self-Revealed Findings Cornerstones: Mitigating Systems

Green.

The NRC identified a finding of very-low safety significance (Green) and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B,

Criterion XVI, Corrective Action, for the licensee failure to correct a Condition Adverse to Quality (CAQ) associated with a seismic piping interaction affecting the Motor Driven Auxiliary Feedwater (MDAFW) system. Specifically, the licensee identified a flange clearance to the Unit 1 MDAFW suction piping was nonconforming and captured it in the Corrective Action Program (CAP) as Action Request (AR) 01684524. However, the licensee closed the AR without correcting the CAQ. The licensee captured the inspectors concern in the CAP as AR 02212810 and performed an evaluation that reasonably concluded the MDAFW remained operable.

The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors 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 screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability determination which concluded the stresses resulting from the seismic interaction would reasonably be bounded by the applicable stress operability limits. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance because the performance deficiency occurred more than 3 years ago. Specifically, the licensee closed AR 01684524 without correcting this CAQ on September 20, 2011. (Section 1R21.3.b(1))

Cornerstones: Initiating Events and Mitigating Systems

Green.

The inspectors identified a finding of very-low safety significance (Green), and an associated (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify a condition adverse to quality. Specifically, after receiving and reviewing the Flowserve 10 CFR Part 21 report, the licensee misunderstood the information provided and failed to identify 36 safety-related valves that were nonconforming. Of these 36 valves, 14 were identified as being susceptible to pin failure based on their torque setting, 6 of which had open or close safety functions.

The licensee captured the inspectors concern in the CAP as AR 02212531, and AR 02212915. In addition, the licensee performed operability evaluations that concluded the affected valves remained operable.

The performance deficiency was more-than-minor because it was associated with the equipment performance attribute of the Mitigating System and Initiating Event cornerstones, and adversely affected the cornerstone individual objectives. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding screened as of very-low safety significance (Green) by answering No to the questions contained in Exhibit 1, and in accordance with Exhibit 2, it did not result in the loss of operability or functionality of mitigating systems. The team did not identify a cross-cutting aspect associated with this finding because the most significant cause for the error was not reflective of current performance. Specifically, the Part 21 report and associated review by the licensee occurred in February 2013. (Section 1R21.5.b(1))

REPORT DETAILS

1R21 Design Bases Assurance Inspection (Teams)

.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 (SSC) 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 (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 selected risk-significant components and operator actions for review using information contained in the licensees PRA and the Point Beach Nuclear Plant 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 14 samples (5 components, 1 component with Large Early Release Frequency implications, 5 modifications, and 3 operating experience) as defined in Inspection Procedure 71111.21M-02.01.

.3 Component Design

a. Inspection Scope

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

For each of the components selected, the inspectors reviewed the FSAR, Applicable Technical Specifications, 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 6 components (samples) were reviewed:

• 125 Volts Direct Current Bus (D-03) and Associated Distribution Panel (D-31):

The inspectors reviewed 125 Volts Direct Current (Vdc) short circuit calculations and verified the interrupting ratings of the fuses were above the calculated short circuit currents. The 125 Vdc voltage drop 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 reviewed the motor control logic diagrams and the 125 Vdc voltage drop calculation to ensure adequate voltage would be available for the control circuit components under all design basis conditions. The inspectors also reviewed the 125 Vdc short circuit and coordination calculations to assure coordination between various feed breaker open and close control circuit fuses, and 125 Vdc supply fuses and to verify the interrupting ratings of the control circuit fuses and the 125 Vdc control power feed fuses.

• Unit 1 Yellow Instrument Bus Inverter (1DY04): The team 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 including ratings and branch circuit cabling with the upstream protective device. The inspectors reviewed the physical and material condition by visual inspection and reviewed health report documents to verify identification of adverse trends. The inspectors also reviewed voltage drop and minimum voltage calculations. The calculation review verified methodology, design inputs, assumptions, and results.

• Motor-Driven Auxiliary Feedwater Pump (1P-53): The team reviewed the following hydraulic calculations to assess the pump capability to perform its required mitigating functions: hydraulic model, pump minimum required flow, minimum required net positive suction head and vortexing. In addition, the team reviewed measures established by the licensee to manage the potential for steam void formation due to feedwater back-leakage into the pump discharge piping; the sizing of the cavitating venturi; and the manual and automatic pump suction switchover from the condensate storage tank to the service water (SW)system. The team also reviewed test procedures and completed tests, including pump inservice testing, to assess the associated methodology, acceptance criteria, and test results. In addition, the team interviewed licensee personnel from multiple disciplines such as operations, engineering, and maintenance.

The team also reviewed electrical calculations, drawings and equipment modifications to determine whether adequate voltage and current would be available at the pump motor terminals for starting and running under worst case design basis loading, operation on emergency power, and grid voltage conditions. Control Logic was also reviewed for the appropriate operation.

Protective relay settings and cable short circuit current capability were also reviewed as part of the electrical review to determine whether appropriate electrical protection coordination margins had been applied and whether the power supply feeder cables had been properly sized for the maximum available short circuit current capability requirements. The team reviewed modifications made to support the replacement of the pump.

• Minimum Recirculation Check Valve for the 1P-53 Motor-Driven Auxiliary Feedwater Pump (1AF-196): The teams review included installed configuration, system operation, detailed design, system Inservice Testing (IST), and operating experience. In addition, the team interviewed licensee personnel from multiple disciplines such as operations, engineering, and maintenance.

• Reactor Makeup Water to Pressurizer Relief Tank Containment Isolation Valve (1RC-508): The team inspected the component to verify that it was capable of meeting its design basis requirements. The team inspected the component, to determine if the normally closed valve is capable of performing its design basis function to isolate its associated containment penetration (P 30c) when it is opened for the reactor water makeup function. The team reviewed IST stroke time testing, Appendix J testing, periodic Air-Operated Valve (AOV) diagnostic test results to verify acceptance criteria were met. The team evaluated whether the AOV safety functions, performance capability, and design margins were adequately monitored and maintained in accordance with Nexteras AOV Program requirements.

The design, operation, and maintenance of the valve were discussed with the system engineer to evaluate the valves performance history, maintenance, and overall health. The team also conducted a walkdown of the valve and associated equipment to assess the material condition of the equipment and to evaluate whether the installed configuration was consistent with the plant drawings, procedures, and the design bases.

The inspectors also reviewed power and control wiring to this solenoid valve to verify circuit protection fuse size and type. The inspectors verified from the vendor manual the minimum voltage required for this solenoid and compared it to the 125 Vdc voltage drop calculation and verified that sufficient voltage will be available to this valve under design basis conditions. The inspectors also reviewed the 125 Vdc short circuit calculation to verify the calculated short circuit current at this valve is well within the interrupting capacity of the power supply fuse. The inspectors also verified that the licensee replaces this type of solenoid valves every fifth refueling outage for service life concerns.

• Service Water Pump (P-32F): The team inspected the F motor-driven SW pump to verify that it was capable of meeting its design basis requirements. The SW pump safety related function is to provide Lake Michigan cooling water flow to mitigate the consequences of a Loss of Coolant accident in one unit while supporting normal cooling flow to the unaffected unit. The team reviewed system drawings, analyses, procedures, recent modifications, and test results associated with operation of the SW pump under postulated transient, accident, and station blackout conditions. The analyses included considerations for hydraulic performance, hydraulic instability, net positive suction head, required total developed head, pump vibration analysis (shaft critical speed and column natural frequency), pump run-out conditions, operable but non-conforming condition, and potential for vortex formation at the suction source. Seismic design documentation was reviewed to verify pump design was consistent with limiting seismic conditions. The team also evaluated the pump suction alarm setpoint to verify that it had an adequate basis. The IST results were reviewed to verify acceptance criteria were met and performance degradation would be identified, taking into account set-point tolerances and instrument inaccuracies.

Additionally, IN 2007-06 was reviewed for applicability and corrective actions taken. The team interviewed system, test, and design engineers to discuss pump performance and maintenance history to determine the overall condition of the pump.

The team reviewed electrical calculations, drawings and equipment modifications to determine whether adequate voltage and current would be available at the pump motor terminals for starting and running under worst case design basis loading, operation on emergency power, and grid voltage conditions. Control Logic was also reviewed for the appropriate operation. Protective relay settings and cable short circuit current capability were also reviewed as part of the electrical review to determine whether appropriate electrical protection coordination margins had been applied and whether the power supply feeder cables had been properly sized for the maximum available short circuit current capability requirements.

b. Findings

(1) Failure to Correct a Condition Adverse to Quality Associated with a Seismic Interaction of the Motor-Driven Auxiliary Feedwater Piping

Introduction:

The team 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 XVI, Corrective Action, for the licensee failure to correct a Condition Adverse to Quality (CAQ) associated with a seismic piping interaction affecting the Motor Driven Auxiliary Feedwater (MDAFW) system.

Specifically, the licensee identified a flange clearance to the Unit 1 MDAFW suction piping was nonconforming and captured it in the CAP as Action Request (AR) 01684524. However, the licensee closed the AR without correcting the CAQ.

Description:

On September 7, 2011, the licensee identified a small clearance between a blind flange and the MDAFW common suction piping from the condensate storage tanks (CSTs). Specifically, a clearance of approximately 0.2 inches existed between the blind flange located downstream of the unit cross-tie drain/FLEX valve (i.e., valve AF-201)and the piping section upstream of the Unit 1 MDAFW CST supply isolation valve (i.e., valve 1AF-190). The licensee captured this condition in their CAP as AR 01684524 as a CAQ.

On June 28, 2017, the team noted the small clearance during a walkdown and subsequently learned that AR 01684524 was closed without correcting this CAQ on September 20, 2011. As a result of the team questions, the licensee determined that the potential existed for the blind flange to impact the MDAFW suction piping. Specifically, the piping associated with the blind flange would have had a maximum displacement of 2.2 inches during a safe shutdown earthquake. However, the associated impact force had not been evaluated as required by the construction code of the MDAFW piping.

Specifically, Section 101.5.1, Impact, of the 1967 Edition of the USAS B31.1 code stated, Impact forces caused by all external and internal conditions shall be considered in the piping design. The team was concerned because the failure of the pressure boundary at this location would challenge the Units 1 and 2 MDAFW capability to switchover their suction to their safety-related water supply and would drain the CSTs into locations containing safety-related equipment.

The licensee captured the inspectors concern in the CAP as AR 02212810. The immediate corrective action was to perform an evaluation that reasonably concluded the MDAFW remained operable. Specifically, the licensee reasonably estimated the total impact stresses would be less than the applicable operability limits. The proposed corrective action to restore compliance at the time of this inspection was to modify the piping to eliminate the seismic interaction.

Analysis:

The team determined that the failure to correct a CAQ associated with a seismic piping interaction affecting the MDAFW system was contrary to 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, 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 protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct the unanalyzed pipe clearance did not ensure MDAFW system would remain available and capable to provide its accident mitigating function.

The team 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 7, 2016. Because the finding impacted the Mitigating Systems cornerstone, the inspectors screened the finding through IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued on June 19, 2012, using Exhibit 2, Mitigating Systems Screening Questions. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability determination which concluded the stresses resulting from the seismic interaction would reasonably be bounded by the applicable stress operability limits.

The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance because the performance deficiency occurred more than 3 years ago. Specifically, the licensee closed AR 01684524 without correcting this CAQ on September 20, 2011.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, that measures shall be established to assure that CAQs, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances, are promptly identified and corrected.

Contrary to the above, from September 7, 2011, to at least June 28, 2017, the licensee failed to correct a CAQ. Specifically, on September 7, 2011, the licensee identified a flange clearance associated with the Unit 1 MDAFW suction piping was nonconforming and captured it in the CAP as AR 01684524. However, the licensee closed AR 01684524 on September 20, 2011, without correcting this CAQ.

The licensee is still evaluating the planned corrective actions. However, the inspectors determined that the continued non-compliance does not present an immediate safety concern because the licensee performed an evaluation that reasonably concluded the MDAFW system remained operable.

Because this violation was of very-low safety significance and was entered into the licensees CAP as AR 02212810, this violation is being treated as a NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000266/2017007-01; NCV 05000301/2017007-01, Failure to Correct a Condition Adverse to Quality Associated with a Seismic Interaction of the Motor-Driven Auxiliary Feedwater Piping)

.4 Mitigating System Modifications

a. Inspection Scope

The inspectors reviewed 5 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 Auxiliary Feedwater (AFW) System, SW System and station batteries D-05 and D-06. 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:

• Engineering Change (EC)EC 0281270 Revision 2, Addition of 60th Cell to Batteries D-05 and D-06;

• EC 0257927 Revision 1, Replace D-06 Station Battery;

• EC 0258407 Revision 0, Auxiliary Switch Wiring Configuration of SCI Inverter Breaker B2;

• EC 259830 Revision 5, EPU - Replacement MDAFW Pump and Piping Installation Only; and

• EC 0272153 Revision 19, SW System Vertical Pump Replacements.

b. Findings

No findings were identified.

.5 Operating Experience

a. Inspection Scope

The inspectors reviewed 3 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:

• Part 21 Report No. 2013-09-00, Wedge Pin Failure in Anchor Darling Motor Operated Double Disc Gate Valve at Browns Ferry Nuclear Plant Unit 1;

• IN 2000-06, Offsite Power Voltage Inadequacies and

• IN 2016-01, Commercial Grade Dedication of ABB Relays.

b. Findings

(1) Failure to Identify Non-Conforming Conditions after Receipt of Anchor Darling Double Disc Gate Valve Related Part 21 Report

Introduction:

The inspectors identified a finding of very-low safety significance (Green),and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify a condition adverse to quality. Specifically, after receiving and reviewing a Flowserve 10 CFR Part 21 report, the licensee misunderstood the information provided and failed to identify 36 safety-related valves that were nonconforming. Of these 36 valves 14 were identified as being susceptible to pin failure based on their torque setting, six of which had open or close safety functions.

Description:

On January 4, 2013, Tennessee Valley Authority issued a 10 CFR Part 21 report following the failure of an Anchor Darling Double Disk Gate Valve reactor side disc to seat properly due to a broken wedge pin. The report described that the cause of the pin failure was determined to be that the valve stem was not adequately torqued into the upper wedge when assembled by the vendor.

On February 25, 2013, the valve vendor, Flowserve, issued a separate but related 10 CFR Part 21 report for the same Tennessee Valley Authority valve failure. Flowserve described that the pin was designed to ensure that the joint does not loosen due to vibration and other secondary loads and that the pin shearing can allow stem rotation and ultimately result in stem to wedge separation. Similarly, the report identified that the wedge pin failed due to excessive load on the pin because the system operating torque had exceeded the unknown preassembly torque to tighten the stem into the upper wedge combined with the wedge pin material strength. Attachment 1 within this 10 CFR Part 21 report, provided a recommended standard stem preload torque based upon valve size and pressure class. Point Beach was listed in the Part 21 report as a customer for the parts and was sent a copy of the 10 CFR Part 21 report from Flowserve.

On March 4, 2013, the licensee entered this issue into their CAP as AR 01853370. The licensee determined that 44 valves were in-service and that 36 out of the 44 valves were safety-related (Reference: CMP 2.2.7 Engineering Instructions for Performing Valve Operator Checkouts, Revision 14). The licensee evaluated the issue by assuming that the stem was pre-torqued into the wedge at the recommended Attachment 1 value in the Flowserve 10 CFR Part 21 report. Using this pre-load value, the licensee concluded that the maximum operating torque values had not and would not break the wedge pin during valve operation. Therefore, all 36 safety-related valves were dispositioned as conforming to the design requirements and, therefore, no corrective actions were necessary.

The team identified that the licensee had misunderstood the 2013 10 CFR Part 21 report. Specifically, the licensee used the Flowserve Part 21 report Attachment 1 recommended pre-torque table as the known assembly torque instead of the recommended torque to correct the issue. The licensee entered this issue into their CAP as AR 02212531, and AR 02212915. The licensee re-evaluated the issue and identified that 14 of the 44 valves had a maximum operating torque that could challenge the minimum strength of the wedge pin. This issue could lead to overstressing and breaking the wedge pin during valve operations and ultimately cause valve failure.

The 6 of the 14 valves had open or close safety functions. The licensee performed an operability evaluations for these 14 valves and concluded that the valves could perform their specified function when providing an allowance for stem to wedge thread friction when thrust was applied.

Analysis:

The inspectors determined that the failure to identify the 36 safety-related valves with the non-conforming condition, after the receipt of the Flowserve 10 CFR Part 21 report, was a performance deficiency. Specifically, the licensee misunderstood the 10 CFR Part 21 report following receipt and evaluation although the report adequately described the issue and provided the licensee with specific contact information if any licensee had questions or confusion.

The inspectors evaluated if the performance deficiency was more-than-minor by reviewing the safety-related and any nonsafety-related function(s) associated with the 44 valves and reviewing those valves (14 total) in which the operating torque exceeded the minimum wedge pin material strength. The team focused their review on two groups of valves. Specifically:

• Low head safety injection pump suction from the reactor coolant system (RCS)

[Unit 1 and Unit 2 RH-700 trains; two valves total]; and the

• Safety injection accumulator isolation [Unit 1 and Unit 2 SI-841(A) and (B) trains; four valves total].

The inspectors determined that the performance deficiency was more-than-minor because it was associated with the equipment performance attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective to ensure the availability, capability, and reliability of systems that respond to initiating events to prevent core damage. Specifically, the SI-841 valves have a safety function to remain open during a loss of coolant accident to inject highly concentrated boric acid into the RCS. The SI-841 valves are shut in accordance with emergency operating procedures to prevent nitrogen gas from unnecessarily entering the RCS. In addition, the RH-700 valves perform an active safety-related function to close. Specifically, the RH-700 valves must be capable of closure, if open, to allow Residual Heat Removal to be realigned to the Emergency Core Cooling System mode of operation. Also, low head safety injection pump suction from the RCS valves RH-700 perform a nonsafety-related risk-significant mitigating function to place residual heat removal in the RCS cooling mode to reach cold shutdown conditions.

Additionally, the inspectors determined that the performance deficiency was more-than-minor because it was associated with the equipment performance attribute for the Initiating Event cornerstone and adversely 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, RH-700 has a safety function in the closed position to preserve the pressure integrity of the RCS as one of two pressure isolation valves.

The inspectors determined that the finding could be evaluated using the Significance Determination Process in accordance with IMC 0609, Significance Determination Process, dated April 29, 2015, and Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016. Because this finding was associated with the Initiating Event and Mitigating Systems cornerstones, the inspectors screened the finding through IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power, using Exhibits 1, Initiating Events Screening Questions and Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very-low safety significance (Green) by answering No to all the questions contained in Exhibit 1 and answering Yes to question A.1 of Exhibit 2, because the finding was a deficiency that affected the design of mitigating SSCs, however the SSCs maintained their operability and functionality. Specifically, the licensee performed operability evaluations for the affected SSCs and concluded these remained operable.

The team did not identify a cross-cutting aspect associated with this finding because the most significant cause for the error was not reflective of current performance.

Specifically, the Part 21 report and associated review by the licensee occurred in February 2013.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified.

Contrary to the above, from March 4, 2013, through June 27, 2017, the licensee failed to identify a condition adverse to quality. Specifically, on March 4, 2013 after receiving and reviewing the Flowserve 10 CFR Part 21 report, the licensee misunderstood the information provided and failed to identify 36 safety-related valves that were nonconforming to 10 CFR Part 50, Appendix B, Quality Assurance requirements. Of these 36 valves, 14 were identified as being susceptible to pin failure based on their torque setting, 6 of which had open or close safety functions.

The licensee is evaluating planned corrective actions. However, the inspectors determined that the continued non-compliance does not present an immediate safety concern because the licensee performed an evaluation that reasonably concluded the affected valves remained operable.

Because this violation was of very-low safety significance and was entered into the licensees CAP as AR 02212531 and AR 02212915, this violation is being treated as a NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy.

(NCV 05000266/2017007-02, NCV 05000301/2017007-02, Failure to Identify Non-Conforming Conditions after Receipt of Anchor Darling Double Disc Gate Valve Related Part 21 Report)

.6 Operating Procedure Accident Scenarios

a. Inspection Scope

The inspectors performed a detailed reviewed of the general operating procedures for the motor driven AFW pump, SW system, and safety-related station batteries.

Precautions, limitations, and instructions for operating the systems and components were compared to the FSAR, design assumptions, and training materials to determine if the equipment was being operated as assumed within the current licensing basis. In addition, the inspectors observed the station simulator modeling and expected operator response for these systems during the following accidents or transients scenarios:

• Anticipated Transient without a Scram;

• Loss of Feedwater; and

• Station Blackout.

The inspectors performed a margin assessment and detailed review of eight risk-significant, operator actions. These actions were selected from the licensees PRA rankings of human action importance based on risk achievement worth values and other factors. Where possible, margins were determined by the review of the assumed design basis and FSAR response times and performance times documented by the licensee.

For the selected operator actions, the inspectors performed a detailed review and walk through of associated procedures to determine if the selected actions could be performed as assumed within the current licensing basis. The following operator actions were reviewed:

• AFW Suction Swap from the Condensate Storage Tank to SW Supply;

• Locally Shut SW Pump Discharge Valve following SW Pump Trip;

• Isolate Zurn Strainer if not Intact;

• Manually Backwash SW Strainer;

• Manual Control of AFW Pump Discharge Valves;

• Take Local Manual Control of Motor Driven AFW Pump if Control Room is not Accessible;

• Gagged Open AFW Recirculation Valves Following a Loss of Instrument Air; and

• Restoration of Safety-Related Battery Charger Following Loss of AC Power.

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

The inspectors also selected two issues identified during previous Component Design Basis 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:

• NCV 05000266/2015008-01; 05000301/2015008-01, Failure to Promptly Correct Conditions Adverse to Quality Regarding Electrical Power Cable Sizing and Protection;

• NCV 05000266/2011009-02; 05000301/2011009-02, Failure to Incorporate Minimum AFW Flow Rate Requirements Into Emergency Procedures.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On July 13, 2017, the inspectors presented the inspection results to Mr. Coffey, 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

R. Coffey, Site Vice President

R. Craven, Plant General Manager

D. Shepherd, Design Engineering Manager

E. Schultz, Licensing Manager

D. Hofstra, Operations Supervisor

M. Rosseau, Design Engineering Supervisor

K. Locke, Licensing

U.S. Nuclear Regulatory Commission

T. Hartman, Senior Resident Inspector

K. Barclay, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000266/2017007-01, NCV Failure to Correct a Condition Adverse to Quality

05000301/2017007-01 Associated with a Seismic Interaction of the Motor-Driven Auxiliary Feedwater Piping

05000266/2017007-02, NCV Failure to Identify Non-Conforming Conditions after

05000301/2017007-02 Receipt of Anchor Darling Double Disc Gate Valve Related Part 21 Report

Discussed

None