IR 05000254/2017007

NRC Design Bases Assurance Inspection (Teams): Inspection Report 05000254/2017007; 05000265/2017007 (DRS-M.Jones)

ML18012A450
Person / Time
Site:	Quad Cities
Issue date:	01/12/2018
From:	Jeffers M NRC/RGN-III/DRS/EB2
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
IR 2017007
Download: ML18012A450 (19)

Text

UNITED STATES uary 12, 2018

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2- NRC DESIGN BASES ASSURANCE INSPECTION (TEAMS): INSPECTION REPORT 05000254/2017007; 05000265/2017007

Dear Mr. Hansen:

On December 28, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed a Triennial Baseline Design Bases Assurance Inspection (Teams) at your Quad Cities Nuclear Power Station. The enclosed report documents the results of this inspection, which were discussed on December 28, 2018, with Mr. Humphrey, and other members of your staff.

Based on the results of this inspection, no violations of significance were identified.

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. 50-254, 50-265 License Nos. DPR-29;` DPR-30

Enclosure:

IR 05000254/2017007; 05000265/2017007

REGION III==

Docket No: 50-254; 50-265 License No: DPR-29; DPR-30 Report No: 05000254/2017007; 05000265/2017007 Licensee: Exelon Generating Facility: Quad Cities Nuclear Power Station Location: Cordova, IL Dates: November 13-December 28, 2017 Inspectors: M. Jones, Engineering Inspector, Lead A. Dunlop, Senior Engineering Inspector, Mechanical I. Hafeez, Engineering Inspector, Electrical D. Betancourt, Operations Inspector H. Leake, Electrical Contractor W. Sherbin, Mechanical Contractor Approved by: M. Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY

Inspection Report 05000254/2017007; 05000265/2017007, 11/13/2017-12/01/2017; Quad

Cities Nuclear Power Station; Design Bases Assurance Inspection (Teams).

The inspection was a 2-week onsite baseline inspection that focused on the design of components. The inspection was conducted by regional engineering inspectors and two consultants. No findings of significance were identified by the inspectors. The U.S. Nuclear Regulatory Commissions 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 No findings were identified during this inspection.

Licensee-Identified Violations

No findings were identified during this inspection.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

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 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 Quad Cities Nuclear Power Station 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 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 11 samples (5 components, with 1 component associated with Large Early Release Frequency implications, 4 modifications, and 2 operating experience) 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, 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 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 five components (samples), including a component with Large Early Release Frequency were reviewed:

Unit 1/2 Reactor Building Closed Cooling Water Pump (1/2-3701): The team inspected the performance of reactor building closed cooling water (RBCCW)pump 1/2-3701 and the associated potential impact on plant operations (failure of the pumps could lead to a plant transient). The inspection included interviews with system and design engineers and operators, system walkdowns; and reviews of drawings, and normal, alarm response, and abnormal plant procedures. This review focused on the RBCCW systems response to a postulated automatic initiation of stand-by pump 1/2-3701 due to a discharge header low-pressure signal; and operator actions following this initiation. The team reviewed plant procedures to determine whether the operator actions were acceptable to assure reliable operation of the RBCCW system. Additionally, the inspectors reviewed electrical drawings, including one-lines and schematics to verify consistency with UFSAR descriptions and engineering analyses. Loading and voltage calculations were reviewed for pump operation on both offsite and onsite power sources (emergency diesel generators) to verify the adequacy of the motor power supplies. Finally, the team reviewed condition reports, maintenance history, and system health reports to determine the overall health of the pump, and to determine if issues entered into the Corrective Action Program were properly addressed.

Unit 1/2 Emergency Diesel Generator (EDG) Ventilation Fan (1/2-5727): The team reviewed the calculations related to EDG room supply air ventilation requirements, and compared the calculated airflow requirements with fan test data to ensure adequate heat removal capability. The team reviewed failure positions of pneumatic louver operators in the ventilation enclosures to ensure louvers will open on a loss of instrument air. The team also reviewed the control and wiring diagrams for the starting and stopping of the ventilation fan.

Preventive maintenance activities for lubricating the ventilation exhaust fan motor and fan shaft bearings were also reviewed to ensure vendor recommended preventive maintenance activities were being performed. The inspection included interviews with system and design engineers and operators, system walkdowns; and reviews of drawings, and alarm response procedures.

125 Vdc Distribution Panels: The team reviewed load flow calculations to determine whether the panels were applied within their required current ratings. The team reviewed voltage drop calculations to determine whether loads had their required minimum voltage and whether they were applied within their maximum voltage rating during battery equalizing. The team reviewed short circuit and protective device calculations to determine whether equipment was adequately protected and immune from spurious tripping. The team reviewed maintenance schedules, procedures, and maintenance records, including circuit breaker test requirements, to determine whether the panels and their associated circuit breakers were being properly maintained. In addition, the team performed a visual inspection of the 125Vdc Distribution Panels to assess material condition and the presence of hazards.

Alternating Current (AC) Bus Supplying Power to Residual Heat Removal (RHR)

Pumps 1A and 1B (4160V Switchgear 13-1): The team reviewed one-line diagrams, drawings, calculations of loading, short circuit, voltage drop, and protective relay trip setpoints to verify the capability of the switchgear to adequately supply the essential loads when powered by the unit auxiliary transformer, reserve auxiliary transformer, or EDG. The team also verified the maximum short circuit current available at the bus was within the interrupting capacity of the feeder breakers. The team reviewed the fast transfer design of the switchgear from the unit auxiliary transformer (main generator source) to the reserve auxiliary transformer (offsite power source) when the main generator trips. The team verified the feeder cable size and ampacity for the RHR pumps was adequate to carry the maximum load current. Administrative controls were reviewed for mitigating potential conductor and EDG overload conditions identified in the load flow and EDG sizing calculations. The inspectors performed a walkdown of 4160V Switchgear 13-1 to observe its material condition.

Main Steam Isolation Valves (MSIV) (1-203-001(A-D), 1-203-002(A-D)): The team reviewed the design basis of the inboard and outboard MSIVs for Unit 1, the basis for its closure time requirement, and the associated control logic. The team reviewed operating procedures associated with the MSIVs under normal and accident conditions. The air accumulator leakage limits, leak test procedures, air quality, and recent test results were reviewed to verify acceptance criteria were met and performance degradation would be identified.

The team reviewed closure time surveillance procedures and recent results to verify that the test results were representative of the most limiting postulated accident conditions. The (a)(1) action plan for the MSIV timing issues was reviewed to verify the cause(s) were identified and corrective actions were planned or implemented to resolve the timing issue. The team reviewed the testing of the control circuits required to close the MSIVs to ensure that the testing was comprehensive.

.4 Findings

No findings were identified.

.5 Mitigating System Modifications

a. Inspection Scope

The team reviewed 4 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 Unit 1/2 EDG ventilation fan, 1B RHR pump seal Cooler, 4160V switchgear bus 13-1, 125 Vdc distribution panels, and 125 Vdc normal and alternate batteries. The team 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 team 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) 398602, Replace the 1B RHR Pump Seal Cooler; EC 395167, Install Close Torque Switch Bypass Mod in 1-1001-29A to Increase Margin; EC 398044 Revision 1, Unit 1 4kv Bus Transfer Logic Modification for an Open Phase Event Concurrent with a LOCA; and EC 932979, U1 250 VDC MCC Cubicle Bucket Replacement.

b. Findings

No findings were identified.

.6 Operating Experience

a. Inspection Scope

The inspectors reviewed 2 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 2015-13, Main Steam Isolation Valve Failure Events; and Generic Letter 2006-02, Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power.

b. Findings

No findings were identified.

.7 Operating Procedure Accident Scenarios

a. Inspection Scope

The team performed a detailed reviewed of selected procedures associated with the inspections component samples. For the procedures listed time critical operator actions were reviewed for reasonableness, in plant action were walked down with a licensed operator, and any interfaces with other departments were evaluated. The procedures were compared to UFSAR, design assumptions, and training materials to assess their consistency. In addition, operator actions were observed at the stations simulator for two scenarios: Anticipated Transient with a Scram, and Loss of AC to the 125 Volt Direct Current Battery Chargers with a simultaneous loss of auxiliary electrical Power.

The following operating procedures were reviewed in detail:

QGA 101, Reactor Pressure Vessel Control (ATWS);

QCOA 6900-07, Loss of AC Power to 125 VDC Battery Charger with Simultaneous Loss of Auxiliary Electrical Power; QCOP 3700-02, RBCCW System Startup and Operation; QCOA 3700-06, RBCCW Line Break Inside Containment; QCOP 1000-30, Post Accident RHR Operation; and QGOA 6100-03, Loss of Offsite Power.

The inspectors performed a margin assessment and detailed review of four risk-significant and/or time critical operator actions. These actions were selected from the licensees PRA rankings of human action importance based on risk achievement worth values, and where possible, margins were determined by the review of the assumed design basis and UFSAR response times and performance times documented by job performance measures results. For the selected operator actions, the inspectors performed a detailed review and walk through of associated procedures, including observing the performance of some actions in the stations simulator and in the plant for other actions, with an appropriate plant operator to assess operator knowledge level, adequacy of procedures, and availability of special equipment where required.

The following operator actions were reviewed:

Initiation of Torus Cooling during Appendix R scenarios; Manual Start of RHR Containment Cooling Mode of RHR; Initiate Drywell Spray during an Anticipate Transient without Scram; and Load Shed 125 VDC Loads 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 team reviewed a sample of the selected component problems identified by the licensee and entered into the corrective action program. The team 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 team 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:

Non-Cited Violation 5000254/265/2011009-02; Failure to Perform Required In-Service Testing of Shutdown Cooling Suction Valves; and Non-Cited Violation 05000254/2016008-01, Failure to Provide Appropriate Operating Instructions for Aligning a Battery Charger to the Station Black-Out Diesel Generator.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Interim Exit Meeting Summary

On December 1, 2017, the inspectors presented the inspection results to Mr. Kenneth S. Ohr, 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.

Any documents reviewed by the inspectors that were considered proprietary information were either returned to the licensee or handled in accordance with NRC policy on proprietary information. The team had outstanding questions that required additional review and a following exit meeting.

.2 Exit Meeting Summary

On December 28, 2017, the team presented the inspection results to Mr. M. Humphrey and other members of the licensee staff. The licensee acknowledged the issues presented. The team asked the licensee weather any materials examined during the inspection should be considered proprietary. Several documents reviewed by the team 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

K. Ohr, Site Vice President

M. Humphrey, Regulatory Assurance

R. Swart, Engineering Supervisor

T. Bell, Engineering Director

J. Bries, Operations Director

J. Cox, Operations Supervisor

U.S. Nuclear Regulatory Commission

R. Murray, Senior Resident Inspector

K. Carrington, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened, Closed and

Discussed

None LIST OF ACRONYMS USED AC Alternating Current EC Engineering Change EDG Emergency Diesel Generator MSIV Main Steam Isolation Valve NRC U.S. Nuclear Regulatory Commission PRA Probabilistic Risk Assessment RBCCW Reactor Building Closed Cooling Water UFSAR Updated Final Safety Analysis Report