IR 05000456/2018002

NRC Integrated Inspection Report 05000456/2018002 and 05000457/2018002

ML18219C178
Person / Time
Site:	Braidwood
Issue date:	08/07/2018
From:	Eric Duncan Region 3 Branch 3
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
IR 2018002
Download: ML18219C178 (32)

Text

First initial L UNITED STATES August 7, 2018

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2NRC INTEGRATED INSPECTION REPORT 05000456/2018002 AND 05000457/2018002

Dear Mr. Hanson:

On June 30, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Braidwood Station, Units 1 and 2. On July 12, 2018, the NRC inspectors discussed the results of this inspection with Site Vice President, Ms. M. Marchionda, and other members of your staff. The results of this inspection are documented in the enclosed report.

Based on the results of this inspection, the NRC has identified three issues that were evaluated under the risk significance determination process as having very low safety significance (green).

The NRC has also determined that two violations are associated with these issues. Because issue reports were initiated to address these issues, these violations are being treated as Non-Cited Violations (NCVs), consistent with Section 2.3.2 of the Enforcement Policy. These NCVs are described in the subject inspection report.

If you contest the violations or significance of these NCVs, 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 Inspectors Office at Braidwood Station. 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 Inspectors Office at Braidwood Station.

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/

Eric R. Duncan, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77 Enclosure:

IR 05000456/2018002; 05000457/2018002 cc: Distribution via ListServ

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring licensee performance by conducting an integrated quarterly inspection at the Braidwood Nuclear Station Units 1 and 2 in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. Findings and violations being considered in the NRCs assessment are summarized in the table below.

List of Findings and Violations Inadequate Detail in Maintenance Work Instructions Resulted in Failed Gearbox Oil Cooler Head Gasket and Inoperable 2B Auxiliary Feedwater Pump Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green NCV [H.9] - Training 71111.12 Systems05000457/2018002-01 Opened/Closed A self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to have adequate detail within their maintenance work instructions to enable proper reassembly of the 2B auxiliary feedwater (AF)pump gearbox oil cooler. Specifically, during the licensees 19th Unit 2 refueling outage in April 2017, the gearbox oil cooler closure head was reassembled following scheduled maintenance using an excessive amount of room temperature - vulcanizing silicone (RTV) on the joint and an insufficient amount of torque on the closure head bolting. As a result, on March 16, 2018, the closure head joint failed causing several hours of unplanned inoperability and unavailability for the 2B AF Pump.

Work Instruction Error Results in Reactor Coolant System Pressure Transient Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green FIN [H.3] - Change 71111.15 05000456/2018002-02 Management Opened/Closed A self-revealed finding of very low safety significance (i.e., Green) was identified due to the licensees failure to follow work instructions while performing a digital upgrade to plant control systems. Specifically, while performing maintenance on the volume control tank (VCT) level transmitter on April 10, 2018, maintenance personnel failed to properly track the steps being performed while simultaneously working on multiple packages. This resulted in the Unit 1 reactor coolant system (RCS) experiencing a pressure transient and the actuation of a VCT relief valve.

Inadequate Test Activity Coordination Results in Unintended Valve Actuation and Reactor Coolant System Pressure Drop Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green NCV [H.4] - 71111.19 05000456/2018002-03 Teamwork Opened/Closed A self-revealed finding of very low safety significance (i.e., Green) and an associated NCV of Technical Specification 5.4, Procedures, was identified for the licensees failure to have properly coordinated testing activities associated with redundant Unit 1 pressurizer pressure instruments in accordance with the stations procedural requirements governing such testing.

Specifically, during the licensees 20th Unit 1 refueling outage, on April 23, 2018, redundant pressurizer pressure instrumentation channels were inadvertently subjected to simultaneous testing activities. This resulted in the coincidence logic for both of the units pressurizer power-operated relief valves (PORVs) being satisfied and the PORVs opening to depressurize the RCS from approximately 345 pounds per square inch gauge (psig) to approximately 320 psig.

Additional Tracking Items None.

PLANT STATUS

Unit 1 Unit 1 began the inspection period at approximately 96 percent power and in planned power coastdown operations in preparation for its 20th refueling outage. On April 9, 2018, the reactor was shut down at just after midnight to begin the refueling outage. The unit was restarted for its 21st operating cycle on April 28, 2018, and the main electrical generator was synchronized to the grid on April 29, 2018.

At 11:24 a.m. on April 30, 2018, with the unit operating at approximately 48 percent power, the main turbine received a generator motoring trip due to a failed differential steam pressure instrument. An associated automatic reactor trip was also received as designed for the plant operating conditions. The licensee addressed the failed instrument and restarted the reactor at 9:50 p.m. on April 30, 2018. The main electrical generator was synchronized to the grid on May 1, 2018, and the unit reached full power operation on May 3, 2018.

At approximately 9:17 a.m. on June 4, 2018, the 1C Main Feedwater Pump tripped during a scheduled instrument relay test. Operators in the main control room manually tripped the reactor due to lowering steam generator water levels approximately 3 minutes later. Following corrective actions to address the issues that caused the event, the reactor was restarted on June 5, 2018, at approximately 5:08 a.m. The main electrical generator was synchronized to the grid at approximately 3:25 p.m. later that day, and the unit reached full power operation on June 6, 2018. With the exception of minor reductions in power to support scheduled testing activities or small load changes requested by the transmission system operator, the unit remained operating at or near full power for the remainder of the inspection period.

Unit 2 Unit 2 began the inspection period operating at full power. With the exception of minor reductions in power to support scheduled testing activities or small load changes requested by the transmission system operator, the unit remained operating at or near full power for the entire inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed plant status activities described in IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem Identification and Resolution. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

71111.01Adverse Weather Protection Summer Readiness

The inspectors evaluated summer readiness of offsite and alternate alternating current (AC)power systems during the weeks ending June 9 through June 30, 2018.

71111.04 - Equipment Alignment Partial System Alignment Verification

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) The spent fuel pool (SFP) cooling system with the Unit 1 reactor core offloaded into the SFP during the week ending April 21, 2018;

(2) The Unit 1 residual heat removal (RHR) system while reactor core reload was underway during the week ending April 17, 2018; and

(3) The 1B emergency diesel generator (EDG) with the 1A EDG out-of-service for maintenance during the week ending April 21, 2018.

Complete System Alignment Verification (1 Sample)

The inspectors evaluated system configurations during a complete walkdown of the Unit 1 auxiliary feedwater (AF) system during the weeks ending May 6 through June 30, 2018.

71111.05AQ - Fire Protection Annual/Quarterly Quarterly Inspection

The inspectors evaluated fire protection program implementation in the following selected areas:

(1) The Unit 1 Containment, Elevation 377 - 0 (Fire Zones 1.1-1 and 1.2-1) during the week ending April 14, 2018;

(2) The Unit 1 Containment, Elevation 401 - 0 (Fire Zones 1.1-1 and 1.2-1) during the week ending April 14, 2018;

(3) The Unit 1 Containment, Elevation 426 - 0 (Fire Zone 1.3-1) during the week ending April 14, 2018;

(4) The Unit 1 Auxiliary Building, Elevation 346 - 0 - North (Fire Zone 11.2-0) during the week ending April 21, 2018;

(5) The Unit 2 Auxiliary Building, Elevation 346 - 0 - South Center (Fire Zone 11.2-0)during the week ending April 21, 2018;

(6) The Unit 1 Containment, Elevation 412 - 0 (Fire Zones 1.1-1 and 1.2-1) during the week ending April 28, 2018; and

(7) The Unit 2 Auxiliary Building Basement, Elevation 330 - 0 (Fire Zone 11.1B-0),during the week ending May 26, 2018.

71111.07Heat Sink Performance Heat Sink

The inspectors evaluated the licensees planned inspection and testing activities for the Unit 1 component cooling water (CCW) heat exchanger from May 13 through May 26, 2018.

71111.08Inservice Inspection Activities

The inspectors assessed the effectiveness of the licensees programs for monitoring degradation of the reactor coolant system (RCS) boundary, risk-significant piping system boundaries, and the containment boundary by reviewing the following activities from April 9 through April 19, 2018:

(1) Ultrasonic testing (UT) of 4 pipe-elbow chemical volume control line weld; component ID 1CV-21-12;

(2) UT of 4 elbow-pipe chemical volume control line weld; component ID 1CV-21-11;

(3) UT of 6 elbow-pipe RCS weld; component ID 1RC-16-09;

(4) Liquid penetrant examination (PT) of embedded flaw repair on the reactor vessel head, Control Rod Drive (CRD) Nozzle 69;

(5) Bare Metal Visual (BMV) examinations of the reactor vessel upper head penetrations;

(6) Seal welding of reactor coolant pump (RCP) 1B seal injection inlet check valve 1CV8368B; WO 1816677-1;

(7) Welding for the replacement of 1A auxiliary feedwater (AF) pump suction valve 1AF006A; WO 1962980-1;

(8) Review of examination records with relevant indications accepted for continued service; dye penetrant test (PT) of embedded flaw repair for CRD Nozzle 69 (Issue Report (IR) 2723199) and PT of weld on line 1FW03DC-16 (IR 02730225);

(9) Review of boric acid evaluations and corrective action records for body-to-bonnet threaded connection on 1SI121B (IR 2731397), bolted connection for 1A safety injection (SI) pump discharge header main isolation valve (IR 2727539), and pipe cap downstream of valve 1SI045 (IR 3986087);

(10) Magnetic particle examination (MT) on reactor vessel head of component ID 1RV-03-001, head to flange weld;

(11) PT of reactor vessel head component ID 1RV-03-74; weld in peripheral CRD housing;

(12) UT on reactor vessel head head-upper disc weld; component ID 1RV-03-002, head to upper disc weld;

(13) Visual examination (VT-2) of steam generator (SG) A, B, C, and D inlet and outlet primary manway bolting; component IDs 1SG-05-SGB-01/02, 1SG-06-SGB-01/02, 1SG-07-SGB-01/02, and 1SG-08-SGB-01/02;

(14) UT of 4 elbow-elbow chemical volume control line weld; component ID 1CV-21-03;

(15) UT of 4 elbow-pipe chemical volume control line weld; component ID 1CV-21-04;

(16) UT of 3 pipe-elbow safety injection line weld; component ID 1SI-07-06;

(17) UT of 3 elbow-pipe safety injection line weld; component ID 1SI-07-07;

(18) UT)of 3 pipe-elbow safety injection line weld; component ID 1SI-07-08;

(19) UT of 3 pipe-pipe safety injection line weld; component ID 1SI-07-10;

(20) UT of 3 pipe-bend safety injection line weld; component ID 1SI-07-11;

(21) UT of 3 pipe-bend safety injection line weld; component ID 1SI-07-13;

(22) UT of 3 pipe-elbow safety injection line weld; component ID 1SI-07-14;

(23) UT of 8 pipe-elbow safety injection line weld; component ID 1SI-33-15;

(24) UT of 8 elbow-pipe safety injection line weld; component ID 1SI-33-16;

(25) UT of 8 pipe-elbow safety injection line weld; component ID 1SI-33-17; and

(26) UT of 8 elbow-pipe safety injection line weld; component ID 1SI-33-18.

71111.11Licensed Operator Requalification Program and Licensed Operator Performance Operator Requalification

The inspectors observed and evaluated a licensed operator graded simulator scenario on

===May 15, 2018.

Operator Performance (1 Sample)===

The inspectors observed and evaluated various control room activities during the weeks ending April 15 through May 5, 2018.

71111.12Maintenance Effectiveness Routine Maintenance Effectiveness

The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:

(1) Control room ventilation (VC) chiller oil loss issues and recovery during the weeks ending May 12 through June 30, 2018.

Quality Control (1 Sample)

The inspectors evaluated maintenance and quality control activities associated with the following equipment performance issues:

(1) Planned and scheduled maintenance on the 2B AF pump and failure of the pump gearbox oil cooler gasket during the weeks ending May 12 through June 30, 2018.

71111.13Maintenance Risk Assessments and Emergent Work Control

The inspectors evaluated the risk assessments for the following planned and emergent work activities:

(1) Actions to assess and repair a nicked cable in the Unit 1 Lower Cable Spreading Room during the weeks ending April 7 through May 19, 2018; and

(2) Emergent issues and repairs associated with the 1A EDG during the weeks ending April 28 through May 5, 2018.

71111.15Operability Determinations and Functionality Assessments

The inspectors evaluated the following operability determinations and functionality assessments:

(1) Lowering sodium hydroxide concentrations in the Unit 2 containment spray additive tank, as documented in IR 4115832, during the weeks ending April 14 through May 19, 2018;

(2) Damaged Unit 1 reactor head vent connector pins, as documented in IR 4125626, during the weeks ending April 28 through May 19, 2018;

(3) A pressure transient on the reactor coolant system (RCS) during digital control system modification installation, as documented in IR 4124952, during the weeks ending April 14 through May 19, 2018;

(4) Leakage past safety injection valves 1SI8900 A through D, as documented in IR 4124529, during the weeks ending April 14 through April 28, 2018;

(5) Trip of the 1B EDG on an overspeed signal during testing, as documented in IR 4129116, during the weeks ending April 21 through April 28, 2018;

(6) Issues with 2A EDG protection scheme drawings, as documented in IR 4131529, during the weeks ending May 12 through June 30, 2018;

(7) Issues associated with operating experience information provided by Westinghouse Electric Company under 10 CFR Part 21 for control rod drive (CRD) thermal sleeves, as documented in IR 4140602, during the weeks ending May 26 through June 30, 2018; and

(8) Jacket water temperature fluctuations on the 2B EDG, as documented in IR 4140734, during the weeks ending May 26 through June 30, 2018.

71111.18Plant Modifications

The inspectors evaluated the following permanent plant modifications:

(1) Westinghouse Ovation Digital Upgrade for 7300 nuclear steam system supply cabinets 1PA05J, 1PA06J, 1PA07J and 1PA08J, as documented in Engineering Change (EC)

===400434, during the weeks ending April 21 through April 28, 2018; and

(2) Westinghouse Ovation Digital Upgrade for turbine-driven feedwater pump cabinets 1FW36J and 1FW37J, as documented in EC 404360, during the weeks ending April 21 through April 28, 2018.

71111.19Post Maintenance Testing ===

The inspectors evaluated the following post maintenance testing activities:

(1) Testing for the 1C RCP motor following replacement, as documented in WO 1950807, during the weeks ending April 21 through May 12, 2018;

(2) Control rod drop time testing following reactor refueling activities, as documented in WO 1965830, during the weeks ending April 28 through May 12, 2018;

(3) Unit 1 Cycle 21 core reload verification activities, as documented in EC 619556, during the weeks ending April 21 through April 28, 2018;

(4) Unit 1 Cycle 21 low power physics testing following reactor refueling activities, as documented in WO 1962864, during the weeks ending April 28 through May 12, 2018;

(5) Unit 1 power ascension testing for Westinghouse Ovation 7300 digital controls, as documented in multiple WOs, during the weeks ending April 28 through May 12, 2018;

(6) Unit 1 RCS leakage testing at normal operating pressure and temperature following refueling activities, as documented in WO 4584364, during the weeks ending April 28 through May 12, 2018; and

(7) Testing of the 1A EDG following exciter repairs, as documented in WOs 4774857 and

===1897940, during the week ending April 28, 2018.

71111.20Refueling and Other Outage Activities ===

The inspectors evaluated the activities associated with the 20th refueling outage on Unit 1 (A1R20) during the period from April 8 through May 3, 2018.

71111.22Surveillance Testing The inspectors evaluated the following surveillance tests: Routine

(1) Monthly surveillance testing of the 1B EDG during the week ending April 21, 2018; and

(2) Unit 1 651A and 651B relay surveillance testing during the weeks ending

===June 9 through June 30, 2018.

In-service (3 Samples)===

(1) Main steam safety valve setpoint testing, as governed by BwMP 3305-003, during the weeks ending April 7 through April 14, 2018;

(2) Unit 1 emergency core cooling system full flow testing during the weeks ending April 14 through June 9, 2018; and

(3) 2B essential service water (SX) pump testing during the week ending May 26, 2018.

Containment Isolation Valve (1 Sample)

(1) Local leak rate testing for containment isolation valves 1RE9160A/B and 1RE9157, as specified in 1BwOSR 3.6.1.1-10, during the week ending April 21,

RADIATION SAFETY

71124.01Radiological Hazard Assessment and Exposure Controls Radiological Hazard Assessment

The inspectors evaluated radiological hazards assessments and controls.

Instructions to Workers (1 Sample)

The inspectors evaluated worker instructions.

Contamination and Radioactive Material Control (1 Sample)

The inspectors evaluated contamination and radioactive material controls.

Radiological Hazards Control and Work Coverage (1 Sample)

The inspectors evaluated radiological hazards control and work coverage.

High Radiation Area and Very High Radiation Area Controls (1 Sample)

The inspectors evaluated risk-significant high radiation area and very high radiation area controls.

Radiation Worker Performance and Radiation Protection Technician Proficiency (1 Sample)

The inspectors evaluated radiation worker performance and radiation protection technician proficiency.

===71124.02Occupational As Low As Reasonably Achievable Planning and Controls Implementation of As Low As Reasonably Achievable and Radiological Work Controls

=

The inspectors reviewed as low as reasonably achievable (ALARA) practices and radiological work controls by reviewing the following activities:

(1) Radiation Work Permit (RWP) BW-01-18-00502, Containment and Auxiliary Building Inservice Inspection Examinations/Weld Preparations, Revision 0;

(2) RWP BW-01-18-00505, Containment and Auxiliary Building Outage Scaffold, Revision 0;

(3) RWP BW-01-18-00510, Containment and Auxiliary Building Outage Valve Work with Added Controls, Revision 0;

(4) RWP BW-01-18-00613, Reactor Head Disassembly/Reassembly, Revision 2;

(5) RWP BW-01-18-00534, Reactor Head Peening, Revision 0;

(6) RWP BW-01-18-00614, Reactor Head and Upper Internals Move, Revision 1; and

(7) RWP BW-01-18-00623, Reactor Cavity Decontamination, Revision 0.

Radiation Worker Performance (1 Sample)

The inspectors evaluated radiation worker and radiation protection technician performance.

OTHER ACTIVITIES - BASELINE

71151Performance Indicator Verification

The inspectors reviewed the licensee performance indicators submittals listed below:

(1) MS05: Safety System Functional Failures (SSFFs) - 2 samples (1 per unit) from the second quarter of 2017 through the first quarter of 2018;

(2) MS06: Emergency AC Power Systems - 2 samples (1 per unit) from the second quarter of 2017 through the first quarter of 2018;

(3) MS07: High Pressure Injection Systems - 2 samples (1 per unit) from the second quarter of 2017 through the first quarter of 2018;

(4) BI01: RCS Specific Activity - 2 samples (1 per unit) from the first quarter of 2017 through the fourth quarter of 2017; and

(5) OR01: Occupational Exposure Control Effectiveness - 1 sample from the first quarter of 2017 through the fourth quarter of 2017.

71152Problem Identification and Resolution Semiannual Trend Review

The inspectors reviewed the licensees corrective action program (CAP) and other action tracking databases for trends that might be indicative of more significant safety issues.

71153Follow-Up of Events and Notices of Enforcement Discretion

(1) The inspectors evaluated Event Notification (EN) 53347: Undervoltage Actuation of the Engineered Safety Feature Bus, and the licensees response during the weeks ending April 21 through June 30, 2018;

(2) The inspectors evaluated EN 53353: Both Diesel Generators Inoperable, and the licensees response during the weeks ending April 21 through June 30, 2018;

(3) The inspectors evaluated EN 53354: Degraded Reactor Vessel Head, and the licensees response during the weeks ending April 21 through May 12, 2018;

(4) The inspectors evaluated EN 53358: Undervoltage Actuation of the Engineered Safety Feature Bus, and the licensees response during the weeks ending April 28 through June 30, 2018;

(5) The inspectors evaluated EN 53371: Automatic Reactor Trip Following Turbine Trip, and the licensees response during the weeks ending May 2 through June 30, 2018; and

(6) The inspectors evaluated EN 53443: Manual Reactor Trip on Lowering Steam Generator Water Level, and the licensees response during the weeks ending June 9 through June 30,

INSPECTION RESULTS

71111.12Maintenance Effectiveness Inadequate Detail in Maintenance Work Instructions Resulted in Failed Gearbox Oil Cooler Head Gasket and Inoperable 2B Auxiliary Feedwater Pump Cornerstone Significance Cross-Cutting Report Section Aspect Mitigating Systems Green NCV [H.9] - 71111.12 05000457/2018002-01 Training Opened/Closed A self-revealed finding of very low safety significance (i.e., Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to have adequate detail within their maintenance work instructions to enable proper reassembly of the 2B AF pump gearbox oil cooler. Specifically, during the licensees 19th Unit 2 refueling outage in April 2017, the gearbox oil cooler closure head was reassembled following scheduled maintenance using an excessive amount of room temperature - vulcanizing silicone (RTV) on the joint and an insufficient amount of torque on the closure head bolting. As a result, on March 16, 2018, the closure head joint failed resulting in several hours of unplanned inoperability and unavailability for the 2B AF Pump.

Description:

At 10:14 a.m. on March 16, 2018, plant operators successfully completed surveillance 2BwOSR 3.7.5.4-2, Unit 2 Diesel Driven Auxiliary Feedwater Pump Surveillance. At about 4:00 p.m. that same day operators responding to a leak detection alarm in the 2B AF pump room found that the 2B AF pump gearbox oil cooler was leaking essential service water (SX) from the closure head joint at about two to three gallons per minute (gpm). The pump was removed from service to facilitate repairs to the gearbox oil cooler and subsequently returned to service on the morning of March 17, 2018, following successful post-maintenance testing.

The licensee conducted a causal evaluation to investigate the failure and identified that the gearbox oil cooler had been disassembled for periodic maintenance and inspection during the previous Unit 2 refueling outage in April 2017. The licensees investigation further identified that the gearbox oil cooler closure head had been reassembled using an excessive amount of RTV sealant, and that the closure head bolting had likely not been adequately torqued due to the unexpectedly low breakaway torque licensee mechanics found when removing the closure head for repairs. The licensee concluded that these two issues in combination had led to the gearbox oil cooler closure head being reassembled during the previous Unit 2 refueling outage with inadequate clamping force, and susceptibility for the gearbox oil cooler closure head gasket to eventually extrude from the flange.

Following examination of the work instructions in WO1865874; Contractor Mechanical-2A02A-Disassemble, Clean, Inspect, and Reassemble Cooler; dated April 27, 2017, that was used during the April 2017 Unit 2 refueling outage to reassemble the gearbox oil cooler, the licensee concluded that deficiencies existed in the work instructions provided for the task.

Specifically, the work package instructions contained no information related to the use of RTV on the gearbox oil cooler closure head joint. Additionally, although torque values for the gearbox oil cooler closure head bolting were provided, the work package instructions lacked detail concerning the requirements for torqueing in a diametrically opposed pattern and the requirements for final torqueing until no nut movement was observed.

Corrective Actions: The licensee removed the 2B AF pump from service and repaired the leaking closure head joint on the gearbox oil cooler. The pump was successfully tested and restored to an operable state on March 17, 2018. As part of the licensees evaluation into the issue, an extent-of-condition review was performed for all other safety-related oil coolers serviced by the SX system. Additionally, the licensee created actions to add additional details to applicable model work orders regarding torqueing and the use of sealants, such as RTV, with gaskets.

Corrective Action Reference: The licensee entered this issue into their CAP as IR 4115813.

Performance Assessment:

Performance Deficiency: The licensee failed to have the appropriate details documented in their maintenance work instructions to enable the proper reassembly of the 2B AF pump gearbox oil cooler.

Screening: The inspectors determined that the performance deficiency was associated with the Procedures attribute of the Mitigating Systems Cornerstone of Reactor Safety, and that it was of more than minor safety significance because it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it was determined that the licensees planned maintenance on the 2B AF pump gearbox oil cooler during the April 2017 Unit 2 refueling outage improperly reassembled the gearbox cooler closure head using an excessive amount of RTV and less than adequate clamping force on the joint, which eventually permitted the closure head gasket to extrude and the joint to leak excessively.

Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings, dated October 7, 2016, and IMC 0609, Appendix A, The Significance Determination Process (SDP)for Findings At-Power, dated June 19, 2012. The inspectors answered No to all the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, because even though the 2B AF pump was rendered inoperable and unavailable to facilitate repairs, the SX leakage from the gearbox oil cooler was not sufficiently large enough to have prevented the pump from performing its safety function as designed. As a result, the finding screened as having very low safety significance (i.e., Green).

Cross-Cutting Aspect: The inspectors determined that the finding involved the cross-cutting aspect of Training under the area of Human Performance. In their evaluation of the issue, the licensee noted that the work on the 2B AF pump had been performed during the April 2017 Unit 2 refueling outage utilizing contractor craft personnel. It was further noted that the contractor personnel were not trained to the same level as the licensees own in-house maintenance personnel regarding the requirements and techniques for bolted mechanical joints and instead relied on the level of detail within written work instructions to successfully complete their work. [H.9]

Enforcement:

Violation: Appendix B to 10 CFR Part 50, Criterion V, Instructions, Procedures, and Drawings, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Contrary to the above, during the licensees April 2017 Unit 2 refueling outage, the 2B AF pump, a safety-related component subject to the requirements of Appendix B to 10 CFR Part 50, underwent preventative maintenance using written instructions in WO 1865874; Contractor Mechcanical-2A02A-Disassemble, Clean, Inspect, and Reassemble Cooler; dated April 27, 2017, that were not appropriate to the circumstances. Specifically, 2B AF pump gearbox oil cooler documented instructions for reassembly did not contain adequate instructions to permit the contractor craft personnel to properly reinstall the gearbox oil cooler closure head, which ultimately resulted in the creation of a two to three gpm SX leak on March 16, 2018, that rendered the pump inoperable and unavailable for required repairs.

Disposition: This violation is being treated as a NCV consistent with Section 2.3.2 of the Enforcement Policy.

71111.15Operability Determinations and Functionality Assessments Work Instruction Error Results in Reactor Coolant System Pressure Transient Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green FIN [H.3] - Change 71111.15 05000456/2018002-02 Management Opened/Closed A self-revealed finding of very low safety significance (i.e., Green) was identified due to the licensees failure to follow work instructions while performing a digital upgrade to plant control systems. Specifically, while working on the volume control tank (VCT) level transmitter on April 10, 2018, maintenance personnel failed to adequately track the steps being performed while working on multiple work packages. This resulted in the Unit 1 RCS experiencing a pressure transient and the actuation of a VCT relief valve.

Description:

On April 10, 2018, with Unit 1 in Mode 5, instrument maintenance technicians were performing work associated with the Westinghouse Ovation digital upgrade project (WO 4698242-68) on the VCT level instrumentation loops. The work required the coordination of multiple work packages. As part of the work, electrical leads for one of the VCT level transmitters were lifted. Immediately thereafter, the VCT isolation valves unexpectedly closed and charging pump suction was automatically transferred to the refueling water storage tank (RWST). In order to stabilize and maintain RCS pressure, charging flow adjustments were made by on-watch control room operators. Additionally, the instrument maintenance technicians were instructed to restore the electrical leads that had been lifted per their work package. The resulting pressure transient in the RCS resulted in a peak RCS pressure of approximately 427.5 pounds per square inch gauge (psig), causing the VCT relief valve to lift.

The licensee performed an evaluation to determine the cause of the event. The evaluation identified that prior to the electrical leads being lifted, another coincident level channel was already failed low. As a result, when the instrument maintenance technicians lifted the leads for the second VCT level instrument, the coincidence logic for the automatic VCT isolation and charging pump suction transfer to the RWST was satisfied and the valve actuations described above occurred. The licensees evaluation further identified that the instrument maintenance technicians had failed to properly track what steps had been completed on every package, and therefore failed to realize that one VCT level channel was already failed low when they proceeded to lift the electrical leads on a second level channel.

Tracking of completed steps, or placekeeping, while performing work packages is required by Paragraph 4.3.1 of licensee procedure HU-AA-104-101, Procedure Use and Adherence, Revision 5. The paragraph specifically required that placekeeping be applied to the work instructions within work packages. Additionally, the licensees evaluation also identified that multiple changes had been made to the work packages after the instrument maintenance technicians initially reviewed them to prepare for the work. This was a contributing factor to the confusion experienced by the instrument maintenance technicians as to the proper sequence of the work being performed.

As part of their review of the event, the licensee also reviewed system pressure trends and conducted physical inspections of the residual heat removal suction relief valves, pressurizer power-operated relief valves (PORVs), and the VCT relief valves to determine if any had lifted. The review found that only the VCT relief valve had lifted and that the VCT had been in a water-solid condition for about 12 minutes.

Corrective Actions: The licensee restored the plant to a stable condition following the transient and suspended the instrument maintenance work that was in progress. Additionally, the licensee created actions to add additional details to applicable model work orders regarding how to ensure that requirements were met when simultaneously working multiple work packages.

Corrective Action Reference: The licensee entered this issue into their CAP as IR 4124952.

Performance Assessment:

Performance Deficiency: The licensee failed to implement work package step placekeeping as required by procedure HU-AA-104-101, Procedure Use and Adherence.

Screening: The inspectors determined that the performance deficiency was associated with the Human Performance attribute of the Initiating Events Cornerstone of Reactor Safety, and that it was of more than minor safety significance because it adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to properly use work package placekeeping while working with and coordinating multiple packages on the VCT level transmitters resulted in the Unit 1 RCS experiencing a pressure transient, and the VCT relief valve lifting. The inspectors also noted that the issue was similar to Example No. 4.b of IMC 0612, Appendix E, Examples of Minor Issues, in that the failure to follow procedural requirements led to a transient.

Significance: The inspectors assessed the significance of the finding using IMC 0609, 4, Initial Characterization of Findings, dated October 7, 2016, and IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 9, 2014.

The inspectors determined that the finding pertained to the shutdown safety function of Inventory Control, as described in IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Using Exhibit 2, Initiating Events Screening Questions, the inspectors determined that the finding was of very low safety significance (i.e., Green)because it did not impact decay heat removal, and it did not occur while the reactor was in a reduced inventory condition.

Cross-Cutting Aspect: The inspectors determined that the finding involved the cross-cutting aspect of Change Management under the area of Human Performance. In their evaluation of the issue, the licensee noted that multiple changes had been made to the work packages following the instrument maintenance technicians initial review, which led to confusion due to the multiple sets of competing work instructions and different requirements for work package completion. [H.3]

Enforcement:

Because the issue pertained to work instructions and plant components that were not categorized as safety-related, the inspectors did not identify any violations of regulatory requirements associated with this finding.

71111.19Post Maintenance Testing Inadequate Test Activity Coordination Results in Unintended Valve Actuation and Reactor Coolant System Pressure Drop Cornerstone Significance Cross-Cutting Report Aspect Section Initiating Events Green NCV [H.4] - 71111.19 05000456/2018002-03 Teamwork Opened/Closed A self-revealed finding of very low safety significance (i.e., Green) and an associated NCV of Technical Specification 5.4, Procedures, was identified for the licensees failure to have properly coordinated testing activities associated with redundant Unit 1 pressurizer pressure instruments in accordance with the stations procedural requirements governing such testing.

Specifically, during the licensees 20th Unit 1 refueling outage on April 23, 2018, redundant pressurizer pressure instrumentation channels were inadvertently subjected to simultaneous testing activities. This resulted in the coincidence logic for both of the units pressurizer PORVs being unexpectedly satisfied and the PORVs opening to depressurize the RCS from approximately 345 psig to approximately 320 psig.

Description:

At approximately 11:30 p.m. on April 23, 2018, instrumentation maintenance technicians were conducting surveillance BwISR 3.3.1.10-M238, Operational Test and Channel Verification/Calibration for Loop P-0456 - Pressurizer Pressure Protection Channel II Cabinet 2 (PA02J). At the same time, plant personnel responsible for acceptance testing on a newly installed digital control system for the unit were engaged in a separate activity within EC 400434-1, Modification Test for Westinghouse Ovation Digital Upgrade for NSSS

[Nuclear Steam Supply System] and BOP [Balance of Plant] Cabinets 1PA05J - 1PA08J, 1PA20JA, and 1A20JB, that placed a second, redundant pressurizer pressure instrument loop into a testing configuration. The simultaneous testing of redundant pressurizer pressure instrument loops resulted in the coincidence logic for pressurizer high pressure to be satisfied and a demand signal sent to both of the PORVs to open. Both PORVs opened as designed and vented the pressurizer steam space to the primary relief tank (PRT). The unit was in a cold shutdown condition with RCS pressure at about 345 psig and was being maintained by a steam bubble in the pressurizer. Control room operators took manual control of both PORVs, which had been in the low temperature overpressure protection (LTOP) mode, and closed both valves. The LTOP relief path was realigned to credit the residual heat removal system relief valves to maintain LTOP protection for the unit. The RCS pressure dropped to approximately 320 psig during the event.

The licensee conducted an evaluation into the circumstances surrounding the event and identified that the senior reactor operator (SRO) who had been assigned as the Ovation modification test coordinator had failed to inform the on-watch Unit 1 control room crew that acceptance testing for any pressurizer pressure instrument loop was going to take place.

Conversely, the instrumentation maintenance technicians performing BwISR 3.3.1.10-M238 had checked in with the control room crew and briefed the on-watch unit supervisor and reactor operators as to the testing that they were going to perform as well as all expected alarms and control board indications.

Corrective Actions: Immediate actions taken by the on-watch Unit 1 control room crew were to transfer the LTOP relief path to the units residual heat removal (RHR) system relief valves and then manually close both PORVs using their respective control switches. The licensee halted Ovation modification testing and conducted a review of further testing for potential redundant channel conflicts, and reinforced the requirements to coordinate all Ovation modification testing with the on-watch unit supervisor and reactor operators. Additionally, the licensee planned to review their procedures for potential enhancements related to this event, to review their outage schedule for appropriate logic ties and restrictions related to redundant instrument channels, and to review the training for the Ovation modification to address potential knowledge gaps associated with the event.

Corrective Action Reference: The licensee entered this issue into their CAP as IR 4130106.

Performance Assessment:

Performance Deficiency: The licensee failed to properly coordinate testing activities on coincident pressurizer pressure instrument loop channels.

Screening: The inspectors determined that the performance deficiency was associated with the Human Performance attribute of the Initiating Events Cornerstone of Reactor Safety, and that it was of more than minor safety significance because it adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to properly coordinate testing activities on coincident pressurizer pressure instrument loop channels resulted in demand signals for both Unit 1 PORVs being transmitted to the valves, which caused both valves to open and vent the pressurizer steam space to the PRT.

Significance: The inspectors assessed the significance of the finding using IMC 0609, 4, Initial Characterization of Findings, dated October 7, 2016, and IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 9, 2014.

The inspectors determined that the finding pertained to the shutdown safety function of Inventory Control, as described in IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Using Exhibit 2, Initiating Events Screening Questions, the inspectors determined that the finding was of very low safety significance (i.e., Green)because the loss of inventory through the Unit 1 PORVs was such that it would not have resulted in the loss of the operating train of RHR and was assessed by the inspectors to have been self-arresting.

Cross-Cutting Aspect: The inspectors determined that the finding involved the cross-cutting aspect of Teamwork under the area of Human Performance. In their evaluation of the issue, the licensee noted that the personnel assigned to coordinate the Ovation modification test activities had not communicated with the on-watch Unit 1 control room crew and informed them that acceptance testing for pressurizer pressure instrument loops was planned. [H.4]

Enforcement:

Violation: Technical Specification 5.4.1 requires, in part, that written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Section 3 of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, Procedures for Startup, Operation, and Shutdown of Safety-Related PWR [pressurized water reactor]

Systems, requires procedures for the operation of the pressurizer pressure and spray control systems (Subsection 3.j).

On April 23, 2018, plant personnel conducting acceptance testing on the units newly installed Ovation digital control system were testing functions associated with the RCS pressure control system, which included pressurizer pressure and spray control systems. This testing was governed by the requirements of procedure OP-AA-108-110, Evaluation of Special Tests or Evolutions. Paragraph 3.3.2 of this procedure required that the special test coordinator ensure that on-watch shift management be made aware of the status of testing activities.

Contrary to the above requirements, on April 23, 2018, the Ovation modification test coordinator failed to inform the on-watch Unit 1 control room shift crew that acceptance testing for any pressurizer pressure instrument loop was going to take place.

Disposition: This violation is being treated as a NCV consistent with Section 2.3.2 of the Enforcement Policy.

71152Problem Identification and Resolution Observations: Station Chiller Unit Issues 71152Semiannual Trend Review During their review to complete this inspection sample, the inspectors noted several examples of issues involving heating, ventilation, and air conditioning (HVAC) chiller units at the station.

Because many of these components are idle during the cooler winter season months, the inspectors expanded the time period for the sample to include two season changes from cooler months to warmer months where the components would be transitioned from being idle to being in service. Observed items associated with this trend included, but were not limited to the following:

• Multiple separate issues involving several plant systems. The inspectors review identified over 20 issues entered into the licensees CAP during the period under review involving station chillers. The issues were observed in several plant systems, but primarily control room HVAC (VC), a safety-related system, and primary containment ventilation (VP), a nonsafety-related system that supports maintenance of the Technical Specification-related containment temperature parameter. A lesser number of issues were also observed in service building HVAC (VS) and auxiliary building HVAC (VA). Altogether, the inspectors assessed that the sum total number of issues over the time period under review was more than what would typically be encountered at an average dual unit facility over the same time period.

• Multiple issues specifically involving chiller oil inventory. Approximately half of the issues identified in the inspectors review involved difficulties with the maintenance of chiller oil level. Some amount of migration of chiller oil is expected as the heat removal load on a particular unit fluctuates. However, the inspectors assessed that the number and frequency of oil inventory issues was notably higher than what would typically occur at a dual unit facility over the same time period.

Although the issues reviewed by the inspectors individually have not resulted in any significant consequences, in the aggregate they constitute a large and seasonally recurring demand on station maintenance and engineering resources. In the worst case, both the VC and VP chillers have the capability of placing one or both units at the station in extremis with respect to plant Technical Specifications should multiple chillers within the same system experience issues at the same time. The licensee has recognized this vulnerability, and entered the need for additional evaluation of the trend noted by the inspectors into their CAP as IR

EXIT MEETINGS AND DEBRIEFS

The inspectors confirmed that proprietary information was appropriately controlled to protect it from public disclosure. No proprietary information was documented in this report.

• On April 19, 2018, the inspectors presented the inspection results for their review of the licensees inservice inspection program activities to the Plant Manager, Ms. A. Ferko, and other members of the licensee staff;

• On April 20, 2018, the inspectors presented the inspections results for their review of the licensees radiation protection program to the Site Vice President, Ms. M. Marchionda, and other members of the licensee staff; and

• On July 12, 2018, the inspectors presented the quarterly integrated inspection results to the Site Vice President, Ms. M. Marchionda, and other members of the licensee staff.

THIRD PARTY REVIEWS Review of the Operations Training Programs Accreditation Report from the Institute of Nuclear Power Operations As discussed in IMC 0611, Section 13.01, the inspectors completed a review of the report issued by the Institute of Nuclear Power Operations (INPO) on March 14, 2018, for the most recent periodic accreditation of training programs for licensed and non-licensed operators at the Braidwood Station.

DOCUMENTS REVIEWED

71111.01Adverse Weather Protection

Procedures:

- OP-AA-102-102; General Area Checks and Operator Field Rounds; Revision 15

- OP-AA-108-107-1001; Station Response to Grid Capacity Conditions; Revision 7

- OP-AA-108-107-1002; Interface Procedure Between BGE/COMED/PECO and Exelon

Generation (Nuclear/Power) for Transmission Operations; Revision 11

- OP-AA-108-111-1001; Severe Weather and Natural Disaster Guidelines; Revision 17

- OP-BR-102-102-1001; Augmented Operator Field Rounds; Revision 3

- WC-AA-107; Seasonal Readiness; Revision 20

Other:

- Letter from M. Marchionda-Palmer to D. Enright; Certification of Summer Readiness;

06/15/2018

71111.04Equipment Alignment

Procedures:

- BwOP AF-E1; Electrical Lineup - Unit 1 Operating; Revision 15

- BwOP AF-M1; Operating Mechanical Lineup Unit 1; Revision 20

- BwOP DG-M2; Operating Mechanical Lineup Unit 1 1B D/G; Revision 17

- BwOP FC-M1; Operating Mechanical Lineup Unit 1; Revision 10

- BwOP RH-11; Securing the RH System in Shutdown Cooling; Revision 29

- BwOP RH-6; Placing the RH System in Shutdown Cooling; Revision 58

71111.05Fire Protection

Procedures:

- BwAP 1110-1; Fire Protection Program System Requirements; Revision 41

- BwAP 1110-1A3; GOCAR Action Chart Fire Protection Water Suppression Systems;

Revision 8

- BwAP 1110-1A4; GOCAR Required Compensatory Measures Action Response Carbon

Dioxide Fire Suppression Systems; Revision 10

- BwAP 1110-1A5; GOCAR Required Compensatory Measures Action Response Halon Fire

Suppression Systems; Revision 5

- BwAP 1110-3; Plant Barrier Impairment Program; Revision 38

- BwOP PBI-1; Plant Barrier Impairment Program Pre-Evaluated Barrier Matrix; Revision 2

- CC-AA-201; Plant Barrier Control Program; Revision 12

- ER-AA-600-1069; High Risk Fire Area Identification; Revision 4

- ER-BR-600-1069; Site List of High Risk Fire Areas - Braidwood Unit 1 and Unit 2; Revision 0

- OP-AA-201-004; Fire Prevention for Hot Work; Revision 15

- OP-AA-201-005; Fire Brigade Qualification; Revision 9

- OP-AA-201-008; Pre-Fire Plan Manual; Revision 4

- OP-AA-201-009; Control of Transient Combustible Material; Revision 20

- OP-MW-201-007; Fire Protection System Impairment Control; Revision 7

Pre-Fire Plans:

- No. 1, Fire Zone 1.1-1; Containment 377 - 0 Elevation, Unit 1 Containment Missile Shield

Area; Revision 0

- No. 101, Fire Zone 11.2-0; Auxiliary Building 346 - 0 Elevation, Unit 2 Auxiliary Building

General Area (South); Revision 2

- No. 2, Fire Zone 1.1-1; Containment 401 - 0 Elevation, Unit 1 Containment Missile Shield

Area; Revision 0

- No. 5, Fire Zone 1.2-1; Containment 377 - 0 Elevation, Unit 1 Annular Area; Revision 0

- No. 6, Fire Zone 1.2-1; Containment 401 - 0 Elevation, Unit 1 Annular Area; Revision 0

- No. 9, Fire Zone 1.3-1; Containment 426 - 0 Elevation, Unit 1 Containment Upper Area;

Revision 1

- No. 97, Fire Zone 11.1B-0; Auxiliary Building 330 - 0 Elevation, Unit 2 Auxiliary Building

Basement (1B/2B SX); Revision 1

- No. 98, Fire Zone 11.2-0; Auxiliary Building 346 - 0 Elevation, Unit 2 Auxiliary Building

General Area (Center); Revision 1

- No. 99, Fire Zone 11.2-0; Auxiliary Building 346 - 0 Elevation, Unit 1 Auxiliary Building

General Area (North); Revision 0

71111.07Heat Sink

Work Orders:

- 1875163-01; 1CC01A Clean and Inspect; 05/07/2018

71111.08 - Inservice Inspection Activities

Action Requests/Issue Reports:

- 2723199; A Rejectable Indication on CRD Penetration 69 Weld Buildup; 12/01/2016

- 2727539: Boric Acid Leak of Bolted Connection on Component 1SI8816A; 10/17/2016

- 2729016; NDE Linear Indication Detected on 1FW03DC-16 Weld Prep; 10/16/2016

- 2731397; Boric Acid Leak on Threaded Connection for Component 1SI121B; 10/24/2016

- 3986087; Boric Acid Leak on Pipe Cap Downstream of Component 1SI045; 04/13/2017

Procedures:

- ER-AA-335-003; Magnetic Particle (MT) Examination; Revision 8

- ER-AA-335-014-2003; VT-2 Visual Examination in Accordance with ASME 2001 Edition, 2003

Addenda; Revision 2

- ER-AA-335-018; Visual Examination of ASME IWE Class MC and Metallic Liners of IWL Class

CC Components; Revision 13

- ER-AA-355-002; Liquid Penetrant (PT) Examination; Revision 10

- ER-AP-331-1002; Boric Acid Corrosion Control Program Identification, Screening and

Evaluation; Revision 10

- ER-AP-335-001; Bare Metal Visuals for Nickel Alloy Materials; Revision 6

- EXE-PDI-UT-2; Ultrasonic Examination of Austenitic Piping Welds in Accordance with

PDI-UT-2; 02/21/2017

- WDI-SSP-1194; Ultrasonic Examination of Reactor Pressure Vessel Welds in Accordance

With PDI-UT-6; Revision 1

Work Orders:

- 1962980-01; MM - 1AF006A - Cut Out and Replace Valve, 04/04/18

- 1816677-01; TES-1CV8368B-Disasemble and Inspect Check Valve, 04/04/18

Welding Procedure Specification (WPS):

- 1-1-GTSM-PWHT; Gas Tungsten Arc Welding (GTAW) and Shielded Metal Arc Welding

(SMAW) of Carbon Steel (P-1), Fillet, Socket, Groove) With or Without Backing; Revision 2

- 8-8-GTSM-PWHT; Gas Tungsten Arc Welding (GTAW) and Shielded Metal Arc Welding

(SMAW) of Stainless Steel (P-8), Fillet, Socket, Groove) With or Without Backing; Revision 2

71111.11Licensed Operator Requalification Program

Action Requests/Issue Reports:

- 4132290; Unit 2 Emergency Ramp Down 350 MWe Due to Grid Overload Condition;

04/30/2018

- 4132293; 1FW510A Failed Closed Due to Maintenance Work Package; 04/30/2018

Procedures:

- BwOP CX-99; Ovation Temporary Control Panel (TCP) Operations; Revision 0

- OP-AA-101-111-1001; Operations Standards and Expectations; Revision 20

- OP-AA-101-113; Operator Fundamentals; Revision 11

- OP-AA-101-113-1006; 4.0 Crew Critique Guidelines; Revision 9

- OP-AA-103-102; Watch-Standing Practices; Revision 18

- OP-AA-103-102-1001; Strategies for Successful Transient Mitigation; Revision 2

- OP-AA-103-103; Operation of Plant Equipment; Revision 1

- OP-AA-104-101; Communications; Revision 3

- OP-AA-108-107-1002; Interface Procedure Between BGE/COMED/PECO and Exelon

Generation (Nuclear/Power) for Transmission Operations; Revision 11

- OP-AA-111-101; Operating Narrative Logs and Records; Revision 13

- OP-AA-300; Reactivity Management; Revision 12

- TQ-AA-10; Systematic Approach to Training Process Description; Revision 5

- TQ-AA-150; Operator Training Programs; Revision 16

- TQ-AA-155; Conduct of Simulator Training and Evaluation; Revision 8

- TQ-AA-306; Simulator Management; Revision 9

- TQ-BR-201-0113; Braidwood Training Department Simulator Examination Security Actions;

Revision 21

71111.12Maintenance Effectiveness

Action Requests/Issue Reports:

- 4126999; MCC FD BKR 1AP10EF (131X-3A) Did Not Open Immediately, 04/15/2018

Procedures:

- ER-AA-310; Implementation of the Maintenance Rule; Revision 11

- ER-AA-310-1001; Maintenance Rule - Scoping; Revision 4

- ER-AA-310-1002; Maintenance Rule Functions - Safety Significant Classification; Revision 3

- ER-AA-310-1003; Maintenance Rule - Performance Criteria Selection; Revision 5

- ER-AA-310-1004; Maintenance Rule - Performance Monitoring; Revision 14

71111.13Maintenance Risk Assessments and Emergent Work Control

Action Requests/Issue Reports:

- 4128714; OSP-A Loss of ESF Bus 141 During 1A EDG Sequencer Testing; 04/19/2018

- 4129712; 1A Diesel Generator Secured During 1BwOSR 3.8.1.19-1; 04/22/2018

- 4130140; Oil Leak in Turbo Charger and Coming Out Cylinders; 04/24/2018

Procedures:

- ER-AA-330-001;Section XI Pressure Testing; Revision 14

- ER-AA-330-009; ASME Section XI Repair/Replacement Program; Revision 14

- ER-AA-335-015-2003; VT-2 Visual Examination in Accordance with ASME 2001 Edition, 2003

Addenda; Revision 2

- ER-AA-600; Risk Management; Revision 7

- ER-AA-600-1042; On-Line Risk Management; Revision 11

- MA-BR-EM-4-09070; Diesel Generator Electrical Inspection; Revision 16

- OP-AA-108-117; Protected Equipment Program; Revision 5

- WC-AA-101-1006; On-Line Risk Management and Assessment; Revision 2

- WC-AA-104; Integrated Risk Management; Revision 25

Work Orders:

- 4774957; Loss of ESF Bus 141 During 1A EDG Sequencer Testing - Emergency Mode

Master Relay; 04/22/2018

- 4775881; 1A EDG Troubleshoot/Repair/Replace Emergency Stop Pushbutton; 04/22/2018

Drawings/Prints:

- 20E-1-4020A; Relay and Metering Diagram - Diesel Generator 1A-1DG01KA Generator

Control Part 1; Revision V

- 20E-1-4020B; Relay and Metering Diagram - Diesel Generator 1A-1DG01KA Generator

Control Part 2; Revision W

71111.15Operability Evaluations and Functionality Assessments

Action Requests/Issue Reports:

- 4115832; Unexpected Alarm 2-3-C3, Spray Add Tank Level High/Low; 03/16/2018

- 4129116; OSP-A 1B DG Tripped During Monthly Surveillance; 04/20/2018

- 1670422; 2A DG Overspeed Alarm - 2DG01KA; 06/12/2014

- 4126415; OSP-A 1BwOSR 3.4.11.3 Failed; 04/13/2018

- 4124952; RCS Pressure Transient During 7300 TCP Mod; 04/10/2018

- 4140734; 2B DG Jacket Water Temperature Fluctuations; 5/23/2018

- 4124952; RCS Pressure Transient During 7300TCP Mod; 04/10/2018

- 2468574; 1A DG Overspeed Trip During Monthly Surveillance Run; 03/17/2015

- 4124529; OSP-X 04 GPM Measured 1BwOSR 3.4.14.1 Step 10 1SI8900A-D; 04/09/2018

Engineering Changes/Technical Evaluations:

- 402090; Technical Evaluation for the Acceptable Amount of Oil Content in the Refrigerant of

the VC Chillers; Revision 0

Procedures:

- BwOP DG-11; Diesel Generator Startup and Operation; Revision 48

- HU-AA-104-101; Procedure Use and Adherence; Revision 5

- OP-AA-106-101-1006; Operational Decision Making Process; Revision 19

- OP-AA-108-111; Adverse Condition Monitoring and Contingency Planning; Revision 10

- OP-AA-108-115; Operability Determinations (CM-1); Revision 21

- OU-AP-104; Shutdown Safety Management Program; Revision 24

Work Orders:

- 4698242-68; 1L-0112 Loop + Relays LY112BX1; 04/12/2018

- 4594378; OP IST-LT-1SI8900A/B/C/D-Leak Surveillance; 3/23/18

71111.18Plant Modifications

Drawings/Prints:

- 20E-1-3031FW01; Loop Schematic Diagram Steam Generator Wide Range Level Loop 1A

(1LT-0501) Protection Cabinet 1 (1PA01J); Revision K

- 20E-1-4031MS01; Loop Schematic Diagram Steam Pressure Protection Loop 1A & 1D (1PT-

0516, 0546) Protection Cabinet 4 (1PA04J); Revision I

- 20E-1-4031FW11; Loop Schematic Diagram Steam Generator Loop 1A Level (1LT-0517)

Protection Cabinet 4 (1PA04J); Revision G

Procedures:

- BwOP CX-99; Ovation Temporary Control Panel (TCP) Operations; Revision 0

- SPP-18-001; Power Ascension Test of NSSS-BOP Digital Upgrades; Revision 0

Engineering Changes/Technical Evaluations:

- 404358; Westinghouse Ovation Digital Upgrade for Balance of Plant (BOP) Cabinets 1PA20JA

and 1PA20JB (N Outage), Unit 1; Revision 3

- 404360; Westinghouse Ovation Digital Upgrade for Turbine Driven Feedwater Cabinets

1FW36J and 1FW37J (N Outage), Unit 1; Revision 2

- 400434; Westinghouse Ovation Digital Upgrade for 7300 NSSS Cabinets 1PA05J, 1PA06J,

1PA07J, 1PA08J (N Outage), Unit 1; Revision 2

71111.19Post-Maintenance Testing

Action Requests/Issue Reports:

- 4128714; OSP-A Loss of ESF Bus 141 During 1A EDG Sequencer Testing; 04/19/2018

- 4128804; Newton Unable to Map All Core Locations; 04/18/2018

- 4129712; 1A Diesel Generator Secured During 1BwOSR 3.8.1.19-1; 04/22/2018

- 4131607; N-35 and N-36 Require Channel Operational Tests Prior to Low Power Physics

Testing per TS 3.1.8.1; 04/27/2018

Procedures:

- 1BwOSR 3.1.8.2; Special Test Exceptions - Reactor Coolant System TAVE During Physics

Test Surveillance; Revision 1

- 1BwOSR 3.1.8.3; Special Test Exceptions - Physics Tests Thermal Power Hourly

Surveillance; Revision 1

- 1BwOSR 3.1.8.4; Unit 1: Shutdown Margin Verification During Physics Tests; Revision 1

- 1BwOSR 3.8.1.14-1; 1A Diesel Generator 24 Hour Endurance Run; Revision 9

- BwOP DG-1; Diesel Generator Alignment to Standby Condition; Revision 29

- BwOP DG-11; Diesel Generator Startup and Operation; Revision 48

- BwOP DG-12; Diesel Generator Shutdown; Revision 30

- BwVP 200-30; Nondestructive Examination of the Reactor Coolant Pump Flywheels;

Revision 0

- BwVS 500-6; Low Power Physics Test Program; Revision 43

- EC 400434-1; Modification Test for Westinghouse Ovation Digital Upgrade for NSSS and BOP

Cabinets 1PA05J - 1PA08J, 1PA20JA, and 1A20JB; Revision 0

- NF-AA-330-1001; Core Verification Guideline; Revision 12

- NF-AP-100-7008; Reactor Engineering Core Reload Support Activities; Revision 2

- NF-AP-100-7206; Generation of Core Loading Plan (CLP) and CPIX File; Revision 0

- OP-AA-108-110; Evaluation of Special Tests or Evolutions; Revision 3

Work Orders:

- 1950807; 1C RCP - Perform 10 Year Flywheel Inspection; 04/17/2018

- 1962864; Low Power Physics Test Program with Dynamic Rod Worth Measurement;

04/28/2018

- 1965830; U1 Automatic Rod Drop Time Surveillance; 09/05/2018

- 1968421; 1PT-0456 EQ Calibration/Maintenance/Surveillance; 04/18/2018

- 4774957; 1A Diesel Generator Functional Check at 110 Percent Load; 04/21/2018

Engineering Changes/Technical Evaluations:

- 619556; Braidwood Unit 1 Cycle 21 Core Reload Design; Revision 0

71111.20Refueling and Other Outage Activities

Action Requests/Issue Reports:

- 4132293; 1FW510A Failed Closed Due to Maintenance Work Package; 04/30/2018

- 4125334; A1R20 Flux Thimble Eddy Current Results and Recommendations; 04/11/2018

- 4126172; Battery Charger 111 Repair Removed From A1R20; 04/12/2018

- 4126797; Risk to Ovation Modification Acceptance Testing; 04/14/2018

- 4108710; Potential Scope Change for 1A Rod Drive in A1R20; 02/27/2018

- 4126424; OSP-X - Indication Found on Equipment Hatch O-Ring; 04/13/2018

- 4124987; Containment Closure Test Satisfactory; 04/10/2018

- 4126942; Work Stopped on 1PA08J and 1PA09J; 04/14/2018

- 4130160; Clearance Order Boundary Walkdown; 04/24/2018

- 4130236; Completed Surveillance Paperwork Not Recovered; 04/24/2018

- 4131995; Radiation Protection Fatigue Assessment/Waiver; 04/28/2018

Procedures:

- 1BwGP 100-1; Plant Heatup; Revision 33

- 1BwGP 100-2; Plant Startup; Revision 43

- 1BwGP 100-3; Power Ascension 5% to 100%; Revision 74

- 1BwGP 100-4; Power Descention; Revision 43

- 1BwGP 100-5; Plant Shutdown and Cooldown; Revision 51

- 1BwGP 100-6; Refueling Outage; Revision 30

- 1BwOS TRM 2.5.b.1; Unit 1 Containment Loose Debris Inspection; Revision 18

- BwAP 1450-1; Access to Containment; Revision 44

- SSP-18-001; Power Ascension Test of NSSS-BOP Digital Upgrades; Revision 0

Work Orders:

- 4763634-01; 1B DG Operability Monthly; 04/21/2018

71111.22Surveillance Testing

Action Requests/Issue Reports:

- 4101623; Tech Spec Surveillance Test Interval Change Request; 02/07/2018

- 4141035; 2B SX Pump Requires Engineering Evaluation; 05/24/2018

- 4123317; 1MS015B Pre-A1R20 Testing As-Found Outside Acceptance Range; 04/05/2018

- 4123318; 1MS014D Pre-A1R20 Testing As-Found Outside Acceptance Range; 04/05/2018

- 4123320; 1MS014C Pre-A1R20 Testing As-Found Outside Acceptance Range; 04/05/2018

- 4123321; 1MS014B Pre-A1R20 Testing As-Found Outside Acceptance Range; 04/05/2018

- 4123323; 1MS014A Pre-A1R20 Testing As-Found Outside Acceptance Range; 04/05/2018

Procedures:

- 1BwOSR 3.6.1.1-10; Primary Containment Type C Local Leakage Rate Test of Reactor

Building Drains and Vents System; Revision 15

- 1BwOSR 3.8.1.2-2; 1B Diesel Generator Operability Surveillance; Revision 41

- 1BwOSR 5.5.8.RH-6; Residual Heat Removal System Check Valve Stroke Test; Revision 12

- 1BwOSR 5.5.8.SI-11; Comprehensive Inservice Testing (IST) Requirements for Unit 1 Safety

Injection Pumps and Safety Injection System Check Valve Stroke Test; Revision 7

- BwMP 3305-003; Main Steam Safety Valve Testing Using Setpoint Verification Device;

Revision: 0

- BwMSR 3.7.1.1; Main Steam Safety Valves Operability Test (Setpoint Verification Using the

Furmanite Trevitest System); Revision 13

- BwOP DG-1; Diesel Generator Alignment to Standby Condition; Revision 30

- BwOP DG-11; Diesel Generator Startup and Operation; Revision 49

- BwOP DG-12; Diesel Generator Shutdown; Revision 30

- MA-AA-793-044; Inspection/Certification of Portable Pressure Test Equipment; Revision 5

Work Orders:

- 1956311, Parts 01 - 20; Testing of Main Steam Safety Valves 1MS013A - D, 1MS014A - D,

1MS015A - D, 1MS016A - D, and 1MS017A - D; 04/04/2018

- 1963943, Part 01; SI Hot Leg Check Valve Test; 04/17/2018

- 1957377, Part 01; CV Cold Leg Check Valve Test - 1BwOSR 5.5.8.CV-8

- 4763634, Part 01; 1B DG Operability Monthly; 04/21/2018

Engineering Changes/Technical Evaluations:

- 623852; Unit 1 Main Steam Safety Valve (MSSV) Instrument Uncertainty Evaluation;

Revision 0

71124.01Radiological Hazard Assessment and Exposure Controls

Action Requests/Issue Reports:

- 4127450; Mechanic Receives Dose Rate Alarm; 04/16/2018

Licensee Self-Assessments:

- PI-AA-126-1001-F-01; IR 3986556; Radiological Hazard Assessment and Exposure

Control;12/08/2017

- PI-AA-126-1001-F-01; IR 4077797; Pre-NRC Inspection Occupational ALARA Planning and

Control 71124-02; 02/16/2018

Radiation Work Permits (RWPs):

- BW-01-18-00502; CNMT/AUX Building ISI Examinations/Weld Prep; Revision 0

- BW-01-18-00505; CNMT/AUX Building Outage Scaffold; Revision 0

- BW-01-18-00510; CNMT/AUX Building Outage Valve Work With Added Controls; Revision 0

- BW-01-18-00613; Reactor Head Disassembly/Reassembly; Revision 2

- BW-01-18-00534; Reactor Head Peening; Revision 0

- BW-01-18-00614; Reactor Head and Upper Internals Move; Revision 1

- BW-01-18-00623; Reactor Cavity Decontamination; Revision 0

- BW-01-18-00614; Upper Internal Lift to Stand; Revision 1

Radiation Surveys and Maps:

- Survey Map of Reactor Head Peening; Underside of U-1 Reactor Head; 04/12/2018

71124.02Occupational As Low As Reasonably Achievable (ALARA) Planning and Controls

Procedures:

- RP-AA-400; ALARA Program; Revision 15

- RP-AA-401; Operational ALARA Planning and Controls; Revision 24

- RP-AA-460-002; Additional High Radiation Exposure Control; Revision 4

Radiation Protection Department Assessments and Evaluations:

- NCAP Investigation REORT Elevated Dose Rates During A2R19 Due to RPVH Peening

Project; Exceeded Business Plan Goal for Dose for A2R19; 06/07/2017

- RP-AA-400; SASC/SAC Meeting Minutes; 05/08/2017

- RP-AA-401; BW-01-18-00641; PT of Penetration 69 and Associated Activities; 04/10/2018

- Braidwood Refueling Outage RP Performance Summary; 04/19/2018

- RP-AA-401; ALARA Work In Progress Review; Under Rx Head/Remove EDM Tool and

Remove Remaining Fragments; 04/29/2017

- RP-AA-401; ALARA Waiver/Change Form; Rx Head Cavitation Peening and Support;

05/07/2017

- RP-AA-401; ALARA Waiver/Change Form; Rx Head Peening - No Under Head Work; Head

Peening and Decontamination Support; 05/12/2017

- RP-AA-401; ALARA Waiver/Change Form; Rx Head Emergent Removal of Pen 77 CETC

Funnel; 05/12/2017

- RP-AA-401; ALARA Waiver/Change Form; Under Rx Head - Install Funnel on Penetration 67;

RWP BW-02-17-00534; 05/13/2017

- RP-AA-460-002; Approval for Working in an Area Greater 1500 mrem/hr Radiation Field and

or Electronic Dosimeter Accumulated Dose Alarm Greater than 500 mrem; 04/27/2017

ALARA Plans:

- ALARA Briefing Attendance Form for Upper Internal Lift to Stand; 04/17/2018

- BW-01-18-00534; Reactor Head Peening; Revision 0

- BW-01-18-00613; Reactor Head Disassembly/Reassembly; Revision 2

- BW-01-18-00614; Reactor Head and Upper Internals Move; Revision 1

- BW-01-18-00623; Reactor Cavity Decontamination; Revision 0

71151Performance Indicator Verification

Procedures:

- LS-AA-2001; Collecting and Reporting of NRC Performance Indicator Data; Revision 14

- LS-AA-2080; Monthly Data Elements for NRC Safety System Functional Failures; Revision 7

- LS-AA-2090; Monthly Data Elements for NRC Reactor Coolant System (RCS) Specific

Activity; Revision 4

- LS-AA-2140; Monthly Data Elements for NRC Occupational Exposure Control Effectiveness;

Revision 5

- LS-AA-2200; Mitigating System Performance Index Data Acquisition and Reporting;

Revisions 5 and 6

Other:

- Monthly Performance Indicator Data Compiled by the Licensee from January 2017 through

March 2018

71152Identification and Resolution of Problems

Action Requests/Issue Reports:

- 4015366; 1A VP Chiller Will Not Start 1WO01CA; 05/26/2017

- 4015426; 0WO01CA Compressor Oil Level at 10 Percent; 05/26/2017

- 4019008; 0A VC Chiller Oil Loss Trend; 06/05/2017

- 4035732; 1A VP Chiller Failed To Start 1WO01CA; 07/26/2017

- 4040387; Abnormal Oil Usage 1B Containment Chiller; 08/09/2017

- 4049105; 1WO01CA Not Cooling Due to Failed Purge Unit; 09/05/2017

- 4049574; 0A VA Chiller Oil Level Lowering 0WO03CA; 09/06/2017

- 4087890; As Found Condition of 2B VP Chiller Evaporator; 12/28/2017

- 4113705; 0WO01CB Oil Level is at 95 Percent; 03/11/2018

- 4118167; 0B VC Chiller Oil Level 15 Percent; 03/23/2018

- 4121458; 1A VP Chiller Purge Unit Not Working Properly; 03/31/2018

- 4126691; Unit 1 N44 Control Power Fuses; 04/13/2018

- 4140396; 0B VC Chiller High Oil Level; 05/23/2018

- 4141345; 0B VC Chiller Oil Level is 10 Percent; 05/25/2018

- 4141812; 1A VP Chiller Operating in Red Band at 170 Amps; 05/28/2018

- 4142357; Need WR for 2B VP Chiller Purge Unit Replacement; 05/30/2018

- 4147839; Oil Added to 0B VC Chiller; 06/17/2018

- 4148006; Oil Added to 0B VC Chiller; 06/18/2018

- 4150591; 1A VP Chiller Trips; 06/26/2018

- 4152474; Oil Slinger Ring Not Installed on 0A VC Chiller; 07/02/2018

- 4153578; Adverse Trend IR: Station Chiller Issues; 07/06/2018

- 4005307; Oil Level Drop Following 0A VS Chiller Start - 0WO02CA; 05/01/2017

- 4036630; 1A VP Chiller Guide Vane Indicator Replacement and Discovery; 07/28/2017

- 4140074; 2B VP Chiller Compressor Did Not Start During Start Sequence; 05/22/2018

- 4141702; 0B VC Oil Level at 0 Percent, Oil Pressure Oscillating; 05/27/2018

- 4149917; Oil Slinger Ring Not Installed on 0A and 0B VC Chillers; 06/24/2018

Procedures:

- NO-AA-10; Quality Assurance Topical Report; Revision 92

- PI-AA-120; Issue Identification and Screening Process; Revision 8

- PI-AA-125; Corrective Action Program (CAP) Procedure; Revision 6

- PI-AA-125-1001; Root Cause Analysis Manual; Revision 3

71153Follow-Up of Events and Notices of Enforcement Discretion

Action Requests/Issue Reports:

- 4130140; Oil Leak in Turbo Charger and Coming Out Cylinders; 04/24/2018

- 4131491; 1A EDG Under Frequency Trip Wiring Error; 04/27/2018

- 4131529; 2A EDG Under Frequency Trip Wiring Error; 04/27/2018

- 4144070; Unit 1 Manual Reactor Trip; 06/04/2018

- 4144143; Unit 1 Steam Generator PORVs Opened Post Trip; 06/04/2018

- 4144206; Unintended Actuation During 1BwOSR 3.3.2.8-621A; 06/04/2018

- 4144267; Surveillance 1BwOSR 3.3.2.8-621A Not Completed; 06/04/2018

- 4128714; OSP-A - Loss of ESF Bus 141 During 1A EDG Sequencer Testing; 04/19/2018

- 4129240; OSP - 1RC01R NDE Results for 1RV-03-002 Reactor Head Weld; 04/20/2018

- 4129712; 1A Diesel Generator Secured During 1BwOSR 3.8.1.19-1; 04/22/2018

- 4144103; 1SD002F Did Not Auto Close on LO-2 Narrow Range Steam Generator Water

Level; 06/04/2018

- 4144124; Numerous Radiation Monitor Alarms Following Unit 1 Reactor Trip; 06/04/2018

- 4144499; Steam Generator Water Level Control During A1F39 Start-Up; 06/05/2018

- 4144951; Determine the Cause of Divergent Steam Generator Water Levels During

Condensate/Feedwater Runback; 06/06/2018

- 4144298; 7300 Ovation Modification Deficiency Causes Unit 1 Trip; 05/05/2018

Drawings/Prints:

- M-2035, Sheet 8; Main Steam System; 09/14/1976

- M-2035, Sheet 4; Extraction Steam System; 09/14/1976

Procedures:

- 1BwOA ELEC-3; Loss Of 4KV ESF Bus; Revision 102

- 1BwOSR 3.8.1.11-1; 1A Diesel Generator Loss of ESF Bus Voltage With No SI Signal;

Revision 16

- 1BwOSR 3.3.2.8-621a; Unit 1 ESFAS Instrumentation Slave Relay Surveillance (Train A FW

Pump Trip, SG Level HI-HI-K621); Revision1

Work Orders:

- 1897940-01; 1A Diesel Generator 24-Hour Load Test and ECCS Surveillance; 04/01/18

- 4774957-10; Loss of ESF Bus 141 During 1A EDG Sequencer Testing; 04/21/18

- 1776181; U1 Train A Slave Relay SRV 3.3.2.8-621-A; 06/04/2018

- 1776186; U1 Train B Slave Relay SRV 3.3.2.8-621-B; 06/05/2018

29