IR 05000313/2019001
| ML19122A484 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/30/2019 |
| From: | O'Keefe N NRC Region 4 |
| To: | Richard Anderson Entergy Operations |
| References | |
| IR 2019001 | |
| Download: ML19122A484 (40) | |
Text
April 30, 2019
SUBJECT:
ARKANSAS NUCLEAR ONE UNITS 1 AND 2 - NRC INTEGRATED INSPECTION REPORT 05000313/2019001 AND 05000368/2019001
Dear Mr. Anderson:
On March 31, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Arkansas Nuclear One Units 1 and 2. On April 3, 2019, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented four findings of very low safety significance (Green) in this report.
Three of these findings involved violations of NRC requirements.
If you contest the violations or significance of the violations documented in this inspection report, 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 IV; the Director, Office of Enforcement; and the NRC resident inspector at Arkansas Nuclear One.
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 IV; and the NRC resident inspector at Arkansas Nuclear One. 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/
Neil OKeefe, Chief Reactor Projects Branch D
Docket Nos. 05000313 and 05000368 License Nos. DPR-51 and NPF-6
Enclosure:
Inspection Report 05000313/2019001 and 05000368/2019001 w/attachment: Inservice Inspection Document Request
Inspection Report
Docket Numbers:
05000313 and 05000368
License Numbers:
Report Numbers:
05000313/2019001 and 05000368/2019001
Enterprise Identifier: I-2019-001-0005
Licensee:
Entergy Operations, Inc.
Facility:
Arkansas Nuclear One, Units 1 and 2
Location:
Russellville, AR
Inspection Dates:
January 1, 2019, to March 31, 2019
Inspectors:
C. Henderson, Senior Resident Inspector
T. DeBey, Resident Inspector
T. Sullivan, Resident Inspector
M. Tobin, Resident Inspector
J. Drake, Senior Reactor Inspector
Approved By:
Neil OKeefe, Chief
Reactor Projects Branch D
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a quarterly inspection at Arkansas Nuclear One 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, violations, and additional items being considered in the NRCs assessment are summarized in the tables below.
List of Findings and Violations
Failure to Translate the Design Requirements into Instructions for Service Water System Sluice Gate Single Failure Cornerstone Significance Cross-cutting Aspect Report Section Mitigating Systems Green NCV 05000313/2019001-01 Closed None 71111.04 The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate design requirements for the Unit 1 service water system into instructions or procedures. Specifically, the plant was licensed to allow manual operator actions to compensate for a single failure of a service water sluice gate during a design basis event (e.g., loss of Lake Dardanelle), but the complex actions were never translated into a procedure.
Welding Programmatic Issue Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events Green NCV 05000368/2019001-02 Open/Closed
[H.5] - Work Management 71111.08P The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion IX, "Control of Special Processes," for a failure to establish measures to assure that welding and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Failure to Take Timely Corrective Actions for a Previously Identified Deficient Condition Associated with Bussmann HKA Fuse Holder Caps Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events Green FIN 05000313/2019001-03 Closed
[H.5] - Work Management 71152 The inspectors identified a self-revealed, Green finding associated with the licensees failure to take timely corrective actions for potential failures identified through industry operating experience. Specifically, the licensee identified a failure mode of Bussmann Model HKA fuse holder caps through industry operating experience and through the corrective action program, determined that it was applicable to Arkansas Nuclear One, and created work orders to replace the susceptible fuse holder caps. However, these work orders were assigned the lowest possible priority in the work control system. An actual failure of a fuse cap occurred in the integrated control system which went undetected, and later prevented a runback from initiating when two out of three condensate pumps were no longer running. This resulted in the plant tripping instead of running back to 40 percent power.
Failure to Adequately Pre-Plan Maintenance that Affected Safety-Related Equipment Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events Green NCV 05000313/2019001-04 Closed
[H.12] - Avoid Complacency 71153 The inspectors identified a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment.
Specifically, the licensee failed to adequately evaluate the use of voltage test probes installed in the negative phase sequence relay within the unit auxiliary transformer feeder breaker to the A1 bus, which resulted in the loss of the nonsafety-related 4160 volt A1 bus and the automatic start of emergency diesel generator 1 to provide power to the safety-related 4160 volt A3 bus.
Additional Tracking Items
Type Issue number Title Inspection Procedure Status LER 05000313/2018-004-00 Reactor Trip Due to the Loss of a Non-Vital 4160 Volt Bus 71153 -
Follow-up of Events and Notices of Enforcement Discretion Closed
PLANT STATUS
Unit 1 began the inspection period at full power. On February 4, 2019, power was lowered to 68 percent as requested by the transmission system operator to facilitate repairs on the 500 kV Mabelvale Line. Unit 1 was returned to full power on February 5, 2019.
Unit 2 began the inspection period at full power. On February 4, 2019, power was lowered to 75 percent as requested by the transmission system operator to facilitate repairs on the 500 kV Mabelvale Line. Unit 2 was returned to full power on February 5, 2019.
On March 15, 2019, Unit 1 lowered power to 60 percent as requested by the transmission system operator to facilitate repairs on the 500 kV Mabelvale Line. Unit 1 was returned to full power on March 16, 2019.
On March 15, 2019, Unit 2 lowered power to 18 percent to facilitate an oil leak repair of reactor coolant pump C and repairs on the 500 kV Mabelvale Line. Unit 2 was returned to full power on March 16,
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
71111.01 - Adverse Weather Protection
External Flooding Sample (IP Section 03.04) (1 Sample)
The inspectors evaluated readiness to cope with external flooding for the following areas:
- Unit 1 train bay hatch HTC-492 controls during resin transfer evaluations on February 18, 2019.
71111.04 - Equipment Alignment
Partial Walkdown (IP Section 02.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1 and Unit 2 emergency cooling pond on January 25, 2019;
- (2) Unit 1 service water pump B aligned to train A service water system during service water bay A maintenance on January 28, 2019;
- (3) Unit 1 remote shutdown on February 1, 2019;
- (4) Unit 2 chemical volume and control system on February 15, 2019.
71111.04S - Equipment Alignment
Complete Walkdown (IP Section 02.02) (1 Sample)
Unit 1 safety-related and nonsafety-related 4160 volt switchgear on March 31, 2019.
71111.05Q - Fire Protection
Quarterly Inspection (IP Section 03.01) (6 Samples)
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 1 and Unit 2 alternate alternating current (AC) generator building on January 16, 2019;
- (2) Unit 1 intake structure during service water bay outage, Fire Area N, on January 18, 2019;
- (3) Unit 2 north electrical equipment room, Fire Area B-3, Zone 2091-BB, on February 11, 2019;
- (4) Unit 1 fire impairment for train bay hatch HTC-492, Fire Area B-1, Zone 197-X-5, on February 25, 2019;
- (5) Unit 2 2A3 switchgear room, Fire Area II, Zone 2101-AA, on February 25, 2019;
- (6) Unit 2 2A4 switchgear room, Fire Area SS, Zone 2100-Z, on February 25, 2019.
71111.06 - Flood Protection Measures
Inspection Activities - Internal Flooding (IP Section 02.02a.) (1 Sample)
The inspectors evaluated internal flooding mitigation protections in the:
Unit 2 safety-related 4160 volt 2A3 and 2A4 switchgear rooms on March 5, 2019.
===71111.08P - Inservice Inspection Activities (PWR) The inspectors evaluated nondestructive examination activities by observing the following activities or reviewing the following records:
02.01 PWR Non-destructive Examination (NDE) Activities and Welding Activities===
The inspectors evaluated nondestructive examination activities by observing the following activities or reviewing the following records:
The inspectors evaluated pressurized water reactor nondestructive testing by observing and reviewing the following examinations from October 1, 2018, to January 3, 2019:
- (1) Ultrasonic Examination
a) Reactor coolant system,28-018, pipe to elbow weld b) Reactor coolant system,28-019, elbow to pipe weld
- (2) Magnetic Particle Examination
a) Reactor vessel head, 2P-55 flued head b) 2P-63 flued head c) 2P-65 flued head
- (3) Visual Examination
a) Steam generator, 2P-32C, lower scroll weld
- (4) Dye Penetrant Examination
a) Safety injection, 2P-32E, lugs
The inspectors evaluated the licensees boric acid control program performance.
The steam generator tube inspections were not required to be performed in this outage.
The reactor vessel bare metal visual examination was not required this outage.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
Unit 1 and Unit 2 downpower due to grid operator request and return to rated thermal power on February 5, 2019.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)
- (1) Unit 1 licensed operator requalification scenario on March 6, 2019;
- (2) Unit 2 just in time training for reactor plant downpower to 18 percent rated thermal power for reactor coolant pump C oil leak repair on March 10, 2019.
71111.12 - Maintenance Effectiveness
Routine Maintenance Effectiveness Inspection (IP Section 02.01) (4 Samples)
The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:
- (1) Unit 1 and Unit 2 alternate AC diesel generator on January 23, 2019;
- (2) Unit 1 and Unit 2 emergency cooling pond on January 31, 2019;
- (3) Unit 2 chemical volume and control system established goals and monitoring for multiple relay failures on February 4, 2019;
- (4) Unit 1 4160 volt safety-related switchgear on February 8, 2019.
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (8 Samples)
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) Unit 2 emergent work for control element assembly system logic power supply failure on January 14, 2019;
- (2) Unit 1 Yellow risk window for service water pump A and diesel drive fire water pump unavailable for service water bay A maintenance on January 23, 2019;
- (3) Unit 1 and Unit 2 diesel fire water pump maintenance outage on January 31, 2019;
- (4) Unit 2 emergent work for an unplanned entry into an Orange risk window for train B containment spray and high pressure safety injection system recirculation function made unavailable on February 8, 2019;
- (5) Unit 2 planned Yellow risk window for containment spray pump B seal cooler maintenance on February 11, 2019;
- (6) Unit 1 emergent work for high pressure injection pump C failure to start on March 6, 2019;
- (7) Unit 1 and Unit 2 Yellow risk window power reductions to support Unit 2 reactor coolant pump C oil leak repair and 500 kV Mabelvale switching activities for tower repairs on March 15, 2019;
- (8) Unit 2 emergent work for train B service water pipe leak on the emergency control room chiller return header on March 28, 2019.
71111.15 - Operability Determinations and Functionality Assessments
Sample Selection (IP Section 02.01) (7 Samples)
The inspectors evaluated the following operability determinations and functionality assessments:
- (1) Unit 1 and Unit 2 emergency cooling pond operability during Unit 1 sluice gate 1 maintenance on January 23, 2019;
- (2) Unit 1 train A 4160 volt operability determination for an undervoltage relay that failed to actuate on January 23, 2019;
- (3) Unit 1 emergency diesel generator 1 and 2 exhaust fans compensatory measures incorporated into procedures on January 24, 2019;
- (4) Unit 1 and Unit 2 emergency cooling pond spillway operability on January 25, 2019;
- (5) Unit 2 motor driven emergency feedwater pump operability following re-installing leads on January 29, 2019;
- (6) Unit 2 2BS-1B operability following failure to meet surveillance testing acceptance criteria on February 14, 2019;
- (7) Unit 1 service water pump A operability with potential foreign material identified during service water bay A maintenance on February 14, 2019.
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Unit 2 temporary modification for steam generator B snubber reservoir catches inside containment on March 14, 2019;
- (2) Unit 1 and Unit 2 permanent modification for 4160 volt buses A1 and 2A1 from manual to automatic transfer to startup transformer 2 on March 31, 2019.
71111.19 - Post Maintenance Testing
Post Maintenance Test Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the following post maintenance tests:
- (1) Unit 1 and Unit 2 diesel fire pump preventive maintenance outage window on February 8, 2019;
- (2) Unit 1 train bay hatch HTC-492 post maintenance testing requirements following resin transfer on February 25, 2019;
- (3) Unit 1 sluice gate 1 maintenance activities during service water bay A maintenance outage window on March 1, 2019;
- (4) Unit 1 service water pump A maintenance activities during service water bay A maintenance outage window on March 1, 2019;
- (5) Unit 1 high pressure injection pump C post maintenance testing following breaker installation after pump failure to start on March 6, 2019;
- (6) Unit 2 reactor coolant pump C oil leak repair post maintenance testing on March 19, 2019.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance tests:
In Service Testing (IST) (IP Section 03.01)
Unit 1 sluice gate 1 leak rate surveillance testing on February 14, 2019.
Reactor Coolant System (RCS) Leak Detection (IP Section 03.01) (1 Sample)
Unit 1 unidentified reactor coolant system leak rate increasing to Tier 3 action levels on February 19, 2019.
Surveillance Testing (IP Section 03.01) (4 Samples)
- (1) Unit 1 reactor protection system channel C monthly surveillance testing on March 14, 2019;
- (2) Unit 2 axial shape index (ASI) surveillance during unit down power to 18 percent rated thermal power on March 15, 2019;
- (3) Unit 2 2BS-1B check closure inservice surveillance testing on March 17, 2019;
- (4) Unit 2 emergency diesel generator 1 18-month surveillance testing (fast start and 24-hour run) on March 20, 2019.
71114.06 - Drill Evaluation
Drill and/or Simulator-Based Licensed Operator Requalification Training (IP Section 02.01)
(see 71114.06 for sample guidance)
The inspectors evaluated:
Unit 2 anticipated transient without a scram and fuel damage on February 6,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===
- (1) Unit 1 (January 1, 2018, through December 31, 2018)
- (2) Unit 2 (January 1, 2018, through December 31, 2018)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
- (1) Unit 1 (January 1, 2018, through December 31, 2018)
- (2) Unit 2 (January 1, 2018, through December 31, 2018)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
- (1) Unit 1 (January 1, 2018, through December 31, 2018)
- (2) Unit 2 (January 1, 2018, through December 31, 2018)
71152 - Problem Identification and Resolution
Annual Follow-up of Selected Issues (IP Section 02.03) (2 Samples)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Unit 1 and Unit 2 Bussmann fuse holder failures and operating experience on March 6, 2019;
- (2) Unit 1 and Unit 2 important to safety relay failures on March 7, 2019.
71153 - Followup of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event report which can be accessed at https://lersearch.inl.gov/LERSearchCriteria.aspx:
- (1) LER 05000313/2018-004-00, Reactor Trip Due to the Loss of a Non-Vital 4160 Volt Bus, (ADAMS Accession: ML19049A011):
The inspectors concluded that no violation of NRC requirements occurred.
INSPECTION RESULTS
Failure to Translate the Design Requirements into Instructions for Service Water System Sluice Gate Single Failure Cornerstone Significance Cross-cutting Aspect Report Section Mitigating Systems
Green NCV 05000313/2019001-01 Closed None 71111.04 The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate design requirements for the Unit 1 service water system into instructions or procedures. Specifically, the plant was licensed to allow manual operator actions to compensate for a single failure of a service water sluice gate during a design basis event (e.g., loss of Lake Dardanelle), but the complex actions were never translated into a procedure.
Description:
The inspectors reviewed Unit 1 Procedure OP-1203.025, Natural Emergencies, Revision 67, Procedure OP-1203.030, Loss of Service Water, Revision 26, and the current licensing and design basis for the service water system prior to the licensee aligning service water pump B to the train A service water system. From this review, the inspectors identified the following:
The Unit 1 safety analysis report (SAR) states that the service water system was required to operate during and following certain natural phenomenon (e.g. loss of Lake Dardanelle). It further describes the single failure analysis for the service water system which credits and describes the operator actions required if a sluice gate fails to open or close (information is contained in SAR Table 9-11, Single Failure Analysis - Service Water System). However, the inspectors were unable to locate the instructions discussed in SAR Table 9-11 in Procedure OP-1203.025 or Procedure OP-1203.030. When questioned, the licensee concluded that there were no instructions that would implement the sluice gate single failure operator actions contained in SAR Table 9-11. The licensee entered this deficiency into the corrective action program as Condition Report CR-ANO-1-2019-00155. This condition could challenge operators when responding to a sluice gate failure. Inspectors determined that the delay in correctly aligning the seven sluice gates could allow some loss of emergency cooling pond (ECP) water inventory, but operators were still expected to be able to make the alignment change without challenging the safety function.
Corrective Action(s): The immediate action was to update Procedure OP-1203.025, Natural Emergencies, Revision 68, to incorporate the sluice gate single failure design requirements contained in SAR Table 9-11.
Corrective Action Reference(s): CR-ANO-1-2019-00155
Performance Assessment:
Performance Deficiency: The licensees failure to assure that service water system design requirements were translated into instructions, procedures, or drawings is a performance deficiency. Specifically, operator manual actions credited to meet the design requirement to ensure that a single failure of a sluice gate will not prevent the safety function of the service water system were described in the SAR but were not translated into a procedure.
Screening: The performance deficiency was more than minor, and therefore a finding, because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to translate service water system sluice gate single failure design into instructions would have complicated operator response during a design basis event (e.g. loss of Lake Dardanelle) to maintain service water system capability.
Significance: Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it:
- (1) was not a design deficiency;
- (2) did not represent a loss of system and/or function;
- (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and
- (4) did not result in the loss of a high safety-significant, nontechnical specification train.
Cross-cutting Aspect: A cross-cutting aspect was not assigned to this finding because the performance deficiency occurred during initial construction, and therefore, is not indicative of current licensee performance.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, from initial licensing until January 2019, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application for the Unit 1 service water system design, was correctly translated into specifications, drawings, procedures, and instructions.
Specifically, the licensee failed to translate the operator manual actions credited to meet single failure design requirements for the Unit 1 service water system sluice gate into instructions or procedures.
Disposition: This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy, because it was very low safety significance (Green) and was entered into the licensees corrective action program as Condition Report CR-ANO-1-2019-00155.
Welding Programmatic Issue Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events
Green NCV 05000368/2019001-02 Open/Closed
[H.5] - Work Management 71111.08P The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, for a failure to establish measures to assure that welding and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Description:
The inspectors reviewed the licensees Weld Program Document EN-DC-328, Entergy Nuclear Welding Program, Revision 4, and the associated procedures. The inspectors noted that licensee Procedure CEP-WP-003, Qualification and Control of Welders, Revision 2, requires that all welders shall be qualified in accordance with the applicable construction code, qualification code, and any supplemental requirements contained in engineering and licensing documents. In addition, licensee Procedure CEP-WP-004, Control and Documentation of Welding Activities, Revision 4, requires the following of the welding supervisor:
1. Assign qualified welders to welding activities after determining a welders qualification
status by reviewing the Welder Qualification and Continuity Report;
2. Review the welding package with assigned welders during pre-job briefings;
3. Provide an appropriate level of direction and oversight to welders while performing
welding activities;
4. Review the welding package including the referenced welding process specifications
with the assigned welder(s) during the pre-job brief;
5. Ensure that the qualification status of assigned welders is verified during the pre-job
brief using an up-to-date welder.
The welder/lead welder is instructed to maintain cognizance of his/her qualifications using the Welder Qualification and Continuity Report, which is issued by the site welding engineer in accordance with Procedure CEP-WP-003.
In accordance with licensee Procedure CEP-WP-005, Control and Issuance of Welding Material, Revision 1, the site welding coordinator ensures that welding material is controlled and issued in accordance with applicable procedures, and provides training to welders, issue station attendants, and personnel authorized to approve weld material requisitions. This procedure also requires that the welding supervisor assign welders to welding activities and determine the correct welding material to be used based on review of the applicable weld procedure, determine a welder's qualification status prior to requisitioning welding material from the issue station, and initiate weld material requisitions or authorizes welders to initiate their own weld material requisitions for assigned welding activities. The weld material request approver approves weld material requisitions prior to requisition of welding material from issue stations.
Notwithstanding these welding program requirements, the following adverse trends and conditions were noted by inspectors:
- On September 16, 2018, Unit 2 main feedwater train B drain line valve failed at weld FW-9C2, resulting in a steam leak upstream of drain valves 2FW-1005A/B. The valve had been replaced during the recent refueling outage in 2017. The failure resulted in a manual trip and unplanned outage to complete repairs. The licensees investigation determined there was evidence of poor welding practices, including grinding on the pipe and filing directly below weld FW-9C2. Personnel had not followed Fleet General Welding Standard CEP-WP-GWS-1, General Welding Standard ASME/ANSI, Revision 3, since there was less than adequate worker oversight and direction, and a lack of welder proficiency testing.
- On January 24, 2016, Unit 1 reactor coolant loop A pump discharge valve vent line had a failure of the socket weld that was installed during the 2015 refueling outage.
The failure resulted in reactor coolant system pressure boundary leakage more than the technical specification allowed leak rate. The licensee inserted a manual trip and entered a forced outage to complete repairs. The licensee determined that the weld did not meet the requirements on the design drawing and that the quality control inspection should have identified the nonconforming weld.
- On three separate occasions, personnel who were not qualified welders in accordance with the licensees weld program, were assigned to work orders, obtained welding materials from issue stations, and welded on American Society of Mechanical Engineers (ASME) Code components in the plant. The valves had to be removed and reinstalled using qualified welders.
As noted in these examples and contrary to the welding program requirements, personnel that were not qualified as welders in accordance with the licensees weld program procedures, were assigned to weld packages, issued welding material from the issue stations, and welded on ASME code components in the plant on multiple occasions. In addition, quality assurance personnel failed to identify rejectable indications on two welds that resulted in the failure of the welds and shutdown of the reactor to repair. During the licensees root cause analysis for the failed welds, which was performed under Condition Report CR-ANO-C-2018-01846, the licensee noted that maintenance leadership did not recognize the impacts of not having a site welding coordinator/supervisor and that this condition was previously noted in Adverse Condition Analysis (ACA) CR-ANO-2-2017-05335, which contained a corrective action to: Identify and designate a full-time Entergy employee as the Site Welding Coordinator to assume the responsibilities from the currently designated contractor. This was an opportunity for the licensee to detect and avert these deficiencies in the welding program.
Corrective Action(s): The plant was shutdown and repairs were made. A review of the welding program was performed and a program improvement plan was implemented.
Corrective Action Reference(s): CR-ANO-1-2018-03567, CR-ANO-1-2018-03729, CR-ANO-2-2016-01071
Performance Assessment:
Performance Deficiency: The licensees failure to establish measures to assure that welding and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements is a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events Cornerstone. It adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, personnel not qualified as welders in accordance with the licensees weld program procedures were assigned to complete weld packages, issued welding material from the issue stations, and welded on ASME code components in the plant on multiple occasions. In addition, quality assurance personnel failed to identify rejectable indications on two welds that resulted in the failure of the weld and shutdown of the reactor to repair. Furthermore, the inspectors determined that this was a programmatic issue since multiple departments and multiple levels of workers and management were responsible for the inadequate measures, which included procedures, training, and oversight, to assure that welding and nondestructive testing were controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Significance: The inspectors assessed the significance of the finding using Appendix A, Significance Determination of Reactor Inspection Findings for At - Power Situations. The inspectors reviewed the initiating events screening questions in Exhibit 1 and determined the finding to be of very low safety significance (Green) because after a reasonable assessment of degradation the leak would not have exceeded the reactor coolant system leak rate for a small-break loss-of-coolant accident (LOCA) and the leak would not have affected other systems used to mitigate a LOCA (e.g., an interfacing system LOCA).
Cross-cutting Aspect: H.5 - Work Management: The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, requires, in part, that the licensee establish measures to assure that welding and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. Contrary to the above, between March 2017, and October 2018, the licensee failed to maintain measures to assure that welding and nondestructive testing were controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
Specifically, the measures implemented by the licensee failed to prevent personnel who were not qualified as welders in accordance with the licensees weld program procedures from being assigned to weld packages, being issued welding material from the issue stations, and welding on ASME code components in the plant on multiple occasions. In addition, quality assurance personnel failed to identify rejectable indications on two welds that resulted in the failure of the weld and shutdown of the reactor to repair.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Take Timely Corrective Actions for a Previously Identified Deficient Condition Associated with Bussmann HKA Fuse Holder Caps Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events
Green FIN 05000313/2019001-03 Closed
[H.5] - Work Management 71152 The inspectors identified a self-revealed, Green finding associated with the licensees failure to take timely corrective actions for potential failures identified through industry operating experience. Specifically, the licensee identified a failure mode of Bussmann Model HKA fuse holder caps through industry operating experience and through the corrective action program, determined that it was applicable to ANO, and created work orders to replace the susceptible fuse holder caps. However, these work orders were assigned the lowest possible priority in the work control system. An actual failure of a fuse cap occurred in the integrated control system (ICS) which went undetected, and later prevented a runback from initiating when two out of three condensate pumps were no longer running. This resulted in the plant tripping instead of running back to 40 percent power.
Description:
On December 18, 2018, the licensee was conducting Baker testing on the train A main chiller motor (VCH-1A) for Arkansas Nuclear One (ANO) Unit 1. While restoring from the Baker testing, an electric arc occurred while removing a voltage test probe from the negative phase sequence relay in switchgear cubicle A-112 (unit auxiliary transformer feeder breaker to the nonvital 4160 volt A1 bus). The arc caused fuses in this circuit to open to protect the control power transformer for the A1 bus. The resulting under voltage condition caused the unit auxiliary feeder breaker to A1 (A-112) to trip and de-energize the A1 bus.
The loss of the A1 bus caused a loss of power to the 4160 volt safety bus A3, causing the automatic start of the red train emergency diesel generator, which restored power to the A3 bus. The A1 bus did not initially transfer to the startup 1 transformer because the main generator was still on line. When the A1 bus de-energized, the A and C condensate pumps de-energized. With only one condensate pump running, a Loss of 2 Condensate Pump ICS runback should have actuated. However, the ICS runback feature did not actuate due the latent failure of the Bussmann Model HKA fuse cap. The plant ultimately tripped on high reactor coolant system pressure and operators safely stabilized the plant in Mode 3. The inspectors concluded that the plant trip was a direct result of the Bussmann Model HKA fuse cap failure, and not as a result of the loss of power to the A1 bus.
The inspectors noted that Condition Report CR-ANO-1-2017-00241 was generated by the licensee on January 23, 2017, to address the operating experience related to Bussmann HKA fuse holder circuits. This condition report states, Multiple instances have been documented within the nuclear industry regarding plant trips, runbacks, and phantom SASS (Smart Automatic Signal Selector) mis-match events resulting from aged, degraded Bussmann HKA cap/light indicators. The spring in the body of the orange fuse cap/indicating lamp is prone to recession into the body of the cap as it is compressed and relaxed repeatedly throughout its life. As a result, the fuse may become loose within the fuse holder, and intermittently open the circuit that it is used for. The 2017 condition report directed the generation of work orders to address all Bussmann HKA fuse holder cap applications within both ANO Unit 1 and Unit 2 to inspect and/or replace Bussmann HKA fuse holder caps. Although Work Order (WO) 494514 was generated to replace the Bussmann HKA fuse holder cap that failed in the ICS system, the work order was not implemented, and the fuse holder cap was not replaced until after Unit 1 had tripped in December 2018 via WO 514989.
Additionally, the inspectors noted that a failure of a Bussmann HKA fuse holder cap occurred prior to this event and was documented in Condition Report CR-ANO-1-2018-00044, dated January 4, 2018, associated with the TT-2679 B feedwater loop temperature transmitter for non-nuclear instrumentation (NNI) cabinet Y. The Bussmann HKA fuse holder cap failure mechanism for the transmitter was identical to the Bussmann HKA fuse holder cap failure mechanism for the ICS system, and to the operating experience discussed in Condition Report CR-ANO-1-2017-00241 (aged/degraded cap spring). The licensee successfully replaced the Bussmann HKA fuse holder cap for the transmitter via WO 492254-01, restoring the transmitter to service. Although the January 2018 condition report made reference to the January 2017 condition report, no action was taken to consider the timeliness of the previous corrective action plan.
The inspectors determined that none of the work orders generated by Condition Report CR-ANO-1-2017-00241, which addressed all safety-related and nonsafety-related Bussmann HKA fuse holder cap applications within both units at ANO, were implemented prior to December 18, 2018. Instead, these work orders were prioritized with the lowest priority that exists in the work management system.
The inspectors concluded that the licensee failed to recognize the potential risk associated with failure of some of the fuse holders of concern. In particular, fuse holders that could fail such that the failure would not be recognized created a latent condition. In the case of the ICS failure, the latent condition would only be recognized during an event, and the failure would (and did) prevent a design feature from mitigating the severity of the event.
The inspectors identified that Procedure EN-LI-102, Corrective Action Program, paragraph 5.9.1.a provides the following; Responsibility for monitoring and reporting on the status of WOs that have CRs/CAs closed to them resides with the production department.
This monitoring is to ensure timely correction of the originally identified condition.
Periodically, typically at least monthly, report to the PRG the status of WOs with CRs/CAs closed to them. The focus of this report should be the monitoring for timely resolution of those work orders per Procedure EN-WM-100. The PRG may require additional details for any WOs not completed in a timely manner. Procedure EN-WM-100, Work Request (WR)
Generation, Screening and Classification, states the following; Work Requests and Work Orders are assigned a priority based upon the significance of the condition identified.
Contrary to these statements, the licensee failed to take timely corrective actions for a previously identified deficient condition to inspect and/or replace the Bussmann HKA fuse holder cap in the ICS system, and did not assign the appropriate priority based upon the significance of the condition identified. This led to a plant trip and challenged safety-related systems.
Corrective Action(s): The licensee replaced the degraded fuse holder cap within the ICS system. The licensee also checked for proper voltage in the C46-1 panel to ensure the health of the eight fuse holder caps that directly power auxiliary relays for plant runback features. All eight were verified to be providing the appropriate voltage.
Corrective Action Reference(s): CR-ANO-1-2018-05360
Performance Assessment:
Performance Deficiency: The licensees failure to take timely corrective actions for operating experience that was applicable to the plant associated with Bussmann HKA fuse holder caps is a performance deficiency.
Screening: The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Initiating Events Cornerstone to limit 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 replace old Bussmann HKA fuse holder caps with a known age-related failure mechanism in a timely manner resulted in the failure of the Loss of 2 Condensate Pump ICS runback feature that was designed to avoid a plant trip. The loss of power to the A1 bus combined with the nonfunctional runback resulted in a plant trip.
Significance: Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it did cause a reactor trip by itself, but did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Cross-cutting Aspect: The finding had a cross-cutting aspect in the area of human performance associated with work management because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the implementation of the work orders was not prioritized commensurate with the safety significance of the systems within which Bussmann HKA fuse holder caps are used in accordance with Procedure EN-WM-100, Work Request (WR)
Generation, Screening and Classification. In general, those systems which would not failsafe if a Bussmann HKA fuse holder cap failed, should have been identified during the risk evaluation process, the work order prioritization process, and the work management process.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding. Specifically, the failure of the ICS fuse holder and the earlier failures did not involves safety-related or quality systems.
Failure to Adequately Pre-Plan Maintenance that Affected Safety-Related Equipment Cornerstone Significance Cross-cutting Aspect Report Section Initiating Events
Green NCV 05000313/2019001-04 Closed
[H.12] - Avoid Complacency 71153 The inspectors identified a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment.
Specifically, the licensee failed to adequately evaluate the use of voltage test probes installed in the negative phase sequence relay within the A-112 unit auxiliary transformer feeder breaker to the A1 bus, which resulted in the loss of the nonsafety-related 4160 volt A1 bus and the automatic start of emergency diesel generator 1 to provide power to the safety-related 4160 volt A3 bus.
Description:
On December 18, 2018, the licensee was conducting Baker testing on the train A main chiller motor. While restoring from the Baker testing, an electric arc occurred while removing a voltage test probe from the A-112 unit auxiliary transformer feeder breaker to the A1 bus. The current from the arc opened fuses that caused an under voltage protective relay to open the unit auxiliary feeder breaker to A-1 (A-112), de-energizing the A1 bus. The loss of the A1 bus caused a loss of power to the 4160 volt safety bus A3, causing the automatic start of the emergency diesel generator (EDG) 1, which restored power to the A3 bus. The A1 bus did not initially transfer to the startup one transformer because the main generator was still on line.
Due to an unrelated fault in the integrated control system (ICS), the plant did not immediately run-back to 40 percent reactor power as designed and tripped instead. The inspectors determined that the loss of the A1 bus did not directly cause a reactor trip by itself; therefore, the loss of the A1 bus and the plant trip were evaluated as separate events. Other than the fault in the ICS system, the plant responded as designed during the trip, and the operators safely stabilized the plant in Mode 3.
The licensee entered this issue into their corrective action program as Condition Report CR-ANO-1-2018-05311 and performed an apparent cause evaluation. The apparent cause evaluation identified several causal factors which contributed to the loss of the A1 bus.
The individual tasked with removing the voltage test probe was not the individual briefed for installing or removing test equipment from the A1 bus. Also, the pre-job risk review focused mainly on the possibility of losing the opposite train main chiller and the potential for a plant trip due to loss of cooling water to reactor building cooled components and did not consider the possibility that the A1 bus may also be adversely impacted. Additionally, as recently as September 2018, and additionally in prior correspondence, the licensee was evaluating the replacement of terminal block screws with threaded banana jack adapters, which would allow testing and maintenance to be performed more safely and efficiently.
The inspectors identified that the licensees on-line risk assessment Procedure EN-WM-104, On Line Risk Assessment, Revision 18, addresses work with the potential for an electrical transient, including blown fuses. Specifically, Procedure EN-WM-104, Attachment 3, Integrated Risk, step 10, defines work with the potential for an electrical transient to include work being performed on energized conductors, lifting energized electrical leads, active troubleshooting in a live AC/DC circuit where a short, daisy chain, blown fuse, open circuit could cause unwanted system actuation and/or isolation. Although the operations impact statement mentioned work being performed within the A-112 unit auxiliary feeder breaker to A-1, there was no mention of any adverse impact to the A1 bus as a result of the electrical maintenance being performed.
Corrective Action Reference: CR-ANO-1-2018-05311
Corrective Actions: The licensee implemented corrective actions which centered around adding more oversight of all plant electrical work activities, testing, calibration, or trouble shooting that requires use of non-engineered test points, jumper installations, or lifted leads while the system is still in service, and if the actions have a potential to cause a plant trip, plant transient, or entry into a limiting condition for operation. The corrective actions included developing work orders and scheduling the installation of engineering test points during bus outages and safety system panel outage windows.
Performance Assessment:
Performance Deficiency: The licensees failure to properly pre-plan maintenance involving the use of voltage test probes installed in the negative phase sequence relay within the A-112 unit auxiliary transformer feeder breaker to the A1 bus which had the potential to impact safety-related equipment is a performance deficiency.
Screening: The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Initiating Events Cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to recognize that the use of voltage test probes installed in the negative phase sequence relay within the A-112 unit auxiliary transformer feeder breaker to the A1 bus created the potential to cause a plant transient that challenged safety-related equipment.
The loss of the A1 bus resulted in a temporary loss of the safety-related A3 bus, which was reenergized by EDG 1.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, and determined that the finding screened to Green because it did not by itself cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
Cross-cutting Aspect: The finding had a cross-cutting aspect in the area of human performance associated with avoiding complacency because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee relied upon past successful outcomes, which resulted in not adequately evaluating risk or implementing required risk management actions, such as the installation of engineered test points, thorough review of planned activities, and a pre-job brief for all participating personnel.
Enforcement:
Violation: Technical Specification 5.4.1.a for Unit 1 requires, in part, that written procedures be established, implemented, and maintained covering the applicable procedures in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a, states, in part, that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings to the circumstances. Procedure EN-WM-104, On-Line Risk, step 4.4, states the following: Perform integrated risk assessments of emergent equipment out of service/unplanned maintenance conditions as soon as practical utilizing Attachment 3, Integrated Risk. Apply a risk informed approach to mitigation of emergent conditions and prioritize recovery of plant systems utilizing Attachment 3, Integrated Risk. Attachment 3, Integrated Risk, step 10, defines work with the potential for an electrical transient to include work being performed on energized conductors, lifting hot leads, active troubleshooting in a live AC/DC circuit where a short, daisy chain, blown fuse, open circuit could cause unwanted system actuation and/or isolation.
Contrary to the above, on December 18, 2018, the licensee failed to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to adequately evaluate the risk of using voltage test probes installed in the negative phase sequence relay within the A-112 unit auxiliary transformer feeder breaker to the A1 bus as required by Procedure EN-WM-104, Attachment 3. This condition led to a plant transient that challenged safety-related equipment. The loss of the A1 bus, which resulted in a temporary loss of the safety-related A3 bus, was reenergized by EDG 1.
Disposition: This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy, because it was of very low safety significance (Green) and was entered into the licensees corrective action program as Condition Report CR-ANO-1-2018-05311.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On January 4, 2019, the inspector presented the inservice inspection results to Mr. J. Kirkpatrick, General Manager Plant Operations, and other members of the licensee staff.
- On April 3, 2019, the inspector presented the quarterly resident inspector inspection results to Mr. R. Anderson, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
71111.01 - Adverse Weather Protection
Condition Reports (CR-ANO-)
1-2019-00256
1-2019-00420
C-2019-00241
C-2019-00252
Procedures
Number
Title
Revision
OP-1015.052
Passive Barrier Breach Permitting Process
Engineering
Reports
Number
Title
Revision
Review of Opening Hatch HTC-492 during Normal
Operation for HELB and Flooding Concerns Identified in
CR-ANO-C-2015-01888
Calculations
Number
Title
Revision
CALC-95-R-0024-
Basic Requirements for the Component Database on
Station Doors and Hatches
71111.04 - Equipment Alignment
Condition Reports (CR-ANO-)
1-2016-04750
1-2019-00155
2-2019-00316
C-2013-01248
C-2019-00177
Work Orders
51660885
2759974
Procedures
Number
Title
Revision
OP-1104.029
Service Water and Auxiliary Cooling System
119
OP-1203.002
Alternate Shutdown
OP-1203.025
Natural Emergencies
67, 68
OP-1203.029
OP-1203.030
Loss of Service Water
OP-1411.084
Unit 1 Sluice Gate and S.W. Bay Cleaning and
Inspection
OP-2102.010
Plant Cooldown
OP-2104.029
Service Water System Operations
113
OP-2105.014
Safety Parameter Display System
OP-2202.001
Standard Post Trip Actions
OP-2202.010
Standard Attachments
Procedures
Number
Title
Revision
OP-2203.012L
Annunciator 2K12 Corrective Action
OP-2203.013
Natural Circulation Operations
OP-2203.014
Alternate Shutdown
OP-2203.030
OP-2203.036
Loss of Charging
Drawings
Number
Title
Revision
M-209, Sheet 1
Piping and Instrument Diagram Circ. Water, Service
Water, and Fire Water Intake Equipment
115
M-2210
Piping and Instrument Diagram, Service Water System
Miscellaneous
Documents
Number
Title
Revision
ANO-1
DCP 80-2123L
STM 2-76
Safety Parameter Display System
Calculations
Number
Title
Revision
CALC-91-E-0099-
ECP Peak Temperature and Inventory Loss Analysis
Summary
CALC-91-E-0099-
ECP Peak Temperature Analysis and Inventory Loss
Analysis Summary
71111.05 - Fire Protection
Condition Reports (CR-ANO-)
1-2019-00256
2-2018-04398
2-2019-00285
C-2019-00241
C-2019-00252
Procedures
Number
Title
Revision
OP-1000.120
ANO Fire Impairment Program
OP-2203.049
Fires in Areas Affecting Safe Shutdown
OP-2405.016
Unit 2 Penetration Fire Barrier Visual Inspection
Miscellaneous
Documents
Number
Title
Revision
Intake Structure
3-AACGB
Miscellaneous
Documents
Number
Title
Revision
Unit 2 Prefire Plan
71111.06 - Flood Protection Measures
Condition Reports (CR-ANO-)
1-2019-00256
2-2015-02369
2-2019-00476
C-2019-00241
C-2019-00252
C-2019-00621
C-2015-01888
C-2015-01929
Procedures
Number
Title
Revision
OP-1015.052
Passive Barrier Breach Permitting Process
OP-2104.029
Service Water System Operations
113
OP-2104.035
Ventilation System Operations
OP-2203.012E
Annunciator 2K05 Corrective Actions
OP-2203.012F
Annunciator 2K06 Corrective Actions
OP-2203.012K
Annunciator 2K11 Corrective Actions
OP-2203.022
Loss of Service Water
OP-2203.051
Miscellaneous
Documents
Number
Title
Revision
ULD-0-TOP-17
ANO Flooding
Engineering
Reports
Number
Title
Revision
Review of Opening Hatch HTC-492 During Normal
Operation for HELB and Flooding Concerns Identified in
CR-ANO-C-2015-01888
Evaluate Internal Flooding Effects Due to CW Pipe
Break or Rupture on the Unit 1 AND Unit 2 Turbine
Buildings CR-ANO-C-2016-00873
HELB Doors and Hatches Documents Clean-Up
Calculations
Number
Title
Revision
CALC-15-E-0007-
Switchgear Rooms 2100 and 2101 Ponding Evaluation
CALC-83-D-2057-
Corridor 2104 - Maximum Depth of Ponding after
Drainage Upgrade
CALC-83-D-2181-
Summary Calc. - Flooding Depth due to Fire Protection
Calculations
Number
Title
Revision
Discharge and Know Elevations of Safety Related
Electrical Equipment
CALC-95-R-0024-
Basic Requirements for the Component Database on
Station Doors and Hatches
CALC-ANOC-CS-
15-00003
ANO Flood Protection Design Basis
CALC-ANOC-CS-
15-00012
ANO Flood Protection Room Evaluations - Unit 1 and 2
71111.08 - Inservice Inspection Activities
Condition Reports (CR-ANO-)
2-2015-03210
2-2015-04604
2-2017-01631
2-2017-01694
2-2017-03138
2-2017-04838
2-2018-01275
2-2018-02092
2-2018-02225
2-2018-02366
2-2018-02377
2-2018-02498
2-2018-02546
C-2017-01783
C-2018-00634
C-2018-02215
C-2018-03783
Work Orders
00436151
00510132
2770764
2774467
Procedures
Number
Title
Revision
CEP-BAC-001
Boric Acid Corrosion Control (BACC) Program Plan
CEP-CII-003
Visual Examinations Of Class MC Components
307
CEP-NDE-0404
Manual Ultrasonic Examination of Ferritic Piping Welds
(ASME XI)
CEP-NDE-0423
Manual Ultrasonic Examination of Austenitic Piping
CEP-NDE-0478
Manual Ultrasonic Examination of Reactor Pressure
Vessel Welds (Section XI App. VIII)
CEP-NDE-0493
Manual Ultrasonic Examination of Reactor Coolant
Pump Flywheel
CEP-NDE-0641
Liquid Penetrant Examination (FT) for ASME Section XI
CEP-NDE-0731
Magnetic Particle Examination (MT) for ASME Section
XI
CEP-NDE-0901
VT-1 Examination
CEP-NDE-0903
VT-3 Examination
CEP-NDE-0955
Visual Examination (VE) of Bare-Metal Surfaces
305
CEP-WP-002
Qualification, Development, and Control of Welding
Procedure Specifications
CEP-WP-003
Qualification and Control of Welders
CEP-WP-004
Control and Documentation of Welding Activities
Procedures
Number
Title
Revision
CEP-WP-005
Control and Issuance of Welding Material
CEP-WP-006
Review and Approval of Vendor Welding Programs
CEP-WP-GBS-1
General Brazing Standard
CEP-WP-GWS-1
General Welding Standard ASME/ANSI
CEP-WP-GWS-2
General Welding Standard Structural Steel AWS D1.1
CEP-WP-GWS-3
General Welding Standard Structural Stainless Steel,
AWS D1.6
CEP-WP-GWS-4
General Welding Standard Sheet Metal AWS D9.1
CEP-WP-GWS-5
General Welding Standard Stud Welding
CEP-WP-IGP-1
Internal Gas Purging
CEP-WP-PHT-1
Preheat and Postweld Heat Treatment Requirements
CEP-WP-RBMD-1 Repair of Base Material Defects
CEP-WP-WIIR-1
Weld Inprocess Inspection Requirements
Entergy Nuclear Welding Program
Boric Acid Corrosion Control Program (BACCP)
Microbiologically Influenced Corrosion (MIC) Monitoring
Program
Entergy EN-DC-174 Engineering Program Section -
ASME Section XI Repair and Replacement and
Containment Repair and Replacement
SEP-BAC-ANO-
001
Boric Acid Corrosion Control Program Inspection and
Identification of Boric Acid Leaks for ANO-1 and ANO-2
71111.11 - Licensed Operator Requalification Program
Condition Reports (CR-ANO-)
C-2019-00379
Procedures
Number
Title
Revision
OP-1102.004
Power Operations
OP-1104.002
Makeup and Purification System
OP-1203.045
Rapid Plant Shutdown
OP-2104.004
Power Operation
OP-2203.053
Rapid Power Reduction
OP-2203.054
Abnormal Grid
Miscellaneous Documents
Title
Scenarios: Internal ICS Failure and Steam Generator Tube Rupture
71111.12 - Maintenance Effectiveness
Condition Reports (CR-ANO-)
1-2018-02847
1-2018-02855
2-2018-00310
2-2018-00529
2-2018-01480
2-2018-01530
2-2018-01743
2-2018-01785
2-2018-01981
2-2018-02003
C-2013-01248
C-2017-02826
C-2017-03230
C-2017-03293
C-2018-00050
C-2018-00519
C-2018-02179
C-2018-02540
C-2018-02925
C-2018-04735
Work Orders
2811
506082
51660885
Procedures
Number
Title
Revision
Planning
OP-1305.029
Off-Site Power Undervoltage and Protective Relay Test
OP-2104.029
Service Water System Operations
113
OP-2305.001
Integrated Engineering Safeguards Test
Drawings
Number
Title
Revision
M-2210
Piping and Instrument Diagram, Service Water System
Miscellaneous
Documents
Number
Title
Revision
ER-ANO-1999-
1559-004
Load Sequencing Timer Replacement Project
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Condition Reports (CR-ANO-)
1-2019-00085
1-2019-00160
1-2019-00228
1-2019-00484
1-2019-00494
2-2019-00047
2-2019-00328
2-2019-00780
Work Orders
515654
516198
513375
519804
2719206
2759974
2774476
Procedures
Number
Title
Revision
OP-1104.029
Service Water and Auxiliary Cooling System
119
OP-1104.032
Sup 2, Surveillance Test of Diesel Fire Pump (P-6B)
OP-1307.004
Diesel Fire Pump Engine (K-5) Biannual Surveillance
OP-1411.084
Unit 1 Sluice Gate and S.W. Bay Cleaning and
Inspection
OP-2104.036
Emergency Diesel Generator Operations
Procedures
Number
Title
Revision
OP-2105.009
CEDM Control System Operation
OP-2203.012J
Annunciator 2K10 Corrective Action
Drawings
Number
Title
Revision
M-209, Sheet 1
Piping and Instrument Diagram Circ. Water, Service
Water, and Fire Water Intake Equipment
115
M-2210
P&ID Service Water System
Miscellaneous
Documents
Number
Title
Revision
2-ONLI-SW-02
Tagout
COPD-024
Risk Assessment Guidelines
Protected Equipment Postings
On-Line Risk Assessment
71111.15 - Operability Determinations and Functionality Assessments
Condition Reports (CR-ANO-)
1-2018-02847
1-2018-02855
1-2018-04294
1-2018-05007
1-2019-00139
1-2019-00140
1-2019-00141
1-2019-00144
1-2019-00154
1-2019-00157
1-2019-00160
2-2018-04446
2-2019-00328
C-2019-00310
Work Orders
516688
2759974
Procedures
Number
Title
Revision
OP-1104.029
Service Water and Auxiliary Cooling System
119
OP-1104.036
Emergency Diesel Generator Operation
OP-1305.02
Off-site Power Undervoltage and Protective Relay Test
OP-1306.037
Surveillance Testing for Unit 1 Sluice Gates SG-1 and
SG-2
OP-1411.084
Unit 1 Sluice Gate and S.W. Bay Cleaning and
Inspection
ULD-0-TOP-03
ANO Unit 1 and Unit 2 Seismic
Drawings
Number
Title
Revision
E-2294, Sheet 1
Schematic Emergency Feedwater Motor Driven
Pump 2P7B Control
Drawings
Number
Title
Revision
E-2613
4160VAC Switchgear (GE Unit 011) ACB 2A311 EFW
Motor Driven Pump 2P7B
M-209, Sheet 1
Piping and Instrument Diagram Circ. Water, Service
Water, and Fire Water Intake Equipment
115
Miscellaneous
Documents
Number
Title
PIF 2-19-0001
Procedure Improvement Form
Engineering
Reports
Number
Title
Revision
Emergency Cooling Pond Spillway Replacement
Calculations
Number
Title
Revision
CALC-11-E-0006-
CALC 83-D-1007-
Mech Calc for Seismic Updgrade of Firewater Header
Intake Structure
CALC-91-E-0099-
ECP Peak Temperature and Inventory Loss Analysis
Summary
CALC-91-E-0099-
ECP Peak Temperature Analysis and Inventory Loss
Analysis Summary
CALC-M-3600-37-
71111.18 - Plant Modifications
Condition Reports (CR-ANO-)
2-2018-04028
2-2018-04506
2-2019-00673
Work Orders
513000
Procedures
Number
Title
Revision
Control of Scaffolding
OP-1015.036
Containment Closeout
OP-2202.003
Loss of Coolant Accident
OP-2202.010
Standard Attachments
Engineering
Reports
Number
Title
Revision
B Steam Generator Snubber and Snubber Reservoir
Catches Review; Containment Impact Review
0, 1
Calculations
Number
Title
Revision
CALC-A-13
Containment Heat Sinks
CALC-A-22
Containment Net Free Internal Volume
CALC-A-132
Steam Generator Sub Compartment Pressure Analysis
CALC-07-E-0016-
Accountability of Unit 2 Containment Strainers Sacrificial
Surface Area
71111.19 - Post Maintenance Testing
Condition Reports (CR-ANO-)
1-2019-00137
1-2019-00160
1-2019-00228
1-2019-00248
1-2019-00256
1-2019-00420
1-2019-00484
1-2019-00494
2-2019-00701
2-2019-00718
2-2019-03580
C-2019-00241
Work Orders
517696
519804
531000
2719206
2759974
2843223
Procedures
Number
Title
Revision
OP-1015.036
Containment Building Closeout
OP-1015.052
Passive Barrier Breach Permitting Process
OP-1104.032
Sup 2, Surveillance Test of Diesel Fire Pump (P-6B)
OP-1306.037
Surveillance Test for Unit 1 Sluice Gate SG-1 and SG-2
OP-1307.004
Diesel Fire Pump Engine (K-5) Biannual Surveillance
OP-1402.240
Inspection of Watertight Hatches
OP-1411.084
Unit 1 Sluice Gate and Service Water Bay Cleaning and
Inspection
Engineering
Reports
Number
Title
Revision
ANO-2 Service Water Pump 2P-4B SS Wet End
Upgrade
Clarify Dimensions of Isolating Gasket Fit to Discharge
Cone and Pump Bowl and Acceptance Criteria for
Electrical Isolation
Document the Purpose of the Dielectric Gasket Between
the Wet Ends and Columns
71111.22 - Surveillance Testing
Condition Reports (CR-ANO-)
1-2011-00184
1-2011-00585
1-2011-01070
1-2015-01741
1-2018-01970
2-1997-00229
2-2017-03771
2-2017-06022
2-2019-00328
2-2019-00336
Work Orders
2680501
27192016
Procedures
Number
Title
Revision
OP-1015.003B
Unit 2 Operations Logs
OP-1104.029
Service Water and Auxiliary Cooling System
119
OP-1304.039
Reactor Protection System Test
OP-1306.037
Surveillance Testing for Unit 1 Sluice Gates SG-1 and
SG-2
OP-1411.084
Unit 1 Sluice Gate and S.W. Bay Cleaning and
Inspection
OP-2104.005
Drawings
Number
Title
Revision
M-209, Sheet 1
Piping and Instrument Diagram Circ. Water, Service
Water, and Fire Water Intake Equipment
115
Miscellaneous
Documents
Number
Title
Revision
ER-ANO-2002-
285-00
Service Water Boundary Valve Leakage Criteria
PIF 2-18-0017
PMRQ/PMID
50016233
SEP-ANO-1-IST-1 ANO Unit 1 Inservice Testing Bases Document
SEP-ANO-1-IST-2 Inservice Testing Plan Site Engineering Program
Section
SEP-ANO-1-IST-3 Inservice Testing Cross Reference Document
Calculations
Number
Title
Revision
CALC-91-E-0099-
ECP Peak Temperature and Inventory Loss Analysis
Summary
CALC-91-E-0099-
ECP Peak Temperature Analysis and Inventory Loss
Analysis Summary
71152 - Problem Identification and Resolution
Condition Reports (CR-ANO-)
1-2017-00241
1-2018-02847
1-2018-02855
1-2018-05360
1-2019-00484
1-2019-00494
1-2019-00531
Work Orders
514989
20790
Procedures
Number
Title
Revision
OP-1305.029
Off-Site Power Undervoltage and Protective Relay Test
Miscellaneous
Documents
Number
Title
Mechanical Failures of Indicating-Type Fuses
Aging Assessment of Safety Related Fuses Used in Low
and Medium Voltage Applications in Nuclear Power
Plants
Part 21
Cracked Ferrules Accession #9612170255
This letter does not contain new or amended information collection requirements subject to the
Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing information collection
requirements were approved by the Office of Management and Budget, Control Number
31500011. The NRC may not conduct or sponsor, and a person is not required to respond to, a
request for information or an information collection requirement unless the requesting document
displays a currently valid Office of Management and Budget control number.
This letter and its enclosure 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, and Requests for
Withholding.
Information Request
September 17, 2018
Notification of Inspection and Request for Information
Comanche Peak Unit 2
NRC Inspection Report 05000446/2018004
INSERVICE INSPECTION DOCUMENT REQUEST
Inspection Dates: November 26 to December 7, 2018
Inspector: Jim Drake, Senior Reactor Inspector
A.
Information Requested for the In-Office Preparation Week
The following information should be sent to the Region IV office in hard copy or electronic
format (ims.certrec.com preferred), in care of Jim Drake, by November 2, 2018, to facilitate the
selection of specific items that will be reviewed during the onsite inspection week. The inspector
will select specific items from the information requested below and then request from your staff
additional documents needed during the onsite inspection week Please provide requested
documentation electronically if possible. If requested documents are large and only hard copy
formats are available, please inform the inspector(s), and provide subject documentation during
the first day of the onsite inspection.
If you have any questions regarding this information request, please call the inspector as soon
as possible.
On November 26, 2018, a reactor inspector from the Nuclear Regulatory Commissions (NRC)
Region IV office will perform the baseline inservice inspection at Comanche Peak Unit 2, using
NRC Inspection Procedure 71111.08, "Inservice Inspection Activities. Experience has shown
that this inspection is a resource intensive inspection both for the NRC inspector and your staff.
The date of this inspection may change dependent on the outage schedule you provide. In
order to minimize the impact to your onsite resources and to ensure a productive inspection, we
have enclosed a request for documents needed for this inspection. The information identified
on this request (Section A) is to be provided prior to the inspection to ensure that the
inspector(s) are adequately prepared. The section identified as Documents Upon Request is
intended to provide guidance to the type of information an inspector(s) will be requesting to
complete the inspection. It is important that all of these documents are up to date and complete
in order to minimize the number of additional documents requested during the preparation
and/or the onsite portions of the inspection (i.e., condition reports with attachments).
We have discussed the schedule for these inspection activities with your staff and understand
that our regulatory contact for this inspection will be Mr. James Barnette of your licensing
organization. The tentative inspection schedule is as follows:
Preparation week: November 13 - 16, 2018
Onsite weeks: November 26 - December 7, 2018
Our inspection dates are subject to change based on your updated schedule of outage
activities. If there are any questions about this inspection or the material requested, please
contact Jim Drake at (817) 200-1558 (email to: James.Drake@nrc.gov).
A.1
ISI/Welding Programs and Schedule Information
1.
A detailed schedule (including preliminary dates) of:
1.1.
Nondestructive examinations planned for ASME Code Class Components performed as
part of your ASME Section XI, risk informed (if applicable), and augmented inservice inspection
programs during the upcoming outage.
1.2.
Examinations planned for Alloy 82/182/600 components that are not included in the
Section XI scope (If applicable)
1.3.
Examinations planned as part of your boric acid corrosion control program (Mode 3
walkdowns, bolted connection walkdowns, etc.)
1.4.
Welding activities that are scheduled to be completed during the upcoming outage
(ASME Class 1, 2, or 3 structures, systems, or components)
2.
A copy of ASME Section XI Code Relief Requests and associated NRC safety
evaluations applicable to the examinations identified above.
2.1.
A list of ASME Code Cases currently being used to include the system and/or
component the Code Case is being applied to.
3.
A list of nondestructive examination reports which have identified recordable or
rejectable indications on any ASME Code Class components since the beginning of the last
refueling outage. This should include the previousSection XI pressure test(s) conducted during
start up and any evaluations associated with the results of the pressure tests.
4.
A list including a brief description (e.g., system, code class, weld category,
nondestructive examination performed) associated with the repair/replacement activities of any
ASME Code Class component since the beginning of the last outage and/or planned this
refueling outage.
5.
If reactor vessel weld examinations required by the ASME Code are scheduled to occur
during the upcoming outage, provide a detailed description of the welds to be examined and the
extent of the planned examination. Please also provide reference numbers for applicable
procedures that will be used to conduct these examinations.
6.
Copy of any 10 CFR Part 21 reports applicable to structures, systems, or components
within the scope of Section XI of the ASME Code that have been identified since the beginning
of the last refueling outage.
7.
A list of any temporary non-code repairs in service (e.g., pinhole leaks).
8.
A copy of (or ready access to) most current revision of the inservice inspection program
manual and plan for the current interval.
9.
Copy of NDE procedures for NDE that will be used during the outage.
10.
Copy of overaching site procedure for welding.
11.
Provide evaluation of operating experience associated with CRDM thermal sleeve wear
issue.
A.2
Reactor Pressure Vessel Head (If Being Inspected, otherwise N/A)
1.
Provide a detailed scope of the planned bare metal visual examinations (e.g., volume
coverage, limitations, etc.) of the vessel upper head penetrations and/or any nonvisual
nondestructive examination of the reactor vessel head including the examination procedures to
be used.
1.1.
Provide the records recording the extent of inspection for each penetration nozzle
including documents which resolved interference or masking issues that confirm that the extent
of examination meets 10 CFR 50.55a(g)(6)(ii)(D).
1.2.
Provide records that demonstrate that a volumetric or surface leakage path examination
assessment was performed.
2.
Copy of current calculations for EDY, and RIY as defined in Code Case N-729-1 that
establish the volumetric and visual inspection frequency for the reactor vessel head and J-
groove welds.
A.3
Boric Acid Corrosion Control Program
1.
Copy of the procedures that govern the scope, equipment and implementation of the
inspections required to identify boric acid leakage and the procedures for boric acid
leakage/corrosion evaluation.
2.
Please provide a list of leaks (including code class of the components) that have been
identified since the last refueling outage and associated corrective action documentation. If
during the last cycle, the unit was shutdown, please provide documentation of containment
walkdown inspections performed as part of the boric acid corrosion control program.
A.4
Steam Generator Tube Inspections
1.
A detailed schedule of:
Steam generator tube inspection, data analyses, and repair activities for the upcoming
outage.
- Steam generator secondary side inspection activities for the upcoming outage.
2.
Copy of steam generator history documentation given to vendors performing eddy
current (ET) testing of the steam generators during the upcoming outage.
3.
Copy of steam generator condition monitoring and operational assessment of previous
cycle. Also include a copy of the following documents as they become available for the current
cycle:
Degradation assessment
Condition monitoring assessment
4.
Define the planned steam generator ET scope (e.g., 100 percent of unrepaired tubes
with bobbin probe and 20 percent sample of hot leg expansion transition regions with rotating
probe) and identify the scope expansion criteria, which will be applied.
5.
Identify and quantify any steam generator tube leakage experienced during the previous
operating cycle. Also provide documentation identifying which steam generator was leaking and
corrective actions completed and planned for this condition.
6.
Copy of steam generator eddy current data analyst guidelines and site validated eddy
current technique specification sheets. Additionally, please provide a copy of EPRI Appendix H,
Examination Technique Specification Sheets, qualification records.
7.
Copy of the guidance to be followed if a loose part or foreign material is identified in the
8.
Provide past history of the condition and issues pertaining to the secondary side of the
steam generators (including items such as loose parts, fouling, top of tube sheet condition, crud
removal amounts, etc.).
Indicate where the primary, secondary, and resolution analyses are scheduled to take place.
A.5
Additional Information Related to all Inservice Inspection Activities
1.
A list with a brief description of inservice inspection, and boric acid corrosion control
program related issues (e.g., CR) entered into your corrective action program since the
beginning of the last refueling outage. For example, a list based upon data base searches
using key words related to piping such as: inservice inspection, ASME Code,Section XI, NDE,
cracks, wear, thinning, leakage, rust, corrosion, boric acid, or errors in piping examinations.
2.
Provide training (e.g. scaffolding, fall protection, foreign material exclusion, confined
space) if they are required for the activities described in A.1 through A.3.
3.
Please provide names and phone numbers for the following program leads:
Inservice inspection (examination, planning)
Containment exams
Snubbers and supports
Repair and replacement program
Licensing
Site welding engineer
Boric acid corrosion control program
Steam generator inspection activities (site lead and vendor contact)
DOCUMENTS UPON REQUEST
Inservice Inspection / Welding Programs and Schedule Information
1.
Updated schedules for inservice inspection/nondestructive examination activities,
including planned welding activities, and schedule showing contingency repair plans, if
available.
2.
For ASME Code Class welds selected by the inspector please provide copies of the
following documentation (as applicable) for each subject weld:
Weld data sheet (traveler).
Weld configuration and system location.
Applicable welding procedures used to fabricate the welds.
Copies of procedure qualification records (PQRs).
Welders performance qualification records (WPQ).
Nonconformance reports for the selected welds (If applicable).
Radiographs of the selected welds and access to equipment to allow viewing
radiographs (if radiographic testing was performed).
Preservice and inservice examination records for the selected welds.
Readily accessible copies of nondestructive examination personnel qualifications
records for reviewing.
3.
For ultrasonic examination procedures qualified in accordance with ASME Code, Section
XI, Appendix VIII, provide documentation supporting the procedure qualification (e.g. the EPRI
performance demonstration qualification summary sheets). Also, include qualification
documentation of the specific equipment to be used (e.g., ultrasonic unit, cables, and
transducers including serial numbers) and nondestructive examination personnel qualification
records.
Boric Acid Corrosion Control Program
1.
Boric acid walk down inspection results, an updated list of boric acid leaks identified so
far this outage, associated corrective action documentation, and overall status of planned boric
acid inspections.
2.
List of boric acid evaluation and corrective action documents associated with the
leakage.
Steam Generator Tube Inspections
1.
Examination technique specification sheets and associated justification for any revisions.
2.
Copy of the eddy current testing procedures used to perform the steam generator tube
inspections (specifically calibration and flaw characterization/sizing procedures, etc.).
3.
Procedures with guidance/instructions for identifying (e.g. physically locating the tubes
that require plugging) and plugging steam generator tubes.
4.
List of corrective action documents generated by the vendor and/or site with respect to
steam generator inspection activities.
Codes and Standards
1.
Ready access to (i.e., copies provided to the inspector(s) for use during the inspection at
the onsite inspection location, or room number and location where available):
Applicable editions of the ASME Code (Sections V, IX, and XI) for the inservice
inspection program and the repair/replacement program.
2.
Copy of the performance demonstration initiative (PDI) generic procedures with the
latest applicable revisions that support site qualified ultrasonic examinations of piping welds and
components (e.g., PDI-UT-1, PDI-UT-2, PDI-UT-3, PDI-UT-10, etc.).
3.
Boric Acid Corrosion Guidebook Revision 1 - EPRI Technical Report 1000975.
SUNSI Review
ADAMS:
Non-Publicly Available
Non-Sensitive
Keyword:
By: NOKeefe
Yes No
Publicly Available
Sensitive
OFFICE
SRI:DRP/D
RI:DRP/D
RI:DRP/D
RI:DRP/D
BC:DRS/EB1
C:DRS/EB2
NAME
CHenderson
MTobin
TSullivan
TDeBey
VGaddy
JDrake
SIGNATURE
cmh
mct
tms
tmd
vgg
JFD
DATE
4/24/2019
4/24/2019
4/25/2019
4/24/2019
4/29/2015
4/29/2015
OFFICE
C:DRS/OB
C:DNMS/RIB
TL:DRS/IPAT
C:DRS/RCB
BC:DRP/D
NAME
GWerner
GWarnick
RKellar
NMakris
NOKeefe
SIGNATURE
GEW
JFK for
RLK
NM
/RA/
DATE
04/25/2019
4/29/2019
4/25/2019
4/25/2019
4/30/2019