IR 05000298/2018004
| ML19028A445 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/28/2019 |
| From: | Jason Kozal NRC/RGN-IV/DRP/RPB-C |
| To: | Dent J Nebraska Public Power District (NPPD) |
| References | |
| IR 2018004 | |
| Download: ML19028A445 (43) | |
Text
January 28, 2019
SUBJECT:
COOPER NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000298/2018004
Dear Mr. Dent:
On December 31, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Cooper Nuclear Station. On January 17, 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 two findings of very low safety significance (Green) in this report.
Both of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
Further, inspectors documented a licensee-identified violation, which was determined to be of very low safety significance (Green) and Severity Level IV, in this report. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
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 IV; the Director, Office of Enforcement; and the NRC resident inspector at the Cooper Nuclear Station.
If you disagree with a cross-cutting aspect assignment 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 the Cooper Nuclear Station. Jr.
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/
Jason W. Kozal, Chief Project Branch C Division of Reactor Projects
Docket No. 50-298 License No. DPR 46
Enclosure:
Inspection Report 05000298/2018004 w/attachments:
1. Documents Reviewed 2. Request for Information for the Inservice Inspection 3. Request for Information for the O
Inspection Report
Docket Number:
05000298
License Number:
Report Number:
Enterprise Identifier: I-2018-004-0003
Licensee:
Nebraska Public Power District
Facility:
Cooper Nuclear Station
Location:
Brownville, Nebraska
Inspection Dates:
October 1, 2018 to December 31, 2018
Inspectors:
P. Vossmar, Senior Resident Inspector
M. Stafford, Resident Inspector
C. Alldredge, Health Physicist
B. Baca, Health Physicist
D. Proulx, Senior Project Engineer
I. Anchondo, Reactor Inspector, Inservice Inspection Activities
R. Deese, Senior Reactor Analyst
Approved By:
J. Kozal
Chief, Project Branch C
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Cooper Nuclear Station 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. NRC-identified violations and additional items are summarized in the tables below. A licensee-identified non-cited violation is documented in the Inspection Results at the end of this report.
List of Findings and Violations
Failure to Identify and Correct Nonconforming Safety-related Relays Cornerstone Significance Cross-cutting Aspect Inspection Procedure Mitigating Systems Green NCV 05000298/2018004-01 Closed P.2 -
Problem Identification and Resolution -
Evaluation 71111.15 -
Operability Determinations and Functionality Assessments The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality associated with safety-related low pressure coolant injection relays installed in reactor recirculation discharge valves RR-MO-53A and RR-MO-53B. Specifically, between October 8, 2018, and November 12, 2018, the licensee failed to identify that three Allen Bradley relays installed in the plant had serial numbers that were identified in a Nutherm Part 21 Notification, and were subject to infant mortality failure during the first 2 months of energized life. As a result, one of the three relays failed shortly after being installed in valve RR-MO-53B, rendering the low pressure coolant injection subsystem B inoperable, and the remaining two relays were inappropriately justified for continued use without consideration of the Part 21 Notification.
Failure to Manage the Increase in Risk During Shutdown Cooling Maintenance Cornerstone Significance Cross-cutting Aspect Inspection Procedure Initiating Events Green NCV 05000298/2018004-02 Closed H.1 - Human Performance-Resources 71153 -
Follow-up of Events and Notices of Enforcement Discretion The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(4) for the licensees failure to manage the increase in risk that resulted from maintenance activities. Specifically, when the licensee removed shutdown cooling from service for planned maintenance, the licensee failed to protect the reactor equipment cooling system which was being utilized as the credited support system to satisfy the shutdown key safety function of decay heat removal.
This resulted in the licensee failing to recognize the risk of performing maintenance on the reactor equipment cooling system and inadvertently isolating cooling flow to the fuel pool cooling heat exchangers which were removing decay heat from the reactor vessel.
Additional Tracking Items
Type Issue Number Title Inspection Procedure Status LER 05000298/2016-001-01 05000298/2016-001-02 De-Energized High Pressure Coolant Injection Auxiliary Lube Oil Pump Caused by Relay Failure Results in Loss of Safety Function, a Condition Prohibited by Technical Specifications, and a 10 CFR Part 21 Report 71153 -
Follow-up of Events and Notices of Enforcement Discretion Closed LER 05000298/2017-001-00 05000298/2017-001-01 Residual Heat Removal Minimum Flow Valves Out of Position Results in Loss of Safety Function and Condition Prohibited by Technical Specifications 71153 -
Follow-up of Events and Notices of Enforcement Discretion Closed LER 05000298/2017-003-01 Mispositioned Control Room Emergency Filter System Supply Fan Damper Causes Loss of Safety Function 71153 -
Follow-up of Events and Notices of Enforcement Discretion Closed URI 05000298/2016008-01 Possible Failure to Ensure that the Assumptions in the Engineering Analysis Remain Valid 71111.05XT -
Fire Protection-NFPA 805 (Triennial)
Closed
PLANT STATUS
The Cooper Nuclear Station began the inspection period in a shutdown status for Refueling Outage 30, which started on September 29, 2018. On November 15, 2018, the station commenced reactor startup and the reactor was made critical. On November 17, 2018, the station synchronized the main generator to the grid and began power ascension. The plant returned to full power on November 20, 2018, and remained there for the rest of the inspection period, with the exception of minor reductions in power for planned rod pattern adjustments.
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
71111.04Equipment Alignment Partial Walkdown
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Fuel pool cooling while credited for decay heat removal on October 25, 2018
- (2) Core spray B while credited as an inventory control system on October 25, 2018
71111.05AQFire Protection Annual/Quarterly Quarterly Inspection
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Reactor water cleanup (RWCU) heat exchanger room, RWCU pump rooms, and reactor building 931 feet elevation, Fire Areas RB-M and RB-N, Zones 3D and 3E, on October 11, 2018
- (2) Control building basement 882 feet elevation, Fire Area CB-A, Zone 7A, on October 23, 2018
- (3) Fuel pool cooling room, Fire Area RB-P, Zone 4C, on October 25, 2018
- (4) Drywell, Fire Area Reactor Building - Drywell, on November 9, 2018
71111.08Inservice Inspection Activities
The inspectors evaluated boiling water reactor nondestructive testing by observing or reviewing the following examinations from October 1, 2018, to October 5, 2018:
- (1) Ultrasonic Testing a) Reactor recirculation pump (RRP-1B-BG1) studs, Report UT-F18-006
- (2) Visual Testing a) Main steam system, mechanical snubber VRS-21, Report VT-F18-037 b) Main steam system, mechanical snubber VRS-22, Report VT-F18-038 c) Reactor feedwater, constant support RFH-71A, Report VT-F18-048 d) Reactor feedwater, constant support RFH-71, Report VT-F18-047 e) Reactor feedwater, mechanical snubber RFS-11, Report VT-F18-049
- (3) Magnetic Particle Testing a) Main steam system, lugs on support PSA-BK1-19, Report MT-F18-003
The inspector evaluated welding activities by observing the following welding activities or reviewing the following records:
a) Residual heat removal system, welds 2 through 9, Weld Record 18-152
The inspector evaluated a sample of condition reports associated with inservice inspection activities and found the corrective actions were appropriate.
71111.11Licensed Operator Requalification Program and Licensed Operator Performance Operator Requalification
The inspectors observed and evaluated just-in-time training for reactor vessel pressure testing on October 28, 2018.
Operator Performance (1 Sample)
The inspectors observed and evaluated reactor startup from the refueling outage, including pulling rods to bring the reactor critical, on November 14, 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) Drywell fan coil unit D bearing failure on November 30, 2018
- (2) Reactor equipment cooling and fuel pool cooling on December 20, 2018
Quality Control (1 Sample)
The inspectors evaluated maintenance and quality control activities associated with the following equipment performance issues:
- (1) Residual heat removal low-pressure coolant injection on December 20, 2018
71111.13Maintenance Risk Assessments and Emergent Work Control
The inspectors evaluated the risk assessments for the following planned and emergent work activities. The inspectors used information from Operating Experience Smart Sample (OpESS) 2007/03, Crane and Heavy Lift Inspection, Supplemental Guidance to IP 71111.20 and IP 71111.13, to inform baseline inspection sample
- (1) below:
- (1) Reactor pressure vessel head removal during disassembly on October 2, 2018
- (2) Emergency station service transformer Yellow risk window during bus replacement on October 3, 2018
- (3) Shutdown cooling Yellow risk window on October 19, 2018
- (4) Reactor recirculation M-G set B speed control failure during reactor pressure vessel pressure test Yellow risk window on November 8, 2018
71111.15Operability Determinations and Functionality Assessments
The inspectors evaluated the following operability determinations and functionality assessments:
- (1) Residual heat removal pump C casing vent leak on October 4, 2018
- (2) Steam dryer lower guide bracket cracking on October 12, 2018
- (3) Level instrumentation operability following pressure transient on November 2, 2018
- (4) Reactor pressure vessel upper top guide cracking identified during in-vessel visual inspection (IVVI) on November 8, 2018
- (5) Part 21 relays installed in reactor recirculation pump discharge valve RR-MO-53A and RR-MO-53B power circuits on November 10, 2018
- (6) Scram outlet valve drain time following tagging error on October 3, 2018
71111.18Plant Modifications
The inspectors evaluated the following permanent modification:
- (1) Reactor core isolation cooling governor control modification on December 14, 2018
71111.19Post Maintenance Testing
The inspectors evaluated the following post-maintenance tests:
- (1) Service water system post loss-of-coolant accident testing following service water piping disassembly on October 19, 2018
- (2) Low pressure permissive interlock functional testing following relay maintenance on October 21, 2018
- (3) Reactor core isolation cooling relay testing following maintenance on October 24, 2018
- (4) Reactor equipment cooling to reactor recirculation lube oil cooler testing on October 25, 2018
- (5) Reactor building ventilation exhaust outlet isolation damper HV-MO-258 and HV-MO-260 testing on October 30, 2018
- (6) Reactor pressure vessel pressure test following maintenance on multiple reactor coolant system valves on November 6, 2018
- (7) High pressure coolant injection digital controller modification testing on November 15, 2018
71111.20Refueling and Other Outage Activities
The inspectors evaluated Refueling Outage 30 activities from September 29, 2018, to November 17, 2018.
===71111.22Surveillance Testing The inspectors evaluated the following surveillance tests:
Routine===
- (1) Division 2 emergency diesel generator sequential load test on October 19, 2018
- (2) Division 1 emergency diesel generator sequential load test on October 23, 2018
- (3) Control rod scram time testing on November 7, 2018
Containment Isolation Valve (1 Sample)
- (1) Main steam isolation valve local leak-rate test on October 15, 2018
71114.06Drill Evaluation Drill/Training Evolution
The inspectors evaluated a full team drill on December 18,
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 (ALARA) Planning and Controls Implementation of ALARA and Radiological Work Controls
The inspectors reviewed ALARA practices and radiological work controls by reviewing the following activities:
- (1) ALARA Package 2017-01, Spent Fuel Pool Cleanup
- (2) ALARA Package 2018-05, RE30 Reactor Disassemble, Refueling, and Reactor Reassembly
- (3) ALARA Package 2018-17, RE30 Miscellaneous Maintenance, Electrical, Instrumentation and Control work not covered under Turbine Generator Building or other ALARA Packages
- (4) ALARA Package 2018-56, Planned Outage in 2018
Radiation Worker Performance (1 Sample)
The inspectors evaluated radiation worker and radiation protection technician performance.
OTHER ACTIVITIES - BASELINE
71151Performance Indicator Verification
The inspectors verified licensee performance indicators submittals listed below:
- (1) MS06: Emergency AC Power Systems (October 1, 2017 - September 30, 2018)
- (2) MS07: High Pressure Injection Systems (October 1, 2017 - September 30, 2018)
- (3) MS08: Heat Removal Systems (October 1, 2017 - September 30, 2018)
- (4) MS09: Residual Heat Removal Systems (October 1, 2017 - September 30, 2018)
- (5) MS10: Cooling Water Support Systems (October 1, 2017 - September 30, 2018)
- (6) OR01: Occupational Exposure Control Effectiveness Sample (April 1, 2017 - September 30, 2018)
- (7) PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (April 1, 2017 - September 30, 2018)
71152Problem Identification and Resolution Annual Follow-up of Selected Issues
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Emergency diesel generator 1 voltage regulator cabinet on December 21, 2018
- (2) Torus-drywell vacuum breaker failure to close on December 11, 2018
71153Follow-up of Events and Notices of Enforcement Discretion Events
- (1) The inspectors evaluated the loss of reactor recirculation pump B during startup on November 18, 2018
- (2) The inspectors evaluated the loss of fuel pool cooling due to reactor equipment cooling isolation on December 19, 2018
- (3) The inspectors evaluated a Notification of Unusual Event due to toxic gas caused by a fire in the radwaste building on December 29, 2018
Licensee Event Reports (3 Samples)
The inspectors evaluated the following licensee event reports which can be accessed at https://lersearch.inl.gov/LERSearchCriteria.aspx:
- (1) Licensee Event Report 05000298/2016-001-01 and 05000298/2016-001-02, De-Energized High Pressure Coolant Injection Auxiliary Lube Oil Pump Caused by Relay Failure Results in Loss of Safety Function, a Condition Prohibited by Technical Specifications, and a 10 CFR Part 21 Report, on November 28, 2018
- (2) Licensee Event Report 05000298/2017-001-00 and 05000298/2017-001-01, Residual Heat Removal Minimum Flow Valves Out of Position Results in Loss of Safety Function and Condition Prohibited by Technical Specifications, on November 29, 2018
- (3) Licensee Event Report 05000298/2017-003-01, Mispositioned Control Room Emergency Filter System Supply Fan Damper Causes Loss of Safety Function, on November 21,
INSPECTION RESULTS
Unresolved Item (Closed)
Possible Failure to Ensure that the Assumptions in the Engineering Analysis Remain Valid URI 05000298/2016008-01 71111.05XT - Fire Protection - NFPA 805 (Triennial)
Description:
During the NRC triennial fire protection inspection in 2016 (see NRC Inspection Report 0500298/2016008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16270A561), the inspectors reviewed the licensees implementation of the monitoring program required in Section 2.6 of National Fire Protection Association (NFPA) Standard 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants. NFPA 805 requires the following in Section 2.6:
Monitoring. A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained, and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid.
The inspectors reviewed selected samples of equipment monitored by the licensee using Procedure 3-CNS-DC-357, National Fire Protection Association (NFPA) 805 Monitoring Program, Revision 0, to ensure that the licensees program properly implemented the requirements of NFPA 805, Section 2.6. The inspectors also reviewed Engineering Report ER-2015-002, National Fire Protection Association (NFPA) 805 Fire Protection Monitoring Program, Revision 2. The inspectors observed that for many components used in the fire probabilistic risk assessment, the unavailability time for those components were monitored using the existing maintenance rule (10 CFR 50.65) monitoring program.
The inspectors noted that the action levels for availability in the maintenance rule monitoring program were greater than the assumptions in the fire probabilistic risk assessment and questioned whether this met the requirement in NFPA 805 to maintain the assumptions in the engineering analysis.
At the time of the original inspection, clarifications of the monitoring program requirements were being discussed between the industry and the NRCs Office of Nuclear Reactor Regulation during periodic public meetings which discussed Frequently Asked Question (FAQ) 10-0059, National Fire Protection Association (NFPA) 805 Monitoring. With the further clarification pending, the inspectors documented the issue as an unresolved item.
On August 7, 2018, staff from the Office of Nuclear Reactor Regulation found that the proposed changes set forth by the Nuclear Industry Institute and industry contained in FAQ 10-0059, National Fire Protection Association (NFPA) 805 Monitoring, Revision 6, were acceptable for meeting the provisions of National Fire Protection Standard 805. The inspectors reviewed the licensees actions relative to the approved guidance in the FAQ and determined that the licensee met the guidance delineated in the FAQ resolution. This guidance ensured the licensee was conducting effective monitoring by ensuring that the increase in core damage frequency was very small (less than 1.0E-6/year) from their monitoring process and actions were taken in the licensees corrective action program to ensure performance of the fire protection program is maintained over the life of the plant.
Based upon the above review, the NRC determined there was no violation of NRC requirements. This Unresolved Item is closed.
Corrective Action Reference: Condition Report CR-CNS-2016-05109
Licensee-Identified Non-Cited Violation
71111.22 - Surveillance
Testing This violation of very low safety-significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: Title 10 CFR 50.55a(g)(4) requires, in part, that components (including supports)that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME Boiler and Pressure Vessel (BPV) Code to the extent practical. Title 10 CFR 50.55a(z) states, alternatives to the requirements of paragraphs
- (b) through
- (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. ASME Section XI, IWB-5222(a) requires, in part, that the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup.
Contrary to the above, between November 2, 2016, and November 5, 2018, the licensee failed to ensure that components that were classified as ASME Code Class 1 met the requirements set forth in Section XI of editions and addenda of the ASME BPV Code to the extent practical, and failed to ensure a proposed alternative was submitted and authorized prior to implementation. Specifically, the licensee failed to meet component testing requirements of ASME,Section XI, IWB-5222(a), during the 2016 refueling outage reactor pressure vessel (RPV) pressure test, and failed to obtain relief from these requirements prior to performance of the test. In particular, the licensee performed RPV pressure testing with certain Class 1 pressure retaining components in a modified valve lineup rather than in the configuration required for normal reactor operation startup, as required by IWB-5222(a). The relief request had been used and approved in previous ASME Code inservice inspection (ISI)intervals, but had been inadvertently omitted in the submittal to the NRC for the 5th ISI interval that began on April 1, 2016. On November 5, 2018, the licensee requested and received emergent relief for testing activities related to the 2018 refueling outage.
Significance/Severity Level: The performance deficiency was more than minor because it was associated with the reactor coolant system (RCS) equipment and the barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (RCS)protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the inspectors determined the finding was of very low safety significance (Green) in accordance with the Exhibit 4 Barrier Integrity screening questions.
Traditional enforcement also applied to this finding because the failure to submit a relief request to the NRC prior to implementation of an alternative to the ASME Code impacted the regulatory process. The inspectors determined that this issue represented a Severity Level IV violation because it was similar to Enforcement Policy Example 6.1.d.2, Violations of 10 CFR 50.59 result in conditions evaluated as having very low safety significance (i.e., Green) by the SDP.
Corrective Action Reference: CR-CNS-2018-07312
Failure to Identify and Correct Nonconforming Safety-related Relays Cornerstone Significance Cross-cutting Aspect Inspection Procedure Mitigating Systems Green NCV 05000298/2018004-01 Closed P.2 - Problem Identification and Resolution - Evaluation
71111.15 - Operability
Determinations and Functionality Assessments The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify and correct a condition adverse to quality associated with safety-related low pressure coolant injection relays installed in reactor recirculation discharge valves RR-MO-53A and RR-MO-53B. Specifically, between October 8, 2018, and November 12, 2018, the licensee failed to identify that three Allen Bradley relays installed in the plant had serial numbers that were identified in a Nutherm Part 21 Notification, and were subject to infant mortality failure during the first 2 months of energized life. As a result, one of the three relays failed shortly after being installed in valve RR-MO-53B, rendering the low pressure coolant injection subsystem B inoperable, and the remaining two relays were inappropriately justified for continued use without consideration of the Part 21 Notification.
Description:
On October 30, 2018, during the licensees refuel outage, the licensee experienced a failure of B reactor recirculation discharge valve RR-MO-53B. The valve lost power and failed as is due to the failure of a relay coil that had been installed in the power circuit for valve RR-MO-53B for approximately 430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />. This relay was an Allen-Bradley commercial relay that had been purchased by Nutherm, had its 24 Vdc relay coil replaced with a 250 Vdc coil, and then had been dedicated for safety-related applications. Initially, the licensee determined that both valve RR-MO-53B and the A reactor recirculation discharge valve RR-MO-53A were inoperable, due to having the same model of relay installed in both locations within a few days of each other, and due to internal operating experience associated with this model of relay. Because these valves are required to close upon a low pressure coolant injection (LPCI) injection signal, the licensee declared both LPCI A and B inoperable.
Subsequently, the licensee replaced the failed RR-MO-53B relay with another of the same model, and performed an evaluation that concluded that these relays would successfully perform their required functions.
The inspectors reviewed the evaluation and the licensees internal operating experience, noting that in April 2016, the licensee experienced a loss of the high pressure coolant injection (HPCI) system due to failure of the same model of relay. That relay had been energized for 133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br /> before it experienced coil failure. The inspectors also identified an NRC Part 21 report that had been made by Nutherm in 2017. This Part 21 report was issued by Nutherm as a result of Coopers 2016 relay failure, and as a follow-up report to an initial report made by Cooper. The Nutherm Part 21 report included the serial numbers for the relays that remained installed in valves RR-MO-53A and RR-MO-53B, and the serial number for the relay that had just failed. The Part 21 stated that the manufacturing flaw that led to Coopers 2016 event presented an infant mortality vulnerability up until 2 months after the relays have been installed and energized.
On November 11, 2018, the residents challenged the licensees conclusion that the two relays installed in the power circuits for valves RR-MO-53A and RR-MO-53B would function properly and that the LPCI A and B subsystems were operable. Specifically, the inspectors observed that the licensee had experienced two failures of the same model of relay over the course of the last 2 years, despite the fact that a total of only 12 relays of this type had been installed in the plant during the same time span. The inspectors also presented the licensee with the Nutherm Part 21, highlighting the 2 month vulnerability period.
The inspectors observed that the licensees evaluation had not considered the Nutherm Part 21 report, and the licensee was not aware of it. The inspectors observed that the RR-MO-53A and RR-MO-53B relays had been installed on October 20 and October 31, 2018, which was within the period of vulnerability. The licensee reviewed the Part 21 report, determined that the operability of valves RR-MO-53A and RR-MO-53B was not assured, and declared LPCI A and B inoperable for this condition. The licensee took action to replace the vulnerable relays with the previously installed relays, which still had 18 months of environmentally qualified (EQ) life remaining.
Corrective Actions: The licensee replaced the Part 21 nonconforming relays with the previously installed relays in both the RR-MO-53A and B valve control logic cabinets, since these relays had 18 months remaining on their EQ life. The licensee also took action to send the failed RR-MO-53B relay to the vendor for failure mode evaluation.
Corrective Action References: CR-CNS-2018-07146; CR-CNS-2018-07207; CR-CNS-2018-07608; CR-CNS-2018-07620
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to identify and correct a condition adverse to quality associated with LPCI relays installed in the reactor recirculation discharge valves was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the inoperability of low pressure coolant injection A and B subsystems.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014. Using Exhibit 3, Mitigating Systems Screening Questions, the inspectors determined that this finding was of very low safety significance (Green) because the finding:
- (1) was not a deficiency affecting the design or qualification of a mitigating system, structure, or component where operability was maintained;
- (2) did not represent a loss of system and/or function;
- (3) did not represent an actual loss of function for greater than technical specification allowed outage times;
- (4) did not represent an actual loss of function of one or more highly safety-significant nontechnical specification trains for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the cavity flooded;
- (5) did not degrade a functional residual heat removal auto-isolation;
- (6) did not pertain to external events; and
- (7) did not pertain to the fire brigade.
Cross-cutting Aspect: This finding had a problem identification and resolution cross-cutting aspect associated with evaluation, in that the organization failed to thoroughly evaluate premature Allen Bradley relay failures and the applicable Nutherm Part 21 Notification, in order to ensure that the resolution addressed causes and extent of conditions commensurate with their safety significance. [P.2]
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, between October 8, 2018, and November 12, 2018, the licensee failed to establish measures to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances were promptly identified and corrected. Specifically, the licensee failed to identify that three quality-related LPCI relays installed in RR-MO-53A and RR-MO-53B control circuitry had serial numbers that were identified in a Nutherm Part 21 Notification, and were subject to infant mortality failure during the first 2 months of energized life. As a result, one of the three relays failed shortly after being installed in RR-MO-53B, rendering the LPCI B subsystem inoperable, and the remaining two relays were inappropriately justified for continued use without consideration of the Part 21 Notification.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Manage the Increase in Risk During Shutdown Cooling Maintenance Cornerstone Significance Cross-cutting Aspect Inspection Procedure Initiating Events Green NCV 05000298/2018004-02 Closed H.1 - Human Performance - Resources 71153 - Follow-up of Events and Notices of Enforcement Discretion The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(4) for the licensees failure to manage the increase in risk that resulted from maintenance activities. Specifically, when the licensee removed shutdown cooling from service for planned maintenance, the licensee failed to protect the reactor equipment cooling system which was being utilized as the credited support system to satisfy the shutdown key safety function of decay heat removal. This resulted in the licensee failing to recognize the risk of performing maintenance on the reactor equipment cooling system and inadvertently isolating cooling flow to the fuel pool cooling heat exchangers which were removing decay heat from the reactor vessel.
Description:
In order to help manage the increase in risk when a piece of equipment is taken out of service for maintenance, the licensee will protect other important systems or trains per Administrative Procedure 0-PROTECT-EQP, Protected Equipment Program, Revision 44.
Section 4.1.2.1 of this procedure outlines the necessary equipment to be protected during the outage. It states, in part, that the equipment to be protected is based upon protecting the key safety function of decay heat removal.
Administrative Procedure 0.50.5, Outage Shutdown Safety, Revision 39, defines the shutdown key safety functions and what equipment can be credited in order to meet these functions. Attachment 1 of this procedure states, in part, that General Operating Procedure 2.1.20.2, Cycle Specific Fuel Transfer and Alternate Cooling Guideline, Revision 21, provides guidance on alternate system lineups needed to accomplish the decay heat removal safety function. These alternate systems would be utilized when the residual heat removal (RHR) pumps are removed from service or when RHR shutdown cooling (SDC)is removed from service.
Procedure 2.1.20.2, Section 3.2.2, outlines specific equipment lineups, called cases, which have adequate capacity to be credited to remove heat from the reactor vessel. Each of these cases requires reactor equipment cooling (REC) flow to the fuel pool cooling (FPC) heat exchangers. Based on this requirement, and per the protected equipment procedure, the inspectors determined that the licensee should have protected the REC system when RHR SDC was removed from service.
On October 17, 2018, during the recent refueling outage, the licensee removed the RHR SDC subsystem from service for planned maintenance. At this stage of the refueling outage, the licensee can utilize the FPC system as a backup to the RHR SDC subsystem. The FPC system removes decay heat from the reactor vessel through as many as three heat exchangers, all of which are cooled by the REC system. When the RHR SDC subsystem was removed from service the licensee had not protected the REC system.
On October 21, 2018, the licensee authorized post work testing to be performed on a portion of the REC system. During this testing, a low pressure condition in the REC system caused the REC to FPC heat exchangers supply valve to go closed, thereby removing the flow path to remove decay heat from the reactor vessel. The licensee was able to reestablish flow to the heat exchangers in approximately 1 minute. Had this equipment been protected, the licensee could have recognized the risk impacts that this testing had to the plant and postponed the test or taken additional risk mitigation actions to ensure the isolation would not have occurred.
The inspectors determined this non-cited violation was NRC identified because earlier in the outage, around October 7, 2018, the residents challenged the lack of REC protection and the licensee failed to take action to protect the system. Additionally, following the unplanned isolation, the residents pointed out to the licensee that the shutdown risk management process outlined above should have driven the licensee to protect the REC system prior to performing the shutdown cooling maintenance.
Corrective Actions: The licensee protected the REC system for the remainder of the RHR SDC maintenance period.
Corrective Action References: CR-CNS-2018-06797, CR-CNS-2018-07108
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to manage the increase in risk that resulted from maintenance activities associated with removing shutdown cooling from service was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it adversely affected the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to protect the reactor equipment cooling system while shutdown cooling was removed from service led to the licensee inadvertently isolating cooling flow to the fuel pool cooling heat exchangers, thereby changing the shutdown equipment lineup and challenging the shutdown key safety function of decay heat removal.
Significance: The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, Appendix G, Exhibit 2, Initiating Events Screening Questions, dated May 9, 2014. The inspectors determined that this finding was of very low safety significance (Green) because the finding:
- (1) did not increase the likelihood of a shutdown initiating event;
- (2) did not pertain to a loss of inventory initiator;
- (3) did not pertain to a loss of offsite power initiator;
- (4) occurred when the refuel canal/cavity was flooded;
- (4) did not pertain to a loss of level control; and
- (5) did not increase the likelihood of a fire or internal/external flood that could cause a shutdown initiating event.
Cross-cutting Aspect: This finding had a human performance cross-cutting aspect associated with resources, in that the licensee failed to ensure that the protected equipment procedure was adequate to support successful work performance, and thereby, nuclear safety. [H.1]
Enforcement:
Violation: Title 10 CFR 50.65(a)(4) requires, in part, that prior to performing maintenance activities, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities.
Contrary to the above, from October 17, 2018, to October 22, 2018, prior to performing a maintenance activity, the licensee failed to assess and manage the increase in risk that could result from the proposed maintenance activity. Specifically, when the licensee removed shutdown cooling from service for planned maintenance, the licensee failed to protect the reactor equipment cooling system which was being utilized as the credited support system to satisfy the shutdown key safety function of decay heat removal. This resulted in the licensee failing to recognize the risk of performing maintenance on the reactor equipment cooling system and inadvertently isolating cooling flow to the fuel pool cooling heat exchangers which were removing decay heat from the reactor vessel.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
On October 5, 2018, the inspector presented the inservice inspection results to Mr. J. Dent, Jr.,
Site Vice President and Chief Nuclear Officer, and other members of the licensee staff. The inspector verified no proprietary information was retained or documented in this report.
On October 19, 2018, the inspectors presented the occupational radiation safety inspection results to Mr. J. Dent, Jr., Site Vice President and Chief Nuclear Officer, and other members of the licensee staff. The inspectors verified no proprietary information was retained or documented in this report.
On January 17, 2019, the inspectors presented the quarterly resident inspector inspection results to Mr. J. Dent, Jr., Site Vice President and Chief Nuclear Officer, and other members of the licensee staff. The inspectors verified no proprietary information was retained or documented in this report.
DOCUMENTS REVIEWED
71111.04 - Equipment Alignment
Condition Reports (CR-CNS-)
2018-06814
2018-06828
2018-06949
2018-06950
2018-06951
Procedures
Number
Title
Revision
2.2.32
Fuel Pool Cooling and Demineralizer System
2.2.32A
Fuel Pool Cooling and Demineralizer System
Component Checklist
2.2.32B
Fuel Pool Cooling and Demineralizer System Instrument
Valve Checklist
2.2.69.2
RHR System Shutdown Operations
101
2.2A.CS.DIV2
Core Spray Component Checklist
2.2B.CS.DIV2
Core Spray System Instrument Valve Checklist (Div 2)
Drawings
Number
Title
Revision
2045
Flow Diagram - Core Spray System
N58
Miscellaneous Documents
Title
Date
Protected Equipment Tracking Form
October 25,
2018
Protected Equipment Tracking Form - RE30 Div 2 Protected ADHR
October 24,
2018
71111.05 - Fire Protection
Condition Reports (CR-CNS-)
2018-07604
Procedures
Number
Title
Revision
0.7.1
Control of Combustibles
0.23
CNS Fire Protection Plan
0-BARRIER-
MAPS
Barrier Maps
0-BARRIER-MISC Miscellaneous Buildings
2.1.6
Primary Containment Access Preparation and Closeout
Activities
3.6.1
Fire Barrier Control
Drawings
Number
Title
Revision
2016, Sheet 7
Fire Protection System Site Plan
Miscellaneous
Documents
Number
Title
Revision
CNS-FP-218
Reactor Building - Second Floor
AB/11
CNS-FP-224
Control Building Basement Floor
AB/05
NEDC 10-080
Fundamental Fire Protection Program and Design
Element Review EPM Report R1906-002-001
NEDC 11-104
Fire Safety Analysis for Fire Area AB/05
Control Building Basement Floor Elevation 882
71111.08 - Inservice Inspection Activities
Condition Reports (CR-CNS-)
2016-06259
2016-06277
2016-06453
2016-07398
2016-08790
2018-00599
2018-05403
2018-05439
2018-05470
2018-05863
2018-05898
2018-05908
2018-05909
Work Orders
5062934
5153932
5172664
5173010
5173011
5183072
234301
239533
Procedures
Number
Title
Revision
7.2.57
ASME Category F-A Component Supports Examination
and Adjustments
7.7.3.1
General Welding Standard for ASME and ANSI Code
Applications
7.7.50.1
Visual Inspection Procedure for ANSI B31.1
Drawings
Number
Title
Revision
RR-H7-A
Pipe Support - RR System
N00
Miscellaneous
Documents
Number
Title
Revision/Date
Cooper Nuclear Station 5th 10-Year Interval Inservice
Inspection Program and 3rd 10-Year Interval
Containment Inservice Inspection Program
August 27, 2018
EDP-50
Processing ASME Class 2 and 3 Pressure Boundary
Integrity Challenges
Miscellaneous
Documents
Number
Title
Revision/Date
HEH-VT-103
Procedure for VT-3 Examination
NEDC 90-222
Calc. No. 150-88-S-RR-005 for Pipe Support Mark
No. RR-H7A
May 2, 1990
Liquid Penetrant Examination
71111.11 - Licensed Operator Requalification Program
Condition Reports (CR-CNS-)
2018-07742
2018-07798
2018-07830
2018-07831
Procedures
Number
Title
Revision
2.0.3
Conduct of Operations
2.1.1
Startup Procedure
194
2.1.10
Station Power Changes
116
2.2.75
Steam Sealing System
6.MISC.502
ASME Class 1 System Leakage Test
Miscellaneous
Documents
Number
Title
Revision
TS LCO 3.0.4b Risk Evaluation - Mode 2 and 1 Entry
with Drywell Atmospheric Monitoring System Inoperable
November 13,
2018
RMP 31-001
Beginning of Cycle Startup Reactivity Maneuvering Plan 0
71111.12 - Maintenance Effectiveness
Condition Reports (CR-CNS-)
2018-02869
2018-03244
2018-03245
2018-03260
2018-04693
2018-04901
2018-05421
2018-05438
2018-05845
2018-06137
2018-06193
2018-06238
2018-06239
2018-06347
2018-06797
2018-06924
2018-06950
2018-07146
2018-07207
2018-07279
2018-07311
2018-07608
2018-07620
2018-07859
2018-08019
Work Orders
5172338
273392
273393
Procedures
Number
Title
Revision
2.2.40
Drywell Cooling System
3-EN-DC-204
Maintenance Rule Scope and Basis
3C1
3-EN-DC-205
Maintenance Rule Monitoring
5C1
Drawings
Number
Title
Revision
22, Sheet 3
Primary Containment Cooling & Nitrogen Inerting
System
Miscellaneous
Documents
Number
Title
Revision/Date
Maintenance Rule Function FPC-F01 Performance
Criteria Basis
Maintenance Rule Function HV-F09 Performance
Criteria Basis
Maintenance Rule Function REC-F01 Performance
Criteria Basis
Maintenance Rule Functional Failure Evaluations for
HV-F09
2004 through
2018
Nutherm Part 21 Evaluation - Allen Bradley Model
700DC Control Relay
August 30, 2017
2A1271AC
Drywell Cooling System
CNS-HV-39
Reactor Building Drywell Cooling Developed Flow
Diagram w/ Measurement & Damper Locations
NEDC 89-1439
Drywell Air Cooling System Upgrade Air Flow Derivation 3
Industry Guideline for Monitoring The Effectiveness of
Maintenance at Nuclear Power Plants
4F
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Condition Reports (CR-CNS-)
2018-05835
2018-05853
2018-05864
2018-06308
2018-06309
2018-06313
2018-06694
2018-07300
2018-07359
Work Orders
5173242
273261
Procedures
Number
Title
Revision
0.50.5
Outage Shutdown Safety
0-PROTECT-EQP Protected Equipment Program
2.2.69.2
RHR System Shutdown Operations
101
6.RR.601B
RRMG B Positioner Maintenance and Setting Electrical
and Mechanical Stops
7.4Disassembly
Reactor Vessel Disassembly
Drawings
Number
Title
Revision
045107-01
Interconnect Wiring SM-5360/AD-9120 Relay Panel
N04
730E197BB,
Sheet 4
Variable Speed Recirc Pump & MG Set Recirculation
System
AM/31
730E197BB,
Sheet 6
Variable Speed Recirc Pump & MG Set Recirculation
System
N14
Miscellaneous
Documents
Number
Title
Revision/Date
Outage System Status Report
October 3, 2018
Outage System Status Report
October 17-18,
2018
Protected Equipment Tracking Form - RE30 Initial
Protected Equip with ESST Unavailable
September 11,
2018
Protected Equipment Tracking Form - RE30 Initial
Protected Equip with ESST Unavailable
October 2, 2018
11564325
Notification
EE 01-088
Assessment of the Possibility of Fuel Damage From a
Drop of the Concrete Shield Plugs, Drywell Head, RPV
Head Insulation, RPV Head, Bellows Shield, Steam,
Dryer, MS Line Plugs, and Steam Separator During
Lifting or Movement of these Structures
EE 14-030
Evaluation in Support of Reactor Vessel Head Drop
Analysis
NEDC 94-033
Evaluation/Structural Analysis of Cooper RPV Head,
Steam Dryer, Shroud Head/Steam Separator Assembly
Drop Conditions
RE30-0007
ESST and SDG Unavailable Prior to Backfeed
September 29,
2018
RE30-0009
R101/R102 Containment Closure Plan
September 29,
2018
71111.15 - Operability Determinations and Functionality Assessments
Condition Reports (CR-CNS-)
2018-05786
2018-05817
2018-05820
2018-05866
2018-05879
2018-05894
2018-05965
2018-05987
2018-06420
2018-07134
2018-07146
2018-07207
2018-07608
2018-07620
Work Orders
5172338
270466
273392
273393
Procedures
Number
Title
Revision
0.50.5
Outage Shutdown Safety
6.LOG.602
Daily Surveillance Log - Modes 4 and 5
6.1CSCS.301
CSCS Water Level Channel Calibration (Div 1)
Drawings
Number
Title
Revision
26, Sheet 1
Flow Diagram Reactor Vessel Instrumentation
AC/68
115D6009
Rack 25-5
N09
CNS-NBI-10
Reactor Water Level Indication Correlation
N06
Miscellaneous
Documents
Number
Title
Date
Nutherm Part 21 Evaluation - Allen Bradley Model
700DC Control Relay
August 30, 2017
Safety Evaluation by the Office of Nuclear Reactor
Regulation Related to Amendment No. 260 to Renewed
Facility Operating License No. DPR-46 Nebraska Public
Power District Cooper Nuclear Station Docket
No. 50-298
August 1, 2018
11556454
Notification
Stop Drill for Steam Dryer Skirt Crack Mitigation Near
180 Degrees
71111.18 - Post Maintenance Testing
Procedures
Number
Title
Revision
6.RCIC.102
6.RCIC.105
RCIC Turbine Overspeed Functional Test
6.RCIC.502
ASME Section XI System Leakage Test of the Class 2
Reactor Core Isolation Cooling (RCIC) System in Steam
Tunnel
Special Procedure
SP18-001
RCIC Turbine Control Upgrade Test
Miscellaneous
Documents
Number
Title
28320
Change Evaluation Document, RCIC Turbine Governor
Control System Replacement
71111.19 - Post Maintenance Testing
Condition Reports (CR-CNS-)
2018-06622
2018-06664
2018-06666
2018-06902
2018-06903
2018-07064
2018-07184
2018-07280
2018-07301
2018-07312
2018-07314
2018-07319
2018-07348
2018-07349
2018-07366
2018-07372
2018-07444
2018-07827
2018-07828
Work Orders
5039053
5149734
5172612
5172674
5172870
5173263
5173264
238288
271305
Procedures
Number
Title
Revision
6.CSCS.302
CS, RR, and RHR Valve Low Pressure Permissive
Interlock Functional Test
6.HPCI.313
HPCI (<165 PSIG) Beginning of Cycle Test
6.MISC.502
ASME Class 1 System Leakage Test
53, 54
6.SC.201
Secondary Containment (Reactor Building H&V) Valve
Operability Test
6.SW.102
Service Water System Post-LOCA Flow Verification
7.0.5
CNS Post-Maintenance Testing
7.3.1.16
GE CFD (87) Relay Testing and Maintenance
Miscellaneous Documents
Title
Date
RE30 Fall 2018 6.MISC.502 System Leakage Test PWT List
November 2,
2018
Drawings
Number
Title
Revision
2031, Sheet 3
Flow Diagram Reactor Building - Closed Cooling Water
System
AB/34
791E264, Sheet 3 RCIC Control Power
71111.20 - Refueling and Other Outage Activities
Condition Reports (CR-CNS-)
2018-05791
Procedures
Number
Title
Revision
0.50.5
Outage Shutdown Safety
2.2.32
Fuel Pool Cooling and Demineralizer System
100
2.2.69.2
RHR System Shutdown Operations
101
71111.22 - Surveillance Testing
Condition Reports (CR-CNS-)
2018-05904
2018-05907
2018-06714
2018-06728
2018-06737
2018-06759
2018-07312
2018-07395
2018-07404
2018-07405
2018-07406
2018-07408
2018-07425
2018-07431
Work Orders
5166554
5166628
Procedures
Number
Title
Revision
2.1.5
Reactor Scram
6.PC.513
Main Steam Local Leak Rate Tests
6.1DG.302
Undervoltage Logic Functional, Load Shedding, and
Sequential Loading Test (DIV 1)
6.2DG.302
Undervoltage Logic Functional, Load Shedding, and
Sequential Loading Test (DIV 2)
10.9
Control Rod Scram Time Evaluation
Miscellaneous
Documents
Number
Title
2018-034
Engineering Report, RE30 Past Operability of MSIVs
and MS Pathway
Drawings
Number
Title
Revision
2041
Flow Diagram Reactor Building Main Steam System
71114.06 - Drill Evaluation
Condition Reports (CR-CNS-)
2018-08527
2018-08528
2018-08530
Procedures
Number
Title
Revision
EPIP 5.7.6
Notification
Miscellaneous Documents
Title
Date
EP Drill Scenario Guide
December 18,
2018
71124.01 - Radiological Hazard Assessment and Exposure Controls
Condition Reports (CR-CNS-)
2017-03176
2017-03218
2017-03279
2017-03285
2017-04128
2017-04132
2017-04187
2017-05412
2017-05538
2017-05683
2017-05895
2017-05924
2017-06407
2017-07042
2017-07355
2018-01234
2018-01296
2018-01378
2018-01772
2018-02328
2018-02376
2018-03125
2018-03954
2018-03990
2018-04488
2018-04795
Building
Elevation
Area
RWP/SWP
Date
Reactor
859
NW Quad
2017-001
May 31, 2017
ARW
903
HIC Pit
2017-101
June 14, 2017
Radwaste
903
Filter Demin Valve Room
2017-001
June 21, 2017
Reactor
903
2017-001
June 27, 2017
Reactor
859
E Sump
2017-051
July 10, 2017
Reactor
903
TIP Room
2017-052
July 12, 2017
Reactor
859
HPCI Quad
2018-051
January 17, 2018
ARW
903
HIC Pit
2018-001
February 21, 2018
Reactor
859
2018-051
March 12, 2018
Reactor
976
HV-Fan (EF-R-1A) Ducting
2018-051
April 10, 2018
ARW
903
HIC Pit
2018-102
August 15, 2018
Reactor
1001
Spent Fuel Pool
2018-060
August 24, 2018
Audits and Self-
Assessments
Number
Title
Date
Radiation Protection Program Annual Report (2017)
QA Audit 18-05
Radiological Controls
August 23, 2018
Procedures
Number
Title
Revision
7.4.32
Work Over, Near, or In Reactor Vessel, Dyer/Separator,
Storage Pool, or Spent Fuel Storage Pool
Radiation Worker Expectations
Access Control for Radiologically Controlled Areas
Personnel Contamination
Radiation Protection Posting and Labeling
Response to Contaminated Spills/Leaks
Radioactive Material Control
Radiological Controls for Highly Radioactive Objects
Radiological Control
Procedures
Number
Title
Revision
Radiological Survey Documentation
9.ENN-RP-106-1
Radiation and Contamination Surveys
9.RADOP.1
Radiation Protection at CNS
9.RADOP.2
Radiation Safety Standards and Limits
9.RADOP.5
Airborne Radioactivity Sampling
9.RADOP.10
Radioactive Sources Control and Accountability
9.RADOP.21
Radiological Control of Systems with Potential for
Changing Radiological Conditions
Miscellaneous Documents
Title
Date
Personnel Contamination Monitor (PCM) Alarm Report: October 4 - 16,
2018
October 16, 2018
Source Report - Inventory Only
October 17, 2018
Source Report - Leak Test Source
October 17, 2018
Radiation Work
Permits
Number
Title
Revision
2018-529
Under Vessel Misc (TIP Tubing, Connectors,
RPIS I&C, Cable Repairs, LPRMs, Decon, IRM/SRM
(High Risk)
2018-537
Misc - Maint., Electrical, I/C Support in SWP Areas
(NOT for TG Bldg. Work)
2018-547
Refuel Floor Misc. - SWP
2018-549
Cavity and DS Pit Decon
2018-556
RWCU-AOV-13A/B and RWCU-AOV-14A/B
Replacements in RWCU Valve Room
Radiation Surveys
71124.02 - Occupational As Low As Reasonably Achievable (ALARA) Planning and Controls
Condition Reports (CR-CNS-)
2016-07888
2017-04132
2017-04187
2017-04201
2017-04288
2017-05332
2017-06352
2017-06638
2017-07355
2018-01725
2018-03954
2018-06530
ALARA Planning, In-Progress Reviews, and Post-Job Reviews
Number
Title
2017-01
Spent Fuel Pool Cleanup
2018-05
RE30 Reactor Disassemble, Refueling, and Reactor Reassembly
2018-17
RE30 Misc. Maintenance, Electrical, I/C work not covered under TG
Building or other ALARA Packages
2018-56
Planned Outage in 2018
Audits and Self-
Assessments
Number
Title
Date
Year CRE Reduction Plan 2017-2021
Radiation Protection Program Annual Report (2017)
FO-1802
Lessons Learned
September 6,
2018
QA Audit 18-05
Radiological Controls
August 23, 2018
Procedures
Number
Title
Revision
9.ALARA.4
Radiation Work Permits
ALARA Program
9.EN-RP-110-05
ALARA Planning and Controls
Radiological Control
9.RADOP.1
Radiation Protection at CNS
Radiation Work
Permits
Number
Title
Revision
2018-537
Misc - Maint, Electrical, I/C support in SWP areas
2018-547
Refuel Floor Misc. - SWP
71151 - Performance Indicator Verification
Condition Reports (CR-CNS-)
2017-03218
2017-03279
2017-04132
2017-04187
Procedures
Number
Title
Revision
0-EN-LI-114
Performance Indicator Process
5C2
Miscellaneous
Documents
Number
Title
Revision/Date
2017 Annual Radioactive Effluent Release Report
April 26, 2018
CNS Station Logs, October 1, 2017 - September 30,
2017
MSPI Derivation Reports, October 1, 2017 -
September 30, 2018
LO-2018-0155
NRC PI Occupational Exposure Control Effectiveness
July 27, 2018
Regulatory Assessment Performance Indicator
Guideline
71152 - Problem Identification and Resolution
Condition Reports (CR-CNS-)
2017-01168
2017-04980
2017-07489
2017-07490
2018-06694
2018-07068
2018-07069
2018-07072
2018-08166
Work Orders
29311
29312
Miscellaneous
Documents
Number
Title
Date
MB-2007-01
MPR Maintenance Bulletin: Potential for Solder Joint
Cracks on Basler SBSR AVR Cards
December 15,
2017
71153 - Follow-up of Events and Notices of Enforcement Discretion
Condition Reports (CR-CNS-)
2016-02281
2017-00553
2017-00558
2017-03723
2018-06797
2018-07108
2018-07934
2018-07935
2018-08621
2018-08632
2018-08638
2018-08639
2018-08644
2018-08647
2018-08657
2019-00026
Work Orders
4783927
5179313
Procedures
Number
Title
Revision
0.50.5
Outage Shutdown Safety
0-BARRIER-
MAPS
Barrier Maps
0-PROTECT-EQP Protected Equipment Program
2.1.20.2
Cycle Specific Fuel Transfer and Alternate Cooling
Guideline
Procedures
Number
Title
Revision
2.2.32
Fuel Pool Cooling and Demineralizer System
100
2.2.69.2
RHR System Shutdown Operations
103
2.4FPC
Fuel Pool Cooling Abnormal
2.4RR
Reactor Recirculation Abnormal
2.4SDC
Shutdown Cooling Abnormal
5.7.1
Emergency Classification
5.7.1, Att. 4
Cooper Nuclear Station Emergency Action Level Matrix
7.2.78.2
Pipe Penetration Seal Installation Using Gasket
Placement
7.3.30.2
RRMG Set Voltage Regulator Tuning
Miscellaneous
Documents
Number
Title
Date
AV-248
Avanti AV-248 Flexseal Material Safety Data Sheet
May 5, 2010
Drawings
Number
Title
Revision
28, Sheet 1
Structural Augmented Radwaste Building Modifications
to Existing Radwaste Building
44D209845
Voltage Regulator
N04
44D209846
Static Exciter & Voltage Regulator
N03
PAPERWORK REDUCTION ACT STATEMENT
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, 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, and Requests for
withhold.
Information Request
August 16, 2018
Notification of Inspection and Request for Information
Cooper Nuclear Station
NRC Inspection Report 05000298/2018004
INSERVICE INSPECTION DOCUMENT REQUEST
Inspection Dates: October 1st through October 5th, 2018
Inspector: Isaac Anchondo
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 Isaac Anchondo, by September 19, 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
(Section B of this enclosure). We ask that the specific items selected from the lists be
available and ready for review on the first day of inspection. 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 October 1, 2018, a reactor inspector from the Nuclear Regulatory Commissions (NRC)
Region IV office will perform the baseline inservice inspection at Columbia Generating
Station, 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. These documents have been divided into two groups. The first group
(Section A of the enclosure) identified information to be provided prior to the inspection to
ensure that the inspector is adequately prepared. The second group (Section B of the
enclosure) identifies the information the inspector will need upon arrival at the site. 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.
We have discussed the schedule for these inspection activities with your staff and
understand that our regulatory contact for this inspection will be Thomas Forland of your
licensing organization. The tentative inspection schedule is as follows:
Preparation week: September 24, 2018
Onsite week: October 1, 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 Isaac Anchondo at (817) 200-1152. (email to: isaac.anchondo@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. Welding activities that are scheduled to be completed during the upcoming
outage (ASME Class 1, 2, or 3 structures, systems, or components)
2. Copies 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. Also provide reference
numbers for applicable procedures that will be used to conduct these examinations.
6. Copies 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. Copies of the most recent self-assessments for the inservice inspection, welding,
and Alloy 600 programs.
9. Copies of the procedures for welding techniques and NDE that will be used during
the outage.
A.2
Additional Information Related to all Inservice Inspection Activities
1. A list with a brief description of inservice inspection 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, or errors in piping examinations.
2. Provide training (e.g. Scaffolding, Fall Protection, FME, Confined Space) if they are
required for the activities described in A.1.
3. Provide copies of the applicable editions of the ASME Code (Sections V, VIII, IX,
and XI) for the inservice inspection program and the repair/replacement program.
4. 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
B.
Information to be Provided Onsite to the Inspector(s) at the Entrance Meeting
(October 1, 2018):
B.1
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 from the lists provided from
section A of this enclosure, provide copies of the following documentation for each
subject weld:
Weld data sheet (traveler).
Weld configuration and system location.
Applicable Code Edition and Addenda for weldment.
- Applicable Code Edition and Addenda for welding procedures.
Applicable welding procedures used to fabricate the welds.
Copies of procedure qualification records (PQRs) supporting the weld
procedures from B.1.b.v.
Copies of welders performance qualification records (WPQ).
Copies of the 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).
Copies of the preservice examination records for the selected welds.
Readily accessible copies of nondestructive examination personnel
qualifications records for reviewing.
3. For the inservice inspection related corrective action issues selected by the inspector
from Section A of this enclosure, provide copies of the corrective actions and
supporting documentation.
4. For the nondestructive examination reports with relevant conditions on ASME Code
Class components selected by the inspector from Section A above, provide copies of
the examination records, examiner qualification records, and associated corrective
action documents.
5. Copy of (or ready access to) most current revision of the inservice inspection program
manual and plan for the current interval.
6. For the nondestructive examinations selected by the inspector from section A of this
enclosure, provide copies of the nondestructive examination procedures used to
perform the examinations (including calibration and flaw characterization/sizing
procedures). 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.
B.2
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. Copies 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.).
The following items are requested for the
Occupational Radiation Safety Inspection
at Cooper
Dates of Inspection: 10/15/2018 to 10/19/2018
Integrated Report 2018004
Inspection areas are listed in the attachments below.
Please provide the requested information on or before Monday, September 24, 2018.
Please submit this information using the same lettering system as below. For example, all
contacts and phone numbers for Inspection Procedure 71124.01 should be in a file/folder titled
1-A, applicable organization charts in file/folder 1-B, etc.
If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at
least 30 days later than the onsite inspection dates, so the inspectors will have access to the
information while writing the report.
In addition to the corrective action document lists provided for each inspection procedure listed
below, please provide updated lists of corrective action documents at the entrance meeting.
The dates for these lists should range from the end dates of the original lists to the day of the
entrance meeting.
If more than one inspection procedure is to be conducted and the information requests appear
to be redundant, there is no need to provide duplicate copies. Enter a note explaining in which
file the information can be found.
If you have any questions or comments, please contact Bernadette Baca at 817-200-1235 or via
e-mail at Bernadette.Baca@nrc.gov.
PAPERWORK REDUCTION ACT STATEMENT
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 3150-0011.
1.
Radiological Hazard Assessment and Exposure Controls (71124.01) and
Performance Indicator Verification (71151)
Date of Last Inspection:
May 22, 2017
A.
List of contacts and telephone numbers for the Radiation Protection Organization Staff
and Technicians, as well as the Licensing/Regulatory Affairs staff. Please include area
code and prefix. If work cell numbers are appropriate, then please include them as well.
B.
Applicable organization charts including position or job titles. Please include as
appropriate for your site, Site Management, RP, Chemistry, Maintenance (I&C),
Engineering, and Emergency Protection. (Recent pictures are appreciated.)
C.
Copies of audits, self-assessments, LARs, and LERs written since the last inspection
date, related to this inspection area
D.
Procedure indexes for the radiation protection procedures and other related disciplines.
E.
Please provide procedures related to the following areas noted below. Additional
procedures may be requested by number after the inspector reviews the procedure
indexes.
1. Radiation Protection Program
2. Radiation Protection Conduct of Operations, if not included in #1.
3. Personnel Dosimetry
4. Posting of Radiological Areas
5. High Radiation Area Controls
6. RCA Access Controls and Radiation Worker Instructions
7. Conduct of Radiological Surveys
8. Radioactive Source Inventory and Control
9. Fuel Pool Inventory Access and Control
F.
Please provide a list of NRC Regulatory Guides and NUREGs that you are currently
committed to relative to this program. Please include the revision and/or date for the
commitment and where this may be located in your current licensing basis documents.
G.
Please provide a summary list of corrective action documents (including corporate and
sub-tiered systems) since the last inspection date.
1. Initiated by the radiation protection organization
2. Assigned to the radiation protection organization
NOTE: These lists should include a description of the condition that provides sufficient
detail that the inspectors can ascertain the regulatory impact, the significance
level assigned to the condition, the status of the action (e.g., open, working,
closed, etc.) and the search criteria used. Please provide in document formats
which are sortable and searchable so that inspectors can quickly and
efficiently determine appropriate sampling and perform word searches, as
needed. (Excel spreadsheets are the preferred format.) If codes are used,
please provide a legend for each column where a code is used.
H.
List of radiologically significant work activities scheduled to be conducted during the
inspection period. (If the inspection is scheduled during an outage, please also include a
list of work activities greater than 1 rem, scheduled during the outage with the dose
estimate for the work activity.) Please include the radiological risk assigned to each
activity.
I.
Provide a summary of any changes to plant operation that have resulted or could result
in a significant new radiological hazard. For each change, please provide the
assessment conducted on the potential impact and any monitoring done to evaluate it.
J.
List of active radiation work permits and those specifically planned for the on-site
inspection week.
K.
Please provide a list of air samples taken to verify engineering controls and a separate
list for breathing air samples in airborne radiation areas or high contamination work
areas. Please include the RWP the breathing air sampling supports.
L.
Please provide the current radioactive source inventory, listing all radioactive sources
that are required to be leak tested. Indicate which sources are deemed 10 CFR Part 20,
Appendix E, Category 1 or Category 2. Please indicate the radioisotope, initial and
current activity (w/assay date), and storage location for each applicable source.
M.
The last two leak test results for all required/applicable radioactive sources that have
failed its leak test within the last two years. Provide any applicable condition reports.
N.
A list of all non-fuel items stored in the spent fuel pools, and if available, their appropriate
dose rates (Contact / @ 30cm)
O.
A list of radiological controlled area entries greater than 100 millirem, since the last
inspection date. The list should include the date of entry, some form of worker
identification, the radiation work permit used by the worker, dose accrued by the worker,
and the electronic dosimeter dose alarm set-point used during the entry (for
Occupational Radiation Safety Performance Indicator verification in accordance with IP 71151).
P.
A list describing VHRAs and TS HRAs (> 1 rem/hour) that are current and historical.
Include their current status, locations, and control measures.
Q.
Temporary effluent monitor locations and calibrations (AMS-4) used to monitor normally
closed doors or off-normal release points (e.g., equipment hatch or turbine heater bay
doors). Include any CRs associated with this monitoring or instrumentation.
2.
Occupational ALARA Planning and Controls (71124.02)
Date of Last Inspection:
February 12, 2018
A.
List of contacts and telephone numbers for ALARA program personnel, as well as the
Licensing/Regulatory Affairs staff. Please include area code and prefix. If work cell
numbers are appropriate, then please include them as well.
B.
Applicable organization charts including position or job titles. Please include as
appropriate for your site, Site Management, RP, Chemistry, Maintenance (I&C),
Engineering, and Emergency Protection. (Recent pictures are appreciated.)
C.
Copies of audits, self-assessments, LARs, and LERs, written since the date of last
inspection, focusing on ALARA
D.
Procedure index for ALARA Program procedures and other related disciplines.
E.
Please provide specific procedures related to the following areas noted below.
Additional Specific Procedures may be requested by number after the inspector reviews
the procedure indexes.
1. ALARA Program
2. ALARA Planning
3. ALARA Reviews
4. ALARA Committee
5. Radiation Work Permit Preparation
F.
Please provide a list of NRC Regulatory Guides and NUREGs that you are currently
committed to relative to this program. Please include the revision and/or date for the
commitment and where this may be located in your current licensing basis documents.
G.
Please provide a summary list of corrective action documents (including corporate and
sub-tiered systems) written since the date of last inspection, related to the ALARA
program, including exceeding RWP Dose Estimates.
NOTE: These lists should include a description of the condition that provides sufficient
detail that the inspectors can ascertain the regulatory impact, the significance
level assigned to the condition, the status of the action (e.g., open, working,
closed, etc.) and the search criteria used. Please provide in document formats
which are sortable and searchable so that inspectors can quickly and
efficiently determine appropriate sampling and perform word searches, as
needed. (Excel spreadsheets are the preferred format.) If codes are used,
please provide a legend for each column where a code is used.
H.
List of work activities (RWPs) greater than 1 rem, since date of last inspection,
including the original dose estimates and actual doses accrued. (Excel format
preferred). Please provide all revisions/changes, as well as any related RWPs that
support the work activity.
I.
List of active work activities (RWPs) that will be in use while we are onsite, including the
dose and dose rate settings, and if available, the planned dose. Include planning
documents and surveys. Include radiological risk assessments and proposed control
measures.
J.
Site dose totals for the past 3 years (based on dose of record). Also provide the current
year-to-date (YTD) collective radiation exposure (CRE). In addition, please provide
another document that separates the online and outage doses for the past 3 years.
K.
Most recent assessment of your isotopic mix, including the hard-to-detect radionuclides
and alpha hazards. Include a list of new and historical exposure issues (radiological
source term or high exposure areas/activities).
L.
If available, provide a copy of the lessons learned from the most recently completed
outage for each unit. Include a summary list of any associated corrective action
documents and the current status of any corrective actions assigned.
M.
Please provide the methods/reports that are in your process to meet the requirements of
CFR 20.1101(c) for periodic review of your RP program.
N.
Current exposure trends (BRAC dose rates and/or source term information).
SUNSI Review
ADAMS:
Non-Publicly Available
Non-Sensitive
Keyword:
By: JKozal
Yes No
Publicly Available
Sensitive
OFFICE
SRI:DRP/C
RI:DRP/C
BC:DRS/EB1
ABC:DRS/EB2
BC:DRS/OB
BC:DRS/PSB2
NAME
PVossmar
MStafford
VGaddy
JDrake
GWerner
HGepford
SIGNATURE
PJV
MHS
vgg
JD
GEW
hjg
DATE
1/23/2019
1/23/19
1/16/19
1/16/2019
01/21/2019
01/17/2019
OFFICE
ATL:DRS/IPAT
SPE:DRP/C
BC:DRP/C
NAME
RKellar
DProulx
JKozal
SIGNATURE
RLK
DLP
JWK
DATE
1/16/2019
1/16/2019
1/28/2019