IR 05000298/2018004

NRC Integrated Inspection Report 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

ary 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. 2 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: DPR-46 Report Number: 05000298/2018004 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 Enclosure

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 Inspection Aspect Procedure Mitigating Green P.2 - 71111.15 -

Systems NCV 05000298/2018004-01 Problem Operability Closed Identification Determinations and and Resolution - Functionality Evaluation 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 Inspection Aspect Procedure Initiating Green H.1 - Human 71153 -

Events NCV 05000298/2018004-02 Performance Follow-up of Closed - Resources 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 Status Procedure LER 05000298/2016-001-01 De-Energized High Pressure 71153 - Closed 05000298/2016-001-02 Coolant Injection Auxiliary Follow-up of Lube Oil Pump Caused by Events and Relay Failure Results in Loss Notices of of Safety Function, a Enforcement Condition Prohibited by Discretion Technical Specifications, and a 10 CFR Part 21 Report LER 05000298/2017-001-00 Residual Heat Removal 71153 - Closed 05000298/2017-001-01 Minimum Flow Valves Out of Follow-up of Position Results in Loss of Events and Safety Function and Notices of Condition Prohibited by Enforcement Technical Specifications Discretion LER 05000298/2017-003-01 Mispositioned Control Room 71153 - Closed Emergency Filter System Follow-up of Supply Fan Damper Causes Events and Loss of Safety Function Notices of Enforcement Discretion URI 05000298/2016008-01 Possible Failure to Ensure 71111.05XT - Closed that the Assumptions in the Fire Protection Engineering Analysis Remain - NFPA 805 Valid (Triennial)

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:

(1) Gas Tungsten Arc Welding (GTAW))

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 Possible Failure to Ensure that the Assumptions in the 71111.05XT -

(Closed) Engineering Analysis Remain Valid Fire URI 05000298/2016008-01 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 Inspection Aspect Procedure Mitigating Green P.2 -

==71111.15 - Systems NCV 05000298/2018004-01 Problem Operability

==

Closed Identification Determinations and and Resolution - Functionality Evaluation 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 Inspection Aspect Procedure Initiating Green H.1 - Human 71153 -

Events NCV 05000298/2018004-02 Performance - Follow-up of Closed Resources 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 99

2.2.32A Fuel Pool Cooling and Demineralizer System 20

Component Checklist

2.2.32B Fuel Pool Cooling and Demineralizer System Instrument 4

Valve Checklist

2.2.69.2 RHR System Shutdown Operations 101

2.2A.CS.DIV2 Core Spray Component Checklist 3

2.2B.CS.DIV2 Core Spray System Instrument Valve Checklist (Div 2) 0

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 40

0.23 CNS Fire Protection Plan 79

0-BARRIER- Barrier Maps 9

MAPS

0-BARRIER-MISC Miscellaneous Buildings 5

2.1.6 Primary Containment Access Preparation and Closeout 15

Activities

3.6.1 Fire Barrier Control 21

Drawings

Number Title Revision

2016, Sheet 7 Fire Protection System Site Plan 12

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 3

Element Review EPM Report R1906-002-001

NEDC 11-104 Fire Safety Analysis for Fire Area AB/05 4

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 5234301

239533

Procedures

Number Title Revision

7.2.57 ASME Category F-A Component Supports Examination 20

and Adjustments

7.7.3.1 General Welding Standard for ASME and ANSI Code 20

Applications

7.7.50.1 Visual Inspection Procedure for ANSI B31.1 2

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 August 27, 2018

Inspection Program and 3rd 10-Year Interval

Containment Inservice Inspection Program

EDP-50 Processing ASME Class 2 and 3 Pressure Boundary 0

Integrity Challenges

Miscellaneous

Documents

Number Title Revision/Date

HEH-VT-103 Procedure for VT-3 Examination 11

NEDC 90-222 Calc. No. 150-88-S-RR-005 for Pipe Support Mark May 2, 1990

No. RR-H7A

PT.CNS.1224 Liquid Penetrant Examination 0

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 99

2.1.1 Startup Procedure 194

2.1.10 Station Power Changes 116

2.2.75 Steam Sealing System 40

6.MISC.502 ASME Class 1 System Leakage Test 55

Miscellaneous

Documents

Number Title Revision

TS LCO 3.0.4b Risk Evaluation - Mode 2 and 1 Entry November 13,

with Drywell Atmospheric Monitoring System Inoperable 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 5273392 5273393

Procedures

Number Title Revision

2.2.40 Drywell Cooling System 31

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 4

System

Miscellaneous

Documents

Number Title Revision/Date

Maintenance Rule Function FPC-F01 Performance 1

Criteria Basis

Maintenance Rule Function HV-F09 Performance 3

Criteria Basis

Maintenance Rule Function REC-F01 Performance 5

Criteria Basis

Maintenance Rule Functional Failure Evaluations for 2004 through

HV-F09 2018

Nutherm Part 21 Evaluation - Allen Bradley Model August 30, 2017

700DC Control Relay

2A1271AC Drywell Cooling System 0

CNS-HV-39 Reactor Building Drywell Cooling Developed Flow 1

Diagram w/ Measurement & Damper Locations

NEDC 89-1439 Drywell Air Cooling System Upgrade Air Flow Derivation 3

NUMARC 93-01 Industry Guideline for Monitoring The Effectiveness of 4F

Maintenance at Nuclear Power Plants

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 5273261

Procedures

Number Title Revision

0.50.5 Outage Shutdown Safety 38

0-PROTECT-EQP Protected Equipment Program 42

2.2.69.2 RHR System Shutdown Operations 101

6.RR.601B RRMG B Positioner Maintenance and Setting Electrical 13

and Mechanical Stops

7.4Disassembly Reactor Vessel Disassembly 18

Drawings

Number Title Revision

045107-01 Interconnect Wiring SM-5360/AD-9120 Relay Panel N04

730E197BB, Variable Speed Recirc Pump & MG Set Recirculation AM/31

Sheet 4 System

730E197BB, Variable Speed Recirc Pump & MG Set Recirculation N14

Sheet 6 System

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 September 11,

Protected Equip with ESST Unavailable 2018

Protected Equipment Tracking Form - RE30 Initial October 2, 2018

Protected Equip with ESST Unavailable

11564325 Notification

EE 01-088 Assessment of the Possibility of Fuel Damage From a 1

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 0

Analysis

NEDC 94-033 Evaluation/Structural Analysis of Cooper RPV Head, 1

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 5270466 5273392 5273393

Procedures

Number Title Revision

0.50.5 Outage Shutdown Safety 38

6.LOG.602 Daily Surveillance Log - Modes 4 and 5 69

6.1CSCS.301 CSCS Water Level Channel Calibration (Div 1) 13

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 August 30, 2017

700DC Control Relay

Safety Evaluation by the Office of Nuclear Reactor August 1, 2018

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

11556454 Notification

EC 6039901 Stop Drill for Steam Dryer Skirt Crack Mitigation Near 0

180 Degrees

71111.18 - Post Maintenance Testing

Procedures

Number Title Revision

6.RCIC.102 RCIC IST and 92 Day Test 36

6.RCIC.105 RCIC Turbine Overspeed Functional Test 24

6.RCIC.502 ASME Section XI System Leakage Test of the Class 2 7

Reactor Core Isolation Cooling (RCIC) System in Steam

Tunnel

Special Procedure RCIC Turbine Control Upgrade Test 2

SP18-001

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 5271305

Procedures

Number Title Revision

6.CSCS.302 CS, RR, and RHR Valve Low Pressure Permissive 11

Interlock Functional Test

6.HPCI.313 HPCI (<165 PSIG) Beginning of Cycle Test 36

6.MISC.502 ASME Class 1 System Leakage Test 53, 54

6.SC.201 Secondary Containment (Reactor Building H&V) Valve 39

Operability Test

6.SW.102 Service Water System Post-LOCA Flow Verification 51

7.0.5 CNS Post-Maintenance Testing 57

7.3.1.16 GE CFD (87) Relay Testing and Maintenance 6

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 AB/34

System

791E264, Sheet 3 RCIC Control Power 7

71111.20 - Refueling and Other Outage Activities

Condition Reports (CR-CNS-)

2018-05791

Procedures

Number Title Revision

0.50.5 Outage Shutdown Safety 38

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 76

6.PC.513 Main Steam Local Leak Rate Tests 28

6.1DG.302 Undervoltage Logic Functional, Load Shedding, and 93

Sequential Loading Test (DIV 1)

6.2DG.302 Undervoltage Logic Functional, Load Shedding, and 84

Sequential Loading Test (DIV 2)

10.9 Control Rod Scram Time Evaluation 69

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 90

71114.06 - Drill Evaluation

Condition Reports (CR-CNS-)

2018-08527 2018-08528 2018-08530

Procedures

Number Title Revision

EPIP 5.7.6 Notification 74

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

Air Sample Surveys

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 SE CRD Valve Room 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 RCIC Quad Sump 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, 18

Storage Pool, or Spent Fuel Storage Pool

9.EN-RP-100 Radiation Worker Expectations 14

9.EN-RP-101 Access Control for Radiologically Controlled Areas 20

9.EN-RP-104 Personnel Contamination 16

9.EN-RP-108 Radiation Protection Posting and Labeling 15

9.EN-RP-113 Response to Contaminated Spills/Leaks 4

9.EN-RP-121 Radioactive Material Control 4

9.EN-RP-123 Radiological Controls for Highly Radioactive Objects 4

9.ENN-RP-102 Radiological Control 3

Procedures

Number Title Revision

9.ENN-RP-106 Radiological Survey Documentation 12

9.ENN-RP-106-1 Radiation and Contamination Surveys 22

9.RADOP.1 Radiation Protection at CNS 14

9.RADOP.2 Radiation Safety Standards and Limits 18

9.RADOP.5 Airborne Radioactivity Sampling 31

9.RADOP.10 Radioactive Sources Control and Accountability 23

9.RADOP.21 Radiological Control of Systems with Potential for 1

Changing Radiological Conditions

Miscellaneous Documents

Title Date

Personnel Contamination Monitor (PCM) Alarm Report: October 4 - 16, October 16, 2018

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, 0

RPIS I&C, Cable Repairs, LPRMs, Decon, IRM/SRM

(High Risk)

2018-537 Misc - Maint., Electrical, I/C Support in SWP Areas 0

(NOT for TG Bldg. Work)

2018-547 Refuel Floor Misc. - SWP 1

2018-549 Cavity and DS Pit Decon 0

2018-556 RWCU-AOV-13A/B and RWCU-AOV-14A/B 0

Replacements in RWCU Valve Room

Radiation Surveys

CNS-1810-0177 CNS-1810-0178 CNS-1810-0182 CNS-1810-0191 CNS-1810-0194

CNS-1810-0196 CNS-1810-0201

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 23

9.EN-RP-110 ALARA Program 10

9.EN-RP-110-05 ALARA Planning and Controls 5

9.ENN-RP-102 Radiological Control 3

9.RADOP.1 Radiation Protection at CNS 14

Radiation Work

Permits

Number Title Revision

2018-537 Misc - Maint, Electrical, I/C support in SWP areas 0

2018-547 Refuel Floor Misc. - SWP 1

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

NEI 99-02 Regulatory Assessment Performance Indicator 7

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 5229312

Miscellaneous

Documents

Number Title Date

MB-2007-01 MPR Maintenance Bulletin: Potential for Solder Joint December 15,

Cracks on Basler SBSR AVR Cards 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 39

0-BARRIER- Barrier Maps 9

MAPS

0-PROTECT-EQP Protected Equipment Program 44

2.1.20.2 Cycle Specific Fuel Transfer and Alternate Cooling 21

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 36

2.4RR Reactor Recirculation Abnormal 43

2.4SDC Shutdown Cooling Abnormal 16

5.7.1 Emergency Classification 61

5.7.1, Att. 4 Cooper Nuclear Station Emergency Action Level Matrix 17

7.2.78.2 Pipe Penetration Seal Installation Using Gasket 1

Placement

7.3.30.2 RRMG Set Voltage Regulator Tuning 8

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 4

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).

ML19028A445

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By: JKozal Yes No Publicly Available Sensitive NRC-002

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