IR 05000220/2024001

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Integrated Inspection Report 05000220/2024001 and 05000410/2024001
ML24131A005
Person / Time
Site: Nine Mile Point  Constellation icon.png
Issue date: 05/10/2024
From: Erin Carfang
NRC/RGN-I/DORS
To: Rhoades D
Constellation Energy Generation, Constellation Nuclear
References
IR 2024001
Download: ML24131A005 (1)


Text

May 10, 2024

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000220/2024001 AND 05000410/2024001

Dear David Rhoades:

On March 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Nine Mile Point Nuclear Station, Units 1 and 2. On May 2, 2024, the NRC inspectors discussed the results of this inspection with Peter Orphanos, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Resident Inspector at Nine Mile Point 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 I; and the NRC Resident Inspector at Nine Mile Point Nuclear Station. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reactor Safety

Docket Nos. 05000220 and 05000410 License Nos. DPR-63 and NPF-69

Enclosure:

As stated

Inspection Report

Docket Numbers:

05000220 and 05000410

License Numbers:

DPR-63 and NPF-69

Report Numbers:

05000220/2024001 and 05000410/2024001

Enterprise Identifier: I-2024-001-0051

Licensee:

Constellation Energy Generation, LLC

Facility:

Nine Mile Point Nuclear Station, Units 1 and 2

Location:

Oswego, NY

Inspection Dates:

January 1, 2024 to March 31, 2024

Inspectors:

C. Kline, Senior Resident Inspector

C. Borman, Health Physicist

E. Eve, Senior Project Engineer

N. Floyd, Senior Reactor Inspector

M. Hardgrove, Senior Project Engineer

J. Kulp, Senior Reactor Inspector

E. Miller, Senior Resident Inspector

C. Swisher, Resident Inspector

S. Veunephachan, Health Physicist

Approved By:

Erin E. Carfang, Chief

Projects Branch 1

Division of Operating Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Nine Mile Point Nuclear Station, Units 1 and 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. A licensee-identified NCV is documented in report section 7115

List of Findings and Violations

Failure to Promptly Correct a Condition Adverse to Quality Rendered Unit 1 Average Power Range Monitors (APRMs) Inoperable Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000220,05000410/2024001-01 Open/Closed

[H.14] -

Conservative Bias 71152A A self-revealed finding of very low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50 Appendix B, Criterion XVI, Corrective Action, was identified when Constellation failed to establish measures to ensure a condition adverse to quality was promptly corrected. Specifically, a previously identified rise in reactor recirculation motor generator 12 vibrations resulted in the tachometer generator uncoupling, a subsequent trip of the reactor recirculation pump 12, and inoperable APRMs.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000220/2023-001-00 Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria 71153 Closed LER 05000220/2023-001-01 Supplement to Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria 71153 Closed LER 05000220/2023-002-00 Average Power Range Monitors (APRMs) Declared Inoperable Due to Trip of Reactor Recirculation Pump 71153 Closed

PLANT STATUS

Unit 1 began the inspection period at rated thermal power and remained at or near rated thermal power throughout the inspection period.

Unit 2 began the inspection period at rated thermal power. On January 13, 2024, operators reduced power to 72 percent for control rod pattern adjustment. The unit returned to rated thermal power on the same day. On February 3, 2024, operators reduced power to 62 percent for control rod pattern adjustment and feed pump swap. The unit returned to rated thermal power on February 4, 2024. On February 22, 2024, the unit commenced coastdown for refueling outage N2R19. On March 3, 2024, the unit experienced an automatic reactor scram due to turbine trip from 55 percent power. The main turbine trip and automatic reactor scram was caused by low main condenser vacuum due to inadequate gland seal steam supply. The refueling outage commenced following the reactor scram. Startup was commenced on March 25, 2024. Unit 2 reached rated thermal power on March 29, 2024, and remained at rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

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

REACTOR SAFETY

71111.01 - Adverse Weather Protection

Impending Severe Weather (IP Section 03.02) (2 Samples)

(1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather for a high wind advisory on January 8, 2024.
(2) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather for high winds on February 16, 2024.

71111.04 - Equipment Alignment

Partial Walkdown (IP Section 03.01) (5 Samples)

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

(1) Unit 1 liquid poison pump 12 on February 26, 2024
(2) Unit 2 decay heat removal system during refueling outage N2R19 shutdown conditions on March 12, 2024
(3) Unit 2 required electrical power source during shutdown operations on March 13, 2024
(4) Unit 2 division I emergency diesel generator following restoration from loss of offsite power and loss of coolant accident testing on March 20, 2024
(5) Unit 2 instrument air system following installation of an engineering change on March 21, 2024

71111.05 - Fire Protection

Fire Area Walkdown and Inspection (IP Section 03.01) (5 Samples)

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

(1) Unit 2 control building 261', remote shutdown room and division II switchgear, fire area 19, on February 20, 2024
(2) Unit 2 reactor building 289', fire area 5, on February 29, 2024
(3) Unit 2 turbine building heater bays, fire area 50, on March 6, 2024
(4) Unit 2 turbine building condenser bays, fire area 50, on March 7, 2024
(5) Unit 2 control building 288', relay room, fire area 24, on March 14, 2024

71111.08G - Inservice Inspection Activities (BWR)

BWR Inservice Inspection Activities - Nondestructive Examination (NDE) and Welding Activities

(IP Section 03.01)

The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities for the Unit 2 refueling outage N2R19 from March 4 to March 15, 2024.

(1)

2R19 -ISI-VE-010). This examination was performed in accordance with BWRVIP-

75-A, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules.

2R19 -ISI-VE-001). This examination was performed in accordance with BWRVIP-

75-A, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules.

  • Visual examination of the N14A, N14B, N14C, and N14D reactor vessel instrumentation nozzles (NDE Reports 2R19-ISI-VT-002 thru -005). This examination was performed as an augmented enhanced VT-2 in accordance with Nuclear Event Report NC-18-005-Y.
  • Visual examinations of the in-vessel visual inspection of reactor components focused on the jet pump components (work order C93885819).

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)

(1) The inspectors observed and evaluated licensed operator performance in the Unit 1 control room during the 121 core spray surveillance test on February 28, 2024.
(2) The inspectors observed and evaluated licensed operator performance in the Unit 2 control room during reactor plant cooldown on March 4, 2024.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)

(1) The inspectors observed a Unit 2 simulator evaluation that included a design basis earthquake, loss of recirculating system flow control valve control (lowering), loss of condenser vacuum, scram, and an isolable reactor core isolation cooling leak, on January 30, 2024.
(2) The inspectors observed a Unit 1 simulator evaluation that included a feed water pump flow control valve failure, a failure of instrument and control bus 130, and a high power anticipated transient without scram with main steam line break in the steam tunnel, on February 8, 2024.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (2 Samples)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended functions:

(1) Unit 2 control room air conditioning unit 1B following surveillance failure on January 31, 2024
(2) Unit 2 drywell (primary containment) on March 18, 2024

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management (IP Section 03.01) (8 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1) Unit 2 elevated risk due to division II residual heat removal unit cooler maintenance on January 30, 2024
(2) Unit 2 elevated risk due to reactor core isolation cooling snubber replacement on February 23, 2024
(3) Unit 2 elevated risk during residual heat removal heat exchanger relief valve replacement on February 29, 2024
(4) Unit 2 elevated risk due to reduced reactor coolant inventory during reactor cavity floodup on March 5, 2024
(5) Unit 2 elevated risk due to alternate decay heat removal lineup during refueling operations on March 11, 2024
(6) Unit 2 elevated risk due to division I switchgear outage maintenance window on March 13, 2024
(7) Unit 2 elevated risk during reactor cavity draindown on March 20, 2024
(8) Unit 2 elevated risk during Ovation software update on March 25, 2024

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (4 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1) Unit 2 control room air conditioning unit 1B on January 31, 2024
(2) Unit 2 off-normal diesel fire pump engine oil resulting in Alert range on February 1, 2024
(3) Unit 2 automatic depressurization system valve operability on February 21, 2024
(4) Unit 2 source range nuclear instrument 'B' on March 17, 2024

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (3 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) Permanent Modification: ECP-23-000375, Unit 2 Add Manual Isolation Valves for Automatic Depressurization System Makeup Solenoid Operated Valves
(2) Permanent Modification: ECP-18-000372, Unit 2 Division I Emergency Diesel Generator Voltage Regulator Replacement
(3) Permanent Modification: ECP-22-000357, Unit 2 Division I Emergency Diesel Generator Slow Start Mod

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated Unit 2 refueling outage N2R19 activities from March 3 to March 29, 2024.

71111.24 - Testing and Maintenance of Equipment Important to Risk

The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:

Post-Maintenance Testing (PMT) (IP Section 03.01) (9 Samples)

(1) N1-ST-Q6B, Containment Spray System Loop 121 Quarterly Test, following system outage window on February 7, 2024
(2) N2-OSP-CSL-R002, Low Pressure Core Spray Valve Operability Test, following snubber replacements on March 11, 2024
(3) N2-OSP-SLS-R003, SLS Valve Operability Test, following squib valve replacement on March 12, 2024
(4) N2-OSP-CSL-R@001, Low Pressure Core Spray Pressure Isolation Valve Leak Test, following snubber replacements on March 11, 2024
(5) N2-OSP-MSS-CS001, Main Steam Isolation Valve Operability Test, following stroke time adjustments on March 18, 2024
(6) N2-OSP-EGS-R004, Operating Cycle Diesel Generator Simulated Loss of Offsite Power With and Without Emergency Core Cooling System Division I/II, following division I emergency diesel generator engineering change installation on March 18, 2024
(7) N1-PM-M9, Operation of Fire Pumps, following starting air system maintenance on March 18, 2024
(8) N2-OSP-IAS-R001, Instrument Air System Valve Position Indication Operability Test, following 2IAS*SOVY(X)186 replacement on March 23, 2024
(9) N2-OSP-ICS-Q@002, Reactor Core Isolation Cooling Pump and Valve Operability Test and System Integrity Test and ASME XI Functional Test and Analysis, following system outage window on March 27, 2024

Surveillance Testing (IP Section 03.01) (5 Samples)

(1) N1-ST-Q8B, Liquid Poison Pump 12 and Check Valve Operability Test on January 29, 2024
(2) N2-OSP-ICS-Q@002, Reactor Core Isolation Cooling Pump and Valve Operability Test and System Integrity Test on February 5, 2024
(3) N2-OSP-MSS-CS001, Main Steam Isolation Valve Operability Test on March 4, 2024
(4) N2-OSP-MSS-003, Main Steam Isolation Valve Leak Rate Test on March 14, 2024
(5) N2-OSP-CSL-R001, Division I Emergency Core Cooling System Functional Test on March 18, 2024

Inservice Testing (IST) (IP Section 03.01) (1 Sample)

(1) N2-ISP-RRC-R001, Alternate Rod Insertion Function of Redundant Reactivity Control System on March 6, 2024

Containment Isolation Valve Testing (IP Section 03.01) (1 Sample)

(1) N2-OSP-CNT-R003, Containment Isolation Valve Isolation Actuation Test on March 22,

RADIATION SAFETY

71124.01 - Radiological Hazard Assessment and Exposure Controls

Radiological Hazard Assessment (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials, and how the licensee assesses radiological hazards.

Instructions to Workers (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.

Contamination and Radioactive Material Control (IP Section 03.03) (2 Samples)

The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:

(1) Licensee surveys of potentially contaminated material leaving the Unit 2 radiologically controlled area
(2) Workers exiting the Unit 2 radiologically controlled area during a refueling outage

Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)

The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:

(1) Unit 2 motor operated valve MOV18A rework
(2) Unit 2 moisture separator/reheater 'A' welding activities
(3) Unit 2 reactor head vessel lift and staging
(4) Unit 2 drywell shielding installation packages

High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (3 Samples)

The inspectors evaluated licensee controls of the following high radiation areas and very high radiation areas (VHRAs):

(1) Unit 2 locked high radiation area (LHRA) drywell entrance during a refueling outage
(2) Unit 2 LHRA radioactive waste filter backwash tank room 2LWS-TK16
(3) Unit 2 LHRA waste collection tank room 2LWS-TK1A/B/C

Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)

(1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours (IP Section 02.01)===

(1) Unit 1 for the period of January 1, 2023 through December 31, 2023
(2) Unit 2 for the period of January 1, 2023 through December 31, 2023

IE03: Unplanned Power Changes per 7000 Critical Hours (IP Section 02.02) (2 Samples)

(1) Unit 1 for the period of January 1, 2023 through December 31, 2023
(2) Unit 2 for the period of January 1, 2023 through December 31, 2023

IE04: Unplanned Scrams with Complications (IP Section 02.03) (2 Samples)

(1) Unit 1 for the period of January 1, 2023 through December 31, 2023
(2) Unit 2 for the period of January 1, 2023 through December 31, 2023

===71152A - Annual Follow-up Problem Identification and Resolution

Annual Follow-up of Selected Issues (Section 03.03)===

The inspectors reviewed the licensees implementation of its corrective action program related to the following issue:

(1) IR 04711520 - Reactor Recirculation Pump 12 Tripped Due to a Failed Coupling on the Tachometer Generator

71153 - Follow-up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)

(1) The inspectors evaluated a Unit 2 main turbine trip and reactor scram due to low main condenser vacuum and the licensees response on March 3, 2024.

Event Report (IP Section 03.02) (2 Samples)

The inspectors evaluated the following licensee event reports (LERs):

(1) LERs 05000220/2023-001-00 and 05000220-2023-001-01, Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria and Supplement to the Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria (ADAMS Accession Nos. ML23164A272 and ML23227A037, respectively). The inspection conclusions associated with these LERs are documented in this report under Inspection Results. These LERs are closed.
(2) LER 05000220/2023-002, Average Power Range Monitors Declared Inoperable Due to Trip of Reactor Recirculation Pump 12 (ADAMS Accession No. ML23363A151).

The inspection conclusions associated with this LER are documented in this report under Inspection Results. This LER is closed.

INSPECTION RESULTS

Failure to Promptly Correct a Condition Adverse to Quality Rendered Unit 1 Average Power Range Monitors (APRMs) Inoperable Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems

Green NCV 05000220,05000410/2024001-01 Open/Closed

[H.14] -

Conservative Bias 71152A A self-revealed finding of very low safety significance (Green) and associated NCV of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action, was identified when Constellation failed to establish measures to ensure a condition adverse to quality was promptly corrected.

Specifically, a previously identified rise in reactor recirculation motor generator 12 vibrations resulted in the tachometer generator uncoupling, a subsequent trip of the reactor recirculation pump 12, and inoperable APRMs.

Description:

Unit 1 at Nine Mile Point operates with five reactor recirculation loops, each with an associated reactor recirculation pump, and each pump has a dedicated motor generator tachometer. On October 21, 2023, while operating at 100 percent rated thermal power, the reactor recirculation pump 12 unexpectedly tripped, causing a reduction in reactor power to approximately 88 percent. Without the pump running in that loop, the other four loops forced flow in that loop to move in reverse. The reverse flow condition resulted in an unconservative total flow input to the APRM logic resulting in the APRM flow biased trip set point being higher than the allowable technical specification value, causing the APRMs to be declared inoperable. The operators took action to close the reactor recirculation pump 12 discharge block valve, which ceased the reverse flow condition and restored the APRM to operable.

Constellation was required to make an 8-hour report (EN-56810) under 10 CFR 50.72(b)(3)(v)(A), because the unconservative total flow input to the APRM logic was, any event or condition that could have prevented the fulfillment of a safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe condition.

Constellation documented the condition in their corrective action program as IR 04711520 and performed a corrective action program evaluation to determine the cause of the pump trip. The cause that Constellation assigned to the pump trip was the tachometer generator became uncoupled from the motor generator set.

Constellation has documented in their corrective action program multiple occasions of elevated vibrations in the motor generator set tachometer generators, including, in October 2020 for the 12 and 14 tachometer generators (IR 04377127 and IR 04377132), in October 2021 for the 12 and 13 tachometer generators (IR 04456825), and in April 2023 for the 15-tachometer generator (IR 04672153). Constellation implemented a modification (ECP-23-000201) to stiffen the baseplate to reduce vibrations. This modification was only installed for the 15-tachometer generator, although internal operating experience existed for elevated vibrations on other tachometer generators.

The inspectors reviewed Constellations corrective action documents, past equipment issues, plant modifications and plant design and determined a more specific cause is apparent.

Specifically, Constellation procedure PI-AA-125, "Corrective Action Program (CAP)

Procedure" defines a condition adverse to quality as "An all-inclusive term used in reference to any of the following: failures, malfunctions, deficiencies, defective items, and non-conformances." The inspectors concluded that elevated vibrations on recirculation pump motor generator 12 directly contributed to the tachometer becoming uncoupled and constituted a deficiency requiring corrective action. The inspectors noted that Constellation implemented a plant modification, ECP-23-000201, to successfully reduce vibrations on a different tachometer generator. The inspectors determined through Constellations internal operating experience, plant modification results, corrective maintenance, and cause analysis that information was available to Constellation to correct the condition adverse to quality before the 12-reactor recirculation tachometer generator tripped and APRMs became inoperable.

Corrective Actions: After the reactor recirculation pump trip, operators in the control room took action to close the pump discharge isolation valve which terminated the reverse flow in the loop and restored the APRMs to operable. Corrective maintenance was performed to recouple the 12-reactor recirculation tachometer generator and return the unit to 5-loop operation. Constellation initiated IR 04711520 to document the deficiency and performed a corrective action program evaluation. As part of the corrective action program evaluation, Constellation is evaluating installing the modification that was performed to the 15-tachometer generator, to the other tachometer generators.

Corrective Action References: IR 04711520

Performance Assessment:

Performance Deficiency: Constellation failed to promptly correct high vibrations on the tachometer generator, a condition adverse to quality, rendering Unit 1 APRMs inoperable.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with 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 trip of the 12-reactor recirculation pump caused an unconservative flow bias to the APRM trip setpoint and created a value that was higher than allowed by technical specifications.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined this finding to be of very low safety significance (Green) in accordance with Exhibit 2, because it did not affect a single reactor protection system (RPS) trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown (e.g., other automatic RPS trips, alternate rod insertion, or manual reactor trip capacity).

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, Constellation performed a modification to reduce vibrations for one of the tachometer generators, as opposed to correcting the condition adverse to quality applicable to all tachometer generator sets.

Enforcement:

Violation: Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, 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 non-conformances are promptly identified and corrected.

Contrary to the above, from October 2020 to October 2023, Constellation failed to ensure measures were established to assure conditions adverse to quality were promptly corrected prior to the Unit 1 APRMs being rendered inoperable on October 21, 2023.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Observation: Reactor Recirculation Pump 12 Tripped Due to a Failed Coupling on the Tachometer Generator 71152A The inspectors independently reviewed Constellations corrective action documents, corrective action program evaluation, and associated actions regarding an issue where the reactor recirculation pump 12 unexpectedly tripped due to a failed coupling on the tachometer generator.

Unit 1 at Nine Mile Point is a boiling water reactor that is designed with five reactor coolant recirculation loops, each with an associated reactor recirculation pump. On October 21, 2023, the reactor recirculation pump 12 unexpectedly tripped. Because one of the five pumps tripped, reactor power quickly lowered to 88 percent, and coolant flow in the loop with the tripped pump began to flow backwards in the loop. The reverse flow condition resulted in an unconservative total flow input to the APRM logic resulting in the APRM flow biased trip set point being higher than the allowable technical specification value causing the APRMs to be declared inoperable. Approximately 10 minutes after the pump trip, the operators shut the discharge block valve in the tripped loop, which ended the reverse flow, and the APRMs were restored to operable. Constellation was required to make an 8-hour report (EN-56810) under 10 CFR 50.72(b)(3)(v)(A), because the unconservative total flow input to the APRM logic was, any event or condition that could have prevented the fulfillment of a safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe condition.

Constellation determined the most likely cause to be a failed coupling associated with the pumps tachometer generator. Prior to the pump tripping, Constellation identified that multiple reactor recirculation pump motor generator sets were experiencing elevated vibrations in October 2020 for the 12 and 14 tachometer generators (IR 04377127 and IR 04377132), in October 2021 for the 12 and 13 tachometer generators (IR 04456825), and in April 2023 for the 15-tachometer generator (IR 04672153). After the documented issue with the 15-tachometer generator, Constellation implemented an engineering change package (ECP-23-00201) through work order C93909469 to stiffen the baseplate for the reactor recirculation pump 15-tachometer generator. The modification was performed for the reactor recirculation pump 15. Constellation has since generated an action from the corrective action program evaluation that will install the modification to the other tachometer generators.

The inspectors noted that Constellation determined the cause of the pump trip was that the tachometer generator became uncoupled. Constellation assigned a corrective action to repair the uncoupled generator. Constellation had initially assigned vibrations as a contributing cause and assigned a corrective action to install a baseplate stiffening modification to reduce vibrations. As noted in the preceding NCV, the inspectors determined that the initial cause identified by Constellation was credible and that vibrations were not fully evaluated. The inspectors found that Constellation procedure PI-AA-120, Issue Identification and Screening Process, Attachment 6, How the Constellation Corrective Action Process Meets the Requirements of 10 CFR 50, Criterion XVI, Appendix B and the Constellation Quality Assurance Topical Report, states in part,..the Constellation CAP [corrective action program] process is structured to ensure that Corrective Action Program Evaluations (CAPEs) that result in a Corrective Action to the correct cause, not just simply fix the issue, also meets the NRC regulation for addressing CAQs and SCAQs. As noted in the preceding NCV, the inspectors determined that Constellation did not address the cause of the uncoupled generator and simply fixed the issue.

The inspectors independently evaluated this issue of the lack of cause identification for significance in accordance with the guidance in IMC 0612, Appendix B, "Issue Screening,"

and Appendix E, "Examples of Minor Issues." The inspectors determined that none of the conditions were deficiencies of greater than minor significance and, therefore, are not subject to enforcement action in accordance with the NRC's Enforcement Policy.

Licensee-Identified Non-Cited Violation 71153 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: Nine Mile Point Unit 1 Technical Specification 6.4.1.A, Procedures, requires the establishment, implementation and maintenance of procedures recommended in Regulatory Guide 1.33, dated November 3, 1972. Regulatory Guide 1.33, Appendix A, paragraph 8.b(2)(t), Inspection of Reactor Coolant System Pressure Boundary, is listed as one of the procedures.

Procedure WDI-STD-119-A, Generic Procedure for Ultrasonic Examination of Dissimilar Metal Nozzle to Safe End Welds and Dissimilar Metal Piping Welds Using the IntraSpect Automated Imaging System, Revision 2, was used in 2007 to examine the weld joining the N2E reactor pressure vessel nozzle to the recirculation system safe end, which is part of the reactor coolant system pressure boundary. Procedure WDI-STD-119A, Table 1, directed examiners to configure the ultrasonic testing (UT) instrument transducer in dual side by side (D-SBS) mode when performing 60-degree circumferential scans of the N2E nozzle to safe end weld.

Contrary to the above, based upon review of examination data and records from 2007, Constellation staff determined that the UT instrument was configured in the single mode, which was not in accordance with the procedure to configure the instrument in the D-SBS mode. As a result, the data collected on March 30, 2007, by the 60-degree circumferential scans of the N2E nozzle to safe end weld were invalid. Constellation staff also determined it was reasonable to have identified the flaw in 2007 as the review of the examination data showed the flaw to be detectable at that time. Constellation staff applied a leakage barrier weld overlay during the spring 2023 1R27 refueling outage.

Significance/Severity: Green. The inspectors determined this finding to be of very low safety significance (Green) in accordance with IMC 0609, Appendix A, Exhibit 1, because after a reasonable assessment of the degradation, the potential leakage from the degraded weld would not have exceeded the reactor coolant system leak rate for a small loss of coolant accident (LOCA), and it was unlikely that the degraded weld would have affected other systems used to mitigate a LOCA. The inspectors reviewed Constellations flaw evaluation and leakage calculation and found that the evaluation included a reasonable assessment of degradation involving fatigue and stress corrosion cracking phenomena and showed a resulting leak rate would have been within the capability of normal make-up sources. The inspectors further considered a postulated leak in this location would not have affected systems used to mitigate a LOCA.

Corrective Action References: IR 04563178

Observation: Indication on the N2E Dissimilar Metal Weld Exceeding ASME Acceptance Criteria 71153 The inspectors reviewed LER 05000220/2023-001-00 (ML23164A272) and its supplement (ML23227A037) submitted to the NRC to report an axial defect found during the spring 2023 N1R27 refueling outage in the recirculation inlet nozzle N2E safe end-to-nozzle dissimilar metal weld. The condition was reported to the NRC because the defect was not acceptable for continued service in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, IWB-3600. The defect was characterized by phased array ultrasonic examination and found to be an axially oriented flaw and approximately 83 percent through-wall. The flaw was repaired with a weld overlay during the refueling outage and the subject of several relief requests approved by the NRC and listed in the reference section this report.

The inspectors reviewed the LERs and Constellations supporting causal evaluation (IR 4563178) to verify the accuracy and completeness of the information provided and the appropriateness of the corrective actions. Constellation staff completed corrective actions to repair the flaw, address the potential extent of the problem and consider how the flaw grew axially to 83 percent through-wall without prior detection.

Regarding immediate corrective actions, the inspectors reviewed the overlay repair completed during the N1R27 refueling outage as documented in inspection report 05000220/2023001 (ADAMS Accession No. ML23125A005). Additionally, the inspectors previously reviewed extent of condition examinations completed of similar welds during the N1R27 refueling outage and found the scope of exams and results met requirements.

The inspectors noted that Constellation staffs causal evaluation (IR 04563178) was of sufficient depth and detail to determine that the flaw grew slowly due to intergranular stress corrosion cracking (IGSCC) and that problems with the scope and conduct of exams in 2007 and 2013 contributed to not identifying this defect at that time at an earlier stage. However, this information was not included in the associated LER and was germane to the NRC staffs role in studying potentially generic safety problems, assessing trends and patterns of operational experience, monitoring performance, and providing for operational experience to stakeholders.

Specifically, the inspectors found that the causal evaluation identified a missed opportunity to identify the flaw due to improper setup of the UT examination system in 2007 (see Results Section of this report). A second missed opportunity to identify the flaw occurred during a manual phased array ultrasonic testing exam in 2013 that documented no recordable indications. The inspectors noted that manual exams do not record data, but only overall results, and therefore there was not the capability for the licensee to review the data to determine specifically why the exam did not identify the flaw.

An additional contributing factor was that between 1988 and 2005, the licensee mischaracterized the welds under their response to GL 88-01, NRC Position On IGSCC In BWR Austenitic Stainless Steel Piping. During safe end replacement in 1982, a thin layer of the original alloy 182 weld material was left on the reactor pressure vessel nozzles to alleviate the need to post weld heat treat the nozzle. Subsequently, the reactor recirculation system nozzle to safe end welds were categorized as Category A (Resistant Materials)instead of Category D (Non-Resistant Materials, No Stress Improvement). This resulted in an improper examination scheduling requirement and loss of three additional opportunities to identify the flaw.

Finally, the inspectors identified that Constellation staff did not provide the basis to support the conclusion made in LER 05000220/2023-001-00 and -01 that, The flaw identified in reactor penetration N2E did not have any impact to the health or safety of the public.

Specifically, Constellation staff did not include an assessment of the event under reasonable and credible alternative conditions such as seismic conditions. In response to inspector questions, Constellation staff provided an evaluation of the consequences of a seismic event on the as found condition that showed structural integrity was maintained.

The inspectors found, on a sampling basis, that Constellation staffs corrective actions to repair the flaw and address the potential scope of the problem met standards.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

  • On March 15, 2024, the inspectors presented the Unit 2 inservice inspection results to Peter Orphanos, Site Vice President, and other members of the licensee staff.
  • On April 5, 2024, the inspectors presented the radiological hazards inspection results to Carl Crawford, Plant Manager, and other members of the licensee staff.
  • On May 2, 2024, the inspectors presented the integrated inspection results to Peter Orphanos, Site Vice President, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.01

Procedures

N1-OP-64

Meteorological Monitoring

21

N2-OP-102

Meteorological Monitoring

28

71111.04

Drawings

PID-19D

Piping & Instrumentation Diagram, Service Air

25

Engineering

Changes

ECP-23-000375

Add Manual Isolation Valves for ADS Makeup SOVs

Procedures

N2-OP-100A

Standby Diesel Generators

035

N2-OP-34-

LINEUPS

Nuclear Boiler Automatic Depressurization Safety Relief

Valves - Lineups

003

71111.05

Fire Plans

N2-FPI-PFP-0201 Unit 2 Pre-Fire Plans

007

71111.08G Corrective Action

Documents

04484543

04756074

Engineering

Evaluations

ECP-22-000144

Qualitative Functionality Assessment for Jet Pump 6 Slip

Joint Clamp Ratchet Keeper

Miscellaneous

ER-NM-330-2001

NMP2 Inservice Inspection Program Plan for the Fourth Ten-

Year Interval

ER-NM-330-2004

NMP2 Risk-Informed Inservice Inspection Program Report

71111.12

Procedures

N2-MSP-CNT-

R005

Primary Containment Structural Integrity Inspection and

Suppression Pool Cleaning

013

S-MRM-REL-

0102

Structural Monitoring Program

010

Work Orders

C93788531

71111.13

Procedures

N2-PM-082

RPV Flood Up/Draindown

23

OP-AA-107

Integrated Risk Management

008

OP-AA-107-F-01

Risk Screening / Mitigation Plan

003

71111.15

Corrective Action

Documents

04701455

04739812

04740620

Procedures

N2-MSP-HVC-

2Y002

Performance Evaluation Test for Air Conditioning Units 2

HVC-ACU1B and 2B

2

Work Orders

C93785657

C93788531

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.18

Engineering

Changes

ECP-18-000372

EDG Voltage Regulator Replacement

ECP-22-000357

Division I EDG Slow Start Mod

ECP-23-000375

Add Manual Isolation Valves for ADS Makeup SOVs

71111.24

Corrective Action

Documents

04758780

Procedures

N1-PM-M9

Operation of Fire Pumps

018

N1-ST-Q6B

Containment Spray Loop 121 Quarterly

21

N1-ST-Q8B

Liquid Poison Pump 12 and Check Valve Operability Test

016

N2-MFT-327B

Digital Feedwater Level Control System Power Ascension

Test

2

N2-OSP-CNT-

R003

Containment Isolation Valve Isolation Actuation Surveillance

017

N2-OSP-CSL-

R001

Division 1 ECCS Functional Test

010

N2-OSP-CSL-

R002

LPCS Valve Operability Test

008

N2-OSP-EGS-

R001

Diesel Generator ECCS Start and Load Reject Division 1/2

009

N2-OSP-EGS-

R004

Operating Cycle Diesel Generator Simulated Loss of Offsite

Power With and Without ECCS Division I/II

20

N2-OSP-IAS-

R001

Instrument Air System Valve Position Indication Operability

Test

009

N2-OSP-ICS-

Q@002

RCIC Pump and Valve Operability Test and System integrity

Test

018

N2-OSP-ICS-

R@002

RCIC Pump and Valve Operability Test and System Integrity

Test and ASME XI Functional Test and Analysis

018

N2-OSP-MSS-

003

Main Steam Isolation Valve Leak Rate Test

005

N2-OSP-MSS-

CS001

Main Steam Isolation Valve Operability Test

2

N2-OSP-RPV-

R@003

Reactor Pressure Vessel and Class I Systems Leakage Test

with the RPV Solid

015

N2-OSP-SLS-

R003

SLS Valve Operability Test

007

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

Work Orders

C93654821

C93842758

C93947869

C93948613

C93948613

C93969241

71152A

Corrective Action

Documents

04711520

04723637

71153

Calculations

2400275.301

Flaw Evaluation and Leakage Calculation of Axial Flaw at

N2E Nozzle

Corrective Action

Documents

04563178

04755109

Engineering

Evaluations

2300376.401

Evaluation of a Leakage Barrier Weld Overlay

03/25/2023

2400275.301

NMP-1 Flaw Evaluation for N2E Nozzle

Miscellaneous

NMP1L3547

Supplement to NMP1 Licensee Event Report

05000220/2023-001, Indication on the N2E Dissimilar Metal

Weld Exceeding ASME Acceptance Criteria (ML23227A037)

08/11/2023

NMP1L3513

Submittal of Emergency Relief Request I5R-11 Concerning

the Installation of a Weld Overlay on Reactor Pressure

Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle

Dissimilar Metal Weld (32-WD-208)

03/24/2023

NMP1L3515

Submittal of Emergency Relief Request I5R-11 Concerning

the Installation of a Weld Overlay on Reactor Pressure

Vessel Recirculation Inlet Nozzle N2E Safe End-to-Nozzle

Dissimilar Metal Weld (32-WD-208)

03/27/2023

NMP1L3531

Licensee Event Report 05000220/2023-001, Revision 0,

Indication on the N2E Dissimilar Metal Weld Exceeding

ASME Acceptance Criteria (ML23164A272)

05/12/2023

Procedures

LS-AA-1400

Event Reporting Guidelines 10 CFR 50.72 and 50.73

7