IR 05000220/2022001: Difference between revisions
StriderTol (talk | contribs) (StriderTol Bot insert) |
StriderTol (talk | contribs) (StriderTol Bot change) |
||
| Line 24: | Line 24: | ||
==Dear Mr. Rhoades:== | ==Dear Mr. Rhoades:== | ||
On March 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Nine Mile Point Nuclear Station, Units 1 and 2. On April 21, 2022, the NRC inspectors discussed the results of this inspection with Mr. Peter Orphanos, Site Vice President, | On March 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Nine Mile Point Nuclear Station, Units 1 and 2. On April 21, 2022, the NRC inspectors discussed the results of this inspection with Mr. Peter Orphanos, Site Vice President, a nd other members of your staff. The results of this inspection are documented in the enclosed report. | ||
No findings or violations of more than minor significance were identified during this inspection. | No findings or violations of more than minor significance were identified during this inspection. | ||
This letter, its enclosure, and your response (if any) will be | This letter, its enclosure, and your response (if any) will be m ade available for public inspection and copying at http://www.nrc.gov/reading-rm /adam s.h tm l and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding. | ||
Sincerely, | Sincerely, Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reac to r Safety | ||
Docket Nos. 05000220 and 05000410 License Nos. DPR-63 and NPF-69 | |||
===Enclosure:=== | ===Enclosure:=== | ||
| Line 39: | Line 38: | ||
==Inspection Report== | ==Inspection Report== | ||
Docket Numbers: 05000220 and 05000410 License Numbers: DPR-63 and NPF-69 Report | Docket Numbers: 05000220 and 05000410 | ||
License Numbers: DPR-63 and NPF -69 | |||
Report Num bers: 05000220/2022001 and 05000410/2022001 | |||
Enterprise Identifier: I-2022- 001-0049 | |||
Licensee: Constellation Energy Generation, LLC | |||
Facility: Nine Mile Point Nuclear Station, Units 1 and 2 | |||
Location: Oswego, NY | |||
Inspection Dates: January 1, 2022 to March 31, 2022 | |||
Inspectors: G. Stock, Senior Resident Inspector C. Kline, Resident Inspector B. Sienel, Resident Inspector N. Floyd, Senior Reactor Inspector S. Haney, Senior Project Engineer C. Hobbs, Reactor Inspector S. Wilson, Senior Health Physicist | |||
Approved By: Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reactor Safety | |||
Enclosure | |||
=SUMMARY= | =SUMMARY= | ||
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the | The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licens ees 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 h ttps://www.nrc.gov/reactors/operating/oversight.html for more information. | ||
===List of Findings and Violations=== | ===List of Findings and Violations=== | ||
No findings or violations of more than minor significance were identified. | |||
Additional Tracking Ite ms | |||
None. | None. | ||
| Line 55: | Line 74: | ||
Unit 1 operated at or near rated thermal power for the entire inspection period. | Unit 1 operated at or near rated thermal power for the entire inspection period. | ||
Unit 2 began the inspection period at rated thermal power | Unit 2 began the inspection period at rated thermal power. | ||
On January 28, 2022, the unit began end-of-cycle coastdown. On February 11, 2022, the unit was downpowered to 85 percent to avoid a known oscillation region during end-of-cycle coastdown. On March 7, 2022, the unit was shut down for a planned refueling outage. Startup was commenced on March 25, 2022, and rated thermal power was reached on March 29, 2022. Later that day, the unit was downpowered to 80 percent for a planned rod pattern adjustment. During the downpower, a main turbine control valve malfunction required an additional downpower to 60 percent. The unit returned to rated thermal power on March 31, 2022, and remained at or near rated thermal power for the remainder of the inspection period. | On January 7, 2022, the unit was downpowered to 78 percent to perform planned control r od channel interference testing and a control rod pattern adjustment, and returned to rated thermal power on January 8, 2022. | ||
On January 28, 2022, the unit began end-of-cycle coastdown. | |||
On February 11, 2022, the unit was downpowered to 85 percent to avoid a known oscillation region during end-of-cycle coastdown. On March 7, 2022, the unit was shut down for a planned refueling outage. | |||
Startup was commenced on March 25, 2022, and rated thermal power was reached on March 29, 2022. | |||
Later that day, the unit was downpowered to 80 percent for a planned rod pattern adjustment. During the downpower, a main turbine control valve malfunction required an additional downpower to 60 percent. | |||
The unit returned to rated thermal power on March 31, 2022, and remained at or near rated thermal power for the remainder of the inspection period. | |||
==INSPECTION SCOPES== | ==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, conducted routine reviews using IP 71152, Problem Identification and Resolution, 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. | 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, conducted routine reviews using IP 71152, Problem Identification and Resolution, 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. | |||
On February 1, 2022, the operating license for Nine Mile Point Nuclear Station, held by Exelon Generation Company, LLC, was transferred to Constellation Energy Generation, LLC (Constellation). While some or all of the inspections documented in this report | On February 1, 2022, the operating license for Nine Mile Point Nuclear Station, held by Exelon Generation Company, LLC, was transferred to Constellation Energy Generation, LLC (Constellation). While some or all of the inspections documented in this report w ere performed while the license was held by Exelon Generation Company, LLC, this report will refer to the licensee as Constellation throughout. | ||
==REACTOR SAFETY== | ==REACTOR SAFETY== | ||
==71111.01 - Adverse Weather Protection== | ==71111.01 - Adverse Weather Protection== | ||
===Impending Severe Weather Sample (IP Section 03.02) (1 Sample)=== | ===Impending Severe Weather Sample (IP Section 03.02) (1 Sample)=== | ||
: (1) The inspectors evaluated the adequacy of the overall preparations to protect risk- | : (1) The inspectors evaluated the adequacy of the overall preparations to protect risk - | ||
significant systems due to a winter storm warning on March 11, 2022. | |||
==71111.04 - Equipm ent Alignment== | |||
===Partial Walkdown Sample (IP Section 03.01) (7 Samples)=== | ===Partial Walkdown Sample (IP Section 03.01) (7 Samples)=== | ||
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains: | The inspectors evaluated system configurations during partial walkdowns of the following systems/trains: | ||
| Line 84: | Line 116: | ||
==71111.05 - Fire Protection== | ==71111.05 - Fire Protection== | ||
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (9 Sam ples) | |||
The inspectors evaluated the implementation of the fire protection program by con ducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas: | |||
The inspectors evaluated the implementation of the fire protection program by | |||
: (1) Unit 2 radwaste building 261/265, condensate storage building, fire area 55, on February 1, 2022 | : (1) Unit 2 radwaste building 261/265, condensate storage building, fire area 55, on February 1, 2022 | ||
: (2) Unit 2 reactor building 289, fire areas 34 and 35, on February 10, 2022 | : (2) Unit 2 reactor building 289, fire areas 34 and 35, on February 10, 2022 | ||
| Line 98: | Line 130: | ||
==71111.06 - Flood Protection Measures== | ==71111.06 - Flood Protection Measures== | ||
===Inspection Activities - Internal Flooding (IP Section 03.01) (1 Sample)=== | ===Inspection Activities - Internal Flooding (IP Section 03.01) (1 Sample)=== | ||
: (1) The inspectors evaluated internal flooding mitigation protections in the Unit 1 cable | : (1) The inspectors evaluated internal flooding mitigation protections in the Unit 1 cable spreading room on February 28, 2022. | ||
==71111.07A - Heat Exchanger/Sink Performance== | ==71111.07A - Heat Exchanger/Sink Performance== | ||
===Annual Review (IP Section 03.01) (1 Sample)=== | ===Annual Review (IP Section 03.01) (1 Sample)=== | ||
The inspectors evaluated readiness and performance of: | The inspectors evaluated readiness and performance of: | ||
: (1) Unit 2 'B' residual heat removal heat exchanger | : (1) Unit 2 'B' residual heat removal heat exchanger | ||
==71111.08G - Inservice Inspection Activities (BWR)== | ==71111.08G - Inservice Inspection Activities (BWR)== | ||
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding | BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding | ||
| Line 114: | Line 144: | ||
{{IP sample|IP=IP 71111.08G|count=1}} | {{IP sample|IP=IP 71111.08G|count=1}} | ||
: (1) 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 from March 7 to March 17, 2022: | : (1) 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 from March 7 to March 17, 2022: | ||
* Automated phased array ultrasonic testing of the N1B reactor recirculation nozzle to safe-end dissimilar metal weld, 2RPV-KB02 (NDE Report N2R18-APR-02). | * Automated phased array ultrasonic testing of the N1B reactor recirculation nozzle to safe-end dissimilar metal weld, 2RPV -KB02 (NDE Report N2R18-APR-02). | ||
* Automated phased array ultrasonic testing of the N4E reactor feedwater nozzle to safe-end dissimilar metal weld, 2RPV-KB21 (NDE Report N2R18-APR-09). | * Automated phased array ultrasonic testing of the N4E reactor feedwater nozzle to safe-end dissimilar metal weld, 2RPV -KB21 (NDE Report N2R18-APR-09). | ||
* Manual ultrasonic testing of the reactor water cleanup system pipe-to-pipe and valve-to-pipe welds, | * Manual ultrasonic testing of the reactor water cleanup system pipe-to-pipe and valve-to-pipe welds, 2W CS-09-14-FW039 / - FW040 (NDE Reports | ||
{{a|2R18}} | {{a|2R18}} | ||
==2R18 -ISI-UT-002 / -003). | |||
==2R18 - ISI-UT-002 / -003). | |||
*== | |||
Visual examinations of the containment, including accessible portions of the drywell and suppression chamber metal liner (Work Order [WO] C93672957) | Visual examinations of the containment, including accessible portions of the drywell and suppression chamber metal liner (Work Order [WO] C93672957) | ||
* Welding activities associated with | * Welding activities associated with t he modification of the instrument air check valve, 2IAS*V450, under engineering change ECP -21- 000088 (Work Order | ||
[WO] 93782709). This included liquid penetrant testing of two pipe-to-valve welds, FW-03 and FW-04 (NDE Report BOP-PT-22-004). | |||
* Flaw evaluation of the embedded reflector identified during the spring 2020 refueling outage using automated phased array UT on the N2J reactor recirculation nozzle to safe-end dissimilar metal weld (NDE Report N2R17-APR-06). The flaw was determined to be | * Flaw evaluation of the embedded reflector identified during the spring 2020 refueling outage using automated phased array UT on the N2J reactor recirculation nozzle to safe-end dissimilar metal weld (NDE Report N2R17-APR-06). The flaw was determined to be acceptabl e for continued service. | ||
==71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance== | ==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) | |||
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) | |||
(2 Samples) | |||
: (1) The inspectors observed Unit 1 operations personnel during control rod exercising operability testing on March 5, 2022. | : (1) The inspectors observed Unit 1 operations personnel during control rod exercising operability testing on March 5, 2022. | ||
: (2) The inspectors observed Unit 2 operations personnel during the plant shutdown for refueling outage N2R18 on March 6, 2022. | : (2) The inspectors observed Unit 2 operations personnel during the plant shutdown for refueling outage N2R18 on March 6, 2022. | ||
Licensed Operator Requalification Training /Examinations (IP Section 03.02) (2 Samples) | |||
: (1) The inspectors observed a Unit 2 simulator evaluation that included an instrument air compressor failure, reactor core isolation cooling system inoperability, and a small loss of coolant accident with additional failures that required the depressurization of the reactor on January 25, 2022. | : (1) The inspectors observed a Unit 2 simulator evaluation that included an instrument air compressor failure, reactor core isolation cooling system inoperability, and a small loss of coolant accident with additional failures that required the depressurization of the reactor on January 25, 2022. | ||
: (2) The inspectors observed a Unit 1 simulator evaluation that included the inadvertent opening of an electromatic relief valve, a loss of offsite power, an emergency | : (2) The inspectors observed a Unit 1 simulator evaluation that included the inadvertent opening of an electromatic relief valve, a loss of offsite power, an emergency di esel generator failure to start, and a steam leak in the drywell on February 2, 2022. | ||
==71111.12 - Maintenance Effectiveness== | ==71111.12 - Maintenance Effectiveness== | ||
===Maintenance Effectiveness (IP Section 03.01) (1 Sample)=== | ===Maintenance Effectiveness (IP Section 03.01) (1 Sample)=== | ||
The inspectors evaluated the effectiveness of maintenance to ensure the | The inspectors evaluated the effectiveness of maintenance to ensure the f ollowing structures, systems, and components (SSCs) remain capable of performing their intended function: | ||
: (1) Unit 2 Division I emergency diesel generator jacket water pump | : (1) Unit 2 Division I emergency diesel generator jacket water pump | ||
==71111.13 - Maintenance Risk Assessments and Emergent Work Control== | ==71111.13 - Maintenance Risk Assessments and Emergent Work Control== | ||
Risk Assessment and Management Sample (IP Section 03.01) (9 Sam ples) | |||
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: | 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 during planned 'B' residual heat removal pump maintenance on January 4, 2022 | : (1) Unit 2 elevated risk during planned 'B' residual heat removal pump maintenance on January 4, 2022 | ||
| Line 152: | Line 180: | ||
: (4) Unit 2 elevated risk during planned Division I emergency diesel generator maintenance on February 22, 2022 | : (4) Unit 2 elevated risk during planned Division I emergency diesel generator maintenance on February 22, 2022 | ||
: (5) Unit 2 elevated risk during emergent work on the 'B' service water pump on March 2, 2022 | : (5) Unit 2 elevated risk during emergent work on the 'B' service water pump on March 2, 2022 | ||
: (6) Unit 2 elevated risk during a planned 115-kilovolt Line 5 outage on March 6, 2022 | : (6) Unit 2 elevated risk during a planned 115 -kilovolt Line 5 outage on March 6, 2022 | ||
: (7) Unit 2 elevated risk during a planned reactor cavity flood-up on March 8, 2022 | : (7) Unit 2 elevated risk during a planned reactor cavity flood-up on March 8, 2022 | ||
: (8) Unit 2 elevated risk during planned maintenance on SWP*MOV66B, cooling water to Division II emergency diesel generator, on March 10-12, 2022 | : (8) Unit 2 elevated risk during planned maintenance on SWP*MOV66B, cooling water to Division II emergency diesel generator, on March 10- 12, 2022 | ||
: (9) Unit 2 elevated risk during a planned reactor cavity draindown on March 22, 2022 | : (9) Unit 2 elevated risk during a planned reactor cavity draindown on March 22, 2022 | ||
==71111.15 - Operability Determinations and Functionality | ==71111.15 - Operability Determinations and Functionality Assessme nts== | ||
Operability Determination or Functionality Assessment (IP Section 03.01) (9 Samples) | |||
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments: | The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments: | ||
: (1) Unit 2 Division I emergency diesel generator slow start on January 3, 2022 | : (1) Unit 2 Division I emergency diesel generator slow start on January 3, 2022 | ||
| Line 167: | Line 195: | ||
: (5) Unit 1 emergency diesel generator 103 starting flywheel chipped tooth on February 15, 2022 | : (5) Unit 1 emergency diesel generator 103 starting flywheel chipped tooth on February 15, 2022 | ||
: (6) Unit 1 containment spray raw water 122 rate set valve unable to operate on February 24, 2022 | : (6) Unit 1 containment spray raw water 122 rate set valve unable to operate on February 24, 2022 | ||
: (7) Unit 1 core spray topping pump 111 pump-bearing oil leak on March 2, 2022 | : (7) Unit 1 core spray topping pump 111 pump -bearing oil leak on March 2, 2022 | ||
: (8) Unit 2 Division III diesel under voltage relay failure to reset after testing on March 4, 2022 | : (8) Unit 2 Division III diesel under voltage relay failure to reset after testing on March 4, 2022 | ||
: (9) Unit 2 main steam isolation valve slow fast closure times on March 7, 2022 | : (9) Unit 2 main steam isolation valve slow fast closure times on March 7, 2022 | ||
==71111.18 - Plant Modifications== | ==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: | The inspectors evaluated the following temporary or permanent modifications: | ||
: (1) | : (1) Perm anent Modification: ECP-21-000437, Unit 2 Division II Emergency Diesel Generator Governor Booster | ||
: (2) Permanent Modification: ECP-21-000454, Unit 2 Digital Electro-Hydraulic Control (DEHC) Low Pass Filter Modification | : (2) Permanent Modification: ECP-21-000454, Unit 2 Digital Electro-Hydraulic Control (DEHC) Low Pass Filter Modification | ||
: (3) | : (3) Perm anent Modification: ECP-21-000088, Unit 2 PCIV [primary containment isolation valve] Supply Nitrogen Line Primary Containment Inboard Isolation Re-Design | ||
==71111.19 - Post-Maintenance Testing== | ==71111.19 - Post-Maintenance Testing== | ||
Post-Maintenance Test Sample (IP Section 03.01) (5 Sam ples) | |||
The inspectors evaluated the following post -maintenance testing activities to verify system operability and/or functionality: | |||
The inspectors evaluated the following post-maintenance testing activities to verify system operability and/or functionality: | |||
: (1) Unit 2 Division I emergency diesel generator jacket water pump following replacement on February 23, 2022 | : (1) Unit 2 Division I emergency diesel generator jacket water pump following replacement on February 23, 2022 | ||
: (2) Unit 2 'B' service water pump following a failure to start on March 2, 2022 | : (2) Unit 2 'B' service water pump following a failure to start on March 2, 2022 | ||
: (3) Unit 2 Division II emergency diesel generator following governor oil | : (3) Unit 2 Division II emergency diesel generator following governor oil bo oster installation on March 21, 2022 | ||
: (4) Unit 2 'B' residual heat removal system following heat exchanger | : (4) Unit 2 'B' residual heat removal system following heat exchanger inspectio n on March 25, 2022 | ||
: (5) Unit 2 drywell nitrogen supply system following solenoid operated valve replacement on March 28, 2022 | : (5) Unit 2 drywell nitrogen supply system following solenoid operated valve replacement on March 28, 2022 | ||
==71111.20 - Refueling and Other Outage Activities== | ==71111.20 - Refueling and Other Outage Activities== | ||
===Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)=== | ===Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)=== | ||
: (1) The inspectors evaluated Unit 2 refueling outage N2R18 from March 6 to | : (1) The inspectors evaluated Unit 2 refueling outage N2R18 from March 6 to March 27, 2022. | ||
==71111.22 - Surveillance Testing== | ==71111.22 - Surveillance Testing== | ||
The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality: | The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality: | ||
===Surveillance Tests (other) (IP Section 03.01) (8 Samples)=== | ===Surveillance Tests (other) (IP Section 03.01) (8 Samples)=== | ||
: (1) N1-ST-M4A, Emergency Diesel Generator 102 and PB 102 Operability Test, on January 24, 2022 | : (1) N1-ST-M4A, Emergency Diesel Generator 102 and PB 102 Operability Test, on January 24, 2022 | ||
: (2) N2-OSP-CSL-Q@ | : (2) N2-OSP-CSL-Q@00 2, LPCS [low pressure core spray] Pump and Valve Operability and System Integrity Test, on February 28, 2022 | ||
: (3) N1-ST-Q1A, Core Spray 111 Pump, Valve and Shutdown Cooling Water Seal Check | : (3) N1-ST-Q1A, Core Spray 111 Pump, Valve and Shutdown Cooling Water Seal Check Va lve Operability Test, on March 1, 2022 | ||
: (4) N2-OSP-MSS-CS001, Main Steam Isolation Valve Operability Test, on March 7, 2022 | : (4) N2-OSP-MSS-CS001, Main Steam Isolation Valve Operability Test, on March 7, 2022 | ||
: (5) N2-OSP-RHS-R001, Division II ECCS [emergency core cooling system] Functional Test, on March 9, 2022 | : (5) N2-OSP-RHS-R001, Division II ECCS [emergency core cooling system] Functional Test, on March 9, 2022 | ||
: (6) N2-OSP-SLS-R001, Standby Liquid Control Manual Initiate Actuation and ASME XI Pressure Test, on March 16, 2022 | : (6) N2-OSP-SLS-R001, Standby Liquid Control Manual Initiate Actuation and ASME XI Pressure Test, on March 16, 2022 | ||
: (7) N2-OSP-ADS-R002, ADS [automatic depressurization system] Functional Test and Remote Shutdown System Test, on March 19, 2022 | : (7) N2-OSP-ADS-R002, ADS [automatic depressurization system] Functional Test and Remote Shutdown System Test, on March 19, 2022 | ||
: (8) N2-OSP-EGS-R001, Diesel Generator ECCS Start and Load Reject Division II, on | : (8) N2-OSP-EGS-R001, Diesel Generator ECCS Start and Load Reject Division II, on March 20, 2022 | ||
Inservice Testing (IP Section 03.01) (2 Sam ples) | |||
: (1) N2-OSP- | : (1) N2-OSP-I CS-Q@002, Reactor Core Isolation Cooling Pump and Valve Operability Test and System Integrity Test and ASME XI Functional Test and Analysis, on February 10, 2022 | ||
: (2) N2-ISP-RRC-R001, ARI [alternate rod insertion] Function of RRCS [redundant reactivity control system], on March 7, 2022 | : (2) N2-ISP-RRC-R001, ARI [alternate rod insertion] Function of RRCS [redundant reactivity control system], on March 7, 2022 | ||
| Line 218: | Line 243: | ||
==RADIATION SAFETY== | ==RADIATION SAFETY== | ||
==71124.01 - Radiological Hazard Assessment and Exposure Controls== | ==71124.01 - Radiological Hazard Assessment and Exposure Controls== | ||
===Radiological Hazard Assessment (IP Section 03.01) (1 Sample)=== | ===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 Constellation assesses radiological hazards. | : (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 Constellation assesses radiological hazards. | ||
| Line 227: | Line 250: | ||
: (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. | : (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 Sam p les) | |||
The inspectors observed/evaluated the following licensee | |||
The inspectors observed/evaluated the following licensee process es for monitoring and controlling contamination and radioactive material: | |||
: (1) Workers exiting the Unit 2 radiologically controlled area during refueling outage N2R18 | : (1) Workers exiting the Unit 2 radiologically controlled area during refueling outage N2R18 | ||
: (2) Licensee surveys of contaminated equipment on the refuel floor during refueling outage N2R18 | : (2) Licensee surveys of contaminated equipment on the refuel floor during refueling outage N2R18 | ||
| Line 238: | Line 262: | ||
: (3) Unit 2 refuel floor equipment decontamination activities | : (3) Unit 2 refuel floor equipment decontamination activities | ||
: (4) Unit 2 drywell feedwater nozzle inspection activities | : (4) Unit 2 drywell feedwater nozzle inspection activities | ||
: (5) Unit 2 safety relief valve maintenance in the drywell High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (2 Samples) | : (5) Unit 2 safety relief valve maintenance in the drywell | ||
High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (2 Samples) | |||
The inspectors evaluated licensee controls of the following High Radiation Areas and Very High Radiation Areas: | The inspectors evaluated licensee controls of the following High Radiation Areas and Very High Radiation Areas: | ||
: (1) High Radiation Area in the Unit 2 reactor building valve pit, 196' elevation | : (1) High Radiation Area in the Unit 2 reactor building valve pit, 196' elevation | ||
: (2) Locked | : (2) Locked Hig h Radiation Areas in the Unit 2 drywell | ||
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. | : (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements. | ||
==OTHER ACTIVITIES - BASELINE== | ==OTHER ACTIVITIES - BASELINE== | ||
===71151 - Performance Indicator Verification | |||
The inspectors verified licensee performance indicators submittals listed below: | |||
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) (2 Sam ples) | |||
IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01) | |||
: (1) Unit 1 (January 1, 2021 through December 31, 2021) | : (1) Unit 1 (January 1, 2021 through December 31, 2021) | ||
: (2) Unit 2 (January 1, 2021 through December 31, 2021) | : (2) Unit 2 (January 1, 2021 through December 31, 2021) | ||
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) | IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) | ||
=== | |||
{{IP sample|IP=IP 71151|count=2}} | |||
: (1) Unit 1 (January 1, 2021 through December 31, 2021) | : (1) Unit 1 (January 1, 2021 through December 31, 2021) | ||
: (2) Unit 2 (January 1, 2021 through December 31, 2021) | : (2) Unit 2 (January 1, 2021 through December 31, 2021) | ||
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples) | |||
: (1) Unit 1 (January 1, 2021 through December 31, 2021) | : (1) Unit 1 (January 1, 2021 through December 31, 2021) | ||
: (2) Unit 2 (January 1, 2021 through December 31, 2021) | : (2) Unit 2 (January 1, 2021 through December 31, 2021) | ||
===71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03) === | ===71152A - Annual Follow -up Problem Identification and Resolution Annual Follow -up of Selected Issues (Section 03.03)=== | ||
{{IP sample|IP=IP 71152|count=1}} | {{IP sample|IP=IP 71152|count=1}} | ||
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues: | The inspectors reviewed the licensees implementation of its corrective action program related to the following issues: | ||
| Line 266: | Line 299: | ||
==INSPECTION RESULTS== | ==INSPECTION RESULTS== | ||
Observation: Nine Mile Point Control Rod Blade | Observation: Nine Mile Point Control Rod Blade Techni cal Evaluation 71152A On November 23, 2020, General Electric Hitachi (GEH) released Safety Communication 20 - | ||
06 (SC 20- 06) Revision 0, describing the discovery that Boron 10 (B -10) depletion had been underpredicted in the top 6-inch node of the neutron absorber section of control rod blades currently in use in boiling water reactors in the United States. Nine Mile Point Nuclear Station was on the list of plants affected by this issue. | |||
In Revision 0 of the Safety Communication, the population of control rod blades affected was restricted to Original Equipment Manufacturer (OEM) series D-100 control rod blades, still in operation from the 1970s and 1980s. On February 26, 2021, GEH issued Revision 1 to SC 20- 06, in which it was determined that the population of control rod blades affected by the underprediction of B -10 in the tip segm ent included all control rod blades that did not contain Hafnium (Hf) in the tips, not just OEM D-100 control rod blades. This greatly increased the population of control rod blades affected by this issue. | |||
Control rod blades deplete the B-10 isotope as they absorb neutrons when they are inserted into an operating reactor core. Control rod blades are typically fully withdrawn from the core when the reactor is at full power. However, the tips of the control rod blades still experience some thermal neutron flux. The B-10 depletion in the top node of the control rod blade is accounted for in engineering analysis by adding a "tip adder" factor in the B-10 depletion calculation for each control rod blade. Safety Communication SC 20-06 stated that the tip adder factor was much larger than previously calculated, for control rod blades without Hf in the tips. General Electric manufactured control rod blades with Hf tips for a period of time in the 1980s and 1990s that are excluded from the tip adder issue described in SC 20-06. All other control rod blade models, including the OEM blades, are affected by the tip adder issue described in SC 20-06. The higher depletion values in the control rod blade tips for multiple control rod blade types may cause the control rod blade to exceed its effective neutron absorbing capability before the end of the operating fuel cycle. This has the potential to decrease the overall shutdown margin (SDM), which is the ability of all the control rods, except for the most reactive control | Control rod blades deplete the B -10 isotope as they absorb neutrons when they are inserted into an operating reactor core. Control rod blades are typically fully withdrawn from the core when the reactor is at full power. However, the tips of the control rod blades still experience some thermal neutron flux. The B -10 depletion in the top node of the control rod blade is accounted for in engineering analysis by adding a "tip adder" factor in the B -10 depletion calculation for each control rod blade. Safety Communication SC 20- 06 stated that the tip adder factor was much larger than previously calculated, for control rod blades without Hf in the tips. General Electric manufactured control rod blades with Hf tips for a period of time in the 1980s and 1990s that are excluded from the tip adder issue described in SC 20 -06. All other control rod blade models, including the OEM blades, are affected by the tip adder issue described in SC 20- 06. The higher depletion values in the control rod blade tips for multiple control rod blade types may cause the control rod blade to exceed its effective neutron absorbing capability before the end of the operating fuel cycle. This has the potential to decrease the overall shutdown margin (SDM), which is the ability of all the control rods, except for the most reactive control r od, to shut down the reactor core in all anticipated normal and accident scenarios. | ||
In Revision 1 to SC 20-06, GEH recommended all customers ensure that adequate SDM was available in the current operating fuel cycle until the end of the fuel cycle, once the | In Revision 1 to SC 20- 06, GEH recommended all customers ensure that adequate SDM was available in the current operating fuel cycle until the end of the fuel cycle, once the am ount of B-10 depletion was determined in all control rod blade tips. GEH recommended this be done for all General Electric control rod blades that did not contain Hf tips, as well as alternate vendor control rod blades. GEH also recommended all customers assess the impact of control rod blades exceeding their nuclear end of life (NEOL) criteria - the ability of the control rod blade to effectively absorb neutrons - fo r future fuel cycles beyond the current operating fuel cycle. | ||
Following issuance of GEH SC 20-06, Revision 1, Nine Mile Point Unit 1 entered a refueling outage in March 2021. A control rod blade depletion engineering evaluation was performed, taking into account the new tip adder calculation in SC 20-06. It was determined that two control rod blades that had been located in higher power locations in the previous fuel cycle would be shuffled to two lower power locations in the core for the current Unit 1 operating fuel cycle. No control rod blades in the previous or current fuel cycle have exceeded their NEOL criteria, thus requiring replacement. | Following issuance of GEH SC 20-06, Revision 1, Nine Mile Point Unit 1 entered a refueling outage in March 2021. A control rod blade depletion engineering evaluation was performed, taking into account the new tip adder calculation in SC 20- 06. It was determined that two control rod blades that had been located in higher power locations in the previous fuel cycle would be shuffled to two lower power locations in the core for the current Unit 1 operating fuel cycle. No control rod blades in the previous or current fuel cycle have exceeded their NEOL criteria, thus requiring replacement. | ||
In July 2021, Exelon Nuclear Fuels validated that adequate SDM existed for the remainder of the current fuel cycle for Nine Mile Point Unit 2, and that thermal limit margins were not | In July 2021, Exelon Nuclear Fuels validated that adequate SDM existed for the remainder of the current fuel cycle for Nine Mile Point Unit 2, and that thermal limit margins were not im pa cted. One control rod blade on the core periphery was predicted to exceed its NEOL criteria due to the tip adder issue described in SC 20- 06, before the end of the Unit 2 fuel cycle in March 2022. This blade, along with six other control rod blades, are scheduled for replacement in the upcoming refueling outage. In September 2021, General Electric Global Nuclear Fuels completed an analysis to confirm that SDM would be maintained above the technical specification limit for the remainder of the Unit 2 fuel cycle. Exelon procedure NF - | ||
AB-135-1410, "BWR Control Blade Lifetime Management," contains guidance for calculating B-10 depletion in all models of control rod blades currently in operation in the nuclear fleet. | |||
This procedure will require an update to | This procedure will require an update to incorpora te the new guidance from SC 20- 06, for estimating the tip adder factor for non-Hf tipped control rod blades. Issue Report 04242262 is tracking completion of this procedural update. | ||
The inspectors interviewed engineering staff from Exelon Nuclear | The inspectors interviewed engineering staff from Exelon Nuclear F uels and Nine Mile Point Reactor Engineering to discuss corrective actions taken in response to GEH SC 20-06, and reviewed shutdown margin and control rod blade depletion engineering evaluations. The inspectors also reviewed the corrective actions taken and planned for responding to GEH SC 20- 06 entered into the corrective action system. No performance deficiencies were identified. | ||
Corrective Action References: 04242262, 04386300, | Corrective Action References: 04242262, 04386300, | ||
| Line 284: | Line 321: | ||
==EXIT MEETINGS AND DEBRIEFS== | ==EXIT MEETINGS AND DEBRIEFS== | ||
The inspectors verified no proprietary information was retained or documented in this report. | The inspectors verified no proprietary information was retained or documented in this report. | ||
* On April 21, 2022, the | * On April 21, 2022, the inspector s presented the integrated inspection results to Mr. Peter Orphanos, Site Vice Preside n t, and other members of the licensee staff. | ||
* On February 8, 2022, the inspectors presented the control rod blade technical evaluation problem identification and resolution inspection results to Philip Nichols, Manager Reactor Engineering, and other members of the licensee staff. | * On February 8, 2022, the inspectors presented the control rod blade technical evaluation problem identification and resolution inspection results to Philip Nichols, Manager Reactor Engineering, and other members of the licensee staff. | ||
* On March 17, 2022, the | * On March 17, 2022, the inspector s presented the Unit 2 i nservice inspection results to Mr. Pe ter Orphanos, Site Vice President, and other members of the licensee staff. | ||
* On March 17, 2022, the | * On March 17, 2022, the inspector s presented the radiological hazard assessment and exposure controls inspection results to Mr. Peter Orphanos, Site Vice President, and other members of the licensee staff. | ||
=DOCUMENTS REVIEWED= | =DOCUMENTS REVIEWED= | ||
Inspection Type | Inspection Type Designation Description or Title Revision or | ||
Procedure | Procedure Date | ||
71111.01 | 71111.01 Procedures N1-OP-64 Meteorological Monitoring 02100 | ||
N2-OP-102 | N2-OP-102 Meteorological Monitoring 02800 | ||
OP-AA-108-111- | OP-AA-108-111-Severe Weather and Natural Disaster Guidelines 24 | ||
1001 | 1001 | ||
71111.04 | 71111.04 Drawings C-18022-C Piping & Instrumentation Diagram, Reactor Building Closed 55 | ||
Loop Cooling System | Loop Cooling System | ||
PID-031A, B, E | PID-031A, B, E Piping & Instrumentation Diagram Residual Heat Removal 27 | ||
System | System | ||
PID-31G | PID-31G Piping & Instrumentation Diagram Residual Heat Removal 15 | ||
PID-35A | PID-35A Piping & Instrumentation Diagram Reactor Core Isolation 17 | ||
Cooling | Cooling | ||
PID-35B | PID-35B Piping & Instrumentation Diagram Reactor Core Isolation 13 | ||
Cooling | Cooling | ||
PID-38B | PID-38B Piping & Instrumentation Diagram Fuel Pool Cooling & 15 | ||
Cleanup | Cleanup | ||
PID-38C | PID-38C Piping & Instrumentation Diagram Fuel Pool Cooling & 17 | ||
Cleanup | Cleanup | ||
Procedures | Procedures N1-OP-11 Reactor Building Closed Loop Cooling System 03400 | ||
N2-OP-100A | N2-OP-100A Standby Diesel Generators 03200 | ||
N2-OP-100A- | N2-OP-100A-Standby Diesel Generators - LINEUPS 00500 | ||
LINEUPS | LINEUPS | ||
N2-OP-31 | N2-OP-31 Residual Heat Removal System 03700 | ||
N2-OP-31- | N2-OP-31-Residual Heat Removal System 003 | ||
LINEUPS | LINEUPS | ||
N2-OP-38 | N2-OP-38 Spent Fuel Cooling and Cleanup System 2700 | ||
71111.05 | 71111.05 Corrective Action 04477281 | ||
Docum ents | |||
Drawings | Drawings B-40143-C Fire Zones Reactor Building - Fl. El. 261' Turbine Building - 10 | ||
Fl. El. 261' Fire Rated Walls and Slabs | Fl. El. 261' Fire Rated Walls and Slabs | ||
Fire Plans | Fire Plans N2-FPI-PFP-0201 Unit 2 Pre-Fire Plans 06 | ||
71111.06 | 71111.06 Corrective Action 04416206 | ||
Docum ents 04477927 | |||
Inspection Type | Inspection Type Designation Description or Title Revision or | ||
Procedure | Procedure Date | ||
04661043 | 04661043 | ||
71111.07A | 71111.07A Corrective Action 04484882 | ||
Docum ents 04486864 | |||
Procedures | Procedures ER-AA-340-1002 Service Water Heat Exchanger Inspection Guide 11 | ||
S-TDP-REL-0102 Service Water Heat Exchanger and Component Inspection | S-TDP-REL-0102 Service Water Heat Exchanger and Component Inspection 03 | ||
Guide | Guide | ||
Work Orders | Work Orders C93738267 | ||
71111.08G Corrective Action | 71111.08G Corrective Action 04327298 | ||
Docum ents | |||
Corrective Action 04484400 | Corrective Action 04484400 | ||
Docum ents | |||
Resulting from | Resulting from | ||
Inspection | Inspection | ||
Engineering | Engineering ECP-20-000224 Recirculation Inlet Nozzle DMW No. 2RPV-KB11 (N2J) Flaw 03/25/2020 | ||
Evaluations | Evaluations Evaluation | ||
Miscellaneous | Miscellaneous ER-NM-330-2001 ISI Program Plan Fourth Ten-Year Inspection Interval Revision 3 | ||
ER-NM-330-2004 Risk Informed Inservice Inspection Program Fourth Ten- | ER-NM-330-2004 Risk Informed Inservice Inspection Program Fourth Ten-Revision 0 | ||
Year Inspection Interval | Year Inspection Interval | ||
Procedures | Procedures ER-AA-335-018 Visual Examination of ASME IWE Class MC and Metallic Revision 15 | ||
Liners of Class CC Components | Liners of Class CC Components | ||
ER-AA-335-030 | ER-AA-335-030 Ultrasonic Examination of Ferritic Piping Welds Revision 5 | ||
ER-AA-335-1000 Nondestructive Examination (NDE) | ER-AA-335-1000 Nondestructive Examination (NDE) Revision 16 | ||
GEH-UT-254 | GEH-UT-254 Automated Phased Array Ultrasonic Examination of Version 1 | ||
Dissimilar Metal Welds with the TOPAZ | Dissimilar Metal Welds with the TOPAZ | ||
WPS 8-8-GTSM | WPS 8-8-GTSM Welding Procedure Specification Record for Manual GTAW Revision 7 | ||
and SMAW of P-Number 8 to P-Number 8 Base Metal | and SMAW of P-Number 8 to P-Number 8 Base Metal | ||
71111.11Q Procedures | 71111.11Q Procedures N1-ST-W1 Control Rod Exercising Operability Test 02300 | ||
N2-OP-101C | N2-OP-101C Plant Shutdown 04100 | ||
N2-OP-29 | N2-OP-29 Reactor Recirculation System 03400 | ||
N2-OP-31 | N2-OP-31 Residual Heat Remova l S yst em 03700 | ||
71111.12 | 71111.12 Procedures N2-MSP-EGS-Diesel Generator Inspection Division 1 and 2 025 | ||
R001 | R001 | ||
Work Orders | Work Orders C938144677 | ||
71111.13 | 71111.13 Corrective Action 04481795 | ||
Inspection Type | |||
Procedure | Inspection Type Designation Description or Title Revision or | ||
Procedure Date | |||
Procedures | Docum ents | ||
Procedures N2-OP-19-Instrument and Service Air Systems 8 | |||
Lineups | Lineups | ||
N2-OP-70 | N2-OP-70 Station Electrical Feed and 115KV Switchyard 02800 | ||
N2-PM-082 | N2-PM-082 RPV [reactor pressure vessel] Flood-Up/Draindown 01900 | ||
OP-NM-108-117 | OP-NM-108-117 Protected Equipm ent Program at Nine Mile Point 5 | ||
OP-NM-108-117 | OP-NM-108-117 Protected Equipm ent Program at Nine Mile Point 00500 | ||
OU-NM-103-101 | OU-NM-103-101 Shutdown Safety Management Program 0700 | ||
71111.15 | 71111.15 Calculations 002N3714 Nine Mile Point Nuclear Station Unit 1 TRACG-LOCA 0 | ||
Analysis for GNF2 Fuel | Analysis for GNF2 Fuel | ||
Corrective Action 04061889 | Corrective Action 04061889 | ||
Docum ents 04470659 | |||
04475343 | 04475343 | ||
04476776 | 04476776 | ||
| Line 381: | Line 419: | ||
04483059 | 04483059 | ||
04488015 | 04488015 | ||
Drawings | Drawings 0001040209048 Control Diagram Shutdown System 13.00 | ||
Engineering | Engineering ECP-22-000147 Technical Evaluation for MSIV Failed Stroke Time Extent of 0000 | ||
Changes | Changes Condition | ||
Miscellaneous | Miscellaneous NEI 06-09-A Risk-Informed Technical Specifications Initiative 4b: Risk-0 | ||
Managed Technical Specifications (RMTS) Guidelines | Managed Technical Specifications (RMTS) Guidelines | ||
Purchase Order | Purchase Order Service, Repair, Refurbishment of MSIV Actuator Air Pack 2 | ||
00802656 | 00802656 | ||
RICT Record | RICT Record fo r Failed MSIV 7D RPS Testing 04/06/2022 | ||
March 28, 2022 | March 28, 2022 | ||
Procedures | Procedures N1-OP-1 Nuclear Steam Supply System 07700 | ||
N2-OP-100A | N2-OP-100A Standby Diesel Generators 03100 | ||
N2-OSP-EGS- | N2-OSP-EGS-Diesel Generator and Diesel Air Start Valve Operability Test-024T1 | ||
M@001 | M@001 Division I and II | ||
OP-AA-108-118 | OP-AA-108-118 Risk Informed Completion Time 2 | ||
71111.18 | 71111.18 Corrective Action 02547530 | ||
Inspection Type | |||
Procedure | Inspection Type Designation Description or Title Revision or | ||
Procedure Date | |||
Docum ents 02689624 | |||
04359704 | 04359704 | ||
04365824 | 04365824 | ||
04428910 | 04428910 | ||
04482114 | 04482114 | ||
Engineering | Engineering ECP-21-000088 PCIV Supply Nitrogen Line Primary Containment Inboard 0000 | ||
Changes | Changes Isolation Re-Design | ||
ECP-21-000437 EDG Governor Booster | ECP-21-000437 EDG Governor Booster 0000 | ||
ECP-21-000454 Digital Electro-Hydraulic Control (DEHC) Low Pass Filter | ECP-21-000454 Digital Electro-Hydraulic Control (DEHC) Low Pass Filter 0 | ||
Modification | Modification | ||
Miscellaneous | Miscellaneous FAT/SAT Testing for DEHC Low Pass Filter Mod 0000 | ||
Procedures | Procedures N2-OSP-EGS-Diesel Generator and Diesel Air Start Valve Operability Test 024T1 | ||
M@0001 | M@0001 - Division I and II | ||
Work Orders | Work Orders C93782709 | ||
C93809087 | C93809087 | ||
C93813137 | C93813137 | ||
71111.19 | 71111.19 Corrective Action 04480094 | ||
Docum ents 04481795 | |||
04484882 | 04484882 | ||
04486059 | 04486059 | ||
Corrective Action 04486063 | Corrective Action 04486063 | ||
Docum ents | |||
Resulting from | Resulting from | ||
Inspection | Inspection | ||
Procedures | Procedures GAP-HSC-09 System Aging Inspection and Cleanliness Controls 02000 | ||
N2-MSP-EGS- | N2-MSP-EGS-Diesel Generator Inspection Division I and II 02400 | ||
R001 | R001 | ||
N2-OSP-EGS- | N2-OSP-EGS-Diesel Generator and Diesel Air Start Valve Operability Test 02400.01 | ||
M@0001 | M@0001 - Division I and II | ||
N2-OSP-RHS- | N2-OSP-RHS-RHR System Loop B Pump and Valve Operability Test, 01500 | ||
Q@005 | Q@005 System Integrity Test and ASME XI Pressure Test | ||
N2-OSP-RHS- | N2-OSP-RHS-Division II ECCS Functional Test 00900 | ||
R001 | R001 | ||
Inspection Type | |||
Procedure | Inspection Type Designation Description or Title Revision or | ||
S-EPM-GEN-004 | Procedure Date | ||
S-EPM-GEN-004 Insulation of Power, Control, and Instrument Cable 00700 | |||
Connections | Connections | ||
Work Orders | Work Orders C93738267 | ||
C93782709 | C93782709 | ||
C93814677 | C93814677 | ||
C93825512 | C93825512 | ||
71111.20 | 71111.20 Corrective Action 04482986 | ||
Docum ents 04482995 | |||
04483785 | 04483785 | ||
04484026 | 04484026 | ||
| Line 446: | Line 486: | ||
04485270 | 04485270 | ||
Corrective Action 04484400 | Corrective Action 04484400 | ||
Docum ents 04486063 | |||
Resulting from | Resulting from | ||
Inspection | Inspection | ||
Miscellaneous | Miscellaneous NM2C19-SU Reactivity Maneuver Plan 0 | ||
Procedures | Procedures LS-AA-119 Fatigue Management and Work Hour L im it s 15 | ||
N2-FHP-13.3 | N2-FHP-13.3 Core Shuffle 01200 | ||
N2-OP-101A | N2-OP-101A Plant Start-up 05400 | ||
N2-OP-38 | N2-OP-38 Spent Fuel Cooling and Cleanup System 02700 | ||
N2-OSP-NMS- | N2-OSP-NMS-Source Range Monitor Check During Core Offload/Reload 00201 | ||
@002 | |||
OP-AA-109-101 | OP-AA-109-101 Personnel and Equipment Tagout Process 16 | ||
OP-AA-300 | OP-AA-300 Reactivity Management 14 | ||
OP-AA-300-1520 Reactivity Management - Fuel Handling, Storage and | OP-AA-300-1520 Reactivity Management - Fuel Handling, Storage and 7 | ||
Refueling | Refueling | ||
OU-NM-103-101 | OU-NM-103-101 Shutdown Safety Management Program 00700 | ||
OU-NM-4001 | OU-NM-4001 Refueling Operations 00800 | ||
71111.22 | 71111.22 Corrective Action 04481649 | ||
Docum ents 04482027 | |||
04483059 | 04483059 | ||
04483200 | 04483200 | ||
Inspection Type | |||
Procedure | Inspection Type Designation Description or Title Revision or | ||
Procedure Date | |||
04483399 | 04483399 | ||
04485322 | 04485322 | ||
Miscellaneous | Miscellaneous Technical MSIV Leak Rate Tests Approved | ||
Evaluation N2R18 | Evaluation N2R18 03/17/2022 | ||
Procedures | Procedures N1-ST-M4A Emergency Diesel Generator 102 and PB 102 Operability 03000 | ||
Test | Test | ||
N1-ST-Q1A | N1-ST-Q1A CS 111 Pump, Valve and SDC Water Seal Check Valve 02200 | ||
Operability Test | Operability Test | ||
N2-ISP-RRC- | N2-ISP-RRC-ARI [alternate rod insertion] function of RRCS [redundant 00700 | ||
R001 | R001 reactivity control system] | ||
N2-OSP-ADS- | N2-OSP-ADS-ADS System Functional Test and Remote Shutdown System 009 | ||
R002 | R002 Test | ||
N2-OSP-CSL- | N2-OSP-CSL-LPCS Pump and Valve Operability and System Integrity 01500 | ||
Q@002 | Q@002 Test | ||
N2-OSP-EGS- | N2-OSP-EGS-Diesel Generator ECCS Start and Load Reject Division II 9 | ||
R001 | R001 | ||
N2-OSP- | N2-OSP-I CS-RCIC Pump and Valve Operability Test and System Integrity 01600 | ||
Q@002 | Q@002 Test and ASME XI Functional Test and Analysis | ||
N2-OSP-MSS- | N2-OSP-MSS-Main Steam Isolation Valve Leak Rate Test 00300 | ||
003 | 003 | ||
N2-OSP-MSS- | N2-OSP-MSS-Main Steam Isolation Valve Leak Rate Test (Reactor Vessel 00200 | ||
004 | 004 Head Rem oved) | ||
N2-OSP-MSS- | N2-OSP-MSS-Main Steam Isolation Valve Operability Test 01000 and | ||
CS001 | CS001 01100 | ||
N2-OSP-SLS- | N2-OSP-SLS-Standby Liquid Control Manual Initiate Actuation and ASME 01100 | ||
R001 | R001 XI Pressure Test | ||
71151 | 71151 Procedures NEI 99-02 Regulatory Assessment Performance Indicator Guideline 7 | ||
71152A | 71152A Corrective Action 04242262 | ||
Docum ents 04386300 | |||
04436618 | 04436618 | ||
Engineering | Engineering ECP-20-000291 Control Blade Replacement Strategy for Nine Mile Point Unit 0 | ||
Evaluations | Evaluations 1 Cycle 25 (Reload 26) | ||
ECP-21-000248 | ECP-21-000248 Control Blade Replacement Strategy for Nine Mile Point Unit 0 | ||
(Reload 18) | (Reload 18) | ||
Inspection Type | |||
Procedure | Inspection Type Designation Description or Title Revision or | ||
ECP-21-000380 | Procedure Date | ||
ECP-21-000380 SC 20-06: Nine Mile Point Unit 2 Cycle 18 SDM and 0 | |||
MSBWP Evaluation | MSBWP Evaluation | ||
Miscellaneous General Electric Impact of Ex-core Flux on Control Rod Lifetime | Miscellaneous General Electric Impact of Ex-core Flux on Control Rod Lifetime L im it s 02/26/2021 | ||
Hitachi Safety | Hitachi Safety | ||
Communication | Communication | ||
20-06, Rev. 1 | 20-06, Rev. 1 | ||
Procedures | Procedures NF-AB-130-3690 Maximum Subcritical Banked Withdrawal Position 10 | ||
NF-AB-135-1410 | NF-AB-135-1410 BWR Control Blade Lifetime Management 16 | ||
19 | 19 | ||
}} | }} | ||
Latest revision as of 02:35, 18 November 2024
| ML22119A018 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/29/2022 |
| From: | Erin Carfang NRC/RGN-I/DORS |
| To: | Rhoades D Constellation Energy Generation, Constellation Nuclear |
| Carfang E | |
| References | |
| IR 2022001 | |
| Download: ML22119A018 (21) | |
Text
April 29, 2022
SUBJECT:
NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000220/2022001 AND 05000410/2022001
Dear Mr. Rhoades:
On March 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Nine Mile Point Nuclear Station, Units 1 and 2. On April 21, 2022, the NRC inspectors discussed the results of this inspection with Mr. Peter Orphanos, Site Vice President, a nd other members of your staff. The results of this inspection are documented in the enclosed report.
No findings or violations of more than minor significance were identified during this inspection.
This letter, its enclosure, and your response (if any) will be m ade available for public inspection and copying at http://www.nrc.gov/reading-rm /adam s.h tm l and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reac to r 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 Num bers: 05000220/2022001 and 05000410/2022001
Enterprise Identifier: I-2022- 001-0049
Licensee: Constellation Energy Generation, LLC
Facility: Nine Mile Point Nuclear Station, Units 1 and 2
Location: Oswego, NY
Inspection Dates: January 1, 2022 to March 31, 2022
Inspectors: G. Stock, Senior Resident Inspector C. Kline, Resident Inspector B. Sienel, Resident Inspector N. Floyd, Senior Reactor Inspector S. Haney, Senior Project Engineer C. Hobbs, Reactor Inspector S. Wilson, Senior Health Physicist
Approved By: Erin E. Carfang, Chief Projects Branch 1 Division of Operating Reactor Safety
Enclosure
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licens ees 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 h ttps://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
No findings or violations of more than minor significance were identified.
Additional Tracking Ite ms
None.
PLANT STATUS
Unit 1 operated at or near rated thermal power for the entire inspection period.
Unit 2 began the inspection period at rated thermal power.
On January 7, 2022, the unit was downpowered to 78 percent to perform planned control r od channel interference testing and a control rod pattern adjustment, and returned to rated thermal power on January 8, 2022.
On January 28, 2022, the unit began end-of-cycle coastdown.
On February 11, 2022, the unit was downpowered to 85 percent to avoid a known oscillation region during end-of-cycle coastdown. On March 7, 2022, the unit was shut down for a planned refueling outage.
Startup was commenced on March 25, 2022, and rated thermal power was reached on March 29, 2022.
Later that day, the unit was downpowered to 80 percent for a planned rod pattern adjustment. During the downpower, a main turbine control valve malfunction required an additional downpower to 60 percent.
The unit returned to rated thermal power on March 31, 2022, and remained at or near 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, conducted routine reviews using IP 71152, Problem Identification and Resolution, 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.
On February 1, 2022, the operating license for Nine Mile Point Nuclear Station, held by Exelon Generation Company, LLC, was transferred to Constellation Energy Generation, LLC (Constellation). While some or all of the inspections documented in this report w ere performed while the license was held by Exelon Generation Company, LLC, this report will refer to the licensee as Constellation throughout.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the adequacy of the overall preparations to protect risk -
significant systems due to a winter storm warning on March 11, 2022.
71111.04 - Equipm ent Alignment
Partial Walkdown Sample (IP Section 03.01) (7 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2 reactor core isolation cooling system on January 26, 2022
- (2) Unit 1 reactor building closed loop cooling system on February 7, 2022
- (3) Unit 2 'C' residual heat removal system on February 7, 2022
- (4) Unit 2 Division II emergency diesel generator on February 22, 2022
- (5) Unit 2 'B' residual heat removal system in shutdown cooling on March 8, 2022
- (6) Unit 2 Division I emergency diesel generator on March 10, 2022
- (7) Unit 2 'A' spent fuel pool cooling system on March 15, 2022
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (9 Sam ples)
The inspectors evaluated the implementation of the fire protection program by con ducting 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 radwaste building 261/265, condensate storage building, fire area 55, on February 1, 2022
- (2) Unit 2 reactor building 289, fire areas 34 and 35, on February 10, 2022
- (3) Unit 2 reactor building 240 north, fire area 1, on February 14, 2022
- (4) Unit 1 turbine building 261' west, fire area 5, on February 17, 2022
- (5) Unit 2 reactor building 175' north, fire area 1, on February 28, 2022
- (6) Unit 2 reactor building, primary containment steam tunnel, fire area 50, on March 7, 2022
- (7) Unit 2 turbine building, condenser, fire area 50, on March 7, 2022
- (8) Unit 2 turbine building, feedwater heater bays, fire area 50, on March 9, 2022
- (9) Unit 2 reactor building, drywell, fire area 5, on March 16, 2022
71111.06 - Flood Protection Measures
Inspection Activities - Internal Flooding (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated internal flooding mitigation protections in the Unit 1 cable spreading room on February 28, 2022.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
- (1) Unit 2 'B' residual heat removal heat exchanger
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01)
- (1) 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 from March 7 to March 17, 2022:
- Automated phased array ultrasonic testing of the N1B reactor recirculation nozzle to safe-end dissimilar metal weld, 2RPV -KB02 (NDE Report N2R18-APR-02).
- Automated phased array ultrasonic testing of the N4E reactor feedwater nozzle to safe-end dissimilar metal weld, 2RPV -KB21 (NDE Report N2R18-APR-09).
- Manual ultrasonic testing of the reactor water cleanup system pipe-to-pipe and valve-to-pipe welds, 2W CS-09-14-FW039 / - FW040 (NDE Reports
==2R18 - ISI-UT-002 / -003).
- ==
Visual examinations of the containment, including accessible portions of the drywell and suppression chamber metal liner (Work Order [WO] C93672957)
- Welding activities associated with t he modification of the instrument air check valve, 2IAS*V450, under engineering change ECP -21- 000088 (Work Order
[WO] 93782709). This included liquid penetrant testing of two pipe-to-valve welds, FW-03 and FW-04 (NDE Report BOP-PT-22-004).
- Flaw evaluation of the embedded reflector identified during the spring 2020 refueling outage using automated phased array UT on the N2J reactor recirculation nozzle to safe-end dissimilar metal weld (NDE Report N2R17-APR-06). The flaw was determined to be acceptabl e for continued service.
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 Unit 1 operations personnel during control rod exercising operability testing on March 5, 2022.
- (2) The inspectors observed Unit 2 operations personnel during the plant shutdown for refueling outage N2R18 on March 6, 2022.
Licensed Operator Requalification Training /Examinations (IP Section 03.02) (2 Samples)
- (1) The inspectors observed a Unit 2 simulator evaluation that included an instrument air compressor failure, reactor core isolation cooling system inoperability, and a small loss of coolant accident with additional failures that required the depressurization of the reactor on January 25, 2022.
- (2) The inspectors observed a Unit 1 simulator evaluation that included the inadvertent opening of an electromatic relief valve, a loss of offsite power, an emergency di esel generator failure to start, and a steam leak in the drywell on February 2, 2022.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the f ollowing structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Unit 2 Division I emergency diesel generator jacket water pump
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (9 Sam ples)
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 during planned 'B' residual heat removal pump maintenance on January 4, 2022
- (2) Unit 2 elevated risk during emergent work on the Division I emergency diesel generator starting air system on January 20, 2022
- (3) Unit 1 elevated risk during emergent work on the 'C' instrument air compressor on February 16, 2022
- (4) Unit 2 elevated risk during planned Division I emergency diesel generator maintenance on February 22, 2022
- (5) Unit 2 elevated risk during emergent work on the 'B' service water pump on March 2, 2022
- (6) Unit 2 elevated risk during a planned 115 -kilovolt Line 5 outage on March 6, 2022
- (7) Unit 2 elevated risk during a planned reactor cavity flood-up on March 8, 2022
- (8) Unit 2 elevated risk during planned maintenance on SWP*MOV66B, cooling water to Division II emergency diesel generator, on March 10- 12, 2022
- (9) Unit 2 elevated risk during a planned reactor cavity draindown on March 22, 2022
71111.15 - Operability Determinations and Functionality Assessme nts
Operability Determination or Functionality Assessment (IP Section 03.01) (9 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 2 Division I emergency diesel generator slow start on January 3, 2022
- (2) Unit 2 Division I emergency diesel generator starting air compressor 'B' failure on January 18, 2022
- (3) Unit 2 Division I emergency diesel generator emergency start solenoid valve air leaks on January 20, 2022
- (4) Unit 1 safety relief valve elevated discharge temperature indications on February 1, 2022
- (5) Unit 1 emergency diesel generator 103 starting flywheel chipped tooth on February 15, 2022
- (6) Unit 1 containment spray raw water 122 rate set valve unable to operate on February 24, 2022
- (7) Unit 1 core spray topping pump 111 pump -bearing oil leak on March 2, 2022
- (8) Unit 2 Division III diesel under voltage relay failure to reset after testing on March 4, 2022
- (9) Unit 2 main steam isolation valve slow fast closure times on March 7, 2022
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) Perm anent Modification: ECP-21-000437, Unit 2 Division II Emergency Diesel Generator Governor Booster
- (2) Permanent Modification: ECP-21-000454, Unit 2 Digital Electro-Hydraulic Control (DEHC) Low Pass Filter Modification
- (3) Perm anent Modification: ECP-21-000088, Unit 2 PCIV [primary containment isolation valve] Supply Nitrogen Line Primary Containment Inboard Isolation Re-Design
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01) (5 Sam ples)
The inspectors evaluated the following post -maintenance testing activities to verify system operability and/or functionality:
- (1) Unit 2 Division I emergency diesel generator jacket water pump following replacement on February 23, 2022
- (2) Unit 2 'B' service water pump following a failure to start on March 2, 2022
- (3) Unit 2 Division II emergency diesel generator following governor oil bo oster installation on March 21, 2022
- (4) Unit 2 'B' residual heat removal system following heat exchanger inspectio n on March 25, 2022
- (5) Unit 2 drywell nitrogen supply system following solenoid operated valve replacement on March 28, 2022
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 N2R18 from March 6 to March 27, 2022.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality:
Surveillance Tests (other) (IP Section 03.01) (8 Samples)
- (1) N1-ST-M4A, Emergency Diesel Generator 102 and PB 102 Operability Test, on January 24, 2022
- (2) N2-OSP-CSL-Q@00 2, LPCS [low pressure core spray] Pump and Valve Operability and System Integrity Test, on February 28, 2022
- (3) N1-ST-Q1A, Core Spray 111 Pump, Valve and Shutdown Cooling Water Seal Check Va lve Operability Test, on March 1, 2022
- (4) N2-OSP-MSS-CS001, Main Steam Isolation Valve Operability Test, on March 7, 2022
- (5) N2-OSP-RHS-R001, Division II ECCS [emergency core cooling system] Functional Test, on March 9, 2022
- (6) N2-OSP-SLS-R001, Standby Liquid Control Manual Initiate Actuation and ASME XI Pressure Test, on March 16, 2022
- (7) N2-OSP-ADS-R002, ADS [automatic depressurization system] Functional Test and Remote Shutdown System Test, on March 19, 2022
- (8) N2-OSP-EGS-R001, Diesel Generator ECCS Start and Load Reject Division II, on March 20, 2022
Inservice Testing (IP Section 03.01) (2 Sam ples)
- (1) N2-OSP-I CS-Q@002, Reactor Core Isolation Cooling Pump and Valve Operability Test and System Integrity Test and ASME XI Functional Test and Analysis, on February 10, 2022
- (2) N2-ISP-RRC-R001, ARI [alternate rod insertion] Function of RRCS [redundant reactivity control system], on March 7, 2022
Containment Isolation Valve Testing (IP Section 03.01) (2 Samples)
- (1) N2-OSP-MSS-003, Unit 2 Main Steam Isolation Valve Leak Rate Test, on March 7, 2022
- (2) N2-OSP-MSS-004, Unit 2 Main Steam Isolation Valve Leak Rate Test (Reactor Vessel Head Removed), on March 25,
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 Constellation 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 Sam p les)
The inspectors observed/evaluated the following licensee process es for monitoring and controlling contamination and radioactive material:
- (1) Workers exiting the Unit 2 radiologically controlled area during refueling outage N2R18
- (2) Licensee surveys of contaminated equipment on the refuel floor during refueling outage N2R18
Radiological Hazards Control and Work Coverage (IP Section 03.04) (5 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Unit 2 outage refuel floor activities, low power range monitor exchange, and supporting activities
- (2) Unit 2 in-vessel inspection and supporting activities
- (3) Unit 2 refuel floor equipment decontamination activities
- (4) Unit 2 drywell feedwater nozzle inspection activities
- (5) Unit 2 safety relief valve maintenance in the drywell
High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (2 Samples)
The inspectors evaluated licensee controls of the following High Radiation Areas and Very High Radiation Areas:
- (1) High Radiation Area in the Unit 2 reactor building valve pit, 196' elevation
- (2) Locked Hig h Radiation Areas in the Unit 2 drywell
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 Sample (IP Section 02.01) (2 Sam ples)
- (1) Unit 1 (January 1, 2021 through December 31, 2021)
- (2) Unit 2 (January 1, 2021 through December 31, 2021)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02)
=
- (1) Unit 1 (January 1, 2021 through December 31, 2021)
- (2) Unit 2 (January 1, 2021 through December 31, 2021)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
- (1) Unit 1 (January 1, 2021 through December 31, 2021)
- (2) Unit 2 (January 1, 2021 through December 31, 2021)
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 issues:
- (1) IR 04436618 - Nine Mile Point Control Rod Blade Technical Evaluation
INSPECTION RESULTS
Observation: Nine Mile Point Control Rod Blade Techni cal Evaluation 71152A On November 23, 2020, General Electric Hitachi (GEH) released Safety Communication 20 -
06 (SC 20- 06) Revision 0, describing the discovery that Boron 10 (B -10) depletion had been underpredicted in the top 6-inch node of the neutron absorber section of control rod blades currently in use in boiling water reactors in the United States. Nine Mile Point Nuclear Station was on the list of plants affected by this issue.
In Revision 0 of the Safety Communication, the population of control rod blades affected was restricted to Original Equipment Manufacturer (OEM) series D-100 control rod blades, still in operation from the 1970s and 1980s. On February 26, 2021, GEH issued Revision 1 to SC 20- 06, in which it was determined that the population of control rod blades affected by the underprediction of B -10 in the tip segm ent included all control rod blades that did not contain Hafnium (Hf) in the tips, not just OEM D-100 control rod blades. This greatly increased the population of control rod blades affected by this issue.
Control rod blades deplete the B -10 isotope as they absorb neutrons when they are inserted into an operating reactor core. Control rod blades are typically fully withdrawn from the core when the reactor is at full power. However, the tips of the control rod blades still experience some thermal neutron flux. The B -10 depletion in the top node of the control rod blade is accounted for in engineering analysis by adding a "tip adder" factor in the B -10 depletion calculation for each control rod blade. Safety Communication SC 20- 06 stated that the tip adder factor was much larger than previously calculated, for control rod blades without Hf in the tips. General Electric manufactured control rod blades with Hf tips for a period of time in the 1980s and 1990s that are excluded from the tip adder issue described in SC 20 -06. All other control rod blade models, including the OEM blades, are affected by the tip adder issue described in SC 20- 06. The higher depletion values in the control rod blade tips for multiple control rod blade types may cause the control rod blade to exceed its effective neutron absorbing capability before the end of the operating fuel cycle. This has the potential to decrease the overall shutdown margin (SDM), which is the ability of all the control rods, except for the most reactive control r od, to shut down the reactor core in all anticipated normal and accident scenarios.
In Revision 1 to SC 20- 06, GEH recommended all customers ensure that adequate SDM was available in the current operating fuel cycle until the end of the fuel cycle, once the am ount of B-10 depletion was determined in all control rod blade tips. GEH recommended this be done for all General Electric control rod blades that did not contain Hf tips, as well as alternate vendor control rod blades. GEH also recommended all customers assess the impact of control rod blades exceeding their nuclear end of life (NEOL) criteria - the ability of the control rod blade to effectively absorb neutrons - fo r future fuel cycles beyond the current operating fuel cycle.
Following issuance of GEH SC 20-06, Revision 1, Nine Mile Point Unit 1 entered a refueling outage in March 2021. A control rod blade depletion engineering evaluation was performed, taking into account the new tip adder calculation in SC 20- 06. It was determined that two control rod blades that had been located in higher power locations in the previous fuel cycle would be shuffled to two lower power locations in the core for the current Unit 1 operating fuel cycle. No control rod blades in the previous or current fuel cycle have exceeded their NEOL criteria, thus requiring replacement.
In July 2021, Exelon Nuclear Fuels validated that adequate SDM existed for the remainder of the current fuel cycle for Nine Mile Point Unit 2, and that thermal limit margins were not im pa cted. One control rod blade on the core periphery was predicted to exceed its NEOL criteria due to the tip adder issue described in SC 20- 06, before the end of the Unit 2 fuel cycle in March 2022. This blade, along with six other control rod blades, are scheduled for replacement in the upcoming refueling outage. In September 2021, General Electric Global Nuclear Fuels completed an analysis to confirm that SDM would be maintained above the technical specification limit for the remainder of the Unit 2 fuel cycle. Exelon procedure NF -
AB-135-1410, "BWR Control Blade Lifetime Management," contains guidance for calculating B-10 depletion in all models of control rod blades currently in operation in the nuclear fleet.
This procedure will require an update to incorpora te the new guidance from SC 20- 06, for estimating the tip adder factor for non-Hf tipped control rod blades. Issue Report 04242262 is tracking completion of this procedural update.
The inspectors interviewed engineering staff from Exelon Nuclear F uels and Nine Mile Point Reactor Engineering to discuss corrective actions taken in response to GEH SC 20-06, and reviewed shutdown margin and control rod blade depletion engineering evaluations. The inspectors also reviewed the corrective actions taken and planned for responding to GEH SC 20- 06 entered into the corrective action system. No performance deficiencies were identified.
Corrective Action References: 04242262, 04386300,
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 21, 2022, the inspector s presented the integrated inspection results to Mr. Peter Orphanos, Site Vice Preside n t, and other members of the licensee staff.
- On February 8, 2022, the inspectors presented the control rod blade technical evaluation problem identification and resolution inspection results to Philip Nichols, Manager Reactor Engineering, and other members of the licensee staff.
- On March 17, 2022, the inspector s presented the Unit 2 i nservice inspection results to Mr. Pe ter Orphanos, Site Vice President, and other members of the licensee staff.
- On March 17, 2022, the inspector s presented the radiological hazard assessment and exposure controls inspection results to Mr. Peter Orphanos, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.01 Procedures N1-OP-64 Meteorological Monitoring 02100
N2-OP-102 Meteorological Monitoring 02800
OP-AA-108-111-Severe Weather and Natural Disaster Guidelines 24
1001
71111.04 Drawings C-18022-C Piping & Instrumentation Diagram, Reactor Building Closed 55
Loop Cooling System
PID-031A, B, E Piping & Instrumentation Diagram Residual Heat Removal 27
System
PID-31G Piping & Instrumentation Diagram Residual Heat Removal 15
PID-35A Piping & Instrumentation Diagram Reactor Core Isolation 17
Cooling
PID-35B Piping & Instrumentation Diagram Reactor Core Isolation 13
Cooling
PID-38B Piping & Instrumentation Diagram Fuel Pool Cooling & 15
Cleanup
PID-38C Piping & Instrumentation Diagram Fuel Pool Cooling & 17
Cleanup
Procedures N1-OP-11 Reactor Building Closed Loop Cooling System 03400
N2-OP-100A Standby Diesel Generators 03200
N2-OP-100A-Standby Diesel Generators - LINEUPS 00500
LINEUPS
N2-OP-31 Residual Heat Removal System 03700
N2-OP-31-Residual Heat Removal System 003
LINEUPS
N2-OP-38 Spent Fuel Cooling and Cleanup System 2700
71111.05 Corrective Action 04477281
Docum ents
Drawings B-40143-C Fire Zones Reactor Building - Fl. El. 261' Turbine Building - 10
Fl. El. 261' Fire Rated Walls and Slabs
Fire Plans N2-FPI-PFP-0201 Unit 2 Pre-Fire Plans 06
71111.06 Corrective Action 04416206
Docum ents 04477927
Inspection Type Designation Description or Title Revision or
Procedure Date
04661043
71111.07A Corrective Action 04484882
Docum ents 04486864
Procedures ER-AA-340-1002 Service Water Heat Exchanger Inspection Guide 11
S-TDP-REL-0102 Service Water Heat Exchanger and Component Inspection 03
Guide
Work Orders C93738267
71111.08G Corrective Action 04327298
Docum ents
Corrective Action 04484400
Docum ents
Resulting from
Inspection
Engineering ECP-20-000224 Recirculation Inlet Nozzle DMW No. 2RPV-KB11 (N2J) Flaw 03/25/2020
Evaluations Evaluation
Miscellaneous ER-NM-330-2001 ISI Program Plan Fourth Ten-Year Inspection Interval Revision 3
ER-NM-330-2004 Risk Informed Inservice Inspection Program Fourth Ten-Revision 0
Year Inspection Interval
Procedures ER-AA-335-018 Visual Examination of ASME IWE Class MC and Metallic Revision 15
Liners of Class CC Components
ER-AA-335-030 Ultrasonic Examination of Ferritic Piping Welds Revision 5
ER-AA-335-1000 Nondestructive Examination (NDE) Revision 16
GEH-UT-254 Automated Phased Array Ultrasonic Examination of Version 1
Dissimilar Metal Welds with the TOPAZ
WPS 8-8-GTSM Welding Procedure Specification Record for Manual GTAW Revision 7
and SMAW of P-Number 8 to P-Number 8 Base Metal
71111.11Q Procedures N1-ST-W1 Control Rod Exercising Operability Test 02300
N2-OP-101C Plant Shutdown 04100
N2-OP-29 Reactor Recirculation System 03400
N2-OP-31 Residual Heat Remova l S yst em 03700
71111.12 Procedures N2-MSP-EGS-Diesel Generator Inspection Division 1 and 2 025
R001
Work Orders C938144677
71111.13 Corrective Action 04481795
Inspection Type Designation Description or Title Revision or
Procedure Date
Docum ents
Procedures N2-OP-19-Instrument and Service Air Systems 8
Lineups
N2-OP-70 Station Electrical Feed and 115KV Switchyard 02800
N2-PM-082 RPV [reactor pressure vessel] Flood-Up/Draindown 01900
OP-NM-108-117 Protected Equipm ent Program at Nine Mile Point 5
OP-NM-108-117 Protected Equipm ent Program at Nine Mile Point 00500
OU-NM-103-101 Shutdown Safety Management Program 0700
71111.15 Calculations 002N3714 Nine Mile Point Nuclear Station Unit 1 TRACG-LOCA 0
Analysis for GNF2 Fuel
Corrective Action 04061889
Docum ents 04470659
04475343
04476776
04481649
04482642
04483059
04488015
Drawings 0001040209048 Control Diagram Shutdown System 13.00
Engineering ECP-22-000147 Technical Evaluation for MSIV Failed Stroke Time Extent of 0000
Changes Condition
Miscellaneous NEI 06-09-A Risk-Informed Technical Specifications Initiative 4b: Risk-0
Managed Technical Specifications (RMTS) Guidelines
Purchase Order Service, Repair, Refurbishment of MSIV Actuator Air Pack 2
00802656
RICT Record fo r Failed MSIV 7D RPS Testing 04/06/2022
March 28, 2022
Procedures N1-OP-1 Nuclear Steam Supply System 07700
N2-OP-100A Standby Diesel Generators 03100
N2-OSP-EGS-Diesel Generator and Diesel Air Start Valve Operability Test-024T1
M@001 Division I and II
OP-AA-108-118 Risk Informed Completion Time 2
71111.18 Corrective Action 02547530
Inspection Type Designation Description or Title Revision or
Procedure Date
Docum ents 02689624
04359704
04365824
04428910
04482114
Engineering ECP-21-000088 PCIV Supply Nitrogen Line Primary Containment Inboard 0000
Changes Isolation Re-Design
ECP-21-000437 EDG Governor Booster 0000
ECP-21-000454 Digital Electro-Hydraulic Control (DEHC) Low Pass Filter 0
Modification
Miscellaneous FAT/SAT Testing for DEHC Low Pass Filter Mod 0000
Procedures N2-OSP-EGS-Diesel Generator and Diesel Air Start Valve Operability Test 024T1
M@0001 - Division I and II
Work Orders C93782709
C93809087
C93813137
71111.19 Corrective Action 04480094
Docum ents 04481795
04484882
04486059
Corrective Action 04486063
Docum ents
Resulting from
Inspection
Procedures GAP-HSC-09 System Aging Inspection and Cleanliness Controls 02000
N2-MSP-EGS-Diesel Generator Inspection Division I and II 02400
R001
N2-OSP-EGS-Diesel Generator and Diesel Air Start Valve Operability Test 02400.01
M@0001 - Division I and II
N2-OSP-RHS-RHR System Loop B Pump and Valve Operability Test, 01500
Q@005 System Integrity Test and ASME XI Pressure Test
N2-OSP-RHS-Division II ECCS Functional Test 00900
R001
Inspection Type Designation Description or Title Revision or
Procedure Date
S-EPM-GEN-004 Insulation of Power, Control, and Instrument Cable 00700
Connections
Work Orders C93738267
C93782709
C93814677
C93825512
71111.20 Corrective Action 04482986
Docum ents 04482995
04483785
04484026
04484553
04485270
Corrective Action 04484400
Docum ents 04486063
Resulting from
Inspection
Miscellaneous NM2C19-SU Reactivity Maneuver Plan 0
Procedures LS-AA-119 Fatigue Management and Work Hour L im it s 15
N2-FHP-13.3 Core Shuffle 01200
N2-OP-101A Plant Start-up 05400
N2-OP-38 Spent Fuel Cooling and Cleanup System 02700
N2-OSP-NMS-Source Range Monitor Check During Core Offload/Reload 00201
@002
OP-AA-109-101 Personnel and Equipment Tagout Process 16
OP-AA-300 Reactivity Management 14
OP-AA-300-1520 Reactivity Management - Fuel Handling, Storage and 7
Refueling
OU-NM-103-101 Shutdown Safety Management Program 00700
OU-NM-4001 Refueling Operations 00800
71111.22 Corrective Action 04481649
Docum ents 04482027
04483059
04483200
Inspection Type Designation Description or Title Revision or
Procedure Date
04483399
04485322
Miscellaneous Technical MSIV Leak Rate Tests Approved
Evaluation N2R18 03/17/2022
Procedures N1-ST-M4A Emergency Diesel Generator 102 and PB 102 Operability 03000
Test
N1-ST-Q1A CS 111 Pump, Valve and SDC Water Seal Check Valve 02200
Operability Test
N2-ISP-RRC-ARI [alternate rod insertion] function of RRCS [redundant 00700
R001 reactivity control system]
N2-OSP-ADS-ADS System Functional Test and Remote Shutdown System 009
R002 Test
N2-OSP-CSL-LPCS Pump and Valve Operability and System Integrity 01500
Q@002 Test
N2-OSP-EGS-Diesel Generator ECCS Start and Load Reject Division II 9
R001
N2-OSP-I CS-RCIC Pump and Valve Operability Test and System Integrity 01600
Q@002 Test and ASME XI Functional Test and Analysis
N2-OSP-MSS-Main Steam Isolation Valve Leak Rate Test 00300
003
N2-OSP-MSS-Main Steam Isolation Valve Leak Rate Test (Reactor Vessel 00200
004 Head Rem oved)
N2-OSP-MSS-Main Steam Isolation Valve Operability Test 01000 and
CS001 01100
N2-OSP-SLS-Standby Liquid Control Manual Initiate Actuation and ASME 01100
R001 XI Pressure Test
71151 Procedures NEI 99-02 Regulatory Assessment Performance Indicator Guideline 7
71152A Corrective Action 04242262
Docum ents 04386300
04436618
Engineering ECP-20-000291 Control Blade Replacement Strategy for Nine Mile Point Unit 0
Evaluations 1 Cycle 25 (Reload 26)
ECP-21-000248 Control Blade Replacement Strategy for Nine Mile Point Unit 0
(Reload 18)
Inspection Type Designation Description or Title Revision or
Procedure Date
ECP-21-000380 SC 20-06: Nine Mile Point Unit 2 Cycle 18 SDM and 0
MSBWP Evaluation
Miscellaneous General Electric Impact of Ex-core Flux on Control Rod Lifetime L im it s 02/26/2021
Hitachi Safety
Communication
20-06, Rev. 1
Procedures NF-AB-130-3690 Maximum Subcritical Banked Withdrawal Position 10
NF-AB-135-1410 BWR Control Blade Lifetime Management 16
19