ML23041A113

Yankee Atomic Electric Company - Report of 10 CFR 72.48 Changes, Tests, and Experiments

ML23041A113
Person / Time
Site:	Yankee Rowe
Issue date:	01/05/2023
From:	Macdonald J Yankee Atomic Electric Co
To:	Office of Nuclear Material Safety and Safeguards
References
BYR 2023-002
Download: ML23041A113 (1)

Text

YANKEE ATOMIC ELECTRIC COMPANY 49 Yankee Road, Rowe, Massachusetts 01367 January 05, 2023 BYR 2023-002 10 CFR 72.4 and 10 CFR 72.48 ATIN: Document Control Desk Director, Division of Fuel Management, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission Washington, DC 20555 - 0001 Yankee Atomic Electric Company

• Yankee Rowefuclependent Spent Fuel Storage Iristallation

• NRC License No. DPR-3 (NRC Docket No.50-029) 7Z- D3 /

Subject:

Report of 10 CFR 72.48 Changes, Tests, and Experiments In accordance with 10 CFR 72.48(d)(2), Yankee Atomic Electric Company is required to submit to the NRC a brief description of any changes, tests or experiments made pursuant to 10 CFR 72.48( c), including a summary of the evaluation of each. This report covers the pe1iod from January 1, 2021 through December 31, 2022. During that time period, two changes, tests, or experiments were made pursuant to paragraph (c) of 10 CFR 72.48. The following are summaries of those evaluations:

72.48 Evaluation# 22-01 The performance of a baseline inspection on spent fuel cask Vertical Concrete Cask (VCC) 14 /

• Transportable Storage Canister (TSC) 14, including use of scaffolding and insertion of an inspection robot, is in accordance with the licensing basis. This inspection does not require any specific changes to the design, procedures utilized to conduct operations, or any evaluation methodology used in establishing the design basis of the Yankee Atomic Independent Spent Fuel Storage Installation (ISFSI).

Procedures AM-1 and AM-2 are aging management implementation procedures modeled on the Certificate of Compliance (C of C) renewal application Aging Management Procedures (AMPs),

not yet approved by the NRC, for TSC external and VCC internal inspections. These procedures will not direct the baseline inspections, but will be used as a dry run for future aging management inspections. AM-8 provides the requirements and criteria for the various visual inspections used for aging management examinations including eye exam requirements. The only direction these procedures will provide are the eye exam requirements for exterior visual exams of the VCC and ISFSI pad. These procedures do not require changes to the design procedures utilized to conduct operations, or any evaluation methodology used in establishing the design basis of the Yankee Atomic ISFSL

Yankee Atomic Electric Company BYR 2023-002/January 05, 2023/Page 2 The YNPS ISFSI 10 CFR 72.212 Evaluation Report was revised to describe the inspection and new aging management procedures described above.

The inspection did not affect thennal, structural tip over, structural missile impact or shielding accident analyses. The insertion of a robot into the VCC vents or annulus is less than 50%

blockage and does not impact the heat removal function of the VCC. During the inspection, the Concrete Cask Heat Removal System for VCC14/TSC14 remained operable in accordance with NAC-MPC Technical Specification (TS) A 3.1.6. Combustible loading in the Protected Area is controlled per procedure FP-5, therefore a fire is not considered credible.

Based on the above, the activity did not:

• Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the NAC-MPC FSAR;

• Result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC impo1tant to safety previously evaluated in the NAC-MPC FSAR;

• Result in more than a minimal increase in the consequences of an accident previously evaluated in the NAC-MPC FSAR;

• Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the NAC-MPC FSAR;

• Create a possibility for an accident of a different type than any previously evaluated in the NAC-MPC FSAR;

• Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the NAC-MPC FSAR;

• Result in a design basis limit for a fission product barrier as described in the NAC-MPC FSAR being exceeded or altered; and

• Result in a departure from a method of evaluation described in the NAC-MPC FSAR used in establishing the design bases or in the safety analyses.

72.48 Evaluation # 22-02 The performance of a robot recovery on spent fuel cask VCC14/TSC14, including use of a recovery robot with alligator retrieval gripper, two guide tubes, a borescope camera, a cable redirector, rope and the need for personnel to insert their arm into the outlet vent, is in accordance with the licensing basis. This does not require any specific changes to the design, procedures utilized to conduct operations, or any evaluation methodology used in establishing the design basis of the Yankee Atomic ISFSI.

The temporary modifications to VCC 14 do not affect thermal, structural tip over, structural missile impact or shielding accident analyses. The temporary insertion of the recovery equipment and a recovery pers01mel's arm into the VCC vents or aimulus is less than 50% blockage and does not impact the heat removal function of the VCC. During the recovery effort, the Concrete Cask Heat Removal System for VCC14/TSC14 will remain operable in accordance with NAC-MPC TS A 3.1.6.

Yankee Atomic Electric Company BYR 2023-002/January 05, 2023/Page 3 Based on the above, the activity did not:

• Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the NAC-MPC FSAR;

• Result in more than a minimal increase in the likelihood of occmrence of a malfunction of an SSC impo1tant to safety previously evaluated in the NAC-MPC FSAR;

• Result in more than a minimal increase in the consequences of an accident previously evaluated in the NAC-MPC FSAR;

• Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the NAC-MPC FSAR; Create a possibility for an accident of a different type than any previously evaluated in the NAC-MPC FSAR;

• Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the NAC-MPC FSAR;

• Result in a design basis limit for a fission product bani er as described in the NAC-MPC FSAR being exceeded or altered; and

• Result in a departure from a method of evaluation described in the NAC-MPC FSAR used in establishing the design bases or in the safety analyses.

This letter contains no regulatory commitments.

If you have any questions regarding this submittal, please do not hesitate to contact me at (413) 424-5261 ext. 304.

Respectful~ _

¥:~ v1 h1ru.<,:~cl Jom/Macdonald ISFSI Manager cc: R. Lorson, NRC Region I Administrator A. Dimitriadis, Chief, Decommissioning Branch, NRC, Region 1 Y. Diaz-Sanabria, Chief, Division of Fuel Management, Storage and Transportation Licensing Branch J. Viveiros, Senior Nuclear Planner, MEMA J. Cope-Flanagan, Assistant General Counsel, MDPU J. Dinerstein, State of Massachusetts Office of the Attorney General