BVY 17-041, Supplement to Exemption Request from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 to Support the Dry Fuel Loading Campaign

Supplement to Exemption Request from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 to Support the Dry Fuel Loading Campaign

ML17346A685
Person / Time
Site:	Vermont Yankee, Holtec
Issue date:	12/07/2017
From:	Chappell C Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
BVY 17-041, CAC L24979, EPID L-2017-LLE-0005
Download: ML17346A685 (3)

Text

• ~Entergy Entergy Nuclear Operations, Inc. Vermont Yankee 320 Governor Hunt Rd. Vernon, VT 05354 802-257-7711 Coley C. Chappell Manager, Design and Programs 10 CFR 72.7 BVY 17-041 December 7, 2017 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Supplement to Exemption Request from Certain Requirements of 1 O CFR 72.212 and 1 O CFR 72.214 to Support the Dry Fuel Loading Campaign (EPID No. L-2017-LLE-0005)

Vermont Yankee Nuclear Power Station License No. DPR-28 Docket Nos. 50-271, 72-59 and 72-1014

REFERENCES:

1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Exemption Request from Certain Requirements of 1 o CFR 72.212 and 1 o CFR 72.214 to Support the Dry Fuel Loading Campaign," BVY 17-006, dated May 16, 2017 (ML17142A358}

2. Letter, USNRC to Holtec International, "Certificate of Compliance No. 1014, Amendment No. 1 O for the HI-STORM 100 Cask System (CAC No. L24979)," dated May 25, 2016 (ML16144A177) 3. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Response to Request for Additional Information Related to Exemption Request from Certain Requirements of 1 O CFR 72.212 and 1 O CFR 72.214 to Support the Dry Fuel Loading Campaign (CAC No. L25219}," BVY 17-031, dated September 7, 2017 (ML17255A236)

Dear Sir or Madam:

By letter dated May 16, 2017 (Reference 1 ), Entergy Nuclear Operations, Inc. (ENO) submitted a request for exemption from certain requirements of 10 CFR 72.212 and 72.214 for Vermont Yankee Nuclear Power Station (VY). These regulations require, in part, compliance with the terms and conditions of the Holtec International (Holtec) Cask System Certificate of Compliance (CoC) for spent fuel storage at the VY's independent spent fuel storage installation.

Specifically, the requested exemption would allow the loading of selected fuel assemblies with shorter cooling times and higher heat *loads than those specified in CoC Amendment 1 o (Reference

2) into MPC-68M multi-purpose canisters (MPCs), as well as an optional loading pattern for the MPC-68M.

BVY 17-041 / Page 2 of 3 In Reference 3 , VY provided a response to a Request for Additional Information (RAI) from the U.S. Nuclear Regulatory Commission (NRC), supplementing the information in Reference

1. Subsequently , it has been identified that an additional supplement to Reference 1 is warranted. Specifically , ENO has determined that the proposed peak " Fuel Cladding" temperature (PCT) in the revised Table 4.111.7 , as prov i ded in Attachment 2 of Reference 1 , shou l d be revised based upon the following: The analysis of record in CoC Amendment 1 O is based on a total can i ster heat load of 36.9 kW and a maximum per storage location decay heat of 71 O watts (reference Table 2.1.30 of the Cask Final Safety Analysis Report). In comparison , for VY's site specific cask loading plan , the highest heat load cask planned for loading is approximately 24.5 kW and the highest decay heat in any cell location is 912 watts. Although the planned per assembly decay heat is slightly higher than that in the approved Amendment 10 pattern , the total canister heat l oad , wh i ch has a first orde r effect on the PCT rise during a postulated duct blockage event , is substantially lower. Based on this comparison , i t is concluded that the PCT for canisters with VY's site specific heat loads will be lower than the temperature of 722°F in Amendment 10 , providing reasonable assu r ance that the PCT will r emain less than 752°F. Accordingly , the p r oposed changes to Table 4.111.7 as provided in VY's exemption request (Reference

1) is revised as shown below , reverting the Fuel Cladding temperature to the 722°F value reported in CoC Amendment 1 O and adding a new Note 2 (all other previously proposed changes to this table are not affected by this supplement):

Table 4.111.7: Maximum Temperatures and Pressures Under 32-Hour 100% Air Inlets Blockage Accident Component Temperature

(°F) Fuel Cladding 722 Note.! Fuel Basket 818700 Basket Shims 702e2e MPC Shell 639&74 MPC Lid Noce l 599Ma Overpack Inner Shell 531~ Body Concrete (Local Temperature) 5254ae Lid Concrete (Local Temperature) 447~ Pressure (osig) MPC +44-:e116.3 Note 1: Maximum thru thickness section average temperature reported.

Note 2: The fuel cladding maximum temperature reported in this table is from calculations based on a canister heat load limit of 36.9 kW and a maximum per storage decay heat of 710 watts. Figure 2.111.1 optional loading pattern for the MPC-68M allows for higher canister and per cell allowable heat loads , which could result in a higher maximum cladding temperature. However , for any cask planned for loading at VY , the highest heat load is approximately 24.5 kW and the highest decay heat in an y cell location is 912 watts. Although this highest per assembly decay heat is higher than the maximum used in the prev i ous calculation , the total canister heat load, which has a first order effect on the cladding temperature rise during this accident , is substantially lowe r. B ased upon this comparison , it is concluded that the maximum fuel cladding temperature for any cask planned for loading at VY is lower than 722°F.

BVY 17-041 / Page 3 of 3 This letter contains no new regulatory commitments.

Should you have any questions concerning this letter, please contact me at (802) 451-3374.

Sincerely, cc: Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. Jack D. Parrott, Sr. Project Manager Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Mail Stop T-8F5

• Washington, DC 20555 Ms. Yen-Ju Chen, Sr. Project Manager Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Mail Stop T-4B-34 Washington, DC 20555 Ms. June Tierney, Commissioner Vermont Department of Public Service 112 State Street -Drawer 20 Montpelier, Vermont 05602-2601