ML22290A048

Certificate of Compliance No. 9270, Revision 6

ML22290A048
Person / Time
Site:	07109270
Issue date:	10/24/2022
From:	Yoira Diaz-Sanabria Storage and Transportation Licensing Branch
To:	NAC International
NDEVASER NMSS/DFM/STLB 3014155196
Shared Package
ML22290A046	List: ML22290A048 ML22290A049
References
CoC No. 9270, EPID L-2022-RNW-0014
Download: ML22290A048 (19)

Text

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 1 OF 19

2. PREAMBLE

a. This certificate is issued to certify that the package (packaging and contents) described in Item 5 below meets the applicable safety standards set forth in Title 10, Code of Federal Regulations, Part 71, Packaging and Transportation of Radioactive Material.

b. This certificate does not relieve the consignor from compliance with any requirement of the regulations of the U.S. Department of Transportation or other applicable regulatory agencies, including the government of any country through or into which the package will be transported.

3. THIS CERTIFICATE IS ISSUED ON THE BASIS OF A SAFETY ANALYSIS REPORT OF THE PACKAGE DESIGN OR APPLICATION

a. ISSUED TO (Name and Address) b. TITLE AND IDENTIFICATION OF REPORT OR APPLICATION NAC International, Inc. NAC International, Inc. application dated 3930 East Jones Bridge Rd. August 3, 2016, as supplemented.

Norcross, Georgia 30092

4. CONDITIONS This certificate is conditional upon fulfilling the requirements of 10 CFR Part 71, as applicable, and the conditions specified below.

5.

(a) Packaging (1) Model No.: UMS Universal Transport Cask Package (2)

Description:

For descriptive purposes, all dimensions are approximate nominal values. Actual dimensions with tolerances are as indicated on the Drawings.

The UMS is a canister-based system for the storage and transportation of spent nuclear fuel. The transportation component of the UMS system, designated the Universal Transport System, consists of a Universal Transport cask body with a closure lid and energy-absorbing impact limiters loaded with a Transportable Storage Canister (TSC) containing either spent Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) nuclear fuel, or Maine Yankee site specific contents including Greater than Class C (GTCC) waste.

The NAC-UMS is designed to transport up to 24 intact PWR spent fuel assemblies, 56 intact BWR spent fuel assemblies, GTCC waste, or site specific spent nuclear fuel with associated component hardware. Based on the length of the fuel assemblies, PWR fuels are grouped into three classes (Classes 1 through 3), and BWR fuels are grouped into two classes (Classes 4 and 5). Class 1 and 2 PWR fuel assemblies include non-fuel-bearing inserts (components which include thimble plugs and burnable poison rods installed in the guide tubes). Class 4 and 5 BWR fuel assemblies include the zirconium alloy channels. The loading of site specific fuels that include control component hardware may require the use of a TSC that is longer than if the hardware were excluded. The spent fuel is loaded into a TSC which contains a stainless steel grid work referred to as a basket.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 2 OF 19 5.(a)(2) Description (Continued)

The cask body of the UMS is a right-circular cylinder of multi wall construction which consists of 304 stainless steel inner and outer shells separated by lead gamma radiation shielding which is poured in place. The inner and outer shells are welded to a 304 stainless steel top forging which mates to the cask lid. The inner shell is also welded to a 304 stainless steel bottom forging and the outer shell is welded to the bottom plate. The cask bottom consists of the bottom forging and bottom plate with neutron shield material sandwiched between them. Layers of 4.5 inches thick 304 stainless steel ring and two 0.75 inch stainless steel disks are located at the bottom lead annulus between the bottom forging and the outer shell.

Neutron shield material is also placed in an annulus that surrounds the cask outer shell along the length of the cask cavity and is enclosed by a stainless steel shell with top and bottom plates. The neutron shield material is a solid synthetic polymer (NS-4-FR). Twenty-four bonded copper and Type 304 stainless steel fins are located in the radial neutron shield to enhance the heat rejection capability of the cask and to support the neutron shield shell and end plates.

The containment boundary of the UMS consists of the inner shell; bottom forging; top forging; cask lid and lid inner O-ring; vent port cover plate and vent port cover plate inner O-ring; and drain port cover plate and drain port cover plate inner O-ring.

There are five TSCs of different lengths, each to accommodate a different class of PWR or BWR fuel assembly. Each TSC has an outside diameter of about 67 inches and the lengths vary from about 175 to 192 inches long. The TSC assembly consists of a right circular cylindrical shell with a welded bottom plate, a fuel basket, a shield lid, two penetration port covers, and a structural lid. The TSC contains the basket and fuel assemblies or GTCC waste. Spacers are placed below each Class 1, 2, 4 or 5 canisters to locate and support the canister in the cask cavity.

The spacers are free standing structures that are confined in place by the bottom of the canister and the cask bottom inner surface. The spacer(s) ensure that the canister lid is laterally supported by the cask top forging when the cask is horizontal and minimizes axial movement of the canister. Each Class 1 PWR canister is positioned by a stainless steel spacer that is 16.75 inches in length. Each Class 2 PWR canister is positioned by a stainless steel spacer that is 7.65 inches in length. No spacers are used with the Class 3 PWR canister. The Class 4 BWR canister is located by four 1.5 inch aluminum spacers and the Class 5 BWR canister is located with a 1.5 inch aluminum spacer.

The spent fuel basket design uses a series of high strength stainless steel PWR or carbon steel BWR support disks to support the fuel assemblies in stainless steel tubes. The PWR fuel tubes contain neutron absorber on all four sides of the tubes. Three types of fuel tubes are designed to contain the BWR fuel: (1) tubes containing neutron absorber on two sides of the tubes; (2) tubes containing neutron absorber on one side; and (3) tubes containing no neutron absorber. Aluminum heat transfer disks are provided in both the PWR and BWR fuel baskets to enhance thermal performance of the basket. The heat transfer disks are supported by stainless steel tie rods and split spacers that maintain the basket assembly

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 3 OF 19 configuration.

5.(a)(2) Description (Continued)

The GTCC waste canister is essentially identical to the Class 1 TSC, except for the placement of lifting lugs and the placement of a key way within the canister. The GTCC basket is constructed of Type 304 stainless steel and consists primarily of a cylinder with a 3-inch thick wall closed at the bottom end with a 3-inch thick plate. The cylinder is centered in the GTCC waste canister by 14 Type 304 stainless steel support plates along its length.

A 3-inch thick 304 stainless steel separator fixture divides the cylinder into two vertically stacked components, each 77 inches deep with a diameter of 47.8 inches.

The package has impact limiters at each end of the cask body. The impact limiters consist of a combination of redwood and balsa wood encased in Type 304 stainless steel. The impact limiters limit the g-loads acting on the cask during a transport drop load condition due to crushing of the redwood and balsa wood. The upper and lower impact limiters are bolted to the cask body by 16 equally spaced attachment rods with nuts.

The approximate dimensions and weights of the package are as follows:

Overall length (with impact limiters, in) 273.3 Overall length (without impact limiters, in) 209.3 Impact Limiter Outside diameter (in) 124.0 Outside diameter (without impact limiters, in) 92.9 Cavity diameter (in) 67.6 Cavity length (in) 192.5 Cask lid thickness (in) 6.5 Bottom thickness (in) 10.3 Inner shell thickness (in) 2.0 Outer shell thickness (in) 2.75 Gamma shield thickness (in) 2.75 Radial neutron shield thickness (in) 4.50 Transportable Storage Canister Shell thickness (in) 0.625 Shell bottom (in) 1.75 Shield lid thickness (in) 7 Structural lid thickness (in) 3 Outer diameter (in) 67 Internal cavity diameter (in) 65.8 Internal fuel cavity length (in), depending on class 163-180 Overall length (in), depending on class 175-192 Fuel Basket Basket assembly length (in), depending on class 162-180 Basket assembly diameter (in) 65.5

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 4 OF 19 Number of support disks, depending on class 30-41 5.(a)(2) Description (Continued)

Number of heat transfer disks, depending on class 17-33 Total weight (pounds) including cask, basket, impact limiters, fuel, canister with lids, cask lid, and spacers for each fuel class is approximately:

Class 1 (PWR) 251,000 Class 2 (PWR) 252,000 Class 3 (PWR) 249,000 Class 4 (BWR) 256,000 Class 5 (BWR) 255,000 5.(a)(3) Drawings The package is constructed and assembled in accordance with NAC drawings:

790-209, Rev. 1 790-210, Rev. 1 790-500, Rev. 4 790-501, Rev. 3 790-502, Rev. 7 790-503, Rev. 3 790-504, Rev. 2 790-505, Rev. 2 790-508, Rev. 2 790-509, Rev. 3 790-516, Rev. 3 790-519, Rev. 2 790-520, Rev. 2 790-570, Rev. 4 790-571, Rev. 3 790-572, Rev. 4 790-573, Rev. 7 790-574, Rev. 3 790-575, Rev. 10 790-581, Rev. 9 790-582, Rev. 12 790-583, Rev. 8 790-584, Rev. 19 790-585, Rev. 19 790-587, Rev. 1 790-591, Rev. 6 790-592, Rev. 8 790-593, Rev. 7 790-594, Rev. 2 790-595, Rev. 10 790-605, Rev. 11 790-611, Rev. 6 790-612, Rev. 9 412-501, Rev. 4 412-502, Rev. 6 5.(b) Contents (1) Type and Form of Material The package is designed to transport four types of contents as listed below:

i. 24 intact irradiated PWR fuel assemblies within a TSC; ii. 56 intact irradiated BWR fuel assemblies within a TSC; iii. 24 Intact and Damaged PWR assemblies, and Fuel Debris from Maine Yankee within a TSC; or iv. GTCC waste from Maine Yankee within a TSC.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 5 OF 19 5.(b) Contents (Continued)

Each type of package contents is described in detail below.

(i) Intact PWR assemblies The package is designed to transport 24 irradiated intact PWR fuel assemblies within the TSC. An intact fuel assembly is a spent nuclear fuel assembly without known or suspected cladding defects greater than pinhole leaks or hairline cracks.

An empty fuel rod position must be filled with a solid filler rod, fabricated from either zirconium alloy or Type 304 stainless steel, which displaces an equal or greater volume than that occupied by a fuel rod.

The fuel assemblies consist of uranium dioxide pellets with zirconium alloy type cladding. Prior to irradiation, the fuel assemblies must be within the dimensions and specifications of Table 5.(b)(1)(i)-1 below. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(i)-2 below. PWR fuel assemblies may include standard inserts such as guide tube thimble plugs and burnable poison rods.

The minimum and maximum allowable assembly average enrichment for loading is 1.9 wt% 235U and 4.2 wt% 235U respectively. Unenriched fuel assemblies are not authorized for loading into the TSC. The maximum burn up of the spent fuel assemblies is 45,000 MWD/GTU and the minimum cool time is 5 years. The maximum weight of UO2 is 11.53 MTU per cask.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 6 OF 19 Table 5.(b)(1)(i)-1, Intact PWR Fuel Assembly Characteristics TSC Vendor2 Array Max. Max. Max. Max No of Max Min Min Max Max Min Guide Class1 Length Width Assembly MTU Fuel Pitch Rod Clad Pellet Active Tube (in) (in) Weight Rods (in) Dia (in) Thick Dia (in) Length Thickness (in) (in) (in) 1 CE 14x14 157.3 8.11 1292 0.404 1764 0.590 0.438 0.024 0.380 137.0 0.040 1 Ex/ANF 14x14 160.2 7.76 1271 0.369 179 0.556 0.424 0.030 0.351 142.0 0.034 1 WE 14x14 159.8 7.76 1177 0.362 179 0.556 0.400 0.024 0.345 144.0 0.034 1 WE 14x14 159.8 7.76 1302 0.415 179 0.556 0.422 0.022 0.368 145.2 0.034 1 WE, Ex/ANF 15x15 159.8 8.43 1472 0.465 204 0.563 0.422 0.024 0.366 144.0 0.015 1 Ex/ANF 17x17 159.8 8.43 1348 0.413 264 0.496 0.360 0.025 0.303 144.0 0.016 1 WE 17x17 159.8 8.43 1482 0.468 264 0.496 0.374 0.022 0.323 144.0 0.016 1 WE 17x17 160.1 8.43 1373 0.429 264 0.496 0.360 0.022 0.309 144.0 0.016 2 B&W 15x15 165.7 8.54 1515 0.481 208 0.568 0.430 0.026 0.369 144.0 0.016 2 B&W 17x17 165.8 8.54 1505 0.466 264 0.502 0.379 0.024 0.324 143.0 0.017 3 CE 16x16 178.3 8.10 1430 0.442 2364 0.506 0.382 0.023 0.3255 150.0 0.035 1 Ex/ANF3 14x14 160.2 7.76 1215 0.375 179 0.556 0.417 0.030 0.351 144.0 0.036 1 CE3 15x15 147.5 8.20 1360 0.432 216 0.550 0.418 0.026 0.358 132.0 ---

1 Ex/ANF3 15x15 148.9 8.25 1339 0.431 216 0.550 0.417 0.030 0.358 131.8 ---

1 CE 3 16x16 158.2 8.10 1300 0.403 236 4 0.506 0.382 0.023 0.3255 136.7 0.035 1 Minimum and Maximum initial Enrichments are 1.9 wt% 235U and 4.2 wt% 235U, respectively. All fuel rods are zirconium alloy type clad.

2 Vendor ID indicates the source of assembly base parameters. Loading of assemblies meeting dimensional limits is not restricted to the vendor(s) listed.

3 14x14, 15x15, and 16x16 fuel manufactured for Prairie Island, Palisades and St. Lucie 2 cores, respectively. These are not generic fuel assemblies provided to multiple reactors.

4 Some fuel rod positions may be occupied by burnable poison rods or solid filler rods.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 7 OF 19 Table 5.(b)(1)(i)-2, Loading Table for Intact PWR Fuel Minimum Burnup 30 GWD/MTU 30 < Burnup 35 GWD/MTU Initial Minimum Cooling Time (years) Minimum Cooling Time (years)

Enrichment CE 14x14 15x1 16x1 17x1 CE 14x14 15x15 16x16 17x17 wt% 235U (E) 14x14 5 6 7 14x14 1.9 E < 2.1 6 8 8 7 8 8 10 11 9 10 2.1 E < 2.3 6 7 8 6 7 7 10 10 8 10 2.3 E < 2.5 6 7 7 6 7 7 9 10 8 9 2.5 E < 2.7 6 7 7 6 7 7 9 9 7 8 2.7 E < 2.9 6 7 7 6 7 6 8 9 7 8 2.9 E < 3.1 5 7 7 6 6 6 8 8 7 8 3.1 E < 3.3 5 6 7 6 6 6 8 8 7 7 3.3 E < 3.5 5 6 6 6 6 6 7 8 6 7 3.5 E < 3.7 5 6 6 6 6 6 7 7 6 7 3.7 E 4.2 5 6 6 6 6 6 7 7 6 7 Minimum 35 < Burnup 40 GWD/MTU 40 < Burnup 45 GWD/MTU Initial Minimum Cooling Time (years) Minimum Cooling Time (years)

Enrichment CE 14x14 15x1 16x1 17x1 CE 14x14 15x15 16x16 17x17 wt% 235U (E) 14x14 5 6 7 14x14 1.9 E < 2.1 11 15 15 13 15 18 20 21 20 20 2.1 E < 2.3 10 13 14 12 13 15 19 19 18 19 2.3 E < 2.5 9 12 13 11 12 14 17 19 17 17 2.5 E < 2.7 9 12 12 10 11 12 16 18 15 17 2.7 E < 2.9 8 11 11 9 11 11 15 18 14 17 2.9 E < 3.1 8 10 10 9 10 10 14 18 13 15 3.1 E < 3.3 7 10 10 9 10 10 13 17 13 15 3.3 E < 3.5 7 9 10 8 9 9 12 17 13 15 3.5 E < 3.7 7 9 10 8 9 8 11 17 12 15 3.7 E 4.2 7 8 10 8 8 8 11 15 12 14 5.(b)(1)(ii) Intact BWR assemblies The package is designed to transport 56 irradiated intact BWR fuel assembles within the TSC. An intact fuel assembly is a spent nuclear fuel assembly without known or suspected cladding defects greater than pinhole leaks or hairline cracks.

For BWR fuel, the initial enrichment limit (the enrichment of the as-delivered fresh fuel assembly) represents the maximum peak planar-average enrichment allowed for loading into the TSC. The peak planar-average enrichment is defined to be the maximum planar-average enrichment at any height along the axis of the fuel assembly.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 8 OF 19 5.(b)(1)(ii) Intact BWR assemblies (continued)

The fuel assemblies consist of uranium dioxide pellets with zirconium alloy type cladding.

Prior to irradiation, the fuel assemblies must be within the dimension and specifications of Table 5.(b)(1)(ii)-1 below. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(ii)-2.

BWR intact fuel assemblies are authorized with or without channels based on a maximum channel width of 120 mils. The minimum and maximum allowable assembly average enrichment for loading is 1.9 wt% 235U and 4.0 wt% 235U respectively. The maximum burn up of the spent fuel assemblies is 45,000 MWD/GTU and the minimum cool time is six years. The maximum weight of UO2 is 11.08 MTU per cask. Unenriched fuel assemblies are not authorized for loading into the TSC. BWR fuel assemblies with unenriched axial blankets must have an enriched central fuel region and are acceptable for loading into a TSC if the minimum fuel enrichment of the central region is 1.9 wt% 235U. Any empty fuel position must be filled with a solid filler rod fabricated from either zirconium alloy or Type 304 stainless steel.

Table 5.(b)(1)(ii)-1, Intact BWR Fuel Assembly Characteristics Canister Vendor4 Array Max. Max. Max. Max No of Max Pitch Min Min Max Max Class 1,5 Length Assembly Assembly MTU Fuel (in) Rod Clad Pellet Active (in) Width (in)5 Weight Rods Dia (in) Thick Dia (in) Length (lb)6 (in) (in)2 4 Ex/ANF 7x7 171.3 5.51 620 0.196 48 0.738 0.570 0.036 0.490 144 4 Ex/ANF 8x8 171.3 5.51 563 0.177 63 0.641 0.484 0.036 0.405 145.2 4 Ex/ANF 9x9 171.3 5.51 557 0.173 79 0.572 0.424 0.030 0.357 145.2 4 GE 7x7 171.1 5.51 681 0.199 49 0.738 0.570 0.036 0.488 144.0 4 GE 7x7 171.2 5.51 681 0.198 49 0.738 0.563 0.032 0.487 144.0 4 GE 8x8 171.1 5.51 639 0.173 60 0.640 0.484 0.032 0.410 145.2 4 GE 8x8 171.1 5.51 681 0.179 62 0.640 0.483 0.032 0.410 145.2 4 GE 8x8 171.1 5.51 681 0.186 63 0.640 0.493 0.034 0.416 144.0 5 Ex/ANF 8x8 176.1 5.51 588 0180 62 0.641 0.484 0.036 0.405 150.0 5 Ex/ANF 9x9 176.1 5.51 576 0.167 743 0.572 0.424 0.030 0.357 150.0 55 Ex/ANF 9x9 176.1 5.51 576 0.178 793 0.572 0.424 0.030 0.357 150.0 5 GE 7x7 175.9 5.51 683 0.198 49 0.738 0.563 0.032 0.487 144.0 5 GE 8x8 176.1 5.51 665 0.179 60 0.640 0.484 0.032 0.410 150.0 5 GE 8x8 175.9 5.51 681 0.185 62 0.640 0.483 0.032 0.410 150.0 5 GE 8x8 175.9 5.51 681 0.188 63 0.640 0.493 0.034 0.416 146.0 5 GE 9x9 176.1 5.51 646 0.186 743 0.566 0.441 0.028 0.376 150.0 5 GE 9x9 176.1 5.51 646 0.198 79 3 0.566 0.441 0.028 0.376 150.0 1 Maximum Peak Planar Average Enrichment 4.0 wt%235U. Minimum enrichment is 1.9 wt%235U. All fuel rods are zirconium alloy type clad.

2 150 inch active fuel length assemblies contain 6 inch natural uranium blankets on top and bottom.

3 Shortened active fuel length in some rods.

4 Vendor ID indicates the source of assembly base parameters. Loading of assemblies meeting dimensional limits is not restricted to the vendor(s) listed.

5 Assembly width including channel. Unchanneled or channeled may be loaded based on a maximum channel thickness of 120 mils.

6 Exxon/ANF assembly weights are listed without channel.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 9 OF 19 Table 5.(b)(1)(ii)-2, Loading Table for Intact BWR Fuel Minimum Initial Burnup 30 GWD/MTU 30 < Burnup 35 GWD/MTU Enrichment wt% Minimum Cooling Time (years) Minimum Cooling Time (years) 235U (E) 9x9 8x8 7x7 9x9 8x8 7x7 1.9 E < 2.1 8 8 8 14 13 15 2.1 E < 2.3 7 7 8 12 12 13 2.3 E < 2.5 7 7 7 11 10 11 2.5 E < 2.7 7 6 7 9 9 10 2.7 E < 2.9 6 6 6 9 8 9 2.9 E < 3.1 6 6 6 8 8 8 3.1 E < 3.3 6 6 6 7 7 8 3.3 E < 3.5 6 6 6 7 7 7 3.5 E < 3.7 6 6 6 7 7 7 3.7 E 4.0 6 6 6 7 7 7 Minimum Initial 35 < Burnup 40 GWD/MTU 40 < Burnup 45 GWD/MTU Enrichment wt% Minimum Cooling Time (years) Minimum Cooling Time (years) 235U (E) 9x9 8x8 7x7 9x9 8x8 7x7 1.9 E < 2.1 24 23 25 34 33 35 2.1 E < 2.3 21 20 22 31 30 32 2.3 E < 2.5 19 18 20 29 28 29 2.5 E < 2.7 17 16 17 26 25 27 2.7 E < 2.9 14 14 15 24 23 24 2.9 E < 3.1 13 12 13 21 20 22 3.1 E < 3.3 11 11 12 19 18 20 3.3 E < 3.5 10 10 11 17 16 18 3.5 E < 3.7 10 9 10 15 14 16 3.7 E 4.0 10 9 10 14 13 15 5.(b)(1)(iii) Intact and Damaged PWR assemblies, and Fuel Debris from Maine Yankee The package is designed to transport 24 irradiated intact or damaged PWR fuel assemblies, canistered fuel debris, and GTCC waste within the TSC from the Maine Yankee Reactor. The standard Maine Yankee fuel assembly is the intact PWR CE 14x14 (see section 5.(b)(1)(i)).

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 10 OF 19 5.(b)(1)(iii) Intact and Damaged PWR assemblies, and Fuel Debris from Maine Yankee (continued)

In the course of reactor operations, some of the 14x14 assemblies were modified to change the standard configuration. These modifications included a) the removal of fuel rods without replacement; b) the replacement of removed fuel rods or burnable poison rods with rods of a different material, such as stainless steel, or with fuel rods of a different enrichment; and c) the insertion of control elements, or instruments or plug thimbles, in guide tube positions. In addition to the modified fuel assemblies, there are fuel assemblies that were designed with variable enrichment and axial blankets. These fuel assemblies are not modified, but differ from the cask design basis fuel assemblies.

Stainless steel spacers may be used in canisters to axially position PWR intact fuel assemblies that are shorter than the available cavity length. The minimum length of the PWR intact fuel assembly internal structure and bottom end fitting and/or spacers will ensure that the minimum distance to the fuel region for the base of the canister is 3.2 inches.

Unenriched fuel assemblies are not authorized for loading.

The following are the allowable Maine Yankee site specific contents:

5.(b)(1)(iii)(A) Maine Yankees site specific contents not requiring preferential loading patterns:

(1) Standard Irradiated CE 14 x 14 intact PWR fuel assemblies meeting the PWR fuel assembly characteristics in Table 5.(b)(1)(i)-1. The maximum fuel assembly weight, including other associated hardware is 1,515 pounds. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1.

(2) Irradiated Maine Yankee CE 14 x 14 PWR intact fuel assemblies may contain inserted control element assemblies (CEA), in-core instrument (ICI) thimbles or CEA plugs. CEAs or CEA plugs may not be inserted in damaged fuel assemblies, consolidated fuel assemblies or assemblies with irradiated stainless steel replacement rods. Fuel assemblies with a CEA or CEA plug inserted must be loaded in a Class 2 canister and cannot be loaded in a Class 1 canister. Fuel assemblies without an inserted CEA or CEA plug, including those with inserted ICI Thimbles, must be loaded in a Class 1 canister. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1 except for those assemblies containing ICI thimbles which must meet the specifications of Table 5.(b)(1)(iii)(A)-2.

(3) PWR intact fuel assemblies with fuel rods replaced with stainless steel or zirconium alloy rods or with Uranium oxide rods nominally enriched up to 1.95 wt%. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-3.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 11 OF 19 5.(b)(1)(iii)(A) (Continued)

(4) PWR intact fuel assemblies with fuel rods having variable enrichments with a maximum rod enrichment up to 4.21 wt% 235U and that also have a maximum planar average enrichment up to 3.99 wt% 235U. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1.

(5) PWR intact fuel assemblies with annular axial end blanket enrichments up to 2.6 wt% 235U.

The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1.

(6) PWR intact fuel assemblies with burnable poison rods or solid filler rods may occupy up to 16 of 176 fuel rod positions. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1.

(7) PWR intact fuel assemblies with one or more grid spacers missing or damaged such that the unsupported length of the fuel rods does not exceed 60 inches or with end fitting damage, including damaged or missing hold-down springs, as long as the assembly can be handled safely by normal means. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1.

5.(b)(1)(iii)(B) Maine Yankee site-specific allowable contents requiring preferential loading based on shielding, criticality, or thermal constraints (Maine Yankee CE 14 x 14 intact PWR fuel assemblies). A PWR basket fuel diagram can be found on Figure 5.(b)(1)(iii)(B)-1.

(1) Maine Yankee CE 14 x 14 PWR intact fuel assemblies with a burn up between 45,000 and 50,000 MWD/MTU meeting the following requirements for verification of the oxide layer thickness and high burn up fuel requiring preferential loading in the peripheral PWR fuel basket positions:

A verification program is required to determine the oxide layer thickness on high burn up fuel by measurement or by statistical analysis. A fuel assembly having a burn up between 45,000 MWD/MTU and 50,000 MWD/MTU is classified as high burn up. The verification program shall be capable of classifying high burn up fuel as INTACT FUEL or DAMAGED FUEL based on the following criteria:

I. A HIGH BURN UP FUEL assembly may be stored as INTACT FUEL provided that no more than 1% of the fuel rods in the assembly have a peak cladding oxide thickness greater than 80 microns, and that no more than 3% of the fuel rods in the assembly have a peak oxide layer thickness greater than 70 microns, and that the fuel assembly is otherwise INTACT FUEL.

II. A HIGH BURN UP FUEL assembly not meeting the cladding oxide thickness criteria for INTACT FUEL or that has an oxide layer that is detached or spalled from the cladding is classified as DAMAGED FUEL.

The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 12 OF 19 1.

5.(b)(1)(iii)(B) (Continued)

(2) PWR intact fuel assemblies with up to 176 fuel rods missing from the fuel assembly lattice.

The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1. These assemblies must be placed in a corner loading position in the PWR fuel basket.

(3) PWR intact fuel assemblies with burnable poison rods replaced by hollow zirconium alloy rods. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1. These assemblies must be placed in a corner PWR fuel basket loading position.

(4) Intact fuel assemblies with a start-up source in a center guide tube. The assembly must be loaded in a basket corner position and must be loaded in a Class 1 canister. Only one start-up source may be loaded in any fuel assembly or any canister. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1. These assemblies must be placed in a corner PWR fuel basket loading position.

(5) PWR intact fuel assemblies with CEA ends (fingertips) and/or an ICI segment inserted in corner guide tube positions. The assembly must also have a CEA plug installed. The assembly must be loaded in a PWR fuel basket corner position and must be loaded in a Class 2 canister.

The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1. CEA fingertips are not considered as CEAs for determination of minimum cool times.

5.(b)(1)(iii)(C) Maine Yankee CE 14 x 14 PWR fuel enclosed in a Maine Yankee Fuel Can (MYFC).

All Maine Yankee CE 14 x 14 PWR fuel enclosed in an MYFC must be loaded in a Class 1 fuel canister in a corner position of the PWR fuel basket. Up to 4 MYFC may be loaded into a TSC.

Intact Maine Yankee CE 14 x 14 PWR fuel may be loaded into a MYFC. The contents that must be loaded in the MYFC are:

(1) PWR fuel assemblies with up to two intact or damaged fuel rods inserted in each fuel assembly guide tube or with up to two burnable poison rods inserted in each guide tube. The rods inserted in the guide tubes cannot be from a different fuel assembly. The maximum number of rods in the fuel assembly (fuel rods plus inserted rods, including burnable poison rods) is 176. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1 for intact fuel rods and Table 5.(b)(1)(iii)(A)-4 for damaged fuel rods.

(2) A damaged fuel assembly with up to 100% of the fuel rods classified as damaged and/or damaged or missing assembly hardware components. A damaged fuel assembly cannot have an inserted CEA or other non-fuel component. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 13 OF 19 5.(b)(1)(iii)(A)-4 for damaged fuel rods.

5.(b)(1)(iii)(C) (Continued)

(3) Individual intact or damaged fuel rods in a rod type structure, which may be a guide tube, to maintain configuration control. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-1 for intact fuel rods and Table 5.(b)(1)(iii)(A)-4 for damaged fuel rods.

(4) Fuel debris consisting of fuel rods with exposed fuel pellets or individual intact or partial fuel pellets not contained in fuel rods. The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-4 for damaged fuel rods.

(5) Consolidated Fuel lattice and structure with a 17 x 17 array formed by grids and top and bottom end fittings connected by four solid stainless steel rods. Maximum contents are 289 fuel rods having a total lattice weight less than or equal to 2,100 pounds. A consolidated fuel lattice cannot have an inserted CEA or other non-fuel component. Only one consolidated fuel lattice may be stored in any TSC. Fuel must be cooled a minimum of 24 years.

(6) High burn up fuel assemblies not meeting the oxide layer thickness criteria previously defined in Section 5.(b)(1)(iii)(B)(1). The combined maximum average burn up, minimum cool time and maximum and minimum initial 235U enrichments must be within the specifications of Table 5.(b)(1)(iii)(A)-4 for damaged fuel rods.

PWR Basket Fuel Loading Position Diagram, Figure 5.(b)(1)(iii)(B)-1

1. Basket corner positions are positions 3, 6, 19, and 22. Corner positions are also periphery positions.

2. Basket periphery positions are positions 1, 2, 3, 6, 7, 12, 13, 18, 19, 22, 23, and 24. Periphery positions include the corner

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 14 OF 19 positions.

5.(b)(1)(iii)(C) (Continued)

Table 5.(b)(1)(iii)(A)-1, Loading Table for Maine Yankee CE 14x14 Fuel with and without CEA Cooled to Indicated Time Burnup 30 GWD/MTU Minimum Cool Time (Years) for Enrichment No CEA No CEA 5 Yr CEA 10 Yr CEA 15 Yr. CEA 20 Yr. CEA (Class 1) (Class 2) 1.9 E < 2.1 6 6 7 6 6 6 2.1 E < 2.3 6 6 7 6 6 6 2.3 E < 2.5 6 6 6 6 6 6 2.5 E < 2.7 6 6 6 6 6 6 2.7 E < 2.9 6 6 6 6 6 6 2.9 E < 3.1 5 6 6 6 6 6 3.1 E < 3.3 5 5 6 6 6 5 3.3 E < 3.5 5 5 6 6 5 5 3.5 E < 3.7 5 5 6 5 5 5 3.7 E 4.2 5 6 5 5 5 5 Loading Table for Maine Yankee CE 14x14 Fuel with and without CEA Cooled to Indicated Time Burnup 35 GWD/MTU Minimum Cool Time (Years) for Enrichment No CEA No CEA 5 Yr CEA 10 Yr CEA 15 Yr. CEA 20 Yr. CEA (Class 1) (Class 2) 1.9 E < 2.1 8 8 9 8 8 8 2.1 E < 2.3 7 7 9 8 8 8 2.3 E < 2.5 7 7 8 7 7 7 2.5 E < 2.7 7 7 8 7 7 7 2.7 E < 2.9 6 7 7 7 7 7 2.9 E < 3.1 6 6 7 7 6 6 3.1 E < 3.3 6 6 7 6 6 6 3.3 E < 3.5 6 6 7 6 6 6 3.5 E < 3.7 6 6 6 6 6 6 3.7 E 4.2 6 6 6 6 6 6

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 15 OF 19 5.(b)(1)(iii)(C) (Continued)

Table 5.(b)(1)(iii)(A)-1, continued , Loading Table for Maine Yankee CE 14x14 Fuel with and without CEA Cooled to Indicated Time Burnup 40 GWD/MTU Minimum Cool Time (Years) for Enrichment No CEA No CEA 5 Yr CEA 10 Yr CEA 15 Yr. CEA 20 Yr. CEA (Class 1) (Class 2) 1.9 E < 2.1 11 12 14 13 12 12 2.1 E < 2.3 10 10 13 11 11 11 2.3 E < 2.5 9 9 12 10 10 10 2.5 E < 2.7 9 9 10 9 9 9 2.7 E < 2.9 8 8 10 9 8 8 2.9 E < 3.1 8 8 9 8 8 8 3.1 E < 3.3 7 7 8 8 8 8 3.3 E < 3.5 7 7 8 7 7 7 3.5 E < 3.7 7 7 8 7 7 7 3.7 E 4.2 7 7 7 7 7 7 Loading Table for Maine Yankee CE 14x14 Fuel with and without CEA Cooled to Indicated Time Burnup 45 GWD/MTU Minimum Cool Time (Years) for Enrichment No CEA No CEA 5 Yr CEA 10 Yr CEA 15 Yr. CEA 20 Yr. CEA (Class 1) (Class 2) 1.9 E < 2.1 18 18 21 19 18 18 2.1 E < 2.3 15 16 19 17 17 16 2.3 E < 2.5 14 14 18 16 15 15 2.5 E < 2.7 12 13 16 14 14 13 2.7 E < 2.9 11 12 14 13 12 12 2.9 E < 3.1 10 11 13 12 11 11 3.1 E < 3.3 10 10 12 11 10 10 3.3 E < 3.5 9 9 11 10 10 10 3.5 E < 3.7 9 9 10 10 10 10 3.7 E 4.2 9 9 10 10 10 10

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 16 OF 19 5.(b)(1)(iii)(C) (Continued)

Table 5.(b)(1)(iii)(A)-1, continued, Loading Table for Maine Yankee CE 14x14 Fuel with and without CEA Cooled to Indicated Time Burnup 50 GWD/MTU Minimum Cool Time (Years) for Enrichment No CEA No CEA 5 Yr CEA 10 Yr CEA 15 Yr. CEA 20 Yr. CEA (Class 1) (Class 2) 1.9 E < 2.1 27 27 29 27 27 27 2.1 E < 2.3 24 24 27 25 24 24 2.3 E < 2.5 22 22 25 23 22 22 2.5 E < 2.7 19 19 23 21 20 20 2.7 E < 2.9 17 17 21 19 18 18 2.9 E < 3.1 15 16 19 18 18 18 3.1 E < 3.3 15 15 18 17 17 17 3.3 E < 3.5 15 15 17 17 17 17 3.5 E < 3.7 14 14 15 15 15 15 3.7 E 4.2 14 14 15 15 15 15 Table 5.(b)(1)(iii)(A)-2, Loading Table (Years) for Maine Yankee CE 14x14 fuel containing ICI Thimbles Minimum Initial Burnup 30 30 < Burnup 35 < Burnup 40 < Burnup 45 < Burnup Enrichment wt% GWD/MTU 35 GWD/MTU 40 45 50 GWD/MTU 235U (E)

GWD/MTU GWD/MTU 1.9 E < 2.1 6 8 11 18 27 2.1 E < 2.3 6 7 10 16 24 2.3 E < 2.5 6 7 9 14 22 2.5 E < 2.7 6 7 9 13 19 2.7 E < 2.9 6 6 8 11 17 2.9 E < 3.1 5 6 8 10 15 3.1 E < 3.3 5 6 7 10 15 3.3 E < 3.5 5 6 7 9 15 3.5 E < 3.7 5 6 7 9 14 3.7 E 4.2 5 6 7 9 14

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 17 OF 19 5.(b)(1)(iii)(C) (Continued)

Table 5.(b)(1)(iii)(A)-3, Required Cool Time for Maine Yankee Fuel Assemblies with Activated Stainless Steel Replacement Rods Assy Number Burnup Enrichment SSR Source Cool Time Earliest (GWD/MTU) (wt %) (g/s/assy) (years) Transportable N420 45 3.3 2.1602E+13 10 Jan 2001 N842 35 3.3 3.1396E+12 6 Jan 2001 N868 40 3.3 5.2444E+12 7 Jan 2001 R032 45 3.5 1.4550E+13 9 Jan 2005 R439 50 3.5 1.3998E+13 14 Jan 2010 R444 50 3.5 5.5993E+13 19 Jan 2015 Table 5.(b)(1)(iii)(A)-4, Cool time (years) for Maine Yankee CE 14x14 damaged fuel Minimum Initial Burnup 30 30 < Burnup 35 < Burnup 40 < Burnup 45 < Burnup Enrichment wt% GWD/MTU 35 GWD/MTU 40 45 50 GWD/MTU 235U (E)

GWD/MTU GWD/MTU 1.9 E < 2.1 7 11 19 28 37 2.1 E < 2.3 6 9 16 26 34 2.3 E < 2.5 6 8 14 23 32 2.5 E < 2.7 6 8 12 21 30 2.7 E < 2.9 6 7 11 19 27 2.9 E < 3.1 6 7 10 17 25 3.1 E < 3.3 5 7 9 15 23 3.3 E < 3.5 5 6 8 13 21 3.5 E < 3.7 5 6 8 12 19 3.7 E 4.2 5 6 7 11 17 5.(b)(1)(iv) Greater Than Class C Waste from Maine Yankee The package is designed to transport Maine Yankee Greater Than Class C Waste within a TSC.

Maine Yankee GTCC waste consists of solid, irradiated, and contaminated hardware and solid, particulate debris or filter media, provided the quantity of fissile material does not exceed a Type A quantity and does not exceed the mass limits of 10 CFR 71.15. The maximum curie inventory shall not exceed the values shown in Table 5.(b)(1)(iv)-1.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 18 OF 19 5.(b)(1)(iv) Greater Than Class C Waste from Maine Yankee (continued)

Table 5.(b)(1)(iv)-1, Maine Yankee GTCC Curie Inventory Limits per TSC Radionuclide Curie Inventory (Ci)/ TSC H-3 3.00E+02 C-14 1.50E+02 Mn-54 3.50E+02 Fe-55 2.00E+05 Co-58 1.00E+01 Co-60 2.90E+05 Ni-59 8.20E+02 Ni-63 9.00E+04 Nb-94 1.00E+01 Tc-99 1.00E+01 5.(b)(2) Maximum quantity of material per package The maximum weight of the contents shall not exceed 77,500 pounds.

(i) For the contents described in 5.(b)(1)(i) and 5.(b)(1)(iii): 24 PWR fuel assemblies, including standard inserts such as burnable poison rods or guides or guide tube thimble plugs, with a maximum weight of 38,500 pounds and a maximum decay heat limit per package not to exceed the values in Table 5.(b)(2)-1. The individual PWR assembly decay heat is limited to 0.83 kW.

Table 5.(b)(2)-1, PWR Decay Heat Limits Cool Time (Years) PWR Decay Heat Limit (kW)

Burnup (MWD/MTU) 35,000 40,000 45,000 50,0001 5 20.0 20.0 19.9 19.3 6 19.5 19.3 19.2 18.7 7 17.8 17.8 17.7 17.2 10 17.4 17.3 17.2 16.8 15 16.8 16.8 16.7 16.5 1 Maine Yankee PWR fuel assemblies (ii) For the contents described in 5.(b)(1)(ii): 56 BWR assemblies with a maximum weight of 39,000 pounds and a maximum decay heat limit per package of 16 kW. The individual BWR assembly decay heat is limited to 0.29 kW.

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9270 6 71-9270 USA/9270/B(U)F-96 19 OF 19 5.(b)(2) Maximum quantity of material per package (continued)

(iii) For the contents described in 5.(b)(1)(iv): GTCC waste with a maximum weight per package of 20,000 pounds in total or 10,000 pounds per compartment. The maximum decay heat for the GTCC is 4.5 kW per package.

5.(c) Criticality Safety Index 0.0

6. The package must be transported as exclusive use in a closed transport vehicle as per 10 CFR 71.47(b).

7. In addition to the requirements of Subpart G of 10 CFR Part 71:

(a) The package must be prepared for shipment and operated in accordance with the Operating Procedures in Chapter 7 of the application, as supplemented.

(b) Each packaging must be acceptance tested and maintained in accordance with the Acceptance Tests and Maintenance Program in Chapter 8 of the application, as supplemented.

8. The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CFR 71.17.

9. Transport by air of fissile material is not authorized.

10. Expiration date: October 31, 2027.

REFERENCES NAC International, Inc., Application dated August 3, 2016 as supplemented on August 7, 2017 and July 6, 2022.

FOR THE U.S. NUCLEAR REGULATORY COMMISSION Signed by Diaz-Sanabria, Yoira on 10/24/22 Yoira K. Diaz Sanabria, Chief Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Date: See digital signature

ML22290A046; ML22290A048 OFFICE NMSS/DFM/STLB NMSS/DFM/STLB NMSS/DFM/STLB NAME NDevaser ND SFigueroa SF YDiaz-Sanabria YD DATE Oct 17, 2022 Oct 17, 2022 Oct 24, 2022