ML22306A240

CoC 71-9356, Rev. 4

ML22306A240
Person / Time
Site:	07109356
Issue date:	11/03/2022
From:	Yoira Diaz-Sanabria Storage and Transportation Licensing Branch
To:	NAC International
Shared Package
ML22306A238	List: ML22306A239 ML22306A240 ML22306A241
References
Download: ML22306A240 (1)

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 NUMBERc. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBERPAGE PAGES 9356 4 71-9356 USA/9356/B(U)F-96 1 OF 43

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 NAC International, Inc., application dated 3930 East Jones Bridge Road, Suite 200 March 7 2022.

Peachtree Corners, GA 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.: MAGNATRAN

(2)

Description:

The MAGNATRAN package is canister-based and is part of a dual-purpose system for the storage and transportation of spent nuclear fuel for transporting the MAGNASTOR transportable storage canister (TSC). The MAGNATRAN packaging includes the package body, upper and lower impact limiters, and TSC. The package body consists of the inner and outer shells, lead and upper forging, lid, bottom plate, bottom forging and solid neutron shield.

Leakage testing of the package containment seals, in conjunction with the post-fabrication leakage testing of the entire containment boundary, assures that the containment is leaktight. The TSC may be credited for moderator exclusion, thus serving the function of being a special design feature, as required by 10 CFR 71.55(c), that prevents a single packing error from permitting leakage into the fissile material region.

Regardless of credit applied to the TSC confinement boundary to prevent water in leakage, the containment function is retained by the transport package body.

The packaging body is a cylinder with multiwall construction consisting of inner and outer stainless steel shells separated by a lead gamma radiation shielding. The inner and outer stainless steel shells are 1.75 and 2.25 inches thick, respectively. The lead gamma shield is 3.2 inches thick. Welded above the inner and outer steel shells is the upper forging. The upper forging is 7.2 inches thick where it attaches to the inner and outer shells.

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

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5.(a)(2) Description (continued)

The bottom of the package body consists of the bottom inner forging, the bottom outer forging and the bottom plate. The bottom inner forging is cup shaped and welded to the inner shell and the bottom forging. The ring-shaped bottom outer forging is welded to the outer shell and to the bottom plate. The bottom plate is welded onto the outer ring. The bottom inner forging is 5 inches thick and the bottom plate is 8.65 inches thick for a total of 13.65 inches of stainless steel shielding through the bottom.

The package lid is a 7.75-inch-thick stainless steel disk used to close the package. The lid is attached to the top forging by forty-eight, 2-8 UN-2A socket head cap screws. The socket head cap screws screw into the tapped holes in the upper forging. The package lid is sealed by two concentric O-rings, as is the coverplate for the lid port, using inner metallic and outer ethylene propylene diene monomer (EPDM) O-rings. The MAGNATRAN package contains a lid port that is closed by a bolted Type 304/304L stainless steel coverplate with dual O-rings. There are four stainless steel coverplate bolts. The lid port provides access to the port opening and the quick-disconnect fitting for backfilling and sampling the cavity gas during loading and unloading.

The neutron shield is comprised of NS-4-FR encased in stainless steel enclosures. The neutron shield material and its enclosure have two thicknesses, 5.8 inches and 6.4 inches, and is attached onto the outer shell along the length of the active fuel region around the circumference of the package cavity.

Two diametrically opposite lifting trunnions are bolted to the outside of the top forging to lift the transport package. Prior to transport, the lifting trunnions are removed and replaced with trunnion plugs. Two rotation trunnions are located on the outer shell near the bottom of the package to permit rotation between the horizontal and vertical positions and to provide longitudinal tiedown restraint in the aft direction. The rotation trunnions are located approximately 5 inches off the cask centerline to ensure that the cask rotates in the proper direction.

A cavity spacer is used for the short TSCs to locate and support the canister and to minimize excessive longitudinal movement in the transport cask cavity, which is sized to accommodate the long TSC.

The MAGNATRAN package has cup-shaped impact limiters, consisting of a combination of redwood and balsa wood encased in a stainless-steel shell. The impact limiters are bolted over each end of the package to limit the g-loads acting on the package during a package drop event. The impact limiters are attached to the lid and bottom plate via 16 tapped holes for retaining rods and nuts.

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

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5.(a)(2) Description (continued)

The TSC is constructed of a stainless steel cylindrical shell, bottom-end plate, closure lid, closure ring, and redundant port covers. The TSC confines the fuel basket structure and the spent fuel or the Greater-Than-Class C (GTCC) waste basket liner and GTCC waste. The TSC cylindrical shell is dual certified 304/304L stainless steel with a 72-inch diameter and is 1/2 inch thick and either 191.8 or 184.8 inches long, depending on the contents. The bottom end plate is welded onto the lower end of the TSC shell and is 2.75 inches thick. The closure lid is 9 inches thick and is either a solid stainless steel closure lid or stainless steel/carbon steel closure lid. The closure lid is welded onto the upper end of the TSC shell. The dual port covers provide a dual-welded closure system for the vent and drain ports. The GTCC TSC is similar in design and construction to the TSCs for spent fuel, but instead of a basket, it contains a GTCC waste liner.

The PWR fuel basket design is an arrangement of 21 square, stainless steel fuel tubes held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The 21 tubes develop 37 positions within the basket for the PWR spent fuel. Each PWR basket fuel tube has a nominal 8.86-inch square opening. Each developed cell fuel position has a nominal 8.76-inch square opening. The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube. Each neutron absorber sheet is covered by a thin stainless steel sheet to protect the neutron absorber during fuel loading and to keep it in position. The neutron absorber and stainless steel cover are secured to the fuel tube using weld posts distributed across the width and along the length of the fuel tube.

The PWR damaged fuel basket is designed to store up to four damaged fuel cans in the damaged fuel basket assembly in the short TSC. The damaged fuel basket assembly has a capacity of up to 37 undamaged PWR fuel assemblies, which includes the four damaged fuel can locations. A damaged fuel can may be placed in each of the four damaged fuel can basket locations. The arrangement of tubes and fuel positions is the same as in the standard fuel basket, but the design of each of the four corner support weldments is modified with additional structural support to provide an enlarged position for a damaged fuel can at the outermost corners of the fuel basket. Each damaged fuel can location has a nominal 9.80-inch square opening. A damaged fuel can, or an undamaged fuel assembly may be loaded in a damaged fuel can location.

Similar to the PWR basket, the BWR basket consists of 45 stainless steel fuel tubes that develop 87 basket locations for the BWR spent fuel. Each BWR basket fuel tube has a nominal 5.86-inch square opening. Each developed cell fuel position has a nominal 5.77-inch square opening. The BWR basket fuel tubes are held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube for criticality control. Each neutron absorber sheet is covered by a sheet of stainless steel to protect the neutron absorber during fuel loading and to keep it in position. The neutron absorber and stainless steel cover are secured to the fuel tube using weld posts distributed across the width and along the length of the fuel tube.

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 NUMBERc. DOCKET NUMBERd. PACKAGE IDENTIFICATION NUMBERPAGEPAGES 9356 4 71-9356 USA/9356/B(U)F-96 4 OF 43

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

The damaged fuel can confines the fuel material within the can to minimize the potential for dispersal of the fuel material into the TSC cavity. The side plates that form the upper end of the damaged fuel can are 0.15-in thick and the tube body walls are 0.048-in thick (18-gage sheet). The damaged fuel can lid plate and bottom thicknesses total 11/16 inches and the lid overall height is 2.32 inches. The damaged fuel can bottom plate thickness is 5/8 (0.625) inch. The damaged fuel can is designed in two lengths: an overall length of 166.9 inches with a nominal cavity length of 164.0 inches; or an overall length of 171.8 inches with a nominal cavity length of 169.0 inches (shorter fuel assemblies may be accommodated with a fuel assembly spacer to limit axial movement). For the shorter damaged fuel can, a spacer is used in the damaged fuel basket assembly or alternatively fixed to the damaged fuel can bottom plate to provide an overall height of 171.5 inches. The damaged fuel can (DFC) lid and bottom include screened drain holes.

The stainless steel GTCC waste basket liner is designed to hold GTCC waste and dimensionally fit in a TSC. The GTCC waste basket liner is 173 inches long with a 1-inch-thick bottom plate welded onto it. The GTCC liner stainless steel shell is 2 inches thick for structural and gamma shield functions and has lifting lugs welded on the inside diameter of the shell. The liner design also includes an outer ring and a middle support under the bottom plate and drain holes in the bottom plate to facilitate free flow drainage from the liner. The GTCC TSC includes a sump location in the bottom plate and the closure lid includes a drain tube assembly to enable draining and drying of the loaded TSC.

The package has approximate dimensions and weight as follows:

Cavity diameter 72 inches Cavity length 193 inches Package body outer diameter 87 inches Impact limiter diameter 128 inches Package length without impact limiters 214 inches with impact limiters 322 inches

The maximum gross weight of the package is about 312,000 lbs.

5.(a)(3) Drawings

The MAGNATRAN package is constructed and assembled in accordance with NAC drawings:

71160-500, Rev. 6P Shipping Configuration, Transport Cask, MAGNATRAN 71160-501, Rev. 0 Assembly, Transport Cask, MAGNATRAN 71160-502, Rev. 6P Transport Cask Body, MAGNATRAN 71160-504, Rev. 2 Misc. Details, Transport Cask, MAGNATRAN 71160-505, Rev. 6P Lid Assembly, Transport Cask, MAGNATRAN NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(a)(3) Drawings (Continued)

71160-506, Rev. 1 Cask Cavity Spacer, MAGNATRAN 71160-511, Rev. 1 Personnel Barrier, Shipping Configuration, Transport Cask, MAGNATRAN 71160-512, Rev. 1 Nameplate, MAGNATRAN 71160-530, Rev. 1 Misc. Details, Impact Limiter, MAGNATRAN 71160-531, Rev. 2P Impact Limiter, Transport Cask, MAGNATRAN 71160-551, Rev. 10P Fuel Tube Assembly, MAGNASTOR - 37 PWR 71160-559, Rev. 0 Lifting Trunnion, Transport Cask, MAGNATRAN 71160-571, Rev. 10P Details, Neutron Absorber, Retainer, MAGNASTOR - 37 PWR 71160-572, Rev. 9P Details, Neutron Absorber, Retainer, MAGNASTOR - 87 BWR 71160-574, Rev. 6 Basket Support Weldments, MAGNASTOR - 37 PWR 71160-575, Rev. 11P Basket Assembly, MAGNASTOR - 37 PWR 71160-581, Rev. 5 Shell Weldment, TSC, MAGNASTOR 71160-584, Rev. 8 Details, TSC, MAGNASTOR 71160-585, Rev. 13 TSC Assembly, MAGNASTOR 71160-591, Rev. 8P Fuel Tube Assembly, MAGNASTOR - 87 BWR 71160-598, Rev. 7P Basket Support Weldments, MAGNASTOR - 87 BWR 71160-599, Rev. 8P Basket Assembly, MAGNASTOR - 87 BWR 71160-600, Rev. 5P Basket Assembly, MAGNASTOR - 82 BWR 71160-601, Rev. 0 Damaged Fuel Can (DFC), Assembly, MAGNASTOR 71160-602, Rev. 1 Damaged Fuel Can (DFC), Details, MAGNASTOR 71160-620, Rev. 1P Top Fuel Spacer, MAGNASTOR 71160-671, Rev. 2P Details, Neutron Absorber, Retainer, For DF [Damaged Fuel] Corner Weldment, MAGNASTOR - 37 PWR 71160-673, Rev. 1 Damaged Fuel Can (DFC), Spacer, MAGNASTOR 71160-674, Rev. 4P DF Corner Weldment, MAGNASTOR 71160-675, Rev. 3P DF Basket Assembly, 37 Assembly PWR, MAGNASTOR 71160-681, Rev. 1 DF, Shell Weldment, TSC, MAGNASTOR 71160-684, Rev. 2 Details, DF Closure Lid, MAGNASTOR 71160-685, Rev. 8 DF, TSC Assembly, MAGNASTOR 71160-711, Rev. 1 GTCC Waste Basket Liner, MAGNASTOR 71160-781, Rev. 1 Shell Weldment, GTCC TSC, MAGNASTOR 71160-785, Rev. 4 GTCC TSC, Assembly, MAGNASTOR

5.(b) Contents

(1) Type and Form of Material

(i) Undamaged PWR assemblies

Undamaged PWR fuel assemblies within the 37 PWR basket assembly. Undamaged fuel is spent nuclear fuel that does not have any visible deformation other than uniform bowing that occurs in the reactor, assemblies that do not have missing rods, and assemblies with missing rods that are replaced by solid stainless steel or zirconium filler rods that displace a volume equal to or greater than the original rods and assemblies NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

that do not contain structural defects that adversely affect radiological and/or criticality safety and/or result in unsupported fuel rod lengths in excess of 60 inches and that can be handled by normal means. Undamaged PWR fuel is loaded into the short TSC, except for Combustion Engineering (CE) 16x16 fuel assemblies, which may be loaded into either length TSC.

The fuel assemblies consist of uranium dioxide pellets with zirconium alloy-clad fuel rods and zirconium alloy instrument and guide tubes. Empty fuel rod positions are to be filled with a solid filler rod or a solid neutron absorber rod. PWR fuel assemblies containing nonfuel hardware may be loaded in the TSC. Prior to irradiation, the fuel assemblies must be within the dimensions and specifications of the hybrid assemblies listed in Table 1. In addition, the PWR fuel must meet the fuel class assembly specifications listed in Table 2.

The burnup credit loading curve in Table 3, must be used for the 37 assembly loading profile. WE15x15 fuel may use the burnup credit loading curve in Table 4, with the 33, 35 or 36 assembly loading scheme provided the required cell locations for that profile shown in Figure 1 are left empty, at a minimum. Fuel assembly burnup, minimum initial average enrichment1, and cool time requirements are provided in Table 9 and Table 11, for PWR baskets with Type 2 neutron absorbers and Table 12 and Table 14 for baskets with Type 1 neutron absorbers. Burnup credit curves are only applicable to systems not crediting moderator exclusion. Initial enrichment up to 5 weight percent (wt%) in the 235U isotope, with no burnup requirement, is permitted when crediting moderator exclusion.

Unirradiated fuel and unenriched fuel are not authorized for loading, except that unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than 6 inches. An unenriched rod may be used as a replacement rod to return a fuel assembly to an undamaged condition.

Undamaged PWR fuel assemblies may contain nonfuel hardware. Fuel assemblies with an instrument tube tie rod repair shall be loaded with fuel inserts and/or top spacers to ensure proper spacing and support of the fuel assembly. Fuel inserts and/or top spacers are not required when using the extended fuel tube basket because the top nozzle is adequately supported. The nonfuel hardware may be loaded as a complete assembly or as individual components, individual nonfuel rods may be full-length rods or partial-length rods/rodlets. Partial-length rods/rodlets are permitted in guide tubes provided guide tube plug devises are installed. Nonfuel hardware must meet the exposure and cool time or cobalt-60 activity requirements in Tables 6-8. Fuel assemblies loaded with nonfuel hardware must meet the additional cool time requirements in Table 5 (for Type 2 neutron absorbers), and Table 15 (for Type 1 neutron absorbers).

1 Assembly average fuel enrichment is the enrichment value determined by averaging the 235U wt% enrichment over the entire fuel region (UO2) of an individual assembly, including axial blankets, if present.

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

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Hafnium absorber assemblies (HFRA) are only allowed for Westinghouse (WE) assemblies and may have a maximum exposure of 4.0 GWd/MTU and must have a minimum cool time of 16 years. Fuel assemblies may contain any number of unirradiated 5.(b)(1)(i) Contents - Type and Form of Material (continued)

nonfuel solid filler fuel replacement rods. Activated stainless steel fuel replacement rods are limited to 5 steel rods per assembly, 1 assembly per basket, and a maximum steel rod exposure of 32.5 GWd/MTU. Fuel assemblies with activated stainless steel rods must be cooled for either a minimum of 21 years or the loading table minimum cool time (as adjusted for additional cool times for nonfuel hardware, as applicable) plus 1 year, whichever is greater.

Fuel assemblies loaded with in-core instrument thimbles must meet the additional cool time requirements in Table 5 or Table 15, as appropriate, for burnable poison rod assemblies (BPRAs) or guide tube thimble plug devices (GTPDs), whichever is bounding, for Westinghouse and Babcock and Wilcox (B&W) fuel types and for reactor control components (RCCs) for CE fuel types. The additional cool time requirements for assemblies with nonfuel hardware are added to any additional cool time requirements due to damaged fuel also being loaded in the same TSC. RCCs are restricted to fuel storage locations 11, 12, 13, 18, 19, 20, 25, 26 and 27 in Figure 1. Only one neutron source assembly (NSA) is permitted to be loaded in a TSC in fuel storage locations 11, 12, 13, 18, 19, 20, 25, 26 or 27, as shown on Figure 1.

NSAs may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rodlets and/or thimble plug rodlets. NSAs, GTPDs, and BPRAs are not authorized for CE fuel assemblies. In addition, the following un-irradiated nonfuel hardware may be loaded with the fuel assemblies: stainless steel rods inserted to displace guide tube dashpot water, instrument tube tie components, and guide tube anchors or similar devices. Axial power shaping rods are not allowed contents.

Under-burned Westinghouse 15x15 assemblies (assemblies with a maximum enrichment greater than that dictated by the burnup credit loading curve) may be loaded provided that a RCCA is inserted in the assembly, the enrichment is equal to or less than 4.05 wt. % 235U, and the assembly burnup is greater than or equal to 12,000 MWd/MTU.

When loading under-burned fuel, the RCCAs must be full-length Ag-In-Cd RCCAs comprised of stainless steel clad rods constructed with 80% Ag, 15% In and 5% Cd absorber pellets and having an exposure equal to or less than 200,000 MWd/MTU. The basket must include absorber sheets with an effective 10B areal density of 0.036 g/cm2.

Any assemblies loaded without an RCCA inserted must meet the burnup credit loading curve for the applicable assembly loading profile. Burnup credit curves, and the criticality need for RCCA insertion are only applicable to systems not crediting moderator exclusion. Initial enrichment up to 5 wt% 235U, with no burnup requirement, is permitted when crediting moderator exclusion.

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

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 1 PWR Hybrid Fuel Assembly Characteristics No. of No. of Max Min Min Max Max Max Vendor Hybrid Hybrid Array Fuel Guide Pitch Clad Clad Pellet Active Load Assembly Group Rods Tubes (in.) OD Thick. OD Length (MTU)

(See (in.) (in.) (in.) (in.)

Note 1)

BW BW15H1 H1 15x15 208 17 0.5680 0.4300 0.0265 0.3686 144.0 0.4807 BW BW15H2 H2 15x15 208 17 0.5680 0.4300 0.0250 0.3735 144.0 0.4807 BW BW15H3 H3 15x15 208 17 0.5680 0.4280 0.0230 0.3742 144.0 0.4807 BW BW15H4 H4 15x15 208 17 0.5680 0.4140 0.0220 0.3622 144.0 0.4690 BW BW17H1 H1 17x17 264 25 0.5020 0.3770 0.0220 0.3252 144.0 0.4681 CE CE14H1 H1 14x14 176 5 0.5800 0.4400 0.0260 0.3805 137.0 0.4115 CE CE16H1 H1 16x16 236 5 0.5063 0.3820 0.0250 0.3250 150.0 0.4463 WE WE14H1 H1 14x14 179 17 0.5560 0.4000 0.0162 0.3674 145.2 0.4144 WE WE15H1 H1 15x15 204 21 0.5630 0.4220 0.0242 0.3669 144.0 0.4671 WE WE15H2 H2 15x15 204 21 0.5630 0.4170 0.0265 0.3570 144.0 0.4469 WE WE17H1 H1 17x17 264 25 0.4960 0.3720 0.0205 0.3232 144.0 0.4671 WE WE17H2 H2 17x17 264 25 0.4960 0.3600 0.0225 0.3088 144.0 0.4327

Notes:

1. Combined number of guide and instrument tubes.

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

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 2 PWR Fuel Class Assembly Characteristics Characteristic Fuel Class 14x14 14x14 15x15 15x15 16x16 17x17 Base Fuel Type2 CE, SPC WE, SPC WE, SPC BW, FCF CE BW, SPC, WE, FCF Max Initial Enrichment (wt. % 235U)3 5.0 5.0 5.0 5.0 5.0 5.0 Min Initial Enrichment (wt. % 235U)3 1.3 1.3 1.3 1.3 1.3 1.3 Number of Fuel Rods4 176 179 204 208 236 264 Max Assembly Average Burnup 60,000 60,000 60,000 60,000 60,000 60,000 (MWd/MTU)5 Min Cool Time (years) 4 4 4 4 4 4 Max Weight per Storage Location See Note 1 See Note 1 See Note 1 See Note 1 See Note 1 See Note 1 (lbs.)

Max Decay Heat per Fuel Location See Note 2 See Note 2 See Note 2 See Note 2 See Note 2 See Note 2 (Watts)6 Notes:

1.Maximum weight per storage location is 1,765 lbs (weight includes spent fuel assembly, nonfuel hardware, damaged fuel cans and fuel spacers) with a maximum contents weight of 62,160 lbs for the PWR basket and 61,184 lbs for the damaged fuel basket7. The maximum nominal assembly length is 178.3 inches for assemblies in the long TSC and 167.0 inches for assemblies in the short TSC. The maximum nominal fuel width is 8.54 inches.

2.For PWR baskets with Type 2 thermal conductivity neutron absorbers, the maximum heat load is 622 watts per storage location (590.5 watts for maximum assembly average burnup

>45,000 MWd/MTU), and for PWR baskets with Type 1 thermal conductivity neutron absorbers the maximum heat load is 595 watts per storage location (565 watts for maximum assembly average burnup >45,000 MWd/MTU). The heat load includes the contribution from the nonfuel hardware.

2 Indicates assembly and/or nuclear steam supply system vendor/type referenced for fuel input data. Fuel acceptability for loading is not restricted to the indicated vendor provided that the fuel assembly meets the load limits. Abbreviations are as follows: Westinghouse (WE), Combustion Engineering (CE), Siemens Power Corporation (SPC), Babcock and Wilcox (BW), and Framatome Cogema Fuels (FCF).

3 All reported enrichment values are nominal preirradiation fabrication values.

4 Assemblies may contain nonfuel hardware and/or fuel replacement rods (also referred to as filler rods). Filler rods are considered to be a component of spent nuclear fuel assemblies and not nonfuel hardware. Filler rods may be burnable absorber rods, stainless steel rods or zirconium alloy rods.

5 Assembly average burnup is the burnup value determined by averaging the burnup over the entire fuel region (UO2) of an individual assembly, including axial blankets, if present.

6 Maximum uniform heat load per storage location.

7 TSC and maximum contents shall not exceed 104,500 pounds.

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

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 3 Maximum Initial Enrichment Assembly Undamaged Fuel 15 Year Minimum Cool Time Zero (0) Max Initial Enrichment (wt % 235U) 10B Burnup = C4 x Burnup (GWd/MTU) + C5 Assembly Absorber Maximum Burnup 18 Burnup Burnup ID (g/cm2) Enrichment (GWd/MTU) < 18 (GWd/MTU) 30 (GWd/MTU) > 30 (wt %) C4 C5 C4 C5 C4 C5

BW15 1.9 0.0501 1.69 0.0693 1.65 0.0748 1.60 BW17 1.9 0.0502 1.72 0.0687 1.70 0.0742 1.66 CE14 2.1 0.0473 2.04 0.0675 2.03 0.0759 1.93 CE16 0.036 2.1 0.0464 2.03 0.0657 2.06 0.0733 1.99 WE14 2.2 0.0496 2.08 0.0672 2.21 0.0725 2.29 WE15 1.9 0.0494 1.74 0.0683 1.72 0.0742 1.67 WE17 1.9 0.0494 1.71 0.0685 1.68 0.0749 1.61 BW15 1.8 0.0507 1.61 0.0687 1.59 0.0745 1.48 BW17 1.9 0.0503 1.66 0.0683 1.63 0.0733 1.59 CE14 2.1 0.0468 1.95 0.0664 1.97 0.0738 1.90 CE16 0.030 2.1 0.0470 1.95 0.0649 1.99 0.0727 1.90 WE14 2.1 0.0492 2.03 0.0680 2.10 0.0728 2.19 WE15 1.9 0.0503 1.67 0.0675 1.66 0.0747 1.54 WE17 1.9 0.0494 1.64 0.0685 1.58 0.0737 1.53 BW15 1.8 0.0508 1.58 0.0686 1.52 0.0754 1.41 BW17 1.8 0.0503 1.62 0.0683 1.59 0.0748 1.47 CE14 2.1 0.0471 1.92 0.0666 1.92 0.0729 1.87 CE16 0.027 2.1 0.0462 1.93 0.0657 1.92 0.0747 1.75 WE14 2.1 0.0499 1.98 0.0667 2.10 0.0743 2.07 WE15 1.9 0.0503 1.63 0.0677 1.60 0.0749 1.46 WE17 1.9 0.0497 1.60 0.0683 1.54 0.0749 1.41 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 4 Maximum Initial Enrichment - Undamaged Fuel Configuration WE15x15 Optional Configurations - 20 Year Minimum Cool Time Zero (0) Max Initial Enrichment (wt % 235U)

Number of 10B Burnup = C4 x Burnup (GWd/MTU) + C5 Assemblies Absorber Maximum Burnup 18 Burnup Burnup Loaded (g/cm2) Enrichment (GWd/MTU) < 18 (GWd/MTU) 30 (GWd/MTU) > 30 (wt. % 235U) C4 C5 C4 C5 C4 C5 36 2.0 0.0497 1.93 0.0681 1.99 0.0747 2.00 35 0.036 2.1 0.0507 1.97 0.0673 2.08 0.0730 2.12 33 2.2 0.0504 2.12 0.0664 2.29 0.0745 2.32 36 2.0 0.0494 1.87 0.0687 1.90 0.0737 1.93 35 0.030 2.0 0.0499 1.92 0.0688 1.97 0.0740 1.99 33 2.1 0.0497 2.06 0.0686 2.15 0.0724 2.29 36 2.0 0.0501 1.83 0.0677 1.87 0.0741 1.84 35 0.027 2.0 0.0494 1.89 0.0675 1.94 0.0735 1.96 33 2.1 0.0492 2.03 0.0674 2.12 0.0730 2.21

Table 5 Additional Fuel Assembly Cool Time Required to Load Nonfuel Hardware (23kW/Package)

Core Additional Cool Time (Years)

(Assembly) BPRA/HFRA GTPD RCC NSA CE14 -- -- 0.4 --

WE14 1.1 0.1 0.3 1.1 WE15 1.3 0.1 6.9 1.3 BW15 0.1 0.2 0.3 0.2 CE16 -- -- 0.4 --

WE17 1.4 0.2 6.7 1.4 BW17 0.1 0.2 0.3 0.2 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 6 Nonfuel Hardware Max Exposure and Required Cool Times (Years)

Maximum Minimum Cool Time (Years)

Exposure WE 14x14 WE 15x15 B&W 15x15 WE 17x17 B&W 17x17 Hardware (GWd/MTU)

BPRA 70 8.0 8.0 8.0 8.0 8.0 GTPD 180 8.0 8.0 8.0 8.0 8.0

Note: 1. Specified minimum cool times for BPRAs are independent of the required minimum cool times for the fuel assembly containing the BPRA

2. Specified minimum cool times for GTPDs are independent of the required minimum cool times for the fuel assembly containing the GTPD.

3. The maximum exposure and minimum cooling time limits for NSAs without absorber rods are the same as those for GTPDs while the maximum exposure and minimum cooling time limits for NSAs with absorber rods are the same as those for BPRAs.

4. Only GTPDs that do not include absorber, or poison, rods or water displacement rods are allowed contents.

Table 7 Nonfuel Hardware Max 60Co Activity (Ci)

Hardware WE 14x14 WE 15x15 B&W 15x15 WE 17x17 B&W 17x17 BPRA 704.0 901.0 26.0 894.0 27.0 GTPD 60.5 73.1 99.2 93.3 107.8

Table 8 RCC Cool Time (Years)

Minimum Cool Maximum Exposure Time (Years) (GWd/MTU) 10.0 180 14.0 270 20.0 360 Note: 1. Interpolation between exposure - cool time limits is not allowed.

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Figure 1 Undamaged Fuel Basket Loading Profile8

37 assembly loading

36 assembly loading: remove 19

35 assembly loading: remove 19, 18

33 assembly loading: remove 19, 18, 20, 12

Note: The 33, 35 and 36-Assembly patterns also apply to the damaged fuel basket.

8 A short loaded 33, 35 or 36 assembly loading profile may still use the burnup credit curve in Table 4 provided that, at a minimum, the required cell locations for that profile shown in Figure 1 are left empty.

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

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 9Loading Table for PWR Fuel - 23 kW/Package1

Minimum Initial Assembly Average Burnup 30 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 1414 1414 1515 1515 1616 1717 1717 2.1 E < 2.3 5.7 5.8 6.7 6.9 6.3 6.8 6.8 2.3 E < 2.5 5.7 5.8 6.6 6.9 6.2 6.7 6.7 2.5 E < 2.7 5.6 5.7 6.6 6.8 6.1 6.6 6.6 2.7 E < 2.9 5.5 5.6 6.5 6.7 6.0 6.6 6.6 2.9 E < 3.1 5.6 5.6 6.4 6.7 6.0 6.5 6.5 3.1 E < 3.3 5.4 5.6 6.4 6.6 6.0 6.5 6.5 3.3 E < 3.5 5.4 5.5 6.3 6.6 5.9 6.4 6.4 3.5 E < 3.7 5.3 5.5 6.3 6.5 5.9 6.4 6.4 3.7 E < 3.9 5.3 5.4 6.2 6.5 5.9 6.3 6.3 3.9 E < 4.1 5.3 5.4 6.2 6.5 5.8 6.3 6.3 4.1 E < 4.3 5.2 5.4 6.1 6.4 5.8 6.3 6.3 4.3 E < 4.5 5.2 5.3 6.1 6.4 5.8 6.2 6.2 4.5 E < 4.7 5.2 5.3 6.1 6.4 5.8 6.2 6.2 4.7 E < 4.9 5.1 5.3 6.0 6.3 5.7 6.1 6.1 E 4.9 5.1 5.2 6.0 6.3 5.7 6.1 6.1 Minimum Initial 30 < Assembly Average Burnup 35 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 1414 1414 1515 1515 1616 1717 1717 2.1 E < 2.3 - - 9.2 - - - -

2.3 E < 2.5 7.1 7.4 9.0 9.6 8.1 9.1 9.1 2.5 E < 2.7 7.0 7.2 8.9 9.5 8.0 9.0 9.0 2.7 E < 2.9 6.9 7.1 8.8 9.3 7.9 8.9 8.8 2.9 E < 3.1 6.8 7.0 8.7 9.2 7.8 8.8 8.7 3.1 E < 3.3 6.8 7.0 8.6 9.0 7.7 8.6 8.6 3.3 E < 3.5 6.7 6.9 8.5 9.0 7.7 8.6 8.6 3.5 E < 3.7 6.7 6.9 8.4 8.9 7.6 8.5 8.5 3.7 E < 3.9 6.6 6.8 8.3 8.9 7.5 8.4 8.4 3.9 E < 4.1 6.5 6.7 8.2 8.8 7.5 8.4 8.4 4.1 E < 4.3 6.5 6.7 8.2 8.7 7.4 8.3 8.3 4.3 E < 4.5 6.4 6.6 8.1 8.7 7.4 8.2 8.2 4.5 E < 4.7 6.4 6.6 8.1 8.6 7.3 8.2 8.2 4.7 E < 4.9 6.4 6.6 8.0 8.6 7.3 8.1 8.1 E 4.9 6.3 6.5 8.0 8.5 7.2 8.1 8.1

1. - means not allowed NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 9Loading Table for PWR Fuel - 23 kW/Package1 (continued)

Minimum Initial 35 < Assembly Average Burnup 40 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 1414 1414 1515 1515 1616 1717 1717 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 9.7 11.9 13.5 14.7 11.6 13.7 13.7 2.7 E < 2.9 9.5 10.1 13.3 14.4 11.5 13.4 13.4 2.9 E < 3.1 9.3 9.8 13.1 14.1 11.3 13.2 13.2 3.1 E < 3.3 9.1 9.7 12.8 14.0 11.1 13.0 13.0 3.3 E < 3.5 9.0 9.5 12.6 13.8 10.9 12.8 12.8 3.5 E < 3.7 8.9 9.4 12.5 13.6 10.8 12.7 12.6 3.7 E < 3.9 8.8 9.3 12.3 13.5 10.7 12.5 12.5 3.9 E < 4.1 8.7 9.1 12.1 13.3 10.5 12.3 12.3 4.1 E < 4.3 8.6 9.0 12.0 13.2 10.4 12.2 12.2 4.3 E < 4.5 8.5 9.0 11.9 13.1 10.3 12.1 12.1 4.5 E < 4.7 8.4 8.9 11.8 13.0 10.2 12.0 12.0 4.7 E < 4.9 8.4 8.8 11.7 12.8 10.2 12.0 11.9 E 4.9 8.3 8.8 11.7 12.8 10.1 11.9 11.9 Minimum Initial 40 < Assembly Average Burnup 45 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 1414 1414 1515 1515 1616 1717 1717 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 - - - - - - -

2.7 E < 2.9 14.5 19.0 20.0 21.3 17.2 20.0 20.0 2.9 E < 3.1 13.6 16.7 19.3 21.0 16.8 19.6 19.6 3.1 E < 3.3 13.4 14.6 19.0 20.7 16.5 19.4 19.3 3.3 E < 3.5 13.1 14.1 18.7 20.4 16.3 19.1 19.1 3.5 E < 3.7 12.9 13.8 18.6 20.2 16.0 18.8 18.8 3.7 E < 3.9 12.7 13.7 18.3 19.9 15.8 18.7 18.6 3.9 E < 4.1 12.5 13.5 18.1 19.7 15.6 18.4 18.4 4.1 E < 4.3 12.3 13.3 17.9 19.6 15.4 18.3 18.3 4.3 E < 4.5 12.1 13.1 17.7 19.4 15.3 18.1 18.0 4.5 E < 4.7 12.0 13.0 17.6 19.2 15.2 18.0 17.9 4.7 E < 4.9 11.9 12.8 17.4 19.0 15.0 17.7 17.8 E 4.9 11.8 12.7 17.3 19.0 14.9 17.6 17.6

1. - means not allowed NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 10Loading Table for High Burnup PWR Fuel - 21.85 kW/Package1

Minimum Initial 45 < Assembly Average Burnup 50 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W WE B&W wt % 235U (E) 1414 1414 1515 1515 1717 1717 2.1 E < 2.3 - - - - - -

2.3 E < 2.5 - - - - - -

2.5 E < 2.7 - - - - - -

2.7 E < 2.9 - - 29.5 - - -

2.9 E < 3.1 21.7 25.2 28.4 30.3 28.9 28.8 3.1 E < 3.3 21.3 22.9 28.2 30.1 28.7 28.6 3.3 E < 3.5 21.1 22.2 28.0 29.8 28.4 28.4 3.5 E < 3.7 20.7 21.9 27.7 29.7 28.1 28.1 3.7 E < 3.9 20.5 21.7 27.5 29.4 28.0 28.0 3.9 E < 4.1 20.2 21.4 27.3 29.2 27.8 27.7 4.1 E < 4.3 20.0 21.2 27.1 29.0 27.6 27.5 4.3 E < 4.5 19.7 21.0 26.9 28.8 27.4 27.4 4.5 E < 4.7 19.6 20.7 26.7 28.7 27.3 27.2 4.7 E < 4.9 19.4 20.5 26.5 28.5 27.1 27.0 E 4.9 19.2 20.4 26.3 28.3 27.0 26.9 Minimum Initial 50 < Assembly Average Burnup 55 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W WE B&W wt % 235U (E) 1414 1414 1515 1515 1717 1717 2.1 E < 2.3 - - - - - -

2.3 E < 2.5 - - - - - -

2.5 E < 2.7 - - - - - -

2.7 E < 2.9 - - - - - -

2.9 E < 3.1 - - - - - -

3.1 E < 3.3 26.8 31.7 33.8 35.8 34.9 34.9 3.3 E < 3.5 26.4 29.4 33.5 35.5 34.7 34.6 3.5 E < 3.7 26.2 28.1 33.3 35.3 34.5 34.4 3.7 E < 3.9 25.9 27.9 33.1 35.1 34.4 34.2 3.9 E < 4.1 25.7 27.6 32.9 34.9 34.1 34.1 4.1 E < 4.3 25.4 27.4 32.8 34.8 34.0 33.9 4.3 E < 4.5 25.1 27.2 32.5 34.6 33.9 33.8 4.5 E < 4.7 25.0 26.9 32.4 34.5 33.7 33.7 4.7 E < 4.9 24.7 26.7 32.3 34.3 33.5 33.4 E 4.9 24.5 26.6 32.0 34.2 33.4 33.3

1. - means not allowed NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 10Loading Table for High Burnup PWR Fuel - 21.85 kW/Package1 (continued)

Minimum Initial 55 < Assembly Average Burnup 60 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W WE B&W wt % 235U (E) 1414 1414 1515 1515 1717 1717 2.1 E < 2.3 - - - - - -

2.3 E < 2.5 - - - - - -

2.5 E < 2.7 - - - - - -

2.7 E < 2.9 - - - - - -

2.9 E < 3.1 - - - - - -

3.1 E < 3.3 - - 41.3 - - -

3.3 E < 3.5 32.5 37.6 39.1 41.2 39.8 39.7 3.5 E < 3.7 31.7 35.4 38.9 41.0 39.6 39.5 3.7 E < 3.9 31.4 33.6 38.7 41.0 39.4 39.3 3.9 E < 4.1 31.1 33.4 38.6 40.8 39.3 39.1 4.1 E < 4.3 30.9 33.2 38.4 40.6 39.1 39.0 4.3 E < 4.5 30.7 33.0 38.3 40.4 38.9 38.9 4.5 E < 4.7 30.5 32.8 38.1 40.3 38.8 38.7 4.7 E < 4.9 30.3 32.6 38.0 40.1 38.7 38.6 E 4.9 30.0 32.4 37.9 40.1 38.6 38.5

1. - means not allowed

Table 11Low Burnup PWR Fuel Loading Table - 23 kW/Package

Max.

Assembly Min. Assembly Minimum Avg. Avg. Initial Cool Burnup Enrichment Time

[MWd/MTU] [wt% 235U] [Years]

10,000 1.3 4.0 15,000 1.5 4.0 20,000 1.7 4.4 25,000 1.9 5.5 30,000 2.1 6.9 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 12Loading Table for PWR Fuel - 22 kW/Package1

Minimum Initial Assembly Average Burnup 30 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1 E < 2.3 6.0 6.1 7.1 7.4 6.6 7.2 7.2 2.3 E < 2.5 5.9 6.0 7.0 7.3 6.6 7.0 7.1 2.5 E < 2.7 5.9 6.0 7.0 7.2 6.5 7.0 7.0 2.7 E < 2.9 5.8 5.9 6.9 7.2 6.4 6.9 6.9 2.9 E < 3.1 5.8 5.9 6.8 7.1 6.4 6.9 6.9 3.1 E < 3.3 5.7 5.8 6.8 7.0 6.3 6.9 6.9 3.3 E < 3.5 5.7 5.8 6.8 7.0 6.3 6.8 6.8 3.5 E < 3.7 5.6 5.7 6.7 7.0 6.2 6.8 6.8 3.7 E < 3.9 5.6 5.7 6.7 6.9 6.2 6.7 6.7 3.9 E < 4.1 5.6 5.7 6.6 6.9 6.1 6.7 6.7 4.1 E < 4.3 5.5 5.6 6.6 6.9 6.1 6.7 6.7 4.3 E < 4.5 5.5 5.6 6.6 6.8 6.0 6.6 6.6 4.5 E < 4.7 5.5 5.6 6.5 6.8 6.0 6.6 6.6 4.7 E < 4.9 5.4 5.6 6.5 6.8 6.0 6.6 6.6 E 4.9 5.4 5.5 6.5 6.7 6.0 6.6 6.6 Minimum Initial 30 < Assembly Average Burnup 35 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1 E < 2.3 - - 10.1 - - - -

2.3 E < 2.5 7.7 7.9 9.9 10.7 8.8 10.0 10.0 2.5 E < 2.7 7.5 7.8 9.8 10.6 8.7 9.9 9.9 2.7 E < 2.9 7.4 7.7 9.7 10.4 8.6 9.7 9.7 2.9 E < 3.1 7.3 7.6 9.5 10.2 8.5 9.6 9.6 3.1 E < 3.3 7.2 7.5 9.4 10.1 8.4 9.5 9.5 3.3 E < 3.5 7.2 7.4 9.3 10.0 8.3 9.4 9.4 3.5 E < 3.7 7.1 7.4 9.2 9.9 8.2 9.3 9.3 3.7 E < 3.9 7.0 7.3 9.1 9.8 8.1 9.3 9.2 3.9 E < 4.1 7.0 7.2 9.1 9.7 8.1 9.1 9.2 4.1 E < 4.3 6.9 7.2 9.0 9.6 8.0 9.1 9.1 4.3 E < 4.5 6.9 7.1 9.0 9.6 8.0 9.0 9.0 4.5 E < 4.7 6.9 7.0 8.9 9.5 7.9 9.0 9.0 4.7 E < 4.9 6.8 7.0 8.8 9.5 7.9 9.0 9.0 E 4.9 6.8 7.0 8.8 9.4 7.9 8.9 8.9

1. - means not allowed NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 12Loading Table for PWR Fuel - 22 kW/Package1 (continued)

Minimum Initial 35 < Assembly Average Burnup 40 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 10.7 11.9 15.2 16.6 13.1 15.4 15.4 2.7 E < 2.9 10.5 11.2 14.9 16.2 12.9 15.2 15.1 2.9 E < 3.1 10.3 11.0 14.7 16.0 12.6 14.8 14.8 3.1 E < 3.3 10.1 10.8 14.4 15.8 12.4 14.7 14.7 3.3 E < 3.5 9.9 10.6 14.2 15.6 12.2 14.4 14.5 3.5 E < 3.7 9.8 10.4 14.1 15.4 12.0 14.3 14.2 3.7 E < 3.9 9.7 10.3 13.9 15.3 11.9 14.2 14.1 3.9 E < 4.1 9.6 10.1 13.7 15.1 11.8 14.0 14.0 4.1 E < 4.3 9.5 10.0 13.6 15.0 11.7 13.9 13.9 4.3 E < 4.5 9.4 10.0 13.5 14.8 11.6 13.7 13.8 4.5 E < 4.7 9.3 9.9 13.5 14.8 11.6 13.7 13.6 4.7 E < 4.9 9.2 9.8 13.3 14.6 11.5 13.6 13.6 E 4.9 9.2 9.7 13.3 14.5 11.5 13.5 13.5 Minimum Initial 40 < Assembly Average Burnup 45 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W CE WE B&W wt % 235U (E) 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 - - - - - - -

2.7 E < 2.9 15.7 19.0 21.7 23.5 19.2 22.1 22.1 2.9 E < 3.1 15.3 16.7 21.4 23.2 18.8 21.8 21.8 3.1 E < 3.3 15.0 16.2 21.1 22.9 18.6 21.5 21.5 3.3 E < 3.5 14.8 15.9 20.9 22.6 18.3 21.3 21.3 3.5 E < 3.7 14.5 15.7 20.7 22.4 18.0 21.1 21.0 3.7 E < 3.9 14.2 15.5 20.4 22.2 17.8 20.8 20.8 3.9 E < 4.1 14.0 15.3 20.2 22.0 17.6 20.6 20.6 4.1 E < 4.3 13.9 15.0 20.0 21.8 17.5 20.5 20.4 4.3 E < 4.5 13.7 14.8 19.8 21.6 17.3 20.3 20.3 4.5 E < 4.7 13.6 14.7 19.7 21.5 17.1 20.1 20.1 4.7 E < 4.9 13.5 14.5 19.6 21.3 17.0 20.0 19.9 E 4.9 13.4 14.4 19.5 21.2 16.9 19.8 19.9

1. - means not allowed 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 NUMBERc. DOCKET NUMBERd. PACKAGE IDENTIFICATION NUMBERPAGE PAGES 9356 4 71-9356 USA/9356/B(U)F-96 20 OF 43

5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 13Loading Table for High Burnup PWR Fuel - 20.9 kW/Package 1

Minimum Initial 45 < Assembly Average Burnup 50 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W WE B&W wt % 235U (E) 14X14 14X14 15x15 15x15 17X17 17X17 2.1 E < 2.3 - - - - - -

2.3 E < 2.5 - - - - - -

2.5 E < 2.7 - - - - - -

2.7 E < 2.9 - - 31.0 - - -

2.9 E < 3.1 23.8 25.2 30.7 32.7 31.3 31.2 3.1 E < 3.3 23.5 24.7 30.5 32.5 31.0 31.0 3.3 E < 3.5 23.2 24.4 30.2 32.2 30.8 30.8 3.5 E < 3.7 22.9 24.1 30.0 32.1 30.6 30.5 3.7 E < 3.9 22.6 23.9 29.8 31.9 30.4 30.3 3.9 E < 4.1 22.4 23.6 29.6 31.7 30.2 30.1 4.1 E < 4.3 22.2 23.4 29.4 31.5 30.0 29.9 4.3 E < 4.5 22.0 23.2 29.3 31.3 29.9 29.8 4.5 E < 4.7 21.8 23.0 29.1 31.2 29.7 29.6 4.7 E < 4.9 21.6 22.8 28.9 31.0 29.6 29.5 E 4.9 21.4 22.7 28.7 30.8 29.4 29.3 Minimum Initial 50 < Assembly Average Burnup 55 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W WE B&W wt % 235U (E) 14X14 14X14 15x15 15x15 17X17 17X17 2.1 E < 2.3 - - - - - -

2.3 E < 2.5 - - - - - -

2.5 E < 2.7 - - - - - -

2.7 E < 2.9 - - - - - -

2.9 E < 3.1 - - - - - -

3.1 E < 3.3 28.9 31.7 36.1 38.1 37.3 37.2 3.3 E < 3.5 28.7 30.7 35.8 38.0 37.1 37.0 3.5 E < 3.7 28.3 30.4 35.7 37.8 36.9 36.8 3.7 E < 3.9 28.1 30.2 35.4 37.6 36.8 36.6 3.9 E < 4.1 27.9 29.9 35.2 37.4 36.6 36.5 4.1 E < 4.3 27.6 29.7 35.1 37.3 36.4 36.3 4.3 E < 4.5 27.4 29.5 34.8 37.1 36.3 36.2 4.5 E < 4.7 27.2 29.3 34.7 37.0 36.2 36.1 4.7 E < 4.9 27.1 29.1 34.6 36.8 36.0 35.9 E 4.9 26.9 28.9 34.4 36.7 35.8 35.7

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 13Loading Table for High Burnup PWR Fuel - 20.9 kW/Package1 (continued)

Minimum Initial 55 < Assembly Average Burnup 60 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment CE WE WE B&W WE B&W wt % 235U (E) 14X14 14X14 15x15 15x15 17X17 17X17

2.1 E < 2.3 - - - - - -

2.3 E < 2.5 - - - - - -

2.5 E < 2.7 - - - - - -

2.7 E < 2.9 - - - - - -

2.9 E < 3.1 - - - - - -

3.1 E < 3.3 - - 41.6 - - -

3.3 E < 3.5 34.1 37.6 41.4 43.7 42.2 42.1 3.5 E < 3.7 33.8 36.1 41.3 43.5 42.0 41.9 3.7 E < 3.9 33.6 35.9 41.1 43.4 41.8 41.7 3.9 E < 4.1 33.4 35.7 41.0 43.3 41.8 41.6 4.1 E < 4.3 33.2 35.5 40.8 43.1 41.6 41.5 4.3 E < 4.5 33.0 35.3 40.7 43.0 41.4 41.3 4.5 E < 4.7 32.9 35.1 40.5 42.9 41.3 41.2 4.7 E < 4.9 32.7 35.0 40.4 42.8 41.2 41.1 E 4.9 32.5 34.8 40.3 42.7 41.1 41.0

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5.(b)(1)(i) Contents - Type and Form of Material (continued)

Table 14Low Burnup PWR Fuel Loading Table - 22 kW/Package

Max. Min. Assembly Assembly Avg. Initial Minimum Avg. Enrichment Cool Time Burnup [wt% 235U] [Years]

[MWd/MTU]

10,000 1.3 4.0 15,000 1.5 4.0 20,000 1.7 4.5 25,000 1.9 5.7 30,000 2.1 7.4

Table 15Additional Fuel Assembly Cool Time to Load Nonfuel Hardware (Reduced Heat Load - 22kW PWR - Configuration)

BPRA/

Assembly HFRA GTPD RCC NSA CE 14x14 - - 0.4 -

WE 14x14 1.1 0.1 0.3 1.1 WE 15x15 1.5 0.2 7.6 1.5 BW 15x15 0.1 0.2 0.3 0.2 CE 16x16 - - 0.4 -

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5.(b)(1) Contents - Type and Form of Material (continued)

(ii) Undamaged and damaged PWR assemblies

A combination of damaged and undamaged PWR fuel assemblies in the 37 PWR damaged fuel basket, shown in Figure 2, in a short TSC. Undamaged, low burnup fuel assemblies must meet the description for PWR fuel in 5.(b)(1)(i). Up to four damaged fuel assemblies, or fuel material that is less than, or equivalent to, one undamaged PWR fuel assembly must be placed in a damaged fuel can and must be placed in locations 4, 8, 30 and 34 in the PWR damaged fuel basket. Undamaged, low burnup fuel may also be placed in the 4 damaged fuel locations, without the use of a damaged fuel can. Prior to irradiation, the damaged and undamaged fuel assemblies must be within the dimensions and specifications of the hybrid assemblies listed in Table 1 and meet the fuel class assembly specifications listed in Table 2.

For the 33 non-damaged fuel can locations in the damaged fuel basket, the fuel must meet the class enrichment, post-irradiation cooling time, burnup credit loading curves, and the TSC neutron absorber sheet 10B density in Table 16. For the loading profiles up to the 33, 35 and 36 assembly loading pattern, the PWR fuel must meet the burnup loading curves in Table 17. A short-loaded 33, 35 or 36 assembly loading profile may still use the burnup credit curve in Table 17 provided the required cell locations for that profile shown in Figure 1 are left empty, at a minimum. For a TSC with a damaged fuel basket assembly that does not contain any damaged fuel, the fuel class enrichment, post-irradiation cooling time, burnup credit loading curves, and the TSC neutron absorber sheet 10B density in Tables 3 and 4 may be used for all locations (burnup credit curves are only applicable to systems not crediting moderator exclusion).

Fuel assembly burnup, minimum initial average enrichment9, and cool time requirements are provided in Tables 9-11 for PWR baskets with Type 2 neutron absorbers and Tables 12-14 for baskets with Type 1 neutron absorbers. For TSCs containing damaged fuel, all fuel assemblies in the TSC must meet the additional cool time requirements in Table 18 for the assembly type that is loaded in the damaged fuel can. If two types of fuel assemblies are loaded in different damaged fuel cans in a single TSC, the longest additional fuel cooling time applies to all fuel assemblies in the TSC. The additional cool time requirements in Table 18 apply to assemblies loaded in TSC baskets with Type 1 or Type 2 neutron absorbers. Damaged CE 16x16 fuel assemblies are not authorized for shipment.

The fuel assemblies consist of uranium dioxide pellets with zirconium alloy-clad fuel rods and zirconium alloy instrument and guide tubes. Empty fuel rod positions for undamaged fuel assemblies are to be filled with a solid filler rod or a solid neutron absorber rod that displaces a volume equal to or greater than the original rod. PWR fuel assemblies containing nonfuel hardware may be loaded in the TSC.

9 Assembly average fuel enrichment is the enrichment value determined by averaging the 235U wt% enrichment over the entire fuel region (UO2) of an individual assembly, including axial blankets, if present.

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5.(b)(1)(ii) Contents - Type and Form of Material (continued)

Unirradiated fuel and unenriched fuel are not authorized for loading, except that unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than 6 inches. An unenriched rod may be used as a replacement rod to return a fuel assembly to an undamaged condition. Damaged fuel located in a damaged fuel can location in the damaged fuel basket must have a minimum burnup of 5 GWd/MTU, a maximum enrichment of 4.05 wt. % 235U, and a minimum cool time of 15 years. PWR fuel assemblies loaded in a damaged fuel can must not contain nonfuel hardware with the exception of instrument tube tie components, guide tube anchors or similar devices, and steel inserts. Application of moderator exclusion allows increasing the maximum initial enrichment to 5 wt% 235U, with no burnup requirement.

Undamaged PWR fuel assemblies may contain nonfuel hardware, while damaged PWR fuel assemblies shall not, with the exception of the following unirradiated nonfuel hardware:

instrument tube tie components, guide tube anchors or similar devices, and steel inserts.

The nonfuel hardware may be loaded as a complete assembly or as individual components, individual nonfuel rods may be full-length rods or partial-length rods/rodlets. Partial-length rods/rodlets are permitted in guide tubes provided guide tube plug devises are installed. Fuel assemblies with an instrument tube tie rod repair shall be loaded with fuel inserts and/or top spacers to ensure proper spacing and support of the fuel assembly. Fuel inserts and/or top spacers are not required when using the extended fuel tube basket because the top nozzle is adequately supported. Nonfuel hardware must meet the exposure and cool time or cobalt-60 activity requirements listed in Tables 6-8.

HFRAs are only allowed for Westinghouse assemblies and may have a maximum exposure of 4.0 GWd/MTU and must have a minimum cool time of 16 years. Fuel assemblies loaded with nonfuel hardware must meet the additional cool time requirements of Table 5 (for Type 2 neutron absorbers), and Table 15 (for Type 1 neutron absorbers). Fuel assemblies may contain any number of unirradiated nonfuel solid filler fuel replacement rods. Activated stainless steel fuel replacement rods are limited to 5 steel rods per assembly, 1 assembly per basket, and a maximum steel rod exposure of 32.5 GWd/MTU. Fuel assemblies with activated stainless steel rods must be cooled for either a minimum of 21 years or the loading table minimum cool time (as adjusted for additional cool times for nonfuel hardware and the presence of damaged fuel in the TSC, as applicable) plus 1 year, whichever is greater.

Fuel assemblies loaded with in-core instrument thimbles must meet the additional cool time requirements in Table 5 or Table 15, as appropriate, for BPRAs or GTPDs, whichever is bounding, for Westinghouse and B&W fuel types and for RCCs for CE fuel types. The additional cool time requirements for assemblies with nonfuel hardware are added to any additional cool time requirements due to damaged fuel also being loaded in the same TSC. RCCs are restricted to fuel storage locations 11, 12, 13, 18, 19, 20, 25, 26 and 27 in Figure 1.

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5.(b)(1)(ii) Contents - Type and Form of Material (continued)

One NSA is permitted to be loaded in a TSC in fuel storage locations 11, 12, 13, 18, 19, 20, 25, 26 or 27 in Figure1. NSAs may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rodlets and/or thimble plug rodlets. NSAs, GTPDs, and BPRAs for CE fuel types are not allowed contents. In addition, the following unirradiated, nonfuel hardware may be loaded with the fuel assemblies: stainless steel rods inserted to displace guide tube dashpot water, instrument tube tie components, and guide tube anchors or similar devices. Axial power shaping rods are not allowed contents.

Under-burned Westinghouse 15x15 assemblies (assemblies with a maximum enrichment greater than that dictated by the burnup credit loading curve) may be loaded provided that an RCCA is inserted in the assembly, the enrichment is equal to or less than 4.05 wt. % 235U, and the assembly burnup is greater than or equal to 12,000 MWd/MTU. When loading under-burned fuel, the RCCAs must be full length Ag-In-Cd RCCAs comprised of stainless steel clad rods constructed with 80% Ag, 15% In and 5% Cd absorber pellets and having an exposure equal to or less than 200,000 MWd/MTU. The basket must include absorber sheets with an effective 10B areal density of 0.036 g/cm2. Any assemblies loaded without an RCCA inserted must meet the burnup credit loading curve for the applicable assembly loading profile. Burnup credit curves, and the criticality requirement for RCCA insertion, are only applicable to systems not crediting moderator exclusion. Initial enrichment up to 5 wt% 235U, with no burnup or RCCA requirement, is permitted when crediting moderator exclusion.

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5.(b)(1)(ii) Contents - Type and Form of Material (continued)

Table 16Maximum Initial Enrichment - 37 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt % 235U) 10B Zero (0) = C4 x Burnup (GWd/MTU) + C5 Assembly Absorber Burnup Burnup 18 Burnup 30 < Burnup 50 < Burnup ID (g/cm2) Max. Enr. (GWd/MTU) < (GWd/MTU) 30 (GWd/MTU) 50 (GWd/MTU)

(wt %) 18 C4 C5 C4 C5 C4 C5 C4 C5 BW15 1.6 0.0453 1.42 0.0681 1.29 0.0750 1.03 0.0750 0.736 BW17 1.6 0.0476 1.45 0.0668 1.37 0.0712 1.17 0.0712 0.891 CE14 1.9 0.0504 1.79 0.0696 1.75 0.0751 1.60 0.0751 1.60 CE16 0.036 1.9 0.0484 1.79 0.0679 1.74 0.0758 1.52 0.0758 1.52 WE14 1.9 0.0542 1.85 0.0729 1.85 0.0794 1.75 0.0794 1.75 WE15 1.6 0.0482 1.43 0.0692 1.27 0.0738 1.08 0.0738 0.767 WE17 1.6 0.0439 1.45 0.0657 1.35 0.0732 1.00 0.0732 0.700 BW15 1.5 0.0487 1.31 0.0660 1.26 0.0740 0.896 0.0740 0.614 BW17 1.5 0.0470 1.37 0.0673 1.29 0.0745 0.937 0.0745 0.655 CE14 1.8 0.0494 1.71 0.0705 1.64 0.0781 1.37 0.0781 1.37 CE16 0.030 1.8 0.0489 1.71 0.0679 1.68 0.0724 1.52 0.0724 1.52 WE14 1.9 0.0533 1.82 0.0725 1.76 0.0821 1.50 0.0821 1.50 WE15 1.6 0.0475 1.35 0.0661 1.29 0.0746 0.859 0.0746 0.575 WE17 1.6 0.0448 1.38 0.0646 1.26 0.0710 0.968 0.0710 0.691 BW15 1.5 0.0471 1.30 0.0666 1.19 0.0725 0.857 0.0725 0.581 BW17 1.5 0.0474 1.36 0.0652 1.27 0.0724 0.918 0.0724 0.639 CE14 1.8 0.0486 1.68 0.0696 1.61 0.0778 1.32 0.0778 1.32 CE16 0.027 1.8 0.0493 1.66 0.0660 1.64 0.0761 1.33 0.0761 1.33 WE14 1.8 0.0535 1.71 0.0694 1.75 0.0805 1.52 0.0805 1.52 WE15 1.5 0.0465 1.33 0.0664 1.24 0.0710 0.968 0.0710 0.685 WE17 1.5 0.0447 1.31 0.0647 1.25 0.0714 0.846 0.0714 0.564 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(ii) Contents - Type and Form of Material (continued)

Table 17Maximum Initial Enrichment - WE 15x15 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt % 235U) 10B Zero (0) = C4 x Burnup (GWd/MTU) + C5 Number of Absorber Burnup Burnup 18 Burnup 30 < Burnup 50 <

Assemblies (g/cm2) Max. Enr. (GWd/MTU) (GWd/MTU) (GWd/MTU) Burnup (wt %) < 18 30 50 (GWd/MTU)

C4 C5 C4 C5 C4 C5 C4 C5 36 1.6 0.0483 1.53 0.0721 1.35 0.0750 1.17 0.0750 0.851 35 0.036 1.7 0.0532 1.51 0.0722 1.45 0.0778 1.14 0.0778 1.14 33 1.7 0.0524 1.60 0.0734 1.52 0.0791 1.22 0.0791 1.22 36 1.6 0.0483 1.48 0.0707 1.32 0.0739 1.15 0.0739 0.811 35 0.030 1.6 0.0499 1.48 0.0722 1.34 0.0733 1.20 0.0733 0.847 33 1.7 0.0523 1.52 0.0728 1.40 0.0780 1.19 0.0780 1.19 36 1.6 0.0473 1.42 0.0668 1.33 0.0731 1.02 0.0731 0.693 35 0.027 1.6 0.0477 1.46 0.0736 1.27 0.0738 1.13 0.0738 0.775 33 1.7 0.0491 1.51 0.0718 1.41 0.0784 1.09 0.0784 1.09

Table 18Additional Cool Time Required for Damaged PWR Fuel Contents Max Assembly Min. Initial CE 14x14 WE 14x14 WE 15x15 WE 17x17 Average Burnup Assembly Average Cool Time Cool Time Cool Time Cool Time (GWd/MTU) Enrichment (wt% (years) (years) (years) (years) 235U) 35 2.1 N/A N/A 2.5 N/A 2.3 0.0 0.6 0.8 0.3 2.5 1.9 2.4 3.3 2.8 40 2.7 0.3 2.1 1.2 0.8 2.9 0.0 0.6 0.0 0.0 2.7 3.9 2.6 4.5 4.2 45 2.9 2.5 2.6 2.7 2.2 3.1 0.6 2.5 0.7 0.1 3.3 0.0 1.0 0.0 0.0 2.7 N/A N/A 4.8 N/A 50 2.9 3.6 2.8 3.5 2.8 3.1 1.7 2.8 1.2 0.5 3.3 0.0 1.2 0.0 0.0 3.1 4.2 2.9 4.0 3.6 55 3.3 2.2 3.0 1.9 1.5 3.5 0.2 2.0 0.0 0.0 3.1 N/A N/A 5.0 N/A 3.3 4.6 3.0 4.9 4.1 60 3.5 3.1 3.1 2.9 2.1 3.7 1.3 2.8 0.8 0.0 3.9 0.0 0.9 0.0 0.0 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(ii) Contents - Type and Form of Material (continued)

Figure 2Damaged Fuel Basket Loading Profile

DFC designated locations may contain a loaded DFC or an undamaged PWR fuel assembly.

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5.(b)(1) Type and Form of Material (continued)

(iii) Undamaged BWR assemblies

Undamaged BWR fuel assemblies within the 87 BWR basket assembly shown in Figure

3. Undamaged fuel is spent nuclear fuel that does not have any visible deformation other than uniform bowing that occurs in the reactor, assemblies that do not have missing rods, and assemblies with missing rods that are replaced by solid stainless steel or zirconium filler rods that displace a volume equal to or greater than the original rods and assemblies that do not contain structural defects that adversely affect radiological and/or criticality safety and/or result in unsupported fuel rod lengths in excess of 60 inches and that can be handled by normal means. BWR/2-3 assemblies are to be loaded into short TSCs, and BWR/4-6 assemblies are to be loaded into long TSCs.

The fuel assemblies consist of uranium dioxide pellets with zirconium alloy-clad fuel rods and zirconium alloy-clad water rods or holes. Empty rod positions must be filled with solid, unirradiated, nonfuel filler rods that displace a volume equal to, or greater than, that of the fuel rod that the filler rod replaces. Prior to irradiation, the fuel assemblies must be within the dimensions and specifications of the hybrid assemblies listed in Table 19. In addition, the BWR fuel must meet the fuel class assembly specifications listed in Table 20. Fuel assembly burnup, minimum initial average enrichment10, and cool time requirements are provided in Table 23, Table 24, and Table 25.

Undamaged BWR fuel must meet the hybrid fuel assembly enrichment and the TSC neutron absorber sheet 10B density for loading up to the 87 and 82 assembly loading patterns for fuel with axial blankets in Table 21 and fuel without axial blankets in Table

22. Spacers may be used to axially position fuel assemblies to limit their axial movement in the TSC. Unenriched and unirradiated fuel is not authorized for loading, except that unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than 6 inches.

For a TSC that is less than fully loaded, empty fuel storage locations shall begin with location 44, followed by locations 43, 45, 33, 55, 32, 56, 34, 54 and continuing outward, as required, in an approximately symmetric pattern as shown in Figure 3. Allowable fuel assembly locations for the 82 assembly BWR fuel assembly basket configuration are shown in Figure 4. Prior to use of the 82 assembly configuration, the center cell weldment and upper weldments of nonfuel locations must be physically blocked (fuel storage locations 44, 32, 34, 54, 56 shown as in Figure 4).

BWR fuel assemblies may be unchanneled, or channeled with zirconium-based alloy channels. BWR fuel assemblies with stainless steel channels are not authorized.

10 Assembly average fuel enrichment is the enrichment value determined by averaging the 235U wt% enrichment over the entire fuel region (UO2) of an individual assembly, including axial blankets, if present.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

The 82-Assembly configuration is the result of criticality constraints on maximum enrichment. When crediting moderator exclusion, the 82-Assembly configuration is not required, as full capacity (87-Assembly) is permitted at an initial enrichment up to 5 wt%

235U.

Table 19BWR Hybrid Fuel Assembly Characteristics Geometry2,3 Number Number Max Min Min Max Max Max Assembly of Fuel of Partial Pitch Clad Clad Pellet Active Loading Type Rods Length (inch) OD Thick. OD Length (MTU)

Rods1 (inch) (inch) (inch) (inch)

B7_48A 48 N/A 0.7380 0.5700 0.03600 0.4900 144.0 0.1981 B7_49A 49 N/A 0.7380 0.5630 0.03200 0.4880 146.0 0.2034 B7_49B 49 N/A 0.7380 0.5630 0.03200 0.4910 150.0 0.2037 B8_59A 59 N/A 0.6400 0.4930 0.03400 0.4160 150.0 0.1828 B8_60A 60 N/A 0.6417 0.4840 0.03150 0.4110 150.0 0.1815 B8_60B 60 N/A 0.6400 0.4830 0.03000 0.4140 150.0 0.1841 B8_61B 61 N/A 0.6400 0.4830 0.03000 0.4140 150.0 0.1872 B8_62A 62 N/A 0.6417 0.4830 0.02900 0.4160 150.0 0.1921 B8_63A 63 N/A 0.6420 0.4840 0.02725 0.4195 150.0 0.1985 B8_64A 64 N/A 0.6420 0.4840 0.02725 0.4195 150.0 0.1996 B8_64B4 64 N/A 0.6090 0.4576 0.02900 0.3913 150.0 0.1755 B9_72A 72 N/A 0.5720 0.4330 0.02600 0.3740 150.0 0.1803 B9_74A 741 8 0.5720 0.4240 0.02390 0.3760 150.0 0.1873 B9_76A 76 N/A 0.5720 0.4170 0.02090 0.3750 150.0 0.1914 B9_79A 79 N/A 0.5720 0.4240 0.02390 0.3760 150.0 0.1979 B9_80A 80 N/A 0.5720 0.4230 0.02950 0.3565 150.0 0.1821 B10_91A 911 8 0.5100 0.3957 0.02385 0.3420 150.0 0.1906 B10_92A 921 14 0.5100 0.4040 0.02600 0.3455 150.0 0.1946 B10_96A4 961 12 0.4880 0.3780 0.02430 0.3224 150.0 0.1787 B10_100A4 100 N/A 0.4880 0.3780 0.02430 0.3224 150.0 0.1861

Notes:

1 Assemblies may contain partial-length fuel rods.

2 Assembly characteristics represent cold, unirradiated, nominal configurations.

3 Maximum channel thickness allowed is 120 mils (nominal).

4 Composed of four subchannel clusters.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 20BWR Fuel Class Assembly Characteristics Characteristic Fuel Class 7x7 8x8 9x9 10x10 Base Fuel Type11 SPC, GE SPC, GE SPC, GE SPC, GE, ABB Max Initial Enrichment (wt% 235U) 4.5 4.5 4.5 4.5 59 72 60 7413 9113 Number of Fuel Rods 48 61 76 9213 49 62 79 9613, 14 63 80 10014 6412 Max Assembly Average Burnup (MWd/MTU) 60,000 60,000 60,000 60,000 Min Cool Time (years) 4 4 4 4 Min Average Enrichment (wt% 235U)15 1.3 1.3 1.3 1.3 Max Weight (lb) per Storage Location See Note 1 See Note 1 See Note 1 See Note 1 Max Decay Heat (Watts) per Fuel Location 253 253 253 253

Notes:

1. Maximum weight per storage location is 739 lbs (including fuel spacers and channel) with a maximum contents weight of 62,656 lbs. The maximum nominal assembly length is 176.2 inches for BWR/4-6 assemblies and 171 inches for the BWR/2-3 assemblies, and the maximum nominal assembly width is 5.52 inches.

2. Fuel assembly weight includes the weight of the channel.

3. Maximum initial enrichment is the peak planar-average enrichment.

4. Water rods may occupy more than one fuel lattice location. Fuel assembly to contain nominal number of water rods for the specific assembly design.

5. All enrichment values are nominal pre-irradiation fabrication values.

6. Spacers may be used to axially position fuel assemblies to facilitate handling.

7. Each BWR fuel assembly may have a zirconium-based alloy channel 0.120 inches thick.

11 Indicates assembly vendor/type referenced for fuel input data. Fuel acceptability for loading is not restricted to the indicated vendor/type provided that the fuel assembly meets the limits listed in Table 6.2.1-1. Table 6.2.1-2 contains vendor information by fuel rod array. Abbreviations are as follows: General Electric/Global Nuclear Fuels (GE), Exxon/Advanced Nuclear Fuels/Siemens Power Corporation (SPC).

12 May be composed of four subchannel clusters.

13 Assemblies may contain partial-length fuel rods.

14 Composed of four subchannel clusters 15 Assembly average burnup is the burnup value determined by averaging the burnup over the entire fuel region (UO2) of an individual assembly, including axial blankets, if present.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 21Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel With Axial Blankets)

Max. Initial Enrichment16 (wt % 235U)

Fuel Absorber17 0.027 10B g/cm2 Absorber17 0.0225 10B g/cm2 Absorber17 0.02 10B g/cm2 Type 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Basket Basket Basket Basket Basket Basket B7_48A 4.0% 4.5% 3.7% 4.5% 3.6% 4.4%

B7_49A 3.8% 4.5% 3.6% 4.4% 3.5% 4.3%

B7_49B 3.8% 4.5% 3.6% 4.4% 3.5% 4.2%

B8_59A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%

B8_60A 3.8% 4.5% 3.7% 4.4% 3.5% 4.2%

B8_60B 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%

B8_61B 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%

B8_62A 3.8% 4.5% 3.6% 4.3% 3.5% 4.1%

B8_63A 3.8% 4.5% 3.6% 4.3% 3.4% 4.2%

B8_64A 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%

B8_64B 3.6% 4.3% 3.4% 4.1% 3.3% 4.0%

B9_72A 3.8% 4.5% 3.6% 4.3% 3.4% 4.1%

B9_74A 3.7% 4.3% 3.4% 4.1% 3.4% 4.0%

B9_76A 3.5% 4.2% 3.4% 4.0% 3.3% 3.9%

B9_79A 3.7% 4.4% 3.4% 4.2% 3.3% 4.0%

B9_80A 3.8% 4.5% 3.6% 4.3% 3.5% 4.2%

B10_91A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%

B10_92A 3.8% 4.5% 3.6% 4.3% 3.5% 4.1%

B10_96A 3.7% 4.3% 3.5% 4.1% 3.4% 4.0%

B10_100A 3.6% 4.4% 3.5% 4.1% 3.4% 4.0%

Note: When crediting moderator exclusion, the maximum allowed initial enrichment is 5 wt% 235U for all basket/absorber combinations.

16 Maximum planar average.

17 Borated aluminum neutron absorber sheet effective areal 10B density.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 22Undamaged BWR Fuel Assembly Loading Criteria (Enrichment Limits for Fuel Without Axial Blankets)

Max. Initial Enrichment18 ( wt % 235U)

Fuel Absorber19 0.027 10B g/cm2 Absorber19 0.0225 10B g/cm2 Absorber19 0.02 10B g/cm2 Type 87-Assy 82-Assy 87-Assy 82-Assy 87-Assy 82-Assy Basket Basket Basket Basket Basket Basket B7_48A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3%

B7_49A 3.7% 4.5% 3.6% 4.3% 3.4% 4.1%

B7_49B 3.7% 4.5% 3.6% 4.3% 3.5% 4.2%

B8_59A 3.8% 4.5% 3.7% 4.4% 3.5% 4.3%

B8_60A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%

B8_60B 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%

B8_61B 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%

B8_62A 3.6% 4.4% 3.5% 4.2% 3.4% 4.1%

B8_63A 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%

B8_64A 3.7% 4.5% 3.5% 4.3% 3.4% 4.1%

B8_64B 3.6% 4.2% 3.4% 4.1% 3.3% 4.0%

B9_72A 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%

B9_74A 3.6% 4.2% 3.4% 4.1% 3.3% 4.0%

B9_76A 3.5% 4.1% 3.3% 4.0% 3.2% 3.8%

B9_79A 3.5% 4.2% 3.4% 4.1% 3.2% 3.9%

B9_80A 3.7% 4.5% 3.6% 4.3% 3.5% 4.1%

B10_91A 3.7% 4.4% 3.5% 4.2% 3.4% 4.1%

B10_92A 3.7% 4.4% 3.6% 4.2% 3.4% 4.1%

B10_96A 3.6% 4.2% 3.4% 4.1% 3.4% 4.0%

B10_100A 3.6% 4.3% 3.4% 4.0% 3.3% 3.9%

Note: When crediting moderator exclusion, the maximum allowed initial enrichment is 5 wt% 235U for all basket/absorber combinations.

18 Maximum planar average.

19 Borated aluminum neutron absorber sheet effective areal 10B density.

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Figure 3Undamaged Fuel Basket 87 Assembly Loading Profile

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Figure 4Undamaged Fuel Basket 82 Assembly Loading Profile

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 23Loading Table for BWR Fuel - 22kW/Package 1

Minimum Initial Assembly Average Burnup 30 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235U (E) 77 77 88 88 99 99 1010 2.1 E < 2.3 6.5 12.3 5.8 13.7 5.3 13.0 13.5 2.3 E < 2.5 6.3 11.6 5.7 13.0 5.2 12.3 12.8 2.5 E < 2.7 6.3 11.0 5.7 12.3 5.1 11.7 12.2 2.7 E < 2.9 6.2 10.3 5.6 11.8 5.1 11.1 11.6 2.9 E < 3.1 6.1 9.8 5.6 11.2 5.0 10.5 11.1 3.1 E < 3.3 6.0 9.3 5.5 10.7 5.0 10.0 10.6 3.3 E < 3.5 6.0 8.8 5.5 10.2 4.9 9.6 10.0 3.5 E < 3.7 6.0 8.4 5.4 9.8 4.9 9.1 9.6 3.7 E < 3.9 5.9 8.0 5.4 9.4 4.9 8.8 9.2 3.9 E < 4.1 5.9 7.7 5.3 9.0 4.8 8.4 8.9 4.1 E < 4.3 5.9 7.4 5.3 8.7 4.8 8.0 8.5 4.3 E < 4.5 5.8 7.0 5.3 8.4 4.8 7.7 8.2 4.5 E < 4.7 5.8 6.8 5.2 8.1 4.7 7.5 7.9 4.7 E < 4.9 5.8 6.5 5.2 7.8 4.7 7.2 7.6 E 4.9 5.7 6.3 5.2 7.6 4.7 6.9 7.4 Minimum Initial 30 < Assembly Average Burnup 35 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235U (E) 77 77 88 88 99 99 1010 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 8.9 14.3 7.6 15.6 6.6 15.0 15.5 2.5 E < 2.7 8.8 13.5 7.5 14.8 6.5 14.1 14.6 2.7 E < 2.9 8.6 12.7 7.3 14.0 6.4 13.4 13.9 2.9 E < 3.1 8.5 12.0 7.2 13.4 6.3 12.7 13.2 3.1 E < 3.3 8.4 11.4 7.2 12.8 6.3 12.1 12.6 3.3 E < 3.5 8.3 10.8 7.1 12.2 6.2 11.5 12.0 3.5 E < 3.7 8.2 10.3 7.0 11.7 6.1 11.0 11.5 3.7 E < 3.9 8.1 9.8 6.9 11.2 6.0 10.6 11.0 3.9 E < 4.1 8.0 9.3 6.9 10.8 6.0 10.1 10.6 4.1 E < 4.3 8.0 8.9 6.9 10.4 6.0 9.7 10.1 4.3 E < 4.5 8.0 8.7 6.8 10.0 6.0 9.3 9.8 4.5 E < 4.7 7.9 8.6 6.8 9.6 5.9 8.9 9.4 4.7 E < 4.9 7.8 8.6 6.7 9.3 5.9 8.6 9.1 E 4.9 7.8 8.6 6.7 9.0 5.9 8.3 8.8

1. - means not allowed NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 23Loading Table for BWR Fuel - 22kW/Package 1 (continued)

Minimum Initial 35 < Assembly Average Burnup 40 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235U (E) 77 77 88 88 99 99 1010 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 14.6 16.9 12.2 18.0 10.0 17.3 17.7 2.7 E < 2.9 13.3 15.8 10.7 17.0 8.7 16.3 16.7 2.9 E < 3.1 13.1 14.9 10.5 16.0 8.5 15.4 15.8 3.1 E < 3.3 12.9 14.1 10.3 15.2 8.4 14.6 15.0 3.3 E < 3.5 12.6 13.9 10.1 14.5 8.3 13.8 14.3 3.5 E < 3.7 12.5 13.7 10.0 13.8 8.2 13.2 13.6 3.7 E < 3.9 12.4 13.6 9.9 13.3 8.0 12.6 13.0 3.9 E < 4.1 12.2 13.5 9.8 12.7 8.0 12.1 12.5 4.1 E < 4.3 12.0 13.4 9.7 12.3 7.9 12.0 11.9 4.3 E < 4.5 11.9 13.3 9.6 12.2 7.9 12.0 11.5 4.5 E < 4.7 11.9 13.1 9.6 12.1 7.8 11.9 11.2 4.7 E < 4.9 11.8 13.0 9.5 12.0 7.8 11.8 11.2 E 4.9 11.8 13.0 9.4 12.0 7.7 11.8 11.1 Minimum Initial 40 < Assembly Average Burnup 45 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235U (E) 77 77 88 88 99 99 1010 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 - - - - - - -

2.7 E < 2.9 22.3 24.0 19.9 22.9 17.4 22.7 21.5 2.9 E < 3.1 19.7 21.4 17.2 20.3 14.8 20.0 19.4 3.1 E < 3.3 18.9 20.5 15.4 19.1 12.3 18.8 18.2 3.3 E < 3.5 18.7 20.2 15.2 18.8 11.9 18.6 17.4 3.5 E < 3.7 18.5 20.0 15.0 18.7 11.7 18.3 17.2 3.7 E < 3.9 18.2 19.9 14.7 18.5 11.5 18.0 17.1 3.9 E < 4.1 18.1 19.6 14.6 18.2 11.4 17.9 16.9 4.1 E < 4.3 17.8 19.5 14.3 18.1 11.3 17.7 16.7 4.3 E < 4.5 17.8 19.4 14.3 18.0 11.2 17.7 16.5 4.5 E < 4.7 17.6 19.2 14.1 17.8 11.1 17.5 16.5 4.7 E < 4.9 17.4 19.0 14.0 17.8 11.0 17.4 16.3 E 4.9 17.3 18.9 13.8 17.7 10.9 17.3 16.2

1. - means not allowed NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 24Loading Table for BWR Fuel - 20.9kW/Package 1

Minimum Initial 45 < Assembly Average Burnup 50 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235U (E) 77 77 88 88 99 99 1010 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 - - - - - - -

2.7 E < 2.9 - - - - - - -

2.9 E < 3.1 29.6 31.5 27.3 30.2 24.9 30.2 29.0 3.1 E < 3.3 27.8 29.6 24.7 27.9 22.2 27.6 26.3 3.3 E < 3.5 27.6 29.3 23.6 27.7 19.6 27.4 26.1 3.5 E < 3.7 27.4 29.0 23.2 27.4 19.0 27.1 25.9 3.7 E < 3.9 27.2 28.9 23.0 27.3 18.7 26.9 25.6 3.9 E < 4.1 26.9 28.6 22.8 27.0 18.5 26.7 25.5 4.1 E < 4.3 26.8 28.6 22.6 27.0 18.4 26.5 25.2 4.3 E < 4.5 26.6 28.3 22.3 26.8 18.2 26.5 25.1 4.5 E < 4.7 26.4 28.1 22.3 26.6 17.9 26.3 25.0 4.7 E < 4.9 26.2 28.0 22.1 26.4 17.9 26.1 24.8 E 4.9 26.0 27.8 22.0 26.4 17.9 25.9 24.7 Minimum Initial 50 < Assembly Average Burnup 55 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235U (E) 77 77 88 88 99 99 1010 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 - - - - - - -

2.7 E < 2.9 - - - - - - -

2.9 E < 3.1 - - - - - - -

3.1 E < 3.3 36.4 38.4 34.1 37.2 31.8 37.2 35.9 3.3 E < 3.5 34.0 358 31.7 34.6 29.2 34.6 33.4 3.5 E < 3.7 33.3 35.0 29.1 33.4 26.6 33.1 31.8 3.7 E < 3.9 33.1 34.8 28.8 33.3 24.3 32.8 31.4 3.9 E < 4.1 32.9 34.6 28.6 33.1 24.0 32.7 31.4 4.1 E < 4.3 32.7 34.5 28.5 32.9 23.9 32.5 31.1 4.3 E < 4.5 32.5 34.3 28.2 32.7 23.6 32.4 30.9 4.5 E < 4.7 32.5 34.2 28.0 32.6 23.5 32.2 30.8 4.7 E < 4.9 32.3 34.0 27.8 32.4 23.3 32.0 30.6 E 4.9 32.1 33.9 27.6 32.4 23.2 31.8 30.5

1. - means not allowed NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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5.(b)(1)(iii) Contents - Type and Form of Material (continued)

Table 24Loading Table for BWR Fuel - 20.9kW/Package 1

Minimum Initial 55 < Assembly Average Burnup 60 GWd/MTU Assembly Avg. Minimum Cooling Time (years)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % 235U (E) 77 77 88 88 99 99 1010 2.1 E < 2.3 - - - - - - -

2.3 E < 2.5 - - - - - - -

2.5 E < 2.7 - - - - - - -

2.7 E < 2.9 - - - - - - -

2.9 E < 3.1 - - - - - - -

3.1 E < 3.3 - - - - - - -

3.3 E < 3.5 42.3 44.6 40.0 43.4 38.2 43.5 42.3 3.5 E < 3.7 39.8 42.3 37.7 41.0 357 41.1 39.8 3.7 E < 3.9 38.3 40.0 35.3 38.7 33.3 38.7 37.5 3.9 E < 4.1 38.1 40.2 33.7 38.3 30.9 38.0 36.5 4.1 E < 4.3 37.8 40.0 33.6 38.2 29.0 37.8 36.4 4.3 E < 4.5 37.8 39.8 33.4 38.2 28.9 37.8 36.4 4.5 E < 4.7 37.7 39.6 33.2 38.0 28.7 37.7 36.2 4.7 E < 4.9 37.6 39.5 33.1 37.9 28.4 37.5 36.0 E 4.9 37.4 39.4 32.9 37.8 28.4 37.4 35.8

1. - means not allowed

Table 25Low Burnup BWR Fuel Loading Table - 22 kW/Package

Max.

Assembly Min. Assembly Minimum Avg. Avg. Initial Cool Burnup Enrichment Time

[MWd/MTU] [wt% 235U] [Years]

10,000 1.3 6.3 15,000 1.5 8.6 20,000 1.7 10.3 25,000 1.9 11.9 30,000 2.1 13.7 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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(iv) Greater Than Class C Waste

GTCC waste consisting of solid, irradiated, and contaminated hardware, 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, within a GTCC waste basket liner transported in a GTCC TSC with a welded closure lid. The specific Curie content source of the GTCC waste shall be limited to a maximum specific activity of 2.7 Ci 60Co/lb averaged over the GTCC waste, with a maximum localized peak specific activity of 16.1 Ci 60Co/lb and a total 60Co activity of 85,760 Ci at transport. The maximum allowed weight of this waste is 55,000 lbs.

5.(b)(2) Maximum quantity of material per package

(i) For the contents described in Item 5.(b)(1)(i): Up to 37 undamaged PWR fuel assemblies, including nonfuel hardware and spacers, with a maximum weight of 62,160 pounds and a maximum decay heat limit per fuel location not to exceed the values in Table 2.

(ii) For the contents described in Item 5.(b)(1)(ii): Up to 37 undamaged PWR fuel assemblies, which may include up to 4 damaged fuel assemblies in damaged fuel cans, nonfuel hardware and spacers, with a maximum weight of 61,184 pounds (TSC and maximum contents shall not exceed 104,500 pounds) and a maximum decay heat limit per fuel location not to exceed the values in Table 2.

(iii) For the contents described in Item 5.(b)(1)(iii): 87 undamaged BWR fuel assemblies, including channels and spacers, with a maximum weight of 62,656 pounds and a maximum decay heat limit not to exceed the values in Table 20.

(iv) For the contents described in item 5.(b)(1)(iv): GTCC waste with a maximum weight per package of 55,000 pounds in total. The maximum decay heat for the GTCC waste is 1.7 kW per package.

5.(c) Criticality Safety Index

Undamaged PWR and BWR Fuel 0.00

Damaged PWR Fuel 100.00 NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES

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6. 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, except that the minimum component thicknesses for the mockup in Section 8.1.6.1 and the minimum shielding effectiveness configuration for calculating the dose rates used as acceptance criteria for the tests in Sections 8.1.6.3 and 8.2.3 are defined by the component dimensions and tolerances in the drawings listed in Condition 5.(a)(3).

(c) Only TSCs that are within their initial term for storage (i.e., in storage up to 20 years) or are new and havent been loaded and placed into storage are authorized for shipment using the moderator exclusion option of this certificate.

7. Prior to transport by rail, the Association of American Railroads must have evaluated and approved the railcar and the system used to support and secure the package during transport.

8. Prior to marine or barge transport, the National Cargo Bureau, Inc., must have evaluated and approved the system used to support and secure the package to the barge or vessel, and must have certified that package stowage is in accordance with the regulations of the Commandant, United States Coast Guard.

9. Transport by air is not authorized.

10. Transport of fuel assemblies, as described in Drawing No. 71160-685 Revision 8, Assembly No. 99, is not authorized.

11. Zion TSC basket assemblies with serial numbers TSC-21, TSC-22, TSC-24, TSC-25, and TSC-26, are authorized to not have weldment shims installed as required by license drawing 71160-575, Note 4.

12. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code alternative in Table 2.1.4-1, "ASME Code Alternatives for MAGNATRAN Components", page 2.1.4-6 for the "Fuel Basket Assembly" the description for "Alternative, Justification and Compensatory Measures" is revised to: "Fuel basket materials fabricated from Carbon Steel with a thickness

>5/8 inch, may optionally have impact tests performed on specimens oriented in a direction parallel to the principal rolling direction of the plate, provided that the results from the tests are scaled down to 67% of the measured values before comparing to acceptance criteria in The American Society of Mechanical Engineers Boiler and Pressure Vessel Code Table NG-2331(a)(1) at the Lowest Service Temperature (LST) of -40°F."

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

U.S. NUCLEAR REGULATORY COMMISSION NRC FORM 618 (8-2000) 10 CFR 71CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIAL PACKAGES 1.a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 4 71-9356 USA/9356/B(U)F-96 43 OF 43

14. Expiration date: April 30, 2024 REFERENCES NAC International, Inc., Application dated March 7, 2022.

FOR THE U.S. NUCLEAR REGULATORY COMMISSION 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 Signed by Diaz-Sanabria, Yoira on 11/03/22