Text

NAC-LWT Legal Weight Truck Cask System Safety Analysis Report Initial Submittal Iron Clad Fuel
	ML23244A178
Person / Time
Site:	07109225
Issue date:	08/31/2023
From:	; NAC International
To:	; Office of Nuclear Material Safety and Safeguards
Shared Package
ML23244A176	List: ML23244A177; ML23244A178;
References
	ED20230121
	Download: ML23244A178 (1)
	v • d • e

August 2023

Revision 23C NAC-LWT Legal Weight Truck Cask System ANALSA F E T YYSIS REPORT Initial Submittal Iron Clad Fuel

NON-PROPRIETARY VERSION

Docket No. 71-9225

Atlanta Corporate Headquarters: 3930 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

Enclosure 1

No. 71-9225 for NAC-LWT Cask

Proposed Changes for Revision 74 of Certificate of Compliance

Iron Clad Fuel Rods

NAC-LWT SAR, Revision 23C

August 2023

ED20230121 Page 1 of 3 CoC Sections (revised)

CoC Page 14 of 34

5.(b)(1) Type and form of material (continued)

(viii) PWR rods, consisting of uranium dioxide pellets within zirconium alloy type or Non-zircaloy cladding (FeCrAl based alloy) cladding. The maximum uranium enrichment is 5 weight percent 235U, the maximum active fuel length is 150 inches, and the maximum pellet diameter is 0.3765 inches. The maximum burnup is 80,000 MWd/MTU, and the minimum cool time is 150 days. Non-zircaloy clad fuel rods, and non-fueled non-zirconium rod sections, require an additional 90-day cool time beyond the indicated zircaloy based value.

CoC Page 15 of 34

5.(b)(1) Type and form of material (continue d)

(ix) BWR rods, consisting of uranium dioxide pellets within zirconium alloy type or Non-zircaloy cladding (FeCrAl based alloy). The maximum uranium enrichment is 5 weight percent 235U, the maximu m active fuel length is 150 inches, and the maximum pellet diameter is 0.490 inch. The maximum burnup is 80,000 MWd/MTU and the minimum cool time is between 150 - 270 days, as specified in the table below:

BWR Fuel Type Burnup, b Minimum Cool Time Array Size (GWd/MTU) (days)2

7 x 7 b 60 210 60 < b 70 240 70 < b 80 270

8 x 81 b 80 150 Note 1: Includes rods from all larger BWR assembly arrays (e.g., 9 x 9, 10 x 10)

Note 2: Non-zircaloy clad fuel rods, and non-fueled non-zirconium rod sections, require an additional 90-day cool time beyond the indicated zircaloy based value.

ED20230121 Page 2 of 3 CoC Page 24 of 34

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

(ix) For PWR fuel rods, as described in Item 5.(b)(1)(viii): up to 25 fuel rods.

Maximum decay heat not to exceed 2.3 kilowatts per packa ge.

Intact individual rods may be placed either in an irradiated or unirradiated fuel assembly lattice (skeleton) or in a fuel rod insert. The PWR fuel assembly lattice must be transported in the PWR basket.

CoC Page 25 of 34

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

Up to 14 of the 25 fuel rods may be classified as damaged. Damaged fuel rods may include fuel debris, particles, loose pellets, and fragmented rods or unfueled rods/rod segments in addition to the fuel rods. Fuel rods may be composed of segments. Damaged fuel rods must be placed in a fuel rod insert. Damaged fuel rods may also be placed in individual failed fuel rod capsules, as shown in Figure 1.2.3-11 of the application, prior to placement in the fuel rod insert. Guide/instrument tubes and tube segments may be placed in the fuel rod insert. The fuel rod insert must be transported in a PWR/BWR transport canister, which is positioned in the PWR insert in the PWR basket.

(x) For BWR fuel rods, as described in Item 5.(b)(1)(ix): up to 25 fuel rods.

Maximum decay heat not to exceed 2.1 kilowatts per package.

Intact individual rods may be placed either in a fuel assembly lattice or in a fuel rod insert. The BWR fuel assembly lattice must be transported in the PWR insert in the PWR basket.

Up to 14 of the 25 fuel rods may be classified as damaged. Damaged fu el rods may include fuel debris, particles, loose pellets, and fragmented rods or unfueled rods/rod segments in addition to the fuel rods. Fuel rods may be composed of segments. Damaged fuel rods must be placed in a fuel rod insert. Damaged fuel rods may also be placed in individual failed fuel rod capsules, as shown in Figure 1.2.3-11 of the application, prior to placement in the fuel rod insert. Water rods and inert rods may be placed in the fuel ro d insert. The fuel rod insert must be transported in a PWR/BW R transport canister, which is positioned in the PWR insert in the PWR basket.

ED20230121 Page 3 of 3

Enclosure 2

No. 71-9225 for NAC-LWT Cask

List of Calculations

NAC-LWT SAR, Revision 23C

August 2023

ED20230121 Page 1 of 2 List of Calculations, NAC-LWT SAR, Revision 23C

Contents:

1. 50082-5001 R0

Calculations Withheld in Entirety Per 10 CFR 2.390

ED20230121 Page 1 of 2 ED20230121 - Enclosure 3 Page 1 of 3

Enclosure 3

No. 71-9225 for NAC-LWT Cask

List of SAR Changes

NAC-LWT SAR,

Initial Submittal, Iron Clad Fuel

Revision 23C

August 2023 ED20230121 - Enclosure 3 Page 2 of 3 List of SAR Changes, NAC-LWT SAR, Revision 23C

Chapter 1

Page 1-1, modified text in the fifth bullet and footnote 2 where indicated.

Page 1.1-1, modified text where indicated.

Page 1.2-14, modified text in second paragraph of Section 1.2.3.7 where indicated.

Page 1.2-15 thru 1.2-22, text flow changes.

Page 1.2-48, update Table 1.2-4 where indicated.

Chapter 2

No changes.

Chapter 3

Page 3.4-17, added paragraph at the top of the page in Section 3.4.1.7.1

Page 3.4-18, text flow changes.

Page 3.4-19, added paragraph to the end of Section 3.4.1.7.3 including the addition of an embedded figure.

Page 3.4-20, text flow changes.

Page 3.4-21 thru 3.4-110, text flow to the end of the section (these pages are not included this submittal).

Chapter 4

No changes.

Chapter 5

Page 5-i, 5-viii, 5-xv and 5-xvi, updated TOC where indicated for the addition of new Section 5.3.25

Page 5-2 thru 5-3, added paragraph at the bottom of page 5-2 and top of page 5-3 where indicated.

Page 5-4, text flow changes.

Page5.1.1-8, modified text where indicated.

Page5.1.1-16, modified text in Table 5.1.1-2 where indicated.

Page 5.3.25-1 thru 5.3.25-15, added section 5.3.25

Chapter 6

Page 6.2.4-1, added second paragraph to Section 6.2.4.2 ED20230033 - Enclosure 2 Page 3 of 3

Chapter 7

No changes.

Chapter 8

No changes.

Chapter 9

No changes.

ED20230121 - Enclosure 4 Page 1 of 1

Enclosure 4

No. 71-9225 for NAC-LWT Cask

List of Effective pages and

SAR Changed Pages, Revision 23C

August 2023

Revision 23C NAC-LWT Legal Weight Truck Cask System ANALSAFETYYSIS REPORT

Volume 1 of 3 NON-PROPRIETARY VERSION

Docket No. 71-9225

Atlanta Corporate Headquarters: 3930 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

NAC-LWT Cask SAR August 2023 Revision 23C

LIST OF EFFECTIVE PAGES

Chapter 1 2.4.5-1......................................... Revision 46 1-i thru 1-vi................................ Revision 46 2.4.6-1......................................... Revision 46 1-1............................................ Revision 23C 2.5.1-1 thru 2.5.1-11................... Revision 46 1-2 thru 1-10.............................. Revision 46 2.5.2-1 thru 2.5.2-17................... Revision 46 1.1-1......................................... Revision 23C 2.6.1-1 thru 2.6.1-7..................... Revision 46 1.1-2 thru 1.1-4.......................... Revision 46 2.6.2-1 thru 2.6.2-7..................... Revision 46 1.2-1 thru 1.2-13........................ Revision 46 2.6.3-1......................................... Revision 46 1.2-14 thru 1.2-22.................... Revision 23C 2.6.4-1......................................... Revision 46 1.2-23 thru 1.2-47...................... Revision 46 2.6.5-1 thru 2.6.5-2..................... Revision 46 1.2-48....................................... Revision 23C 2.6.6-1......................................... Revision 46 1.2-49 thru 1.2-64...................... Revision 46 2.6.7-1 thru 2.6.7-137................. Revision 46 2.6.8-1......................................... Revision 46 1.3-1........................................... Revision 46 2.6.9-1......................................... Revision 46 1.4-1........................................... Revision 46 2.6.10-1 thru 2.6.10-15............... Revision 46 1.5-1........................................... Revision 46 2.6.11-1 thru 2.6.11-12............... Revision 46 2.6.12-1 thru 2.6.12-140............. Revision 46 88 drawings in the 2.7-1............................................ Revision 46 Chapter 1 List of Drawings 2.7.1-1 thru 2.7.1-117................. Revision 46 2.7.2-1 thru 2.7.2-23................... Revision 46 Chapter 1 Appendices 1-A 2.7.3-1 thru 2.7.3-5..................... Revision 46 through 1-G 2.7.4-1......................................... Revision 46 2.7.5-1 thru 2.7.5-5..................... Revision 46 Chapter 2 2.7.6-1 thru 2.7.6-4..................... Revision 46 2-i thru 2-xxv............................. Revision 46 2.7.7-1 thru 2.7.7-104................. Revision 46 2-1.............................................. Revision 46 2.8-1............................................ Revision 46 2.1.1-1 thru 2.1.1-2.................... Revision 46 2.9-1 thru 2.9-24......................... Revision 46 2.1.2-1 thru 2.1.2-3.................... Revision 46 2.10.1-1 thru 2.10.1-3................. Revision 46 2.1.3-1 thru 2.1.3-8.................... Revision 46 2.10.2-1 thru 2.10.2-49............... Revision 46 2.2.1-1 thru 2.2.1-5.................... Revision 46 2.10.3-1 thru 2.10.3-18............... Revision 46 2.3-1........................................... Revision 46 2.10.4-1 thru 2.10.4-11............... Revision 46 2.3.1-1 thru 2.3.1-13.................. Revision 46 2.10.5-1....................................... Revision 46 2.4-1........................................... Revision 46 2.10.6-1 thru 2.10.6-19............... Revision 46 2.4.1-1........................................ Revision 46 2.10.7-1 thru 2.10.7-66............... Revision 46 2.4.2-1........................................ Revision 46 2.10.8-1 thru 2.10.8-67............... Revision 46 2.4.3-1........................................ Revision 46 2.10.9-1 thru 2.10.9-9................. Revision 46 2.4.4-1........................................ Revision 46

Page 1 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

LIST OF EFFECTIVE PAGES (Continued)

2.10.10-1 thru 2.10.10-97.......... Revision 46 5.1.1-8...................................... Revision 23C 2.10.11-1 thru 2.10.11-10.......... Revision 46 5.1.1-9 thru 5.1.1-15................... Revision 46 2.10.12-1 thru 2.10.12-31.......... Revision 46 5.1.1-16.................................... Revision 23C 2.10.13-1 thru 2.10.13-17.......... Revision 46 5.1.1-17 thru 5.1.1-22................. Revision 46 2.10.14-1 thru 2.10.14-38.......... Revision 46 5.2.1-1 thru 5.2.1-7..................... Revision 46 2.10.15-1 thru 2.10.15-10.......... Revision 46 5.3.1-1 thru 5.3.1-2..................... Revision 46 2.10.16-1 thru 2.10.16-5............ Revision 46 5.3.2-1......................................... Revision 46 5.3.3-1 thru 5.3.3-8..................... Revision 46 Chapter 3 5.3.4-1 thru 5.3.4-27................... Revision 46 3-i thru 3-v................................. Revision 46 5.3.5-1 thru 5.3.5-4..................... Revision 46 3.1-1 thru 3.1-3.......................... Revision 46 5.3.6-1 thru 5.3.6-22................... Revision 46 3.2-1 thru 3.2-11........................ Revision 46 5.3.7-1 thru 5.3.7-19................... Revision 46 3.3-1........................................... Revision 46 5.3.8-1 thru 5.3.8-25................... Revision 46 3.4-1 thru 3.4-16........................ Revision 46 5.3.9-1 thru 5.3.9-26................... Revision 46 3.4-17 thru 3.4-110.................. Revision 23C 5.3.10-1 thru 5.3.10-14............... Revision 46 3.5-1 thru 3.5-43........................ Revision 46 5.3.11-1 thru 5.3.11-47............... Revision 46 3.6-1 thru 3.6-12........................ Revision 46 5.3.12-1 thru 5.3.12-26............... Revision 46 5.3.13-1 thru 5.3.13-18............... Revision 46 Chapter 4 5.3.14-1 thru 5.3.14-22............... Revision 46 4-i thru 4-iii................................ Revision 46 5.3.15-1 thru 5.3.15-9................. Revision 46 4.1-1 thru 4.1-4.......................... Revision 46 5.3.16-1 thru 5.3.16-5................. Revision 46 4.2-1 thru 4.2-4.......................... Revision 46 5.3.17-1 thru 5.3.17-43............... Revision 46 4.3-1 thru 4.3-4.......................... Revision 46 5.3.18-1 thru 5.3.18-2................. Revision 46 4.4-1........................................... Revision 46 5.3.19-1 thru 5.3.19-9................. Revision 46 4.5-1 thru 4.5-43........................ Revision 46 5.3.20-1 thru 5.3.20-29............... Revision 46 5.3.21-1 thru 5.3.21-45............... Revision 46 Chapter 5 5.3.22-1 thru 5.3.22-34............... Revision 46 5-i............................................. Revision 23C 5.3.23-1 thru 5.3.23-49............... Revision 46 5-ii thru 5-vi............................... Revision 46 5.3.24-1 thru 5.3.24-7................. Revision 46 5-vii.......................................... Revision 23C 5.3.25-1 thru 5.3.24-15............ Revision 23C 5-viii thru 5-xiv.......................... Revision 46 5.4.1-1 thru 5.4.1-6..................... Revision 46 5-xv thru 5-xvi......................... Revision 23C 5-1.............................................. Revision 46 Chapter 6 5-2 thru 5-4.............................. Revision 23C 6-i thru 6-xix............................... Revision 46 5.1.1-1 thru 5.1.1-7.................... Revision 46 6-1 thru 6-2................................. Revision 46

Page 2 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

LIST OF EFFECTIVE PAGES (Continued)

6.1-1 thru 6.1-6.......................... Revision 46 6.5.1-1 thru 6.5.1-13................... Revision 46 6.2-1........................................... Revision 46 6.5.2-1 thru 6.5.2-4..................... Revision 46 6.2.1-1 thru 6.2.1-3.................... Revision 46 6.5.3-1 thru 6.5.3-2..................... Revision 46 6.2.2-1 thru 6.2.2-3.................... Revision 46 6.5.4-1 thru 6.5.4-46................... Revision 46 6.2.3-1 thru 6.2.3-7.................... Revision 46 6.5.5-1 thru 6.5.5-15................... Revision 46 6.2.4-1...................................... Revision 23C 6.5.6-1 thru 6.5.6-20................... Revision 46 6.2.5-1 thru 6.2.5-5.................... Revision 46 6.5.7-1 thru 6.5.7-18................... Revision 46 6.2.6-1 thru 6.2.6-3.................... Revision 46 6.7.1-1 thru 6.7.1-19................... Revision 46 6.2.7-1 thru 6.2.7-2.................... Revision 46 6.7.2-1 thru 6.7.2-16................... Revision 46 6.2.8-1 thru 6.2.8-3.................... Revision 46 6.7.3-1 thru 6.7.3-39................... Revision 46 6.2.9-1 thru 6.2.9-4.................... Revision 46 6.7.4-1 thru 6.7.4-28................... Revision 46 6.2.10-1 thru 6.2.10-3................ Revision 46 6.7.5-1 thru 6.7.5-16................... Revision 46 6.2.11-1 thru 6.2.11-3................ Revision 46 6.7.6-1 thru 6.7.6-22................... Revision 46 6.2.12-1 thru 6.2.12-4................ Revision 46 6.3.1-1 thru 6.3.1-6.................... Revision 46 Appendix 6.6 6.3.2-1 thru 6.3.2-4.................... Revision 46 6.6-i thru 6.6-iii.......................... Revision 46 6.3.3-1 thru 6.3.3-9.................... Revision 46 6.6-1............................................ Revision 46 6.3.4-1 thru 6.3.4-10.................. Revision 46 6.6.1-1 thru 6.6.1-111................. Revision 46 6.3.5-1 thru 6.3.5-12.................. Revision 46 6.6.2-1 thru 6.6.2-56................... Revision 46 6.3.6-1 thru 6.3.6-9.................... Revision 46 6.6.3-1 thru 6.6.3-73................... Revision 46 6.3.7-1 thru 6.3.7-4.................... Revision 46 6.6.4.-1 thru 6.6.4-77.................. Revision 46 6.3.8-1 thru 6.3.8-7.................... Revision 46 6.6.5-1 thru 6.6.5-101................. Revision 46 6.3.9-1 thru 6.3.9-7.................... Revision 46 6.6.6-1 thru 6.6.6-158................. Revision 46 6.3.10-1 thru 6.3.10-2................ Revision 46 6.6.7-1 thru 6.6.7-84................... Revision 46 6.4.1-1 thru 6.4.1-10.................. Revision 46 6.6.8-1 thru 6.6.8-183................. Revision 46 6.4.2-1 thru 6.4.2-10.................. Revision 46 6.6.9-1 thru 6.6.9-53................... Revision 46 6.4.3-1 thru 6.4.3-35.................. Revision 46 6.6.10-1 thru 6.6.10-38............... Revision 46 6.4.4-1 thru 6.4.4-24.................. Revision 46 6.6.11-1 thru 6.6.11-53............... Revision 46 6.4.5-1 thru 6.4.5-51.................. Revision 46 6.6.12-1 thru 6.6.12-20............... Revision 46 6.4.6-1 thru 6.4.6-22.................. Revision 46 6.6.13-1 thru 6.6.13-22............... Revision 46 6.4.7-1 thru 6.4.7-13.................. Revision 46 6.6.14-1 thru 6.6.14-7................. Revision 46 6.4.8-1 thru 6.4.8-14.................. Revision 46 6.6.15-1 thru 6.6.15-................... Revision 46 6.4.9-1 thru 6.4.9-9.................... Revision 46 6.6.16-1 thru 6.6.16-30............... Revision 46 6.4.10-1 thru 6.4.10-18.............. Revision 46 6.6.17-1 thru 6.6.17-7................. Revision 46 6.4.11-1 thru 6.4.11-7................ Revision 46 6.6.18-1 thru 6.6.18-34............... Revision 46

Page 3 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

LIST OF EFFECTIVE PAGES (Continued)

6.6.19-1 thru 6.6.19-3................ Revision 46

Chapter 7 7-i thru 7-iii................................ Revision 46 7.1-1 thru 7.1-86........................ Revision 46 7.2-1 thru 7.2-17........................ Revision 46

Chapter 8 8-i............................................... Revision 46 8.1-1 thru 8.1-15........................ Revision 46 8.2-1 thru 8.2-6.......................... Revision 46 8.3-1 thru 8.3-4.......................... Revision 46

Chapter 9 9-i............................................... Revision 46 9-1 thru 9-11.............................. Revision 46

Page 4 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

1 GENERAL INFORMATION This chapter of the NAC International, Legal Weight Truck spent fuel shipping cask (NAC-LWT) Safety Analysis Report (SAR) presents a general introduction to, and description of, the NAC-LWT cask. Terminology used throughout this report is presented in Table 1.1-1.

Shipment of the NAC-LWT cask by truck, ISO container, and/or by railcar, as a Type B(U)F-96 package, as defined in 10 CFR 71.4, is authorized for the following contents:

PWR and BWR fuel assemblies1;

MTR fuel assemblies and plates;

DIDO fuel assemblies;

metallic fuel rods;

25 high burnup PWR and BWR fuel rods (including up to 14 fuel rods classified as damaged, or having FeCrAl based alloy cladding)2;

16 PWR MOX fuel rods (or mixed load of up to 16 PWR MOX and UO2 PWR fuel rods) and up to 9 burnable poison rods (BPRs);

TRIGA fuel elements and TRIGA fuel cluster rods;

General Atomics (GA) High-Temperature Gas-Cooled Reactor (HTGR) and Reduced-Enrichment Research and Test Reactor (RERTR) Irradiated Fuel Materials (IFM);

up to 700 PULSTAR fuel elements;

spiral fuel assemblies;

MOATA plate bundles;

up to eight (8) SLOWPOKE Fuel Canisters;

up to eighteen (18) NRU or NRX Fuel Assemblies (or equivalent number of fuel rods);

up to eighteen (18) NRU or NRX caddies loaded with EFN rods (Enriched Fast Neutron) rods, Booster rods and Mo-99 (Moly Targets);

HEUNL; and

One SLOWPOKE Fuel Core.

The authorized contents previously listed, except for HEUNL, include both irradiated and unirradiated forms of the materials.

Irradiated hardware is also authorized to be shipped in the NAC-LWT cask by truck, ISO container, and/or by railcar, as a Type B(U)F-96 package, as defined in 10 CFR 71.4. Irradiated hardware is defined as solid, irradiated and cont aminated fuel assembly structural or reactor internal component hardware, which may include fissile material, provided the quantity of fissile material does not exceed a Type A quantity and does not exceed the exemptions of 10 CFR 71.15, paragraphs (a), (b) and (c).

1 NAC-LWT casks containing PWR and BWR fuel assemblies are to be transported on an open trailer with a personnel barrier.

2 PWR and BWR fuel rods may be transported in either a fuel assembly lattice (skeleton) or in a fuel rod insert. The fuel rod insert may contain PWR instrument/guide tubes, BWR water/inert rods, or unfueled rods/rod segments in addition to the fuel rods. Fuel rods may be composed of segments.

NAC International 1-1 NAC-LWT Cask SAR October 2020 Revision 46

Shipment of the NAC-LWT cask by truck, ISO container, and/or by railcar, as a Type B(M)-96 package, as defined in 10 CFR 71.4, is also authorized for the following contents:

up to 300 Tritium Producing Burnable Absorber Rods (TPBARs), of which two can be prefailed; and

up to 55 TPBARs segmented during post-irradiation examination (PIE), including segmentation debris.

In accordance with 10 CFR 71.59, the NAC-LWT cask is assigned a Criticality Safety Index (CSI) for criticality control of the approved contents as follows:

Approved Contents CSI PWR fuel assemblies 100 BWR fuel assemblies 5.0 MTR fuel elements 0.0 Metallic fuel rods 0.0 TRIGA fuel elements (in poisoned TRIGA fuel baskets) 0.0 TRIGA fuel elements (in nonpoisoned TRIGA fuel baskets) 12.5 TRIGA fuel cluster rods 0.0 High burnup PWR rods 0.0 High burnup BWR rods 0.0 PWR MOX rods 0.0 DIDO fuel elements 12.5 General Atomic Irradiated Fuel Material (GA IFM) 0.0 TPBARS and segmented TPBARS 0.0 Intact (uncanned) PULSTAR fuel 0.0 Canned PULSTAR fuel 33.4 ANSTO fuel 0.0 Solid irradiated hardware 0.0 ANSTO-DIDO fuel combination 0.0 SLOWPOKE Fuel Rods in Fuel Canisters 0.0 NRU / NRX Fuel Assemblies, EFN rods, Booster rods, and Moly 100 targets HEUNL containers 0.0 SLOWPOKE Fuel Core 100

TPBARs do not contain fissile material and criticality assessments are not required. Solid, irradiated and contaminated hardware contents could include fissile material not exceeding a

NAC International 1-2 NAC-LWT Cask SAR August 2023 Revision 23C

1.1 Introduction The NAC-LWT spent-fuel shipping cask has been developed by NAC International (NAC) as a safe means of transporting radioactive materials authorized as approved contents. The cask design is optimized for legal weight over the road transport, with a gross weight of less than 80,000 pounds. The cask provides maximum safety during the loading, transport, and unloading operations required for spent-fuel shipment. The NAC-LWT cask assembly is composed of a package that provides a containment vessel that prevents the releas e of radioactive material. The actual containment boundary provided by the package consists of a 4.0-inch thick bottom plate, a 0.75-inch thick, 13.375-inch inner diameter shell, an upper ring forging, and an 11.3-inch thick closure lid. The cask lid closure is accomplished using twelve, 1-inch diameter bolts. The cask has an outer shell, 1.20 inches thick, to protect the containment shell and also to enclose the 5.75-inch thick lead gamma shield. Neutron shielding is provid ed by a 5.0-inch thick neutron shield tank with a 0.24-inch (6mm) thick outer wall, containing a water/ethylene glycol mixture and 1.0 minimum weight percent (wt %) boron. The neutron shield tank system includes an expansion tank to permit the expansion and contraction of the shield tank liquid without compromising the shielding or overstressing the shield tank structure. Aluminum honeycomb impact limiters are attached to each end of the cask to absorb kinetic energy developed during a cask drop, and limit the consequences of normal operations and hypothetical accident events.

The NAC-LWT is a legal weight truck cask designed to transport the following contents:

1 PWR assembly;

up to 2 BWR assemblies;

up to 15 sound metallic fuel rods;

up to 42 MTR fuel elements;

up to 42 DIDO fuel assemblies;

up to 25 high burnup PWR fuel rods (including up to 14 rods classified as damaged, or having FeCrAl based alloy cladding)1;

up to 25 high burnup BWR fuel rods (including up to 14 rods classified as damaged, or having FeCrAl based alloy cladding)1;

up to 16 PWR MOX fuel rods (or a combination of 16 PWR MOX and UO2 PWR rods) and up to 9 BPRs

up to 9 damaged metallic fuel rods;

up to 3 severely damaged metallic fuel rods in filters;

up to 140 TRIGA intact or damaged fuel elements/fuel debris (TRIGA is a Trademark of General Atomics);

up to 560 TRIGA intact or damaged fuel cluster rods/fuel debris;

2 GA IFM packages;

up to 300 TPBARs (of which two can be prefailed) in a consolidation canister; 1 PWR and BWR fuel rods may be transported in either a fuel assembly lattice (skeleton) or in a fuel rod insert. The fuel rod insert may contain PWR instrument/guide tubes, BWR water/inert rods, or unfueled rods/rod segments in addition to the fuel rods. Fuel rods may be composed of segments

NAC International 1.1-1 NAC-LWT Cask SAR October 2020 Revision 46

up to 25 TPBARs (of which two can be prefailed) in a rod holder;

up to 55 TPBARs segmented during post-irradiation examination (PIE), including segmentation debris;

up to 700 PULSTAR fuel elements (intact or damaged);

up to 42 spiral fuel assemblies;

up to 42 MOATA plate bundles;

up to 800 SLOWPOKE undamaged and/or damaged fuel rods contained in up to eight (8)

SLOWPOKE fuel canisters (up to 100 fuel rods each);

up to 18 NRU or NRX undamaged or damaged fuel assemblies (one per flow tube) or the equivalent number of loose rods as an assembly per basket tube (12 rods for NRU or 7 rods for NRX);

up to eighteen (18) NRU/NRX caddies loaded with EFN rods, Booster rods, or Moly targets;

4 HEUNL containers (empty or filled such that a minimum under filled cavity void of one gallon exists);

One SLOWPOKE fuel core containing up to 298 undamaged SLOWPOKE fuel rods; or

up to 4,000 lbs of solid, irradiated and contaminated hardware, which may include fissile material less than a Type A quantity and meeting the exemptions of 10 CFR 71.15, paragraphs (a), (b) and (c). Total allowed ma ss includes the weight of spacers, shoring and dunnage.

PWR or BWR fuel rods may be placed in a fuel rod insert (also referred to as a rod holder) or in a fuel assembly lattice. The fuel rod holder is composed of a 4x4 or a 5x5 rod array. An alternate 5x5 rod holder is designed to contain an oversize nonfuel-bearing component (e.g., CE guide tube or BWR water rod). The alternative configuration reduces fuel-bearing capacity to a maximum of 21 fuel rods. The lattice may be irra diated or unirradiated. Up to 14 of the fuel rods may be classified as damaged. Damaged fuel rods must be placed in a rod holder.

Damaged fuel rods or rod sections may be encapsulated to facilitate handling prior to placement in the rod holder. PWR rods may include Integral Fuel Burnable Absorber (IFBA) rods.

PWR MOX fuel rods (or a combination of PWR MOX and UO2 PWR fuel rods) are required to be loaded in a screened or free flow PWR/BWR Rod Transport Canister with a 5x5 insert. PWR MOX/UO2 rods may include Integral Fuel Burnable Absorber (IFBA) rods.

Damaged TRIGA fuel elements, cluster rods and fuel debris are required to be loaded in a sealed damaged fuel canister (DFC).

PULSTAR fuel elements may be configured as int act fuel assemblies, may be placed into a fuel rod insert, i.e., a 4x4 rod holder (intact elements only), or may be loaded into one of two can designs, designated as the PULSTAR screened fuel can or the PULSTAR failed fuel can.

Damaged PULSTAR fuel elements and nonfuel components of PULS TAR fuel assemblies must be loaded into cans. PULSTAR fuel cans may only be loaded into the to p or base module of the 28 MTR basket assembly. Intact PULSTAR fuel assemblies and intact PULSTAR fuel elements in a TRIGA fuel rod insert may be loaded in any basket module.

NAC International 1.1-2 NAC-LWT Cask SAR October 2020 Revision 46 1.2.3.5 BWR Fuel Assembly

The NAC-LWT cask is analyzed for the BWR fuel assemblies listed in Table 1.2-6. This table provides the dimensional constraints for the BWR fuel. The en richment, burnup and decay heat limits are specified in Table 1.2-4.

The BWR fuel rod cladding is a zirconium alloy type (Zircaloy-2, Zircaloy-4, Zirlo, M-5, etc.).

Minor variations of alloy composition have no impact on performance of cladding material.

1.2.3.6 TPBARs

The NAC-LWT cask is analyzed for the transport of three separate Tritium Producing Burnable Absorber Rod (TPBAR) content configurations. For the transport of production TPBARs from the reactor facility to the DOE processing facility, an open (i.e., unsealed) stainless steel consolidation canister is utilized to contain up to 300 TPBARs, two of which can be prefailed.

The characteristics of the production TPBARs are listed in Table 1.2-8. The consolidation canister assembly is shown in Figure 1.2.3-10. Up to 25 TPBARs may also be transported within the 5x5 rod holder located within the PWR/BWR Rod Transport Canister. For a TPBAR shipment, the transport canister is located w ithin the TPBAR basket. Up to two of the 25 TPBARs located within the rod holder may be classified as prefailed.

The third transport configuration is for the shipment of segmented TPBARs, following post-irradiation examination (PIE), contained in a welded stainless steel waste container containing segments and debris from up to 55 TPBARs. The characteristics of the TPBAR PIE segments are provided in Table 1.2-12. The waste container and extension weldment assembly is shown in Figure 1.2.3-16.

TPBARs are similar in size and nuclear characteristics to standard, commercial PWR, stainless steel-clad burnable absorber rods. The exterior of a typical TPBAR is a stainless steel clad tube.

The internal components of the TPBAR are designed and selected to produce and retain tritium.

Internal configurations differ for various TP BAR designs (see DOE reports provided in the Chapter 1 Appendices). The internal components of a typical TPBAR include a plenum spacer tube (getter tube), a spring clip or a plenum (compression) spring, pellet stack assemblies (pencils), and a bottom spacer tube. A pencil co nsists of a zirconium alloy liner around which lithium aluminate absorber pellets are stacked and then confined in a getter tube as shown in Figure 1.2.3-9. The unclassified design details of the various TPBAR designs are provided in the unclassified DOE documents and drawings provided in the Chapter 1 Appendices.

The transport assembly arrangements for the cons olidation canister and waste container TPBAR content configurations are identical and include a closure lid spacer assembly, a TPBAR basket and Alternate B port covers with bolting installed. The detailed requirements for the NAC-LWT

NAC International 1.2-13 NAC-LWT Cask SAR August 2023 Revision 23C

assembly are provided in license drawing 315-40-128 in Section 1.4. The overall payload arrangement for the NAC-LWT with the consolidation canister and waste container are shown in Figure 1.2.3-12 and Figure 1.2.3-17, respectively. For the transport of fewer than 300 TPBARs in the consolidation canister, stainless steel dunnage may be used to align and protect the contents. The weight and volume of the dunnage and the reduced TPBAR contents of the consolidation canister must be less than, or equal to, the weight and volume of 300 TPBARs. Up to 25 TPBPAR rods may also be transported in a PWR/BWR Rod Transport Canister in a NAC-LWT assembly as shown in License Drawing No. 315-40-104 (assembly 95).

The TPBAR content conditions are analyzed a nd evaluated for compliance with structural, thermal, containment and shielding conditions of the NAC-LWT in the appropriate SAR chapters. TPBARs do not contain fissile material and, therefore, criticality evaluations have not been performed. The operating procedures for the wet and dry loading and dry unloading of the TPBAR contents are provided in Chapter 7. The special leakage and pressure testing requirements for NAC-LWT casks intended for the transport of TP BAR contents are provided in Chapter 8.

1.2.3.7 PWR/BWR Fuel Rods PWR and BWR fuel rods are transported within the fuel lattice (skeleton) or 4x4 or 5x5 inserts (rod holder). The rod holder is located within a free flow, screened or sealed PWR/BWR transport canister. (The rod holder may also contain a nonfuel-bearing irradiated hardware component [e.g., BWR water rod, PWR instrument/guide tube].)

The PWR and BWR fuel rod cladding is of Zirconium alloy type (Zircaloy-2, Zircaloy-4, Zirlo, M-5, etc.). Minor variations of alloy composition have no impact on performance of cladding material. No clad integrity is required for damaged fuel rods, which are limited to 14 per PWR/BWR transport canister. Sufficient clad stru ctural properties may not be available for non-zirconium alloy-based cladding to justify that fuel rods will maintain undamaged configuration through all normal and accident condition. Therefore, up to 14 fuel rods of Iron-Chromium-Aluminum (FeCrAl) alloy based clad may be loaded in the PWR/BWR transport canister as they can be considered to fail during transport while maintaining evaluated safety margins (note the 14 limit represents the total of damaged zirconium alloy clad fuel rods plus the non-zirconium clad rods).

1.2.3.8 PULSTAR Fuel Element and Transport Configuration Description

PULSTAR fuel elements are transported in the NAC-LWT in the 28 MTR fuel basket assembly, which contains four modules with seven cells per module. The basket assembly is composed of

NAC International 1.2-14 NAC-LWT Cask SAR August 2023 Revision 23C a top module, a base module, and two intermediate modules (Dwgs 315-40-051, -049, and -050, respectively).

PULSTAR fuel elements may be loaded into the module cells in one of four configurations:

a) intact PULSTAR fuel assemblies b) intact PULSTAR fuel elements loaded into the 4 x4 TRIGA fuel rod insert (Dwg. 315-40-096); c) intact or damaged PULSTAR fuel elements, fuel debris and nonfuel-bearing components of PULSTAR fuel assemblies in the PULSTAR screened can (Dwg. 315-40-135); or d) intact or damaged PULSTAR fuel elements, fuel debris and nonfuel-bearing components of PULSTAR fuel assemblies in the PULSTAR sealed can (Dwg.

315-40-130). The contents of either can type are restricted to a quantity of fissile material and a total volume of material equivalent to 25 PULSTAR fuel elements. The sealed cask contents are restricted to the displaced volume of 25 intact PULSTAR fuel elements. The total cask payload shall not exceed 700 PULSTAR fuel elements. Loading of modules with mixed PULSTAR payload configurations is allowed, but PULSTAR cans, either screened or sealed, are restricted to loading in the base and top modules.

PULSTAR fuel elements are low enriched (< 7 wt %) uranium oxide rods, with zirconium alloy cladding. During reactor operation, 25 PULSTAR fuel elements are arranged in a rectangular 5x5 lattice, surrounded by a zirconium alloy box, and capped by top-and bottom-end fittings to form a PULSTAR fuel assembly. The nonfuel components of a PULSTAR fuel assembly are primarily aluminum and zirconium alloy and do not contain a significant activation source. A sketch of a PULSTAR fuel assembly is provided in Figure 1.2.3-13. Key physical, radiation protection and thermal characteristics of the PULSTAR fuel assembly/elements are listed in Table 1.2-9.

The sealed and screened PULSTAR cans are stainless steel containers that: a) minimize the dispersal of gross fuel particles that may escape from damaged fuel element cladding and/or fuel debris; b) facilitate retrieval of the contents from the transportation cask; and c) confine damaged fuel and/or debris within a known volume to facilitate criticality control, maintain dose limits, and control thermal loads within the cask. PULSTAR fuel pellets, pieces, and debris may be placed in an encapsulating rod for handling purposes prior to placement into either a sealed or screened can. The encapsulating rod is not required and has no safety significance. In addition to fuel elements, the cans may contain fuel asse mbly hardware up to the total content weight limit specified in Table 1.2-9. For operational/retrievability purposes, stainless steel rod inserts may be used to position the PULSTAR fuel elements within the fuel rod insert. Total content weight shall not exceed the total weight limit specified in Table 1.2-9. The fuel rod insert is composed of a 4x4 grid of 0.75-inch OD x 0.065-inch wall stainless steel tubes. The tubes provide structural support for individual intact PULSTAR fuel elements during transport in the NAC-LWT.

NAC International 1.2-15 NAC-LWT Cask SAR August 2023 Revision 23C Spacers may be used to axially position PULSTAR fuel contents near the top of the module for ease of loading and unloading operations. The spacers are provided for ease of operations and do not provide a safety function.

1.2.3.9 ANSTO Basket and Payload Description

Three basic fuel types are to be transported in the ANSTO baskets within the NAC-LWT cask:

spiral fuel assemblies, MOATA plate bundles and DIDO elements. Spiral fuel assemblies are composed of cylindrical aluminum inner and outer shells connected by curved metallic fuel plates. Further detail on the spiral fuel assemblies is provided in Section 1.2.3.9.1. MOATA plate bundles are comprised of up to 14 MTR fuel plates. Further detail on the plate bundles is provided in Section 1.2.3.9.2. DIDO elements are described in Section 1.2.3.2. The spiral fuel assemblies, MOATA plate bundles and DIDO elements may be intact or may have degraded cladding and be disassembled. Note that spiral assemblies may be cropped by removing nonfuel-bearing hardware to fit within the basket tubes. Cropped spiral fuel assemblies are classified as intact fuel. Spiral, MOATA and DIDO having degraded cladding or disassembled may be placed in aluminum DFCs as shown on Figure 1.2.3-18 prior to loading into the top ANSTO basket module or assembly to facilitate handling during loading and unloading operations. DFCs containing fuel elements shall be limited to lo ading in the 7 cells of the top ANSTO basket module of a standard six-module ANSTO basket assembly or in the top ANSTO basket module of the ANSTO-DIDO combination basket assembly. The ANSTO-DIDO combination basket is an assembly of a top ANSTO module, four intermediate DIDO modules and one base DIDO module. As the interfaces, weight and overall dimensions of the ANSTO and DIDO basket modules are essentially identical, the combination basket assembly is bounded by the structural, thermal and criticality analyses for the se parate DIDO and ANSTO basket assemblies.

Up to 42 spiral fuel assemblies or 42 MOATA plate bundles may be loaded in an ANSTO basket assembly. Up to 7 intact DIDO, spiral or MOATA plate elements/bundles and/or degraded clad DIDO, spiral or MOATA plate elements/bundles in DFCs may be loaded in an ANSTO top module in either the standard ANSTO basket assembly or in the ANSTO-DIDO combination basket assembly. DIDO fuel elements loaded into an ANSTO top module in either the ANSTO basket assembly or the ANSTO-DIDO combination basket assembly will be limited to a maximum decay heat load of 10 W per element wh ether with or without a DFC. Spiral fuel elements loaded into DFCs shall be limited to a maximum decay heat load of 10 W per element.

MOATA plate bundles loaded into DFCs shall be limited to a maximum decay heat of 1 W per bundle. The remaining 35 cells of the ANSTO-DIDO combination basket may be loaded with up to 35 intact DIDO fuel elements.

NAC International 1.2-16 NAC-LWT Cask SAR August 2023 Revision 23C A full cask load of either the ANSTO basket assembly or the ANSTO-DIDO combination basket assembly contains 6 baskets of up to 7 fuel assemblies or plate bundles per basket. The mixed loading of ANSTO and ANSTO-DIDO combination basket assemblies as described previously containing spiral fuel assemblies, MOATA plate bundles or DIDO elements is authorized.

1.2.3.9.1 Spiral Fuel Assemblies

The design basis characteristics of spiral fuel assemblies are presented in Table 1.2-10. The fuel material in spiral fuel assembly plates is a solid, homogeneous mixture of uranium-aluminum alloy, i.e., a metal alloy fuel. The fuel meat of each plate is clad in aluminum. A set of 10 curved fuel plates is located between an inner and outer cylindrical aluminum shell. Fuel elements are cropped to fit axially within the basket envelope. Fuel material is not cut during the cropping operation. The fuel plates are located in a spiral pattern, maintaining a constant pitch between fuel plate centers. A sketch of the assembly cross-section is provided in Figure 1.2.3-14.

1.2.3.9.2 MOATA Plate Bundles

The design basis characteristics of MOATA plate bundles are presented in Table 1.2-4. The fuel material in the plate bundle is a solid, homogeneous mixture of uranium-aluminum alloy, i.e., a metal alloy fuel. Each plate is clad in aluminum. A plate bundle is comprised of up to 14 fuel plates. Two thick (0.635 cm) aluminum nonfuel side plates support the fuel plate stack from two sides, making a possible total of 16 plates per bundle. At each axial end, the plates in the stack are connected by a pin. Spacing between plates is maintained by disk spacers placed onto the top and bottom pins between each fuel plate and the al uminum side plates. A sketch of a typical MOATA plate bundle is provided in Figure 1.2.3-15.

1.2.3.10 Solid, Irradiated and Contaminated Hardware

The design basis characteristics of the solid, irradiated and contaminated hardware are provided in Table 1.2-13. As described in the content de finition, the solid, irradiated and contaminated hardware may contain small quantities of fissile materials. Fissile materials in the irradiated hardware contents are acceptable if the quantity of fissile material does not exceed a Type A quantity and does not exceed the exemptions of 10 CFR 71.15, paragraphs (a), (b) and (c).

The irradiated hardware may be directly loaded into the NAC-LWT cask cavity, or may be contained in a secondary container or basket. As needed, appropriate component spacers, dunnage and shoring may be used to limit the movement of the contents during normal and accident conditions of transport.

NAC International 1.2-17 NAC-LWT Cask SAR August 2023 Revision 23C To ensure that the movement of the irradiated hardware contents above the lead shielded length of the NAC-LWT cask body (i.e., the approximately upper 6.25 inches of the cavity length) is precluded, an Irradiated Hardware Lid Spacer as shown on Drawing No. 315-40-145 shall be installed for all irradiated hardware content conf igurations. The total installed height of the spacer is 6.5 inches. Therefore, the available cavity length for the irradiated hardware is approximately 171 inches. The NAC-LWT cask shall be assembled for transport as shown on NAC Drawing No. 315-40-01 with the irradiated hardware spacer installed on the lid.

A comparative shielding evaluation for a conservatively selected irradiated hardware transport configuration (i.e., a single line source with no self-shielding) or consideration of the additional shielding provided by additional spacers, dunnage, inse rts or secondary containers is presented in Chapter 5. The evaluations show that the regulatory dose rate requirements per 10 CFR 71.47 for normal conditions of transport, or 10 CFR 71.51(b) under hypothetical accident conditions, are not exceeded.

1.2.3.11 PWR MOX Fuel Rods

The NAC-LWT cask is analyzed and evaluated for the transport of up to 16 PWR MOX fuel rods (or a combination of up to 16 PWR MOX and UO2 fuel rods) loaded into a 5 x 5 insert placed in a screened or free flow PWR/BWR Rod Transport Canister. The authorized characteristics of the evaluated PWR MOX fuel rods are provided in Table 1.2-4. For mixed PWR MOX and UO2 PWR fuel rod combinations, the UO2 PWR fuel rods may have the identical heat load, burnup and cool time characteristics as the PWR MOX fuel rods.

In addition to the 16 PWR MOX fuel rods (or a combination of PWR MOX and UO2 PWR fuel rods), up to 9 burnable poison rods (BPRs) may be loaded in the remaining openings in the 5 x 5 insert in the PWR/BWR Rod Transport Canister.

1.2.3.12 SLOWPOKE Fuel Rods in a SLOWPOKE Canister

SLOWPOKE fuel rods are transported in the NAC-LWT in the 28 MTR fuel basket assembly, which consists of four modules with seven cells per module. The basket assembly is composed of a top module, a base module, and two intermediate modules. Fuel load is limited to a maximum of four loaded cells per basket module, with fuel only loaded in the top and top intermediate modules. The lower intermediate and bottom modules are used as axial spacers and are not loaded. The center row of three cells w ithin the basket modules containing fuel are not loaded and contain a blocking device in each opening to prevent inadvertent loading. Therefore, a cask load for SLOWPOKE fuel rods is limited to eight loaded cells per cask.

SLOWPOKE fuel rods must be loaded into a can ister. The canister is a screened boundary providing gross particle control for damaged fuel material. Damaged and undamaged fuel may

NAC International 1.2-18 NAC-LWT Cask SAR August 2023 Revision 23C be mixed when loaded into the canister. Canister content is composed of 4x4 or 5x5 aluminum tube arrays that are stacked four high within the canister. Mixed load of 4x4 and 5x5 tube arrays are permitted in a canister. Based on the 5x5 tube arrays and four tube arrays per canister, the maximum content per canister is 100 SLOWPOKE fuel rods (or the equivalent quantity of damaged material).

SLOWPOKE fuel rods are composed of highly enriched (> 90 wt %) uranium-aluminum alloy fuel meat within aluminum cladding. During reactor operation ~300 rods form a reactor core.

Criticality in a SLOWPOKE core is achieved by the use of a thick beryllium neutron reflector surrounding the core. A sketch of a SLOWPOKE fuel rod is provided in Figure 1.2.3-19. Key physical, radiation protection and thermal characteristics of the SLOWPOKE fuel rods are listed in Table 1.2-14.

The SLOWPOKE canister is constructed primarily of aluminum. A limited quantity of stainless steel is located within the canis ter lid structure. The canister is designed to: a) minimize the dispersal of gross fuel particles that may escape from damaged fuel rod cladding and/or fuel debris (note that metallic fuel is not expected to release significant gross particulate even with severe clad damaged); b) facilitate retrieval of the contents fro m the transportation cask; and, c) confine damaged fuel and/or debris within a known volume to facilitate criticality control, maintain dose limits, and control thermal load s within the cask. SLOWPOKE fuel pieces and debris may be placed into an aluminum tube stru cture located within the canister. The aluminum tubes provide structural support for individual fu el rods/pieces during transport in the NAC-LWT but are not required within the analysis to maintain safety limits.

1.2.3.13 NRU/NRX Fuel Assemblies or Fuel Rods

NRU/NRX fuel assemblies and fuel rods are transported in the NAC-LWT in an 18 tube basket.

The basket assembly is composed of 18 fuel tubes arranged in two concentric rings. The basket is spaced towards the top of the ca sk cavity by a bottom basket spacer.

NRX fuel assemblies or loose fuel rods must be loaded into a fuel rod caddy assembly. Loose NRU fuel rods may be loaded into a caddy. Note, the use of the caddy plug is not required for NRU or NRX shipments. Mixed loading of NRU and NRX assemblies in a basket is not permitted. NRX assemblies are composed of (7) fuel rods and the NRU assemblies are composed of (12) fuel rods.

NRU/NRX HEU fuel rods are composed of highly enriched (> 90 wt%) uranium-aluminum alloy fuel meat within aluminum cladding. NRU LEU fuel meat is composed of <20% wt% 235U enriched material composed of uranium-aluminum-silicon. NRU and NRX rods have a fin structure attached to the clad. The NRX rods have spiral fins to retain rod spacing. NRU assemblies in addition to the fins have a set of spacer disks assuring that rod pitch is maintained.

NAC International 1.2-19 NAC-LWT Cask SAR August 2023 Revision 23C A sketch of both NRU and NRX fuel assemblies is provided in Figure 1.2.3-20. Key physical, radiation protection and thermal characteristics of the NRU and NRX fuel assemblies are listed in Table 1.2-15.

NRU/NRX fuel assemblies, loose rods, or rod segments that do not meet the structural or clad integrity requirements of the undamaged NRU or NRX fuel assembly are classified as damaged.

Loose fuel rods defined as damaged, or rod segments, must be loaded in the NRU/NRX caddy prior to placement into the NRU/NRX basket. Damaged NRU assemblies may be loaded into the NRU/NRX basket tube without use of a caddy. Note, the use of the caddy plug is not required for NRU or NRX shipments. Damaged NRX assemblies shall be loaded into the NRU/NRX basket tube with the use of a caddy. NRU/NRX basket and basket lid, including screens, provide the gross material boundary for damaged fuel (i.e., gross material is retained in basket tube). Clad through-wall damage is limited to 5% of the fueled surface area. Clad removed (i.e., clad originally associated with the rod but no longer present) is limited to 2% of the fueled surface area. Clad through-wall damage without loss of cladding from the system may occur due to processes such as clad splitting. Fuel assemblies with exposed fuel material resulting from the cropping process are considered damaged but do not require use of a caddy (NRU only) provided loose, fuel containing, rod segments are not loaded into the same basket opening as a cropped assembly. To address these definitions the up to a fuel assembly quantity of material is further evaluated. Analysis are also included to address loss of cladding (through-wall clad damage).

The NRU/NRX caddy is constructed of aluminum. The aluminum caddy provides geometry constraint to fuel rod movement. Due to the increased reactivity of NRX fuel relative to high enriched NRU fuel, only NRX criticality evaluations cr edited this constraint. Note, the use of the caddy plug is not required for NRU or NRX shipments.

1.2.3.14 EFN Rods, Booster Rods or Moly Targets (Short and Double Length)

EFN Rods, Booster Rods, and Moly Targets are transported in the NAC-LWT in an 18 tube basket (NRU/NRX basket) in NRU/NRX caddies. Short and Double Length Moly targets are simply referred to as Moly targets unless otherwise noted.

Mixed loading of EFN rod caddies and Moly target caddies is perm itted. Booster rod caddies are not allowed for loading with EFN rod caddies or Moly target caddies. Each payload type may not be mixed in a caddy.

EFN Rods, Booster Rods, and Moly Targets are composed of highly enriched (> 90 wt%)

uranium-aluminum alloy fuel meat within alumin um cladding. Rods/targets have a fin structure attached to the clad. The EFN rods and Booster Rods were a component of an assembly prior to disassembly. Moly targets were part of a rod with multiple targets axially connected. Sketches

NAC International 1.2-20 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C of payload type are provided in Figure 1.2.3-22. Key physical, radiation protection, and thermal characteristics of the are listed in Table 1.2-18.

A distinction between undamaged and damaged material is not required as all rods/targets were evaluated considering complete or majority loss of clad with the damaged evaluation bounding undamaged (no significant clad through damage) condition. Segments/fragments of the rods/targets are permitted for loading providing total equivalent quantity limits in Table 1.2-18 are maintained. Rods/targets or segments/fragments thereof smaller than 6 inches in length require the use of the caddy plug.

1.2.3.15 HEUNL Containers

HEUNL material packaged in HEUNL containers may be directly loaded into the NAC-LWT cavity. Four containers must be packaged in the NAC-LWT for transport. The containers may be partially filled.

A sketch of the HEUNL container is provided in Figure 1.2.3-21. The container design is presented in NAC drawing 315-40-181. All hardware indicated on drawing 315-40-181 has been determined to be Important to Safety and has been evaluated, characterized and will be controlled in accordance with NACs QA Program as described in Section 1.3.

HEUNL material consists of a solution of uranyl nitrate, various other nitrates (primarily aluminum nitrate), and water. The solution may contain uranyl nitrates with up to 7.40 g/L 235U.

Key physical, radiation protection, and thermal characteristics of the HEUNL material are provided in Table 1.2.3-16.

NAC International 1.2-21 NAC-LWT Cask SAR August 2023 Revision 23C 1.2.3.16 SLOWPOKE Fuel Core

One SLOWPOKE fuel core containing up to 298 undamaged SLOWPOKE fuel rods may be transported in the NAC-LWT. The SLOWPOKE fuel core is packaged in the SLOWPOKE fuel core basket. A spacer is attached to the SLOWP OKE fuel core basket lid locating the fuel core at the bottom of the basket. The basket is transported with empty intermediate and bottom MTR-42 basket modules to provide axial spacing. The SLOWPOKE fuel core basket is therefore located next to the NAC-LWT cask lid.

The SLOWPOKE fuel core primary components are up to 298 undamaged SLOWPOKE fuel rods, a center tube, and upper and lower plates. SLOWPOKE fuel rods are composed of highly enriched uranium-aluminum alloy fuel meat within aluminum cladding. As discussed in Section 1.2.3.12, criticality in a SLOWPOKE core during reactor operations is achieved by the use of a thick beryllium neutron reflector surrounding the core. The beryllium reflector is not part of the packaged contents. A sketch of a SLOWPOKE fuel rod is provided in Figure 1.2.3-19. Key physical, radiation protection and thermal characteristics of the SLOWPOKE fuel core, i.e.,

parameters documented in the analytical chapters to be safely transported, are listed in Table 1.2-17.

NAC International 1.2-22 NAC-LWT Cask SAR October 2020 Revision 46 Table 1.2-3 Characteristics of Design Basis TRIGA Fuel Cluster Rods

Element Type TRIGA Fuel Cluster Rod

Max. Rod Length (in) 31.0 Max. Active Length (in) 22.5 Clad Material Incoloy 800 Min. Clad Thickness (in) 0.015 Fuel Material U-ZrH Max. Pellet Diameter (in) 0.53 Max. Rod Weight (kg) 0.65 Min. U in U-ZrH (wt %) 43.0 (LEU) or 9.5 (HEU)1 Max. 235U in U (wt %) 19.9 to 93.3 235U Mass (g) 55.0 (LEU) or 46.5 (HEU)

Max. H to Zr Ratio 1.7

1 Equivalent to a maximum zirconium mass of 357 g for LEU fuel and 457 g for HEU fuel material. Lower weight percents are permitted, provided the maximum zirconium mass limits are not exceeded.

NAC International 1.2-47

August 2023

Revision 23C NAC-LWT Legal Weight Truck Cask System ANALSAFETYYSIS REPORT

Volume 2 of 3 NON-PROPRIETARY VERSION

Docket No. 71-9225

Atlanta Corporate Headquarters: 3930 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

NAC-LWT Cask SAR August 2023 Revision 23C

LIST OF EFFECTIVE PAGES

Chapter 1 2.4.5-1......................................... Revision 46 1-i thru 1-vi................................ Revision 46 2.4.6-1......................................... Revision 46 1-1............................................ Revision 23C 2.5.1-1 thru 2.5.1-11................... Revision 46 1-2 thru 1-10.............................. Revision 46 2.5.2-1 thru 2.5.2-17................... Revision 46 1.1-1......................................... Revision 23C 2.6.1-1 thru 2.6.1-7..................... Revision 46 1.1-2 thru 1.1-4.......................... Revision 46 2.6.2-1 thru 2.6.2-7..................... Revision 46 1.2-1 thru 1.2-13........................ Revision 46 2.6.3-1......................................... Revision 46 1.2-14 thru 1.2-22.................... Revision 23C 2.6.4-1......................................... Revision 46 1.2-23 thru 1.2-47...................... Revision 46 2.6.5-1 thru 2.6.5-2..................... Revision 46 1.2-48....................................... Revision 23C 2.6.6-1......................................... Revision 46 1.2-49 thru 1.2-64...................... Revision 46 2.6.7-1 thru 2.6.7-137................. Revision 46 2.6.8-1......................................... Revision 46 1.3-1........................................... Revision 46 2.6.9-1......................................... Revision 46 1.4-1........................................... Revision 46 2.6.10-1 thru 2.6.10-15............... Revision 46 1.5-1........................................... Revision 46 2.6.11-1 thru 2.6.11-12............... Revision 46 2.6.12-1 thru 2.6.12-140............. Revision 46 88 drawings in the 2.7-1............................................ Revision 46 Chapter 1 List of Drawings 2.7.1-1 thru 2.7.1-117................. Revision 46 2.7.2-1 thru 2.7.2-23................... Revision 46 Chapter 1 Appendices 1-A 2.7.3-1 thru 2.7.3-5..................... Revision 46 through 1-G 2.7.4-1......................................... Revision 46 2.7.5-1 thru 2.7.5-5..................... Revision 46 Chapter 2 2.7.6-1 thru 2.7.6-4..................... Revision 46 2-i thru 2-xxv............................. Revision 46 2.7.7-1 thru 2.7.7-104................. Revision 46 2-1.............................................. Revision 46 2.8-1............................................ Revision 46 2.1.1-1 thru 2.1.1-2.................... Revision 46 2.9-1 thru 2.9-24......................... Revision 46 2.1.2-1 thru 2.1.2-3.................... Revision 46 2.10.1-1 thru 2.10.1-3................. Revision 46 2.1.3-1 thru 2.1.3-8.................... Revision 46 2.10.2-1 thru 2.10.2-49............... Revision 46 2.2.1-1 thru 2.2.1-5.................... Revision 46 2.10.3-1 thru 2.10.3-18............... Revision 46 2.3-1........................................... Revision 46 2.10.4-1 thru 2.10.4-11............... Revision 46 2.3.1-1 thru 2.3.1-13.................. Revision 46 2.10.5-1....................................... Revision 46 2.4-1........................................... Revision 46 2.10.6-1 thru 2.10.6-19............... Revision 46 2.4.1-1........................................ Revision 46 2.10.7-1 thru 2.10.7-66............... Revision 46 2.4.2-1........................................ Revision 46 2.10.8-1 thru 2.10.8-67............... Revision 46 2.4.3-1........................................ Revision 46 2.10.9-1 thru 2.10.9-9................. Revision 46 2.4.4-1........................................ Revision 46

Page 1 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

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2.10.10-1 thru 2.10.10-97.......... Revision 46 5.1.1-8...................................... Revision 23C 2.10.11-1 thru 2.10.11-10.......... Revision 46 5.1.1-9 thru 5.1.1-15................... Revision 46 2.10.12-1 thru 2.10.12-31.......... Revision 46 5.1.1-16.................................... Revision 23C 2.10.13-1 thru 2.10.13-17.......... Revision 46 5.1.1-17 thru 5.1.1-22................. Revision 46 2.10.14-1 thru 2.10.14-38.......... Revision 46 5.2.1-1 thru 5.2.1-7..................... Revision 46 2.10.15-1 thru 2.10.15-10.......... Revision 46 5.3.1-1 thru 5.3.1-2..................... Revision 46 2.10.16-1 thru 2.10.16-5............ Revision 46 5.3.2-1......................................... Revision 46 5.3.3-1 thru 5.3.3-8..................... Revision 46 Chapter 3 5.3.4-1 thru 5.3.4-27................... Revision 46 3-i thru 3-v................................. Revision 46 5.3.5-1 thru 5.3.5-4..................... Revision 46 3.1-1 thru 3.1-3.......................... Revision 46 5.3.6-1 thru 5.3.6-22................... Revision 46 3.2-1 thru 3.2-11........................ Revision 46 5.3.7-1 thru 5.3.7-19................... Revision 46 3.3-1........................................... Revision 46 5.3.8-1 thru 5.3.8-25................... Revision 46 3.4-1 thru 3.4-16........................ Revision 46 5.3.9-1 thru 5.3.9-26................... Revision 46 3.4-17 thru 3.4-110.................. Revision 23C 5.3.10-1 thru 5.3.10-14............... Revision 46 3.5-1 thru 3.5-43........................ Revision 46 5.3.11-1 thru 5.3.11-47............... Revision 46 3.6-1 thru 3.6-12........................ Revision 46 5.3.12-1 thru 5.3.12-26............... Revision 46 5.3.13-1 thru 5.3.13-18............... Revision 46 Chapter 4 5.3.14-1 thru 5.3.14-22............... Revision 46 4-i thru 4-iii................................ Revision 46 5.3.15-1 thru 5.3.15-9................. Revision 46 4.1-1 thru 4.1-4.......................... Revision 46 5.3.16-1 thru 5.3.16-5................. Revision 46 4.2-1 thru 4.2-4.......................... Revision 46 5.3.17-1 thru 5.3.17-43............... Revision 46 4.3-1 thru 4.3-4.......................... Revision 46 5.3.18-1 thru 5.3.18-2................. Revision 46 4.4-1........................................... Revision 46 5.3.19-1 thru 5.3.19-9................. Revision 46 4.5-1 thru 4.5-43........................ Revision 46 5.3.20-1 thru 5.3.20-29............... Revision 46 5.3.21-1 thru 5.3.21-45............... Revision 46 Chapter 5 5.3.22-1 thru 5.3.22-34............... Revision 46 5-i............................................. Revision 23C 5.3.23-1 thru 5.3.23-49............... Revision 46 5-ii thru 5-vi............................... Revision 46 5.3.24-1 thru 5.3.24-7................. Revision 46 5-vii.......................................... Revision 23C 5.3.25-1 thru 5.3.24-15............ Revision 23C 5-viii thru 5-xiv.......................... Revision 46 5.4.1-1 thru 5.4.1-6..................... Revision 46 5-xv thru 5-xvi......................... Revision 23C 5-1.............................................. Revision 46 Chapter 6 5-2 thru 5-4.............................. Revision 23C 6-i thru 6-xix............................... Revision 46 5.1.1-1 thru 5.1.1-7.................... Revision 46 6-1 thru 6-2................................. Revision 46

Page 2 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

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6.1-1 thru 6.1-6.......................... Revision 46 6.5.1-1 thru 6.5.1-13................... Revision 46 6.2-1........................................... Revision 46 6.5.2-1 thru 6.5.2-4..................... Revision 46 6.2.1-1 thru 6.2.1-3.................... Revision 46 6.5.3-1 thru 6.5.3-2..................... Revision 46 6.2.2-1 thru 6.2.2-3.................... Revision 46 6.5.4-1 thru 6.5.4-46................... Revision 46 6.2.3-1 thru 6.2.3-7.................... Revision 46 6.5.5-1 thru 6.5.5-15................... Revision 46 6.2.4-1...................................... Revision 23C 6.5.6-1 thru 6.5.6-20................... Revision 46 6.2.5-1 thru 6.2.5-5.................... Revision 46 6.5.7-1 thru 6.5.7-18................... Revision 46 6.2.6-1 thru 6.2.6-3.................... Revision 46 6.7.1-1 thru 6.7.1-19................... Revision 46 6.2.7-1 thru 6.2.7-2.................... Revision 46 6.7.2-1 thru 6.7.2-16................... Revision 46 6.2.8-1 thru 6.2.8-3.................... Revision 46 6.7.3-1 thru 6.7.3-39................... Revision 46 6.2.9-1 thru 6.2.9-4.................... Revision 46 6.7.4-1 thru 6.7.4-28................... Revision 46 6.2.10-1 thru 6.2.10-3................ Revision 46 6.7.5-1 thru 6.7.5-16................... Revision 46 6.2.11-1 thru 6.2.11-3................ Revision 46 6.7.6-1 thru 6.7.6-22................... Revision 46 6.2.12-1 thru 6.2.12-4................ Revision 46 6.3.1-1 thru 6.3.1-6.................... Revision 46 Appendix 6.6 6.3.2-1 thru 6.3.2-4.................... Revision 46 6.6-i thru 6.6-iii.......................... Revision 46 6.3.3-1 thru 6.3.3-9.................... Revision 46 6.6-1............................................ Revision 46 6.3.4-1 thru 6.3.4-10.................. Revision 46 6.6.1-1 thru 6.6.1-111................. Revision 46 6.3.5-1 thru 6.3.5-12.................. Revision 46 6.6.2-1 thru 6.6.2-56................... Revision 46 6.3.6-1 thru 6.3.6-9.................... Revision 46 6.6.3-1 thru 6.6.3-73................... Revision 46 6.3.7-1 thru 6.3.7-4.................... Revision 46 6.6.4.-1 thru 6.6.4-77.................. Revision 46 6.3.8-1 thru 6.3.8-7.................... Revision 46 6.6.5-1 thru 6.6.5-101................. Revision 46 6.3.9-1 thru 6.3.9-7.................... Revision 46 6.6.6-1 thru 6.6.6-158................. Revision 46 6.3.10-1 thru 6.3.10-2................ Revision 46 6.6.7-1 thru 6.6.7-84................... Revision 46 6.4.1-1 thru 6.4.1-10.................. Revision 46 6.6.8-1 thru 6.6.8-183................. Revision 46 6.4.2-1 thru 6.4.2-10.................. Revision 46 6.6.9-1 thru 6.6.9-53................... Revision 46 6.4.3-1 thru 6.4.3-35.................. Revision 46 6.6.10-1 thru 6.6.10-38............... Revision 46 6.4.4-1 thru 6.4.4-24.................. Revision 46 6.6.11-1 thru 6.6.11-53............... Revision 46 6.4.5-1 thru 6.4.5-51.................. Revision 46 6.6.12-1 thru 6.6.12-20............... Revision 46 6.4.6-1 thru 6.4.6-22.................. Revision 46 6.6.13-1 thru 6.6.13-22............... Revision 46 6.4.7-1 thru 6.4.7-13.................. Revision 46 6.6.14-1 thru 6.6.14-7................. Revision 46 6.4.8-1 thru 6.4.8-14.................. Revision 46 6.6.15-1 thru 6.6.15-................... Revision 46 6.4.9-1 thru 6.4.9-9.................... Revision 46 6.6.16-1 thru 6.6.16-30............... Revision 46 6.4.10-1 thru 6.4.10-18.............. Revision 46 6.6.17-1 thru 6.6.17-7................. Revision 46 6.4.11-1 thru 6.4.11-7................ Revision 46 6.6.18-1 thru 6.6.18-34............... Revision 46

Page 3 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

LIST OF EFFECTIVE PAGES (Continued)

6.6.19-1 thru 6.6.19-3................ Revision 46

Chapter 7 7-i thru 7-iii................................ Revision 46 7.1-1 thru 7.1-86........................ Revision 46 7.2-1 thru 7.2-17........................ Revision 46

Chapter 8 8-i............................................... Revision 46 8.1-1 thru 8.1-15........................ Revision 46 8.2-1 thru 8.2-6.......................... Revision 46 8.3-1 thru 8.3-4.......................... Revision 46

Chapter 9 9-i............................................... Revision 46 9-1 thru 9-11.............................. Revision 46

Page 4 of 4 NAC-LWT Cask SAR August 2023 Revision 23C

1979. Thermal conductivity for 6061-T651 aluminum alloy is based on ASME Code,Section II, Part D, Table TCD.

A limited number, up to 14, non-zirconium alloy clad fuel rods, in particular rods with a FeCrAl base as the alloy, may be loaded into the 25 rod tube structure. While not evaluated using ANSYS the impact on maximum reported temperature it discussed in Section 3.4.1.7.3. The impact disposition is also applicable to the damaged fuel evaluion described in Section 3.4.1.11 and accident results discussed in Section 3.5. As no maximum allowable clad temperature is available for the non-zircaloy fuel clad this material is considered to fall within the damaged allowable contents of the 25 rod shipment configuration (up to 14 damaged rods may be loaded with no clad credit applied).

The finite element model for the condition 1 is shown in Figure 3.4-10. The fuel cladding and the inner surface of the pin tube are considered to be in point-to-point contact. The outer surface of the fuel cladding only contacts the pin tube in one point in the model. The pin tubes are conservatively considered separated and a gap of 0.0005 inch between pin tubes is modeled.

This condition neglects any pin tube contact due to dead weight lo ading of the contents. One of the can weldment sides is modeled in contact with the aluminum insert. For the other three sides, a gap 0.042/0.084/0.042 inch between the aluminum insert and the tube of the can weldment is modeled. The details of this mo deling are shown in Figure 3.4-11. Likewise, only one surface between the PWR aluminum insert and the PWR basket is considered to be in contact.

Conduction (through helium) and radiation (using emissivity of stainless steel for both surfaces) are modeled from the inner shel l of the cask to the basket.

The heat transfer analysis model uses conduction in the remaining volume of the cask cavity.

The conductivity of this material corresponds to helium. (see Table 3.2-7). The properties for the remaining materials are contained in Table 3.2-1 through Table 3.2-8.

The air space between the NAC-LWT cask and the ISO container is modeled using air with an effective conductivity. This effective conductivity (Incropera) is:

1/ 4 k eff RaPr*1/4

= 0.386 ()c k 0.861+Pr

[]ln (D / D 4

0 i Ra c Ra=L3(D +3/5D3/5)5L i 0 g (T T ) 3 Ra = i o L L

NAC International 3.4-17 NAC-LWT Cask SAR August 2023 Revision 23C where:

Pr = Prandtl number (Krieth)

= kinematic viscosity (Krieth)

= thermal diffusivity (Krieth)

=1/Tf Tf = (Ti +To)/2

Ti = inner surface temperature To = outer surface temperature Di = inner diameter (cask surface)

Do = outer diameter (height of the ISO container)

L = (Do -Di )/2

3.4.1.7.2 High Burnup PWR and BWR Fuel Rods Thermal Model of the NAC-LWT (Transported via Truck Trailer)

Thermal analyses of the NAC-LWT cask for Condition 2 are performed using a half-symmetry planar cross-sectional model of the cask in which the inner surface of the inner shell is the boundary of the model. The maximum temperature of 274°F (PWR design basis fuel with 2.5 kW heat load and a peaking factor of 1.2 under normal transport condition [Table 3.4-2]) is applied to the boundary of the model. The modeling of the normal steady state condition of the NAC-LWT from the center of the cask to the inner surface of the i nner shell is identical to the model described in Section 3.4.1.7.1 with the following exceptions:

1. The gas in the NAC-LWT cask cavity is considered to be air.

2. The constant temperature of 274°F is applied to the outer surface of the model, which corresponds to the inner surface of the cask inner shell. This temperature corresponds to the condition, which imposes solar insolance and convection/radiation boundary at the outer shell of the expansion tank. This is also described in Section 3.4.1.1.

The Condition 2 model of the NAC-LWT cask with high burnup PWR and BWR fuel rods is shown in Figure 3.4-12. This model is also used to calculate both normal and accident condition temperatures for the cask.

NAC International 3.4-18 NAC-LWT Cask SAR August 2023 Revision 23C 3.4.1.7.3 High Burnup PWR and BWR Fuel Rods Heat Transfer Analyses Results The thermal analysis is performed to demonstr ate that the component temperature of NAC-LWT cask loaded with high burnup PWR and BWR rods is maintained within acceptable limits.

Maximum temperatures for p ackage components with the NAC-LWT configured for high burnup PWR and BWR rods are summarized in Table 3.4-10. As shown in Table 3.4-10, component temperatures are all maintained within their allowable temperatures.

The shipment of FeCrAl type clad rods will use the LWT 5x5 rod holder. A detailed view of the finite element model for the 5x5 rod holder is shown below. In the view shown the elements representing the cavity gas are not shown. Since radiation in the model is neglected between the clad and the tube surface, the pellet heat in the model is rejected through the pellet material and the clad. The peak clad temper ature would occur opposite the side modeling the single point of contact. The pellet heat away from the contact w ould primarily be through the pellet material and not the clad due to the larger average cro ss-sectional area and reduced length of the heat transfer path in the pellet. The pellet material properties do not change even though the clad properties will be different. A reduction in the clad properties will not contribute significantly to the peak clad temperature due to the small clad radial thickness.

NAC International 3.4-19 NAC-LWT Cask SAR August 2023 Revision 23C 3.4.1.8 Thermal Evaluation for DIDO Fuel

3.4.1.8.1 Analytical Models for the DIDO Fuel Contents Heat transfer analysis of the NAC-LWT containing DIDO fuel is performed using a two-dimensional planar finite element analysis and the general purpose ANSYS computer code.

Two transport conditions are evaluated:

Condition 1:

The NAC-LWT is supported in an ISO container with solar insolance on the surface of the ISO container, and the NAC-LWT is considered to be insulated from the environment (only for the normal conditions of transport steady state conditions). The gas inside the ISO container is air. The cavity of the NAC-LWT is backfilled with helium as required by operational procedures.

Condition 2:

The NAC-LWT is not located in an ISO container and solar insolation is applied to the NAC-LWT cask surface. For the purpose of performing the thermal analysis, the cavity of the NAC-LWT is considered to be filled with air.

A single fuel configuration is considered for this evaluation. Each DIDO fuel assembly is limited to having a heat load of 25 W per assembly. The total contents of the NAC-LWT for the DIDO fuel are limited to having six basket modules and each module is limited to having seven DIDO fuel assemblies. This limits the heat load of a basket module to 175 W, and a total NAC-LWT heat load of 1.05 kW. The 1.05 kW total heat load is enveloped by the 1.26 kW total heat load for the NAC-LWT MTR fuel contents contained in Section 3.4.1.3. Since the NAC-LWT cask ambient conditions are the same for the DIDO fuel as for the MTR fuel, the maximum temperature of all cask body components for the DIDO contents are enveloped by the maximum temperatures for the MTR fuel contents. Therefore, the cask inner shell temperature for the MTR fuel contents bounds the maximum cask inner shell temperature for the DIDO fuel contents. The maximum cask inner shell temperat ure is used as the boundary condition for the finite element model for the DIDO thermal evaluation. For Condition 1 and Condition 2, the maximum inner shell temperatures are 214 °F and 181°F, respectively. These values correspond to the design basis heat load values obtained from Table 3.4-6.

Two finite element models are used in the ev aluation of the DIDO fuel basket and the DIDO fuel assemblies.

The evaluation of the maximum basket component temperatures for these conditions is performed using a finite element model, which is shown in Figure 3.4-14. This model is used to evaluate both conditions. This model corresponds to the 4.01-inch inside diameter stainless

NAC International 3.4-20 NAC-LWT Cask SAR August 2023 Revision 23C

Table of Contents

5 SHIELDING EVALUATION................................................................................. 5-1 5.1 Discussion and Results...................................................................................... 5.1.1-1 5.1.1 NAC-LWT Contents.......................................................................................... 5.1.1-1 5.2 Gamma and Neutron Sources............................................................................ 5.2.1-1 5.2.1 ORIGEN 2......................................................................................................... 5.2.1-1 5.3 Model Specification........................................................................................... 5.3.1-1 5.3.1 Description of Radial and Axial Shielding Configuration................................. 5.3.1-1 5.3.2 Shield Regional Densities.................................................................................. 5.3.2-1 5.3.3 Metallic Fuel Configuration............................................................................... 5.3.3-1 5.3.4 MTR Fuel Configuration................................................................................... 5.3.4-1 5.3.5 25 PWR Fuel Rods Configuration..................................................................... 5.3.5-1 5.3.6 TRIGA Fuel Element Model Specification and Shielding Evaluation.............. 5.3.6-1 5.3.7 TRIGA Fuel Cluster Rod Model Specification and Shielding Evaluation........ 5.3.7-1 5.3.8 High Burnup PWR and BWR Rods Shielding Evaluation................................ 5.3.8-1 5.3.9 DIDO Fuel Configuration.................................................................................. 5.3.9-1 5.3.10 GA IFM Shielding Evaluation......................................................................... 5.3.10-1 5.3.11 High Burnup PWR and BWR Rods in a Fuel Assembly Lattice..................... 5.3.11-1 5.3.12 Damaged High Burnup PWR and BWR Rods in a Rod Holder...................... 5.3.12-1 5.3.13 TPBAR Shielding Evaluation.......................................................................... 5.3.13-1 5.3.14 PULSTAR Fuel Configuration........................................................................ 5.3.14-1 5.3.15 Spiral Fuel Assembly Configuration............................................................... 5.3.15-1 5.3.16 MOATA Plate Bundle Configuration.............................................................. 5.3.16-1 5.3.17 PWR MOX Rod Fuel Configuration............................................................... 5.3.17-1 5.3.18 Mixed ANSTO-DIDO Payload Configuration................................................ 5.3.18-1 5.3.19 Irradiated Hardware Shielding Evaluation....................................................... 5.3.19-1 5.3.20 SLOWPOKE Fuel Configuration.................................................................... 5.3.20-1 5.3.21 NRU and NRX Fuel Assemblies..................................................................... 5.3.21-1 5.3.22 HEUNL............................................................................................................ 5.3.22-1 5.3.23 SLOWPOKE Core Configuration.................................................................... 5.3.23-1 5.3.24 Booster Rods, EFN Rods and Mo-99 Targets................................................. 5.3.24-1 5.3.25 High-Burnup Steel/Iron Clad PWR and BWR Rods in a Rod Holder............ 5.3.25-1 5.4 Shielding Evaluation.......................................................................................... 5.4.1-1 5.4.1 Shielding Evaluation Codes............................................................................... 5.4.1-1

NAC International 5-i NAC-LWT Cask SAR October 2020 Revision 46

List of Figures

Figure 5.3.3-1 Three-Dimensional Radial Model............................................................ 5.3.3-2 Figure 5.3.3-2 End-Fitting Model with Fuel................................................................... 5.3.3-3 Figure 5.3.3-3 Lead Slump Accident - PWR Top End-Fitting....................................... 5.3.3-4 Figure 5.3.3-4 Lead Slump Accident - PWR Bottom End-Fitting.................................. 5.3.3-5 Figure 5.3.3-5 Lead Slump Accident - BWR Bottom End-Fitting................................. 5.3.3-6 Figure 5.3.3-6 One-Dimensional Radial Calculational Model........................................ 5.3.3-7 Figure 5.3.4-1 MTR Fuel Evaluated Configurations........................................................ 5.3.4-8 Figure 5.3.4-2 SAS4 Shielding Model for the MTR Fuel Basket in the NAC-LWT (Upper Half)............................................................................................. 5.3.4-9 Figure 5.3.4-3 Dose Rates 2 Meters from Transport Vehicle (30 W Uniform Loading)................................................................................................. 5.3.4-10 Figure 5.3.4-4 Dose Rate Profile at Radial Surface of LWT Cask - Normal Conditions -

LEU Fuel at 80% Burnup and 40W Uniform Loading.......................... 5.3.4-11 Figure 5.3.4-5 Dose Rate Profile at 2m from Conveyance Radial Surface of LWT Cask -

Normal Conditions - LEU Fuel at 80% Burnup and 40W Uniform Loading.................................................................................................. 5.3.4-11 Figure 5.3.4-6 MTR LEU Low Burnup Dose Rate Profile Comparison....................... 5.3.4-12 Figure 5.3.4-7 MTR MEU Low Burnup Dose Rate Profile Comparison...................... 5.3.4-12 Figure 5.3.4-8 MTR HEU Low Burnup Dose Rate Profile Comparison...................... 5.3.4-13 Figure 5.3.4-9 Assembly Total Neutron Source at Various Burnups - 490 grams 235U LEU Fuel with 40 W Heat Load............................................................ 5.3.4-13 Figure 5.3.4-10 Assembly Total Gamma Source at Various Burnups - 490 grams 235U LEU Fuel with 40 W Heat Load............................................................ 5.3.4-14 Figure 5.3.6-1 TRIGA Fuel Element One-Dimensional Bounding Radial Dose Rate -

Normal Conditions of Transport - Curves and Data Points.................... 5.3.6-8 Figure 5.3.6-2 TRIGA Fuel Element One-Dimensional Bounding Radial Dose Rate -

Accident Condition - Curves and Data Points...................................... 5.3.6-10 Figure 5.3.6-3 TRIGA SAS4A Radial Model Geometry............................................... 5.3.6-12 Figure 5.3.6-4 TRIGA SAS4A Basket Model Geometry.............................................. 5.3.6-13 Figure 5.3.6-5 TRIGA SAS4A Upper Half Model Geometry (Normal Condition -

Shifted Fuel)........................................................................................... 5.3.6-14 Figure 5.3.6-6 TRIGA SAS4A Upper Half Model Geometry (Normal Condition)...... 5.3.6-15 Figure 5.3.6-7 TRIGA SAS4A Lower Half Model Geometry (Normal and Accident Condition).............................................................................................. 5.3.6-16 Figure 5.3.7-1 HEU TRIGA Cluster Fuel Rod SAS2H Sample Input (600 GWd/MTU)..................................................................................... 5.3.7-3 Figure 5.3.7-2 LEU TRIGA Cluster Fuel Rod SAS2H Sample Input (140 GWd/MTU)..................................................................................... 5.3.7-5 Figure 5.3.8-1 PWR Rod SAS2H Model........................................................................ 5.3.8-6 Figure 5.3.8-2 BWR 7x7 SAS2H Model Shown at 80,000 MWd/MTU........................ 5.3.8-6 Figure 5.3.8-3 BWR 8x8 Rod SAS2H Model................................................................. 5.3.8-7 Figure 5.3.8-4 PWR Rods Axial Burnup and Source Profiles........................................ 5.3.8-7 Figure 5.3.8-5 BWR Rods Axial Burnup and Source Profiles........................................ 5.3.8-8

NAC International 5-ii NAC-LWT Cask SAR August 2023 Revision 23C

List of Figures (continued)

Figure 5.3.23-12 Normal Condition Radial Surface Dose Rate Profile by Source Type -

SLOWPOKE Core............................................................................... 5.3.23-40 Figure 5.3.23-13 Normal Condition 2-m + Conveyance Radial Surface Dose Rate Profile by Source Type - SLOWPOKE Core....................................................... 5.3.23-41 Figure 5.3.23-14 Accident Condition Radial 1m Dose Rate Profile by Source Type -

SLOWPOKE Core............................................................................... 5.3.23-42 Figure 5.3.23-15 Accident Condition Radial 1m Dose Rate Azimuthal Profile at Fuel Height -

SLOWPOKE Core............................................................................... 5.3.23-43 Figure 5.3.24-1 EFN, Moly and Booster Rod Illustrations............................................. 5.3.24-5 Figure 5.3.25-1 Normal Condition Axial Surface Dose Rate Profile by Source Type Damaged Fuel Rods with Steel/Iron Clad.............................................................. 5.3.25-6 Figure 5.3.25-2 Normal Condition Radial 2m Dose Rate Profile by Source Type - Damaged Fuel Rods with Steel/Iron Clad.............................................................. 5.3.25-7 Figure 5.3.25-3 Accident Condition Radial 1m Dose Rate Profile by Source Type - Damaged Fuel Rods with Steel/Iron Clad.............................................................. 5.3.25-8

NAC International 5-vii NAC-LWT Cask SAR October 2020 Revision 46

List of Tables

Table 5.1.1-1 Type, Form, Quantity and Potential Sources of Design Basis Fuel......... 5.1.1-8 Table 5.1.1-2 Design Basis Fuel for Shielding Evaluation............................................ 5.1.1-15 Table 5.1.1-3 Nuclear and Thermal Source Parameters................................................. 5.1.1-19 Table 5.1.1-4 Combined Dose Rates for Normal Operations Conditions.................... 5.1.1-20 Table 5.1.1-5 Hypothetical Accident - Loss of Shielding Materials........................... 5.1.1-21 Table 5.1.1-6 Hypothetical Accident - Lead Slump.................................................... 5.1.1-22 Table 5.2.1-1 LOR-2 Input Data 5.2.1-3........................................................................ 5.2.1-3 Table 5.2.1-2 Photon Spectrum for Design Basis Fuel.................................................. 5.2.1-5 Table 5.2.1-3 Fission Product Gas Inventory................................................................. 5.2.1-6 Table 5.2.1-4 Design Basis Fuel Neutron Spectrum...................................................... 5.2.1-7 Table 5.3.3-1 Source Material Compositions................................................................. 5.3.3-8 Table 5.3.3-2 Shield Material Densities and Compositions........................................... 5.3.3-8 Table 5.3.4-1 Design Basis MTR Fuel Assembly Characteristics................................. 5.3.4-15 Table 5.3.4-2 MTR Fuel Element Gamma Source Terms by Thermal Output -

380 grams 235U....................................................................................... 5.3.4-16 Table 5.3.4-3 MTR Fuel Element Neutron Source Terms by Thermal Output - 380 grams 235U.............................................................................................. 5.3.4-17 Table 5.3.4-4 MTR Fuel Element Gamma Source Terms by Thermal Output - 460 grams 235U............................................................................................ 5.3.4-18 Table 5.3.4-5 MTR Fuel Element Neutron Source Terms by Thermal Output - 460 grams 235U.............................................................................................. 5.3.4-19 Table 5.3.4-6 LEU MTR Hardware Source to Fuel Source Comparison.................... 5.3.4-20 Table 5.3.4-7 HEU MTR Hardware Source to Fuel Comparison................................ 5.3.4-21 Table 5.3.4-8 Material Densities for MTR Fuel Shielding Analysis........................... 5.3.4-22 Table 5.3.4-9 LWT Cask Surface Total Dose Rates (Normal Conditions of Transport).............................................................................................. 5.3.4-23 Table 5.3.4-10 LWT Cask Plan of Conveyance Dose Rates (Normal Conditions of Transport).............................................................................................. 5.3.4-23 Table 5.3.4-11 LWT Cask 2 Meter Off The Plane of Conveyance Dose Rates (Normal Conditions of Transport)........................................................................ 5.3.4-24 Table 5.3.4-12 LWT Cask 1 Meter From the Cask Surface Dose Rates (Normal Conditions of Transport)........................................................................ 5.3.4-24 Table 5.3.4-13 Axial Surface Dose Rates at Cask Lid (Normal Conditions of Transport).............................................................................................. 5.3.4-25 Table 5.3.4-14 LWT Cask Dose Rates 5 Meters from the Cask Lid (Back of Tractor Cab) for Normal Conditions of Transport............................................. 5.3.4-25 Table 5.3.4-15 LWT Cask Dose Rates - 1 Meter from the Cask Surface (Hypothetical Accident Conditions)............................................................................. 5.3.4-26 Table 5.3.4-16 LEU MTR Fuel Element Gamma Source Term - 40 W - 490g 235U-80%

Burnup.................................................................................................... 5.3.4-26 Table 5.3.4-17 LEU MTR Fuel Element Neutron Source Term - 40 W - 490g 235U-80%

Burnup.................................................................................................... 5.3.4-27

NAC International 5-viii NAC-LWT Cask SAR August 2023 Revision 23C

List of Tables (continued)

Table 5.3.21-21 Undamaged NRX Fuel Dose Rate Summary....................................... 5.3.21-44 Table 5.3.21-22 Collapsed NRX Fuel Dose Rate Summary.......................................... 5.3.21-45 Table 5.3.21-23 Summarized Maximum Dose Rates for Undamaged Fuel.................. 5.3.21-45 Table 5.3.21-24 Summarized Maximum Dose Rates for Collapsed Fuel...................... 5.3.21-45 Table 5.3.22-1 Composition of Solution Inorganic Chemicals.................................... 5.3.22-22 Table 5.3.22-2 Actinide Concentrations in the Solution.............................................. 5.3.22-22 Table 5.3.22-3 Inventory of Gamma-Emitting Radionuclides..................................... 5.3.22-23 Table 5.3.22-4 Fission Product Content for Defined Radionuclides........................... 5.3.22-24 Table 5.3.22-5 Fission Product Content for Undefined Radionuclides....................... 5.3.22-24 Table 5.3.22-6 Modeled HEUNL Nitrate Contents..................................................... 5.3.22-25 Table 5.3.22-7 Isotopic Contents of Actinides and Light Elements for ORIGEN-S Source Term Calculation..................................................................... 5.3.22-26 Table 5.3.22-8 Neutron Source Terms per Liter.......................................................... 5.3.22-27 Table 5.3.22-9 Gamma Source Terms per Liter........................................................... 5.3.22-28 Table 5.3.22-10 Isotopic Contents of HEUNL Materials for MCNP Shielding Evaluation

.............................................................................................................. 5.3.22-29 Table 5.3.22-11 Cask/Container Material Descriptions for HEUNL............................. 5.3.22-30 Table 5.3.22-12 HEUNL Container Dimensions........................................................... 5.3.22-31 Table 5.3.22-13 ANSI/ANS 6.1.1-1977 Neutron Flux-to-Dose Conversion Factors.... 5.3.22-32 Table 5.3.22-14 ANSI/ANS 6.1.1-1977 Gamma Flux-to-Dose Conversion Factors..... 5.3.22-33 Table 5.3.22-15 HEUNL Dose Rate Summary.............................................................. 5.3.22-34 Table 5.3.22-16 Summarized Maximum Dose Rates for HEUNL Transport................ 5.3.22-34 Table 5.3.23-1 SLOWPOKE Fuel Geometry and Materials........................................ 5.3.23-44 Table 5.3.23-2 Source Term Generation Parameters for SLOWPOKE Fuel............... 5.3.23-44 Table 5.3.23-3 SLOWPOKE Neutron Source Term (per rod)..................................... 5.3.23-45 Table 5.3.23-4 SLOWPOKE Fuel Gamma Source Term (per rod)............................. 5.3.23-46 Table 5.3.23-5 Canister/Basket/Cask Material Descriptions for SLOWPOKE Fuel... 5.3.23-47 Table 5.3.23-6 Modeled SLOWPOKE Core Basket Dimensions................................ 5.3.23-48 Table 5.3.23-7 Maximum Dose Rates for SLOWPOKE Fuel..................................... 5.3.23-49 Table 5.3.23-8 Summarized Maximum Dose Rates for SLOWPOKE Fuel................ 5.3.23-49 Table 5.3.24-1 Additional Payload Fuel Material and Geometry Properties................. 5.3.24-6 Table 5.3.24-2 Additional Payload Caddy Content Material Density........................... 5.3.24-7 Table 5.3.25-1 PWR Rods 80,000 MWd/MTU, 240 Day Cool Time Source Terms.... 5.3.25-9 Table 5.3.25-2 BWR 7x 7 Rods 80,000 MWd/MTU, 300 Day Cool Time Source Terms

.......... 5.3.25-10 Table 5.3.25-3 BWR 8x 8 Rods 80,000 MWd/MTU, 240 Day Cool Time Source Terms

.......... 5.3.25-11 Table 5.3.25-4 Fuel Region Homogenization for PWR Fuel Rods............................. 5.3.25-12 Table 5.3.25-5 Fuel Region Homogenization for BWR 7x7 Fuel Rods...................... 5.3.25-12 Table 5.3.25-6 Region Homogenization for BWR 8x8 Fuel Rods.............................. 5.3.25-13 Table 5.3.25-7 Intact/Damaged Fuel Mixture Composition Determinations............... 5.3.25-13 Table 5.3.25-8 Fuel Region Homogenized Material Description................................ 5.3.25-14 Table 5.3.25-9 Maximum Radial Dose Rates for Damaged PWR and BWR Fuel Rods

.......... 5.3.25-15

NAC International 5-xv NAC-LWT Cask SAR August 2023 Revision 23C

List of Tables (continued)

Table 5.3.25-9 Maximum Axial Dose Rates for Damaged PWR and BWR Fuel Rods

.......... 5.3.25-15 Table 5.4.1-1 Discrete Axial Source Distribution.......................................................... 5.4.1-4 Table 5.4.1-2 Flux to Dose Conversion Factors............................................................. 5.4.1-6

NAC International 5-xvi NAC-LWT Cask SAR October 2020 Revision 46

5 SHIELDING EVALUATION The NAC-LWT cask utilizes a concentric cylindrical arrangement of steel, lead, steel and water to provide gamma shielding for the design basis fuel. The water-glycol solution in the neutron shield tank also provides neutron shielding. The water contains 1 weight per cent (wt %) boron, which absorbs neutrons without producing significant secondary gamma radiation.

The PWR and BWR design basis shielding analysis uses the LOR-2 version of the ORIGEN-2 code to calculate radiation sources. The QAD-CG (Cain) and XSDRNPM (NUREG/CR-0200, Vol, 2, F3) codes are used to calculate the cask dose rates for normal operations and hypothetical accident conditions. The shielding analysis shows that the dose rates are below regulatory limits specified in 10 CFR 71.47 and 71.51 as well as IAEA Transportation Safety Standards (TS-R-1).

The PWR and BWR design basis shielding analyses were performed for a 0.25 inch thick neutron shield tank shell, while the actual fabricated thickness is only 0.24 (6mm). The shell thickness difference of 0.01 inches yields a maximum dose rate increase of only 2.4 percent, which gives lower dose rates than worst case toleran ce analysis in this chapter. The analyses of this chapter, therefore, are valid.

The MTR design basis shielding analysis used the SCALE package. This included SAS2H (Herman) for source terms, and SAS4 (Tang) for three-dimensional shielding analysis. This evaluation is presented in Section 5.3.4. This shielding analysis shows that dose rates are below regulatory limits when the NAC-LWT contains up to 42 design basis MTR fuel elements with less than 210 watts of decay heat per basket.

The MTR shielding analysis explicitly calculated dose rates for LEU, MEU and HEU MTR fuel for a range of burnups and cool times to meet decay heat and dose rate limits. HEU fuel source terms were higher and thus the HEU fuel provides the most limiting dose rates for fixed decay heat limits.

The 25 PWR rod design basis shielding analysis used the SCALE package. This included SAS2H for source terms and SAS1 for one-dimensional radial shielding analysis. This analysis is presented in Section 5.3.5. This shielding analysis shows that the dose rates are below regulatory limits when the NAC-LWT contains up to 25 design basis PWR rods. A shielding evaluation of high burnup PWR and BWR fuel rods in a rod holder is presented in Section 5.3.8.

Up to 25 PWR and BWR fuel rods are evaluated at burnups up to 80,000 MWd/MTU.

The NAC-LWT is evaluated for the transport of up to 140 TRIGA fuel elements or up to 560 TRIGA fuel cluster rods arranged in five (5) basket modules. This shielding evaluation uses the SCALE package with the SAS2H sequence for source term identification, and SAS4, also from the SCALE package, to perform a three-dimens ional shielding analysis. The analysis is

NAC International 5-1 NAC-LWT Cask SAR August 2023 Revision 23

presented in Section 5.3.6. The analysis shows that the dose rates are below the regulatory limits when the cask contains up to 140 TRIGA fuel elements each having a maximum decay heat of 7.5 W, or up to 560 TRIGA fuel cluster rods each having a maximum decay heat of 1.875 W.

There are two TRIGA basket configurations, non-poisoned and poisoned, as described in Section 1.2.3.1.2. Each TRIGA basket module consists of seven cells. The center cell of each non-poisoned basket module is blocked with a stainless steel plate. Consequently, only six (6) cells of each non-poisoned basket module are loaded with fuel. Because the shielding analyses assumes the center cell contains the bounding TRIGA fuel elements or TRIGA fuel cluster rods during the normal and accident conditions of transport; the evaluation of 140 fuel elements or 560 fuel cluster rods bounds the 120 fuel element / 480 fuel cluste r rod configurations.

The DIDO design basis shielding analysis used the SCALE package. This included SAS2H (Herman) for source terms, and SAS4 (Tang) for three-dimensional shielding analysis. This evaluation is presented in Section 5.3.9. This shielding analysis shows that dose rates are below regulatory limits when the NAC-LWT contains up to 42 design basis DIDO fuel assemblies with two allowable heat loads per basket module, either 175 watts or 126 watts, dependent on the axial position of the fuel elements in the top basket.

The DIDO shielding analysis explicitly calculated dose rates for LEU, MEU and HEU DIDO fuel for a range of burnups and cool times to meet decay heat and dose rate limits. HEU fuel source terms were higher and thus the HEU fuel provides the most limiting dose rates for fixed decay heat limits.

The analysis of General Atomics (GA) Irradiated Fuel Material (IFM) used the SCALE package.

The GA IFM consists of two Fuel Handling Units, one containing RERTR (an Incoloy clad TRIGA type fuel) and the other containing HTGR graphite matrix fuel material. The analysis included ORIGEN-S for source terms and SAS1 for one-dimensional radial shielding analysis.

This evaluation is presented in Section 5.3.10. The shielding evaluation shows that dose rates are well below regulatory limits for a combined payload of the two Fuel Handling Units.

Up to 25 high burnup intact PWR or BWR fuel rods loaded into a fuel assembly lattice are analyzed in Section 5.3.11. Source terms were calculated using SAS2H with three-dimensional dose rates calculated using the MCBEND Monte Carlo transport code. Up to 14 high burnup damaged fuel rods may be loaded in a shipment of 25 PWR or BWR fuel rods, as demonstrated in Section 5.3.12. Damaged rods must be loaded in the rod holder. Source terms were calculated using SAS2H with three-dimensional dose rates calculated using the MCBEND Monte Carlo transport code.

Similarly, up to 14 fuel rods with steel/iron cladding, or non-fuel rods sections with the non-zirconium alloy tube material, may be loaded in to a shipment of up 25 PWR or BWR fuel rods,

NAC International 5-2 NAC-LWT Cask SAR August 2023 Revision 23 as demonstrated in Section 5.3.25. The term steel/iron clad is used in this context to describe non-zirconium alloy clad such as FeCrAl clad (variations of the alloy may be loaded). Presence of non-zirconium based clad produces higher gamma source due to presence of higher cobalt levels. Fuel rods and non-fueled sections are evaluated up to the high burnup levels specified for the zirconium alloy clad based rods. Shipment of mixed zirconium and non-zirconium clad materials are permitted. The non-zirconium alloy rods are considered to fail during transport within the context of the shielding evaluation. These rods must be loaded into the rod holder to be bound by the shielding analysis. Three-dimensional dose rates were calculated using the MCNP Monte Carlo transport code.

A combination of up to 16 high burnup undamaged PWR MOX or UO2 fuel rods loaded into a 5x5 rod holder is analyzed in Section 5.3.17. Remaining slots in the 5x5 lattice may be occupied by zirconium alloy-based hardware components such as burnable poison rods (BPRs), provided they are not comprised of highly activated materials (e.g., steel or inconel). The rod lattice is located within a canister placed into an insert located within the NAC-LWT PWR basket.

Source terms were calculated using SCALE 5.0 SAS2H, with three-dimensional dose rates calculated using the MCNP5 Monte Carlo transport code.

An analysis of the content condition of 300 production Tritium Producing Burnable Absorber Rods (TPBARs) in a consolidation canister used the ORIGEN-S module of the SCALE package for source terms and the MCNP code package to calculate three-dimensional dose rates. This evaluation is presented in Section 5.3.13 and shows that dose rates are well below regulatory limits for normal and accident conditions. The second TPBAR content conditi on of 55 segmented TPBARs cooled for a minimum of 90 days is evaluated using the source terms determined by the ORIGEN-S module of the SCALE package. This evaluation readily shows compliance with the previously calculated regulatory dose rates for 300 production TPBARs cooled a minimum of 30 days.

A payload of up to 700 PULSTAR fuel elements is analyzed in Section 5.3.14. Source terms were calculated using SAS2H with three-dimensional dose rates calculated using the MCNP code. PULSTAR fuel elements may be loaded as assemblies in a 5 x5 rectangular array; intact elements in a 4x4 fuel rod insert; or intact or damaged elements and nonfuel components of fuel assemblies in a can. Four 7-element MTR basket modules are stacked to form a 28 MTR basket in the cask cavity. The maximum cell loading is 25 elements.

A payload of up to 42 spiral fuel assemblies or 42 MOATA plate bundles in the ANSTO basket is analyzed in Section 5.3.15. Six 7-element ANSTO basket modules are stacked to form a 42-assembly payload in the cask cavity. Source terms were calculated using SAS2H. Due to similarities in the basket design to the DIDO basket and bounding source terms in the DIDO shielding evaluation, no shielding evaluations are required to dem onstrate regulatory compliance.

NAC International 5-3 NAC-LWT Cask SAR August 2023 Revision 23 A payload of up to 800 SLOWPOKE fuel elements is analyzed in Sect ion 5.3.20. Source terms were calculated using SCALE 6.1 TRITON with three-dimensional dose rates calculated using MCNP5 v1.3. SLOWPOKE fuel elements may be loaded into the canister assembly with either 5x5 or 4x4 canister inserts, with up to four inserts per canister assembly. The canister assemblies may be loaded into the top two 7-element MTR baskets in a 28 MTR basket stack. The SLOWPOKE elements are found to be well within regulatory limits at the minimum cool time of 14 years.

A payload of up to 18 NRU or NRX fuel assemblies is analyzed in Section 5.3.21. Source terms were calculated using TRITON in SCALE 6.1 with three-dimensional dose rates calculated using the MCNP code. NRU HEU and NRX fuel assemblies are found to be within regulatory limits at their respective minimum cool times of 18 and 19 years. NRU LEU fuel assemblies are found to be within regulatory compliance at a minimum cool time of 3 years.

A payload of 4 HEUNL containers is analyzed in Section 5.3.20. Source terms were calculated using an inventory of gamma-emitting radionuclides with the actinide and nitrate contents. The ORIGEN-S control module in SCALE 6.1 is used to calculate source spectra. A maximum payload of 64.3 L (17.0 gal) per container is conservatively applied for the source strength (due to void volume in the container that allows HEUNL thermal expansion, actual container capacity is less). Three-dimensional dose rates are calculated using the MCNP v1.60 code. The HEUNL payload is found to be within regulatory limits.

A payload of a SLOWPOKE core, up to 298 rods with 2.81 g 235U per rod with a minimum enrichment of 90 wt% 235U, is analyzed in Section 5.3.23. Source terms were calculated using TRITON in SCALE 6.1 with three-dimensional dose rates calculated using MCNP5 v1.6. The SLOWPOKE core is found to be within regulatory limits after a minimum cool time of 14 days.

A payload of Booster rods, EFN (Enriched Fast Neutron) rods, or Mo-99 (Moly) targets is analyzed in Section 5.3.24. Source terms were generated using TRITON in SCALE 6.1 for each of the payload types in the NRU/NRX basket and compared to those of the previously evaluated NRU-HEU payload. As demonstrated in Section 5.3.24, the sources and source density associated with the additional payloads are less than those previously evaluated. Dose rate results are therefore bounded by the NRU-HEU configuration.

NAC International 5-4 NAC-LWT Cask SAR October 2020 Revision 46

A payload of 1 SLOWPOKE fuel core is analyzed in Section 5.3.23. The fuel core contains up to 298 SLOWPOKE fuel elements (rods). The core is placed in a SLOWPOKE fuel core basket which is placed on a stack of empty intermediate and bottom MTR-42 basket modules which serve as spacers. A maximum 235U mass of 2.81 g per rod is evaluated. Fuel core 235U content evaluated is 837 gram at a core average depletion of 2.12 % 235U.

A payload of Booster rods, EFN (Enriched Fast Neutron) rods, or Mo-99 (Moly) targets is analyzed in Section 5.3.24. The rods/targets are placed in the NRU/NRX caddy. Small fragments require the use of a caddy. The rods/targets are U-Al alloy and HEU. A 91 wt % U-235 enrichment is evaluated for each rod/target and each has a unique weight, U-235 depletion percentage, and minimum c ool time per rod/target.

The shield materials are selected and arranged to minimize cask weight while maintaining overall shield effectiveness. Lead and steel are chosen as effective gamma radiation shields, and a water tank on the outside of the cask is provided to efficiently moderate and absorb the neutron radiation.

The total neutron and gamma dose rates calculated for the normal operations conditions are shown in Table 5.1.1-4. Note that the maximum dos e rate is on the cask lid surfaces at the top end of the cask and does not exceed the design limit of 200 mrem/hour for the surface of the cask. The 10 CFR 71 limits of 10 mrem/hour at two meters from the cask surface and the design limit of 200 mrem/hour on the cask surface are met. Table 5.1.1-4 contains the total dose rates for the hypothetical accident conditions. These dose rates are well under the 49 CFR 173 limit of 1000 mrem/hour at one meter from the cask surface. The dose rates for the lead slump accident are shown in Table 5.1.1-5. These dose rates show that even with the lead slumped, the hypothetical accident dose rate limits have not been exceeded and the cask is safe for transport.

The cask surface fuel centerline normal operations and hypothetical accident dose rates calculated include neutrons and gammas originating from the fuel, neutrons and gammas scattered from the ground and secondary gammas resulting from neutron capture in the neutron shield. All of the other dose locations also include the contribution from the 60Co in the end-fittings.

NAC International 5.1.1-7 NAC-LWT Cask SAR August 2023 Revision 23C

Table 5.1.1-1 Type, Form, Quantity and Potential Sources of Design Basis Fuel Fuel Type - PWR, Assembly

- 3.7 wt % 235U maximum initial enrichment

- 35,000 MWd/MTU maximum burnup

- 2.5 kW per assembly maximum decay heat

- 2 years (or more) decay time after reactor discharge Fuel Form - Intact assemblies Quantity - 1 design basis fuel assembly Source of Fuel - Commercial PWR nuclear power reactors Transport Index - 35

Fuel Type - BWR, Assembly

- 4.0 wt % 235U maximum initial enrichment

- 30,000 MWd/MTU maximum burnup

- 1.1 kW per assembly maximum decay heat, 2.2 kW per cask for 2 assemblies

- 2 years (or more) decay time after reactor discharge Fuel Form - Intact assemblies Quantity - 2 design basis fuel assemblies Source of Fuel - Commercial BWR nuclear power reactors Transport Index - 35

Fuel Type High Burnup PWR or BWR rods

- 5.0 wt % maximum 235U initial enrichment

- 80,000 MWd/MTU maximum average burnup

- 2.3 kW /cask maximum decay heat

- Minimum cool time dependent on burnup (See Table 5.3.8-29. An additional 90 days cool time is required when loading steel/iron clad, e.g. FeCrAl type alloy, materials)

Fuel Form - Intact rods in a fuel assembly lattice or rod holder and intact rods with up to 14 fuel rods classified as damaged in a rod holder Quantity - Up to 25 Source of Fuel - Commercial PWR or BWR nuclear power reactor Transport Index - 36 (intact rods) 28 (intact rods in a fuel assembly lattice) 37 (intact rods with 14 rods classified as damaged) 37 (intact rods with 14 steel/iron clad rods considered damaged within the analysis)

Fuel Type - Uranium metal fuel rods

- Natural wt % 235U

- 1,600 MWd/MTU maximum burnup

- 0.0357 kW per sound rod maximum decay heat, 0.54 kW per cask for 15 sound fuel rods

- 1 year (or more) decay time after reactor discharge Fuel Form - Intact or encapsulated failed fuel rods Quantity - 15 design basis fuel rods, or 6 design basis failed fuel rods Source of Fuel - Research reactors Transport Index - 25

NAC International 5.1.1-8 NAC-LWT Cask SAR October 2020 Revision 46

Table 5.1.1-2 Design Basis Fuel for Shielding Evaluation MTR Parameter PWR BWR Metallic MTR (HEU) (MEU) MTR (LEU) DIDO Assembly Array 15 x 15 7 x 7 N/A Parallel Plates Parallel Parallel Plates Fuel Tubes Plates Assembly or Element Weight 1650 750 1805 13.0 (max) 13.0 (max) 13.0 (max) 15.0 (max)

(lbs) (15 rods)

Assembly/Element/Rod Length 162 176 120.5 25.235 26.145 26.145 24.6 (in)

Active Fuel Length (in) 144 144 120.0 24.80 25.59 25.59 23.6 No. Rods per Assembly 204 49 N/AN/AN/AN/AN/A No. of Plates per Element N/A N/A N/A 23 23 23 4 Fuel Rod Diameter/Plate 0.422 0.563 1.36 0.050 0.050 0.050 0.059 Thickness (in)

Clad Material Zr-4 Zr-4 Al Al Al Al Al Clad Thickness (in) 0.0243 0.032 0.080 0.0150 0.0150 0.0150 0.0167 Pellet Diameter/Meat Thickness 0.3659 0.487 1.36 0.020 0.020 0.020 0.026 (in)

Fuel Material UO2 UO2 U metal U3O8-Al; U3O8-Al; U3O8-Al; U3O8-Al; U-Al; or U-Al; or U-Al; or U-Al; or U3Si2-Al U3Si2-Al U3Si2-Al U3Si2-Al Percent Theoretical Density 95 95 100 N/A N/A N/A N/A Enrichment (wt % 235 U) 3.7 4.0 Natural 908 408 198 90 (HEU) 400 (MEU) 199 (LEU)

Maximum Average Burnup 35,000 30,000 1,600 Variable up to Variable up Variable up to Variable up to (MWd/MTU) 660,0002,9 to 293,3002 139,3002 577,460 (HEU) 256,650 (MEU) 121,910 (LEU)

Minimum Cool Time 2 Years 2 Years 1 Year Variable down Variable Variable down Variable down to 90 days2 down to 90 to 90 days2 to 180 days10 days2 U Weight (kg/assembly) 475 198 N/AN/AN/AN/AN/A U Weight (kg/element) N/A N/A 54.5 0.422 0.950 3.3684 0.2111 (HEU) 0.511 0.4750 (MEU) 1.0000 (LEU)

UO2 Weight (kg/assembly) 538.9 224.3 N/A N/A N/A N/A N/A

Notes:

1. Up to 2 of the PWR rods may have a maximum average burnup of 65,000 MWd/MTU.

2. Variable cool time down to 90 days using the procedure in Section 7.1.4.

3. Design Basis normal condition source term is for ACPR fuel with 86,100 MWd/MTU (50% 235U depletion) and accident condition source term is for FLIP-LEU-II with 151,100 MWd/MTU (80% 235U depletion).

4. Detailed fuel data is presented in Tables 1.2-1 and 6.2.5-1. The values presented here are the physical values for the bounding source terms of the ACPR and FLIP-LEU-II fuel types.

5. For MTR fuel assemblies, which are cut to remove non-fuel bearing hardware prior to transport, a nominal 0.28 inch of nonfuel hardware will remain above and below the active fuel region to allow for fuel handling operations

6. Minimum cool time varies with burnup such that maximum decay heat is 1.875 watts/rod.

7. Varies with burnup - see Table 5.3.8-29.

8. For the shielding evaluation, lower values are conservatively assumed.

9. Maximum burnup of 660,000 MWd/MTU for 380 g 235U and 577,500 MWd/MTU for 460 g 235U.

10. Variable cool time down to 180 days using the procedure in Section 7.1.4.

NAC International 5.1.1-15 NAC-LWT Cask SAR August 2023 Revision 23C

Table 5.1.1-2 Design Basis Fuel for Shielding Evaluation (continued)

PWR PWR High B/U High B/U BWR MOX/UO2 TRIGA Fuel Parameter Rods PWR Rods Rods Rods TRIGA4 Cluster Rods TPBARs Assembly Array N/A N/AN/AN/A N/A N/A N/A Assembly or Element Weight N/A N/A N/A N/A 8.82 (nominal) 2.655 (lbs) 13.2 (max)

Assembly/Element/Rod Length 162 162 176.1 162 45 31.0 153.035 (in) (pre-irradiation)

Active Fuel Length (in) 144 150 150 153.5 15 22 N/A No. Rods per Assembly per 25 25 25 16 1 1 300 Production Shipment or 55 Segmented No. of Plates per Element N/A N/A N/A N/A N/A N/A N/A Fuel Rod Diameter/Plate 0.422 0.440 0.570 (7x7) 0.440 1.478 0.542 0.381 Thickness (in) 0.4961 (other)

Clad Material Zr-4 Zr-4 Zr-2 Zirc Alloy 304SS Incoloy 800 316 SS Clad Thickness (in) 0.242 0.026 0.036 (7x7) 0.026 0.02 0.016 0.0225 0.0343 (other)

Pellet Diameter/Meat Thickness 0.3659 0.3805 0.4900 (7x7) 0.3805 1.435 (max) 0.510 N/A (in) 0.4213 (other)

Fuel Material UO2 UO2 UO2 UO2 - PuO2/ U-ZrH U-ZrH N/A UO2 Percent Theoretical Density 97 95 95 95 95 95 N/A Enrichment (wt % 235 U) 5.0 5.0 5.0 5.0 (UO2) 20 92 (HEU)N/A 7.0 fissile Pu 19 (LEU)

(MOX))

Maximum Average Burnup 60,0001 80,000 60,000 - 62,500 ACPR 86,100 Variable up to N/A (MWd/MTHM) 80,000 (50% 235U)3 600,000 (HEU)

FLIP-LEU-II Variable up to 151,100 140,000 (LEU)

(80% 235U)3 Minimum Cool Time 150 Varies with Varies with 90 days ACPR 231 Varies with 30 days for days burnup7 burnup7 (Power Grade days burnup6 production MOX - 120 FLIP-LEU-II TPBAR; 90 days days) 908 days for PIE TPBAR U Weight (kg/assembly) 58.2 65.6 108.8 (7x7) N/A N/A N/A N/A 91.3 (other)

HM Weight (kg/element) N/A N/A N/A 2.6311 ACPR 0.280 0.0505 (HEU) N/A FLIP-LEU-II 0.2894 (LEU) 0.824 UO2 Weight (kg/assembly) 66.0 66.0 74.5 N/A N/A N/A N/A

Notes:

1. Up to 2 of the PWR rods may have a maximum average burnup of 65,000 MWd/MTU.

2. Variable cool time down to 90 days using the procedure in Section 7.1.4.

3. Design Basis normal condition source term is for ACPR fuel with 86,100 MWd/MTU (50% 235U depletion) and accident condition source term is for FLIP-LEU-II with 151,100 MWd/MTU (80% 235U depletion).

4. Detailed fuel data is presented in Tables 1.2-1 and 6.2.5-1. The values presented here are the physical values for the bounding source terms of the ACPR and FLIP-LEU-II fuel types.

5. For MTR fuel assemblies, which are cut to remove nonfuel-bearing hardware prior to transport, a nominal 0.28 inch of nonfuel hardware will remain above and below the active fuel region to allow for fuel handling operations.

6. Minimum cool time varies with burnup such that maximum decay heat is 1.875 watts/rod.

7. Varies with burnup - see Table 5.3.8-29. An additional 90 days cool time is required for steel/iron clad rods. Non-zirconium alloy based fuel clad are considered damaged within evaluation space.

8. For the shielding evaluation, lower values are conservatively assumed.

9. Maximum burnup of 660,000 MWd/MTU for 380 g 235U and 577,500 MWd/MTU for 460 g 235U.

10. Variable cool time down to 180 days using the procedure in Section 7.1.4.

11. Heavy metal weight per rod.

NAC International 5.1.1-16 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C

5.3.25 High Burnup Steel/Iron Clad PWR and BWR Rods in a Rod Holder Results of a shielding analysis for up to 25 high burnup PWR or BWR fuel rods with a maximum of 14 non-zirconium clad fuel rods are presented in this section. The non-zirconium clad materials are considered damaged in this evaluation. The 14 damaged fuel rods are analyzed with steel/iron clad The rods have burnups up to 80,000 MWd/MTU. Based on the evaluation of steel/iron clad rods instead of zirconium-based clad rods, the minimum cool times developed in Section 5.3.8 require an extension to ensure that maximum dose rates do not exceed dose rate limits.

. Dose rates are calculated using the MCNP three-dimensional Monte Carlo transport code.

Source Terms Source terms employed in this analysis are identical to those employed in Section 5.3.8 above.

The SAS2H-generated source spectra are rebinned onto 28-group neutron and 22-group gamma schemes as shown in Table 5.3.11-1 and Table 5.3.11-2, respectively.

Source terms are presented in Table 5.3.25-1 through Table 5.3.25-3 for PWR and BWR fuel.

PWR and BWR 88 fuel types are analyzed at 80,000 MWd/MTU and 240 days cool time.

BWR 77 fuel is analyzed at 80,000 MWd/MTU and 300 days cool time. These cool times represent a 90 day addition to the zirconium alloy based high burnup fuel rods minimum cool times.

The effect of subcritical neutron multiplication is computed directly in MCNP.

Axial Source Profile The axial source profiles employed in MCNP for PWR and BWR fuel are identical to those employed in Section 5.3.12.2. Profiles are input by evaluating the fraction of source in each axial bin. A uniform source is applied to the damaged fuel concentrated at the top of the cask.

The nonlinear impact of burnup on the total neutron source strength is accounted for in both the damaged and intact fuel regions.

NAC International 5.3.25-1 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Shielding Model MCNP three-dimensional shielding analysis allows detailed modeling of the fuel, basket, and cask shield configurations. For the fuel rod sources, some fuel rod detail is homogenized in the model to simplify model input and improve computational efficiency. Thus, the three-dimensional models represent the various fuel assembly source regions as homogenized zones within the rod holder, but explicitly model the ax ial extent of the source regions. The basket and cask body details are explicitly modeled, including the axial extents described by the License Drawings.

The geometric description of a MCNP model is based on the combinatorial geometry system embedded in the code. In this system, bodies such as cylinders and rectangular parallelepipeds, and their logical intersections and unions, are used to describe the extent of material zones.

MCNP weight windows are used to accelerate problem convergence.

Fuel Rod Model Based on the fuel parameters provided in Section 5.3.8, and the rod holder cross-sectional detail provided by the License Drawings, homogenized treatments of fuel rod source regions are developed. The homogenized fuel rods are represented in the m odel as a stack of boxes with width equal to the rod holder interior width. The height of each box corresponds to the modeled height of the corresponding source region.

The intact fuel region homogenizations are shown in Table 5.3.25-4 through Table 5.3.25-6, based on a homogenization area of Components of the fuel homogenization are subdivided to account for the various area fractions present in th e homogenized fuel description.

Interstitial refers to the space within the rod holder canister but outside the 5x5 tube array.

Insert void refers to the space inside the rod holder tubes but outside the fuel rods. Gap refers to the pellet to clad gap. All three regi ons are assigned a void material as part of the shielding evaluation since the cask cavity is dry during all transport conditions. Combined with the fuel rod clad, fuel material, and tube materials, the void accounts for the total fuel region volume. The clad region from a material comp osition perspective is set to zirconium alloy (density 6.56 g/cm3) for both PWR and BWR fuel. The non-zirconium alloys are slightly higher in density and therefore provides slightly higher source self-shielding.

NAC International 5.3.25-2 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C

mixture calculations are summarized in Table 5.3.25-7.

Credit is taken for the self-shielding of intact and damaged fuel spanni ng the top of the active fuel region. Similarly, credit is taken for the reduction in active fuel source required for the dispersion of 14 rods. The total source evaluated is 25 rods.

Basket Model For a given fuel type, the MCNP description of the basket elements forms a common sub-model employed in the analysis. The key features of the model are the detailed representation of pin canister, PWR insert, and PWR basket.

NAC-LWT Model The three-dimensional model of the NAC-LWT cask is based on the following features:

Normal conditions:

Radial neutron shield and shield shell

Aluminum impact limiters with (calculated based on the impact limiter weight and dimensions) and diameter equal to the neutron shield shell diameter Accident conditions:

Removal of radial neutron shield and shield shell

Loss of upper and lower impact limiters Common to both the normal and accident conditions models is a

. The elevation of the source regions is controlled such that the offset of the rod holder canister from the bottom of the NAC-LWT

as shown in Figure 5.3.10-6, the least radial shielding is located.

Detailed model parameters used in creating the three-dimensional model are taken directly from the License Drawings. Elevations associated with the three-dimensional features are established with respect to the center bottom of the NAC-LWT cask cavity for th e MCNP combinatorial model. The three-dimensional NAC-LWT models are identical to those shown in Figure 5.3.12-1 and Figure 5.3.12-2.

Shield Regional Densities Based on the homogenization described for the fuel rod model, the resulting fuel regional densities are shown in Table 5.3.25-8. Material compositions for st ructural and shield materials are shown in Table 5.3.17-12.

NAC International 5.3.25-3 NAC-LWT Cask SAR August 2023 Revision 23C Shielding Evaluation Calculational Methods The shielding evaluation is performed using MCNP. As described above, the evaluation includes the effect of fuel burnup peaking on fuel neutron and gamma source terms.

The MCNP shielding model is utilized with the source terms described above to estimate the dose rate profiles at various distances from the side, top, and bottom of the cask for both normal and accident conditions. The method of solution is continuous energy Monte Carlo with weight windows used to accelerate problem convergence.

Significant validation literature is available for MCNP as it is an industry standard tool for spent fuel cask evaluations. Available literature covers a range of shielding penetration problems ranging from slab geometry to spent fuel cask geometries. Confirmatory calculations against other validated shielding codes (SCALE and MCBEND) on NAC casks have further validated the use of MCNP for shielding evaluations.

Flux-to-Dose Rate Conversion Factors The ANSI/ANS 6.1.1-1977 flux-to-dose rate conversion factors are employed in the MCNP analysis. The ANSI/ANS gamma and neutron dose conversion factors are shown in Table 5.3.11-23 and Table 5.3.11-24.

Three-Dimensional Dose Rates for High Burnup Fuel Table 5.3.25-9 and Table 5.3.25-10 summarize the computed dose rates for each fuel type at the tabulated distances and transport conditions (norma l and accident). The highest calculated radial dose rates at the surface and 2-meter locations under normal conditions are for BWR 8x8 and BWR 7x7 rods, respectively. The highest calculated radial dose rates at 1 meter from the cask under accident conditions are for BWR 7x7 rods.

Normal condition radial surface dose rates for all three fuel types are in excess of 200 mrem/hr, necessitating an exclusive use designation fo r the NAC-LWT. The maximum dose rate is dominated by the damaged fuel neutron component, which comprises approximately 77% of the maximum dose rate. The axial elevation of the maximum dose rate is above the neutron shield.

The dose rate profile is shown in Figure 5.3.25-1.

The normal condition maximum radial 2-meter dose rate is 9.6 mrem/hr. At this distance, the damaged fuel neutron component contributes approximately 43% of the maximum. The dose rate profile is skewed towards the top of cask, as shown Figure 5.3.25-2.

Accident condition radial 1-meter dose rates for all three fuel types are well below the 1,000 mrem/hr limit. The maximum dose rate is dominated by the mixture fuel neutron component,

NAC International 5.3.25-4 NAC-LWT Cask SAR August 2023 Revision 23C which contributes approximately 51% towards the maximum. The dose rate profile is shown in Figure 5.3.25-3.

As shown in Table 5.3.25-10, axial surface dose rates are well below limits for all three fuel types. Significant margin is present for th e normal condition 2-meter and accident condition 1-meter dose rate limits.

NAC International 5.3.25-5 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Figure 5.3.25-1 Normal Condition Axial Surface Dose Rate Profile by Source Type -

Damaged Fuel Rods with Steel/Iron Clad

NAC International 5.3.25-6 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Figure 5.3.25-2 Normal Condition Radial 2m Dose Rate Profile by Source Type -

Damaged Fuel Rods with Steel/Iron Clad

NAC International 5.3.25-7 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Figure 5.3.25-3 Accident Condition Radial 1m Dose Rate Profile by Source Type -

Damaged Fuel Rods with Steel/Iron Clad

NAC International 5.3.25-8 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Table 5.3.25-1 PWR Rods 80,000 MWd/MTU, 240 Day Cool Time Source Terms

NAC International 5.3.25-9 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Table 5.3.25-2 BWR 7x7 Rods 80,000 MWd/MTU, 300 Day Cool Time Source Terms

NAC International 5.3.25-10 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Table 5.3.25-3 BWR 8x8 Rods 80,000 MWd/MTU, 240 Day Cool Time Source Terms

NAC International 5.3.25-11 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Table 5.3.25-4 Fuel Region Homogenization for PWR Fuel Rods

Table 5.3.25-5 Fuel Region Homogenization for BWR 7x7 Fuel Rods

NAC International 5.3.25-12 NAC PROPRIETARY INFORMATION REMOVED

NAC-LWT Cask SAR August 2023 Revision 23C Table 5.3.25-6 Region Homogenization for BWR 8x8 Fuel Rods

Table 5.3.25-7 Intact/Damaged Fuel Mixture Composition Determinations

NAC International 5.3.25-13 NAC PROPRIETARY INFORMATION REMOVED

NAC-LWT Cask SAR August 2023 Revision 23C Table 5.3.25-8 Fuel Region Homogenized Material Description

NAC International 5.3.25-14 NAC PROPRIETARY INFORMATION REMOVED NAC-LWT Cask SAR August 2023 Revision 23C Table 5.3.25-9 Maximum Radial Dose Rates for Damaged PWR and BWR Fuel Rods

Table 5.3.25-10 Maximum Axial Dose Rates for Damaged PWR and BWR Fuel Rods

NAC International 5.3.25-15

August 2023

Revision 23C NAC-LWT Legal Weight Truck Cask System ANALSAFETYYSIS REPORT

Volume 3 of 3 NON-PROPRIETARY VERSION

Docket No. 71-9225

Atlanta Corporate Headquarters: 3930 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

NAC-LWT Cask SAR August 2023 Revision 23C

LIST OF EFFECTIVE PAGES

Chapter 1 2.4.5-1......................................... Revision 46 1-i thru 1-vi................................ Revision 46 2.4.6-1......................................... Revision 46 1-1............................................ Revision 23C 2.5.1-1 thru 2.5.1-11................... Revision 46 1-2 thru 1-10.............................. Revision 46 2.5.2-1 thru 2.5.2-17................... Revision 46 1.1-1......................................... Revision 23C 2.6.1-1 thru 2.6.1-7..................... Revision 46 1.1-2 thru 1.1-4.......................... Revision 46 2.6.2-1 thru 2.6.2-7..................... Revision 46 1.2-1 thru 1.2-13........................ Revision 46 2.6.3-1......................................... Revision 46 1.2-14 thru 1.2-22.................... Revision 23C 2.6.4-1......................................... Revision 46 1.2-23 thru 1.2-47...................... Revision 46 2.6.5-1 thru 2.6.5-2..................... Revision 46 1.2-48....................................... Revision 23C 2.6.6-1......................................... Revision 46 1.2-49 thru 1.2-64...................... Revision 46 2.6.7-1 thru 2.6.7-137................. Revision 46 2.6.8-1......................................... Revision 46 1.3-1........................................... Revision 46 2.6.9-1......................................... Revision 46 1.4-1........................................... Revision 46 2.6.10-1 thru 2.6.10-15............... Revision 46 1.5-1........................................... Revision 46 2.6.11-1 thru 2.6.11-12............... Revision 46 2.6.12-1 thru 2.6.12-140............. Revision 46 88 drawings in the 2.7-1............................................ Revision 46 Chapter 1 List of Drawings 2.7.1-1 thru 2.7.1-117................. Revision 46 2.7.2-1 thru 2.7.2-23................... Revision 46 Chapter 1 Appendices 1-A 2.7.3-1 thru 2.7.3-5..................... Revision 46 through 1-G 2.7.4-1......................................... Revision 46 2.7.5-1 thru 2.7.5-5..................... Revision 46 Chapter 2 2.7.6-1 thru 2.7.6-4..................... Revision 46 2-i thru 2-xxv............................. Revision 46 2.7.7-1 thru 2.7.7-104................. Revision 46 2-1.............................................. Revision 46 2.8-1............................................ Revision 46 2.1.1-1 thru 2.1.1-2.................... Revision 46 2.9-1 thru 2.9-24......................... Revision 46 2.1.2-1 thru 2.1.2-3.................... Revision 46 2.10.1-1 thru 2.10.1-3................. Revision 46 2.1.3-1 thru 2.1.3-8.................... Revision 46 2.10.2-1 thru 2.10.2-49............... Revision 46 2.2.1-1 thru 2.2.1-5.................... Revision 46 2.10.3-1 thru 2.10.3-18............... Revision 46 2.3-1........................................... Revision 46 2.10.4-1 thru 2.10.4-11............... Revision 46 2.3.1-1 thru 2.3.1-13.................. Revision 46 2.10.5-1....................................... Revision 46 2.4-1........................................... Revision 46 2.10.6-1 thru 2.10.6-19............... Revision 46 2.4.1-1........................................ Revision 46 2.10.7-1 thru 2.10.7-66............... Revision 46 2.4.2-1........................................ Revision 46 2.10.8-1 thru 2.10.8-67............... Revision 46 2.4.3-1........................................ Revision 46 2.10.9-1 thru 2.10.9-9................. Revision 46 2.4.4-1........................................ Revision 46

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2.10.10-1 thru 2.10.10-97.......... Revision 46 5.1.1-8...................................... Revision 23C 2.10.11-1 thru 2.10.11-10.......... Revision 46 5.1.1-9 thru 5.1.1-15................... Revision 46 2.10.12-1 thru 2.10.12-31.......... Revision 46 5.1.1-16.................................... Revision 23C 2.10.13-1 thru 2.10.13-17.......... Revision 46 5.1.1-17 thru 5.1.1-22................. Revision 46 2.10.14-1 thru 2.10.14-38.......... Revision 46 5.2.1-1 thru 5.2.1-7..................... Revision 46 2.10.15-1 thru 2.10.15-10.......... Revision 46 5.3.1-1 thru 5.3.1-2..................... Revision 46 2.10.16-1 thru 2.10.16-5............ Revision 46 5.3.2-1......................................... Revision 46 5.3.3-1 thru 5.3.3-8..................... Revision 46 Chapter 3 5.3.4-1 thru 5.3.4-27................... Revision 46 3-i thru 3-v................................. Revision 46 5.3.5-1 thru 5.3.5-4..................... Revision 46 3.1-1 thru 3.1-3.......................... Revision 46 5.3.6-1 thru 5.3.6-22................... Revision 46 3.2-1 thru 3.2-11........................ Revision 46 5.3.7-1 thru 5.3.7-19................... Revision 46 3.3-1........................................... Revision 46 5.3.8-1 thru 5.3.8-25................... Revision 46 3.4-1 thru 3.4-16........................ Revision 46 5.3.9-1 thru 5.3.9-26................... Revision 46 3.4-17 thru 3.4-110.................. Revision 23C 5.3.10-1 thru 5.3.10-14............... Revision 46 3.5-1 thru 3.5-43........................ Revision 46 5.3.11-1 thru 5.3.11-47............... Revision 46 3.6-1 thru 3.6-12........................ Revision 46 5.3.12-1 thru 5.3.12-26............... Revision 46 5.3.13-1 thru 5.3.13-18............... Revision 46 Chapter 4 5.3.14-1 thru 5.3.14-22............... Revision 46 4-i thru 4-iii................................ Revision 46 5.3.15-1 thru 5.3.15-9................. Revision 46 4.1-1 thru 4.1-4.......................... Revision 46 5.3.16-1 thru 5.3.16-5................. Revision 46 4.2-1 thru 4.2-4.......................... Revision 46 5.3.17-1 thru 5.3.17-43............... Revision 46 4.3-1 thru 4.3-4.......................... Revision 46 5.3.18-1 thru 5.3.18-2................. Revision 46 4.4-1........................................... Revision 46 5.3.19-1 thru 5.3.19-9................. Revision 46 4.5-1 thru 4.5-43........................ Revision 46 5.3.20-1 thru 5.3.20-29............... Revision 46 5.3.21-1 thru 5.3.21-45............... Revision 46 Chapter 5 5.3.22-1 thru 5.3.22-34............... Revision 46 5-i............................................. Revision 23C 5.3.23-1 thru 5.3.23-49............... Revision 46 5-ii thru 5-vi............................... Revision 46 5.3.24-1 thru 5.3.24-7................. Revision 46 5-vii.......................................... Revision 23C 5.3.25-1 thru 5.3.24-15............ Revision 23C 5-viii thru 5-xiv.......................... Revision 46 5.4.1-1 thru 5.4.1-6..................... Revision 46 5-xv thru 5-xvi......................... Revision 23C 5-1.............................................. Revision 46 Chapter 6 5-2 thru 5-4.............................. Revision 23C 6-i thru 6-xix............................... Revision 46 5.1.1-1 thru 5.1.1-7.................... Revision 46 6-1 thru 6-2................................. Revision 46

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6.1-1 thru 6.1-6.......................... Revision 46 6.5.1-1 thru 6.5.1-13................... Revision 46 6.2-1........................................... Revision 46 6.5.2-1 thru 6.5.2-4..................... Revision 46 6.2.1-1 thru 6.2.1-3.................... Revision 46 6.5.3-1 thru 6.5.3-2..................... Revision 46 6.2.2-1 thru 6.2.2-3.................... Revision 46 6.5.4-1 thru 6.5.4-46................... Revision 46 6.2.3-1 thru 6.2.3-7.................... Revision 46 6.5.5-1 thru 6.5.5-15................... Revision 46 6.2.4-1...................................... Revision 23C 6.5.6-1 thru 6.5.6-20................... Revision 46 6.2.5-1 thru 6.2.5-5.................... Revision 46 6.5.7-1 thru 6.5.7-18................... Revision 46 6.2.6-1 thru 6.2.6-3.................... Revision 46 6.7.1-1 thru 6.7.1-19................... Revision 46 6.2.7-1 thru 6.2.7-2.................... Revision 46 6.7.2-1 thru 6.7.2-16................... Revision 46 6.2.8-1 thru 6.2.8-3.................... Revision 46 6.7.3-1 thru 6.7.3-39................... Revision 46 6.2.9-1 thru 6.2.9-4.................... Revision 46 6.7.4-1 thru 6.7.4-28................... Revision 46 6.2.10-1 thru 6.2.10-3................ Revision 46 6.7.5-1 thru 6.7.5-16................... Revision 46 6.2.11-1 thru 6.2.11-3................ Revision 46 6.7.6-1 thru 6.7.6-22................... Revision 46 6.2.12-1 thru 6.2.12-4................ Revision 46 6.3.1-1 thru 6.3.1-6.................... Revision 46 Appendix 6.6 6.3.2-1 thru 6.3.2-4.................... Revision 46 6.6-i thru 6.6-iii.......................... Revision 46 6.3.3-1 thru 6.3.3-9.................... Revision 46 6.6-1............................................ Revision 46 6.3.4-1 thru 6.3.4-10.................. Revision 46 6.6.1-1 thru 6.6.1-111................. Revision 46 6.3.5-1 thru 6.3.5-12.................. Revision 46 6.6.2-1 thru 6.6.2-56................... Revision 46 6.3.6-1 thru 6.3.6-9.................... Revision 46 6.6.3-1 thru 6.6.3-73................... Revision 46 6.3.7-1 thru 6.3.7-4.................... Revision 46 6.6.4.-1 thru 6.6.4-77.................. Revision 46 6.3.8-1 thru 6.3.8-7.................... Revision 46 6.6.5-1 thru 6.6.5-101................. Revision 46 6.3.9-1 thru 6.3.9-7.................... Revision 46 6.6.6-1 thru 6.6.6-158................. Revision 46 6.3.10-1 thru 6.3.10-2................ Revision 46 6.6.7-1 thru 6.6.7-84................... Revision 46 6.4.1-1 thru 6.4.1-10.................. Revision 46 6.6.8-1 thru 6.6.8-183................. Revision 46 6.4.2-1 thru 6.4.2-10.................. Revision 46 6.6.9-1 thru 6.6.9-53................... Revision 46 6.4.3-1 thru 6.4.3-35.................. Revision 46 6.6.10-1 thru 6.6.10-38............... Revision 46 6.4.4-1 thru 6.4.4-24.................. Revision 46 6.6.11-1 thru 6.6.11-53............... Revision 46 6.4.5-1 thru 6.4.5-51.................. Revision 46 6.6.12-1 thru 6.6.12-20............... Revision 46 6.4.6-1 thru 6.4.6-22.................. Revision 46 6.6.13-1 thru 6.6.13-22............... Revision 46 6.4.7-1 thru 6.4.7-13.................. Revision 46 6.6.14-1 thru 6.6.14-7................. Revision 46 6.4.8-1 thru 6.4.8-14.................. Revision 46 6.6.15-1 thru 6.6.15-................... Revision 46 6.4.9-1 thru 6.4.9-9.................... Revision 46 6.6.16-1 thru 6.6.16-30............... Revision 46 6.4.10-1 thru 6.4.10-18.............. Revision 46 6.6.17-1 thru 6.6.17-7................. Revision 46 6.4.11-1 thru 6.4.11-7................ Revision 46 6.6.18-1 thru 6.6.18-34............... Revision 46

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6.6.19-1 thru 6.6.19-3................ Revision 46

Chapter 7 7-i thru 7-iii................................ Revision 46 7.1-1 thru 7.1-86........................ Revision 46 7.2-1 thru 7.2-17........................ Revision 46

Chapter 8 8-i............................................... Revision 46 8.1-1 thru 8.1-15........................ Revision 46 8.2-1 thru 8.2-6.......................... Revision 46 8.3-1 thru 8.3-4.......................... Revision 46

Chapter 9 9-i............................................... Revision 46 9-1 thru 9-11.............................. Revision 46

Page 4 of 4 NAC-LWT Cask August 2023 SAR Revision 23C

6.2.4 PWR and BWR Rods in a Rod Holder or Fuel Assembly Lattice The NAC-LWT cask may transport up to 25 intact PWR or BWR fuel rods in a fuel rod holder or fuel assembly lattice. Up to 14 of 25 PWR or BWR fuel rods in a fuel rod holder may be classified as damaged.

6.2.4.1 Intact PWR or BWR Rods in a Rod Holder or Fuel Assembly Lattice

To bound all PWR and BWR rods that may be transported in the NAC-LWT cask, rods with a maximum enrichment of 5.0 wt % 235U were analyzed. Characteristics of the design basis PWR rods are presented in Table 6.2.1-1 and Table 6.2.1-2. Characteristics of the design basis BWR rods are presented in Table 6.2.2-1, Table 6.2.2-2 and Table 6.2.2-3. Given an infinite length rod and an enrichment of 5.0 wt % 235U as the basis for this analysis, the most reactive PWR and BWR rod has the greatest fissile mass, i.e. the rod with the largest pellet radius. Therefore, the rod used in the CE 14x14 assembly was chosen as the most reactive PWR fuel rod and the rod used in the Exxon 7x7 assembly was chosen as the most reactive BWR fuel rod. A maximum of 25 PWR or BWR rods were used in the analysis.

6.2.4.2 Damaged PWR or BWR Rods in a Rod Holder

The evaluation of the damaged fuel rods uses the bounding fuel characteristics for the intact fuel rod condition as described in Section 6.2.4.1, but assumes that up to 14 of the fuel rods are classified as damaged. Fuel transported in this configuration must be in a fuel rod holder. The fuel rod used in the CE 14x14 assembly was chosen as the most reactive PWR fuel rod, and the rod used in the Exxon 7x7 assembly was chosen as the most reactive BWR fuel rod.

While the bounding fuel characteristics described in Section 6.2.4.1 are based on zirconium alloy clad fuel the damaged fuel criticality analysis using the rod holder conservatively removes the cladding from the analysis models, see Section 6.4.4.2. As such the evaluation bounds the presence of non-zirconium based clad fuel rods such as FeCrAl alloys.

NAC International 6.2.4-1

ML23244A178

Text

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