ML20114E073

Tn Americas LLC - Application for Revision of Certificate of Compliance No. 9358 for the Model No. TN-LC

ML20114E073
Person / Time
Site:	07109358
Issue date:	04/23/2020
From:	Shaw D Orano TN Americas, TN Americas LLC
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML20114E072	List: ML20114E073
References
E-56457
Download: ML20114E073 (201)

Text

April 23, 2020 E-56457 U. S. Nuclear Regulatory Commission Columbia Office 7135 Minstrel Way Attn: Document Control Desk Columbia, MD 21045 Tel: (410) 910-6900 One White Flint North

@Orano_USA 11555 Rockville Pike Rockville, MD 20852

Subject:

Application for Revision of Certificate of Compliance No. 9358 for the Model No. TN-LC, Docket No. 71-9358

References:

(1) NRC Certificate of Compliance for the Model No. TN-LC, USA/9358/B(U)F-96, Rev 4 (2) Packaging Safety Analysis Report for the Model TN-LC Package, Revision 6 In accordance with 10 CFR 71.38, TN Americas LLC (TN) herewith submits an application to revise Certificate of Compliance (CoC) No. 9358 for the TN-LC packaging. The current CoC, Revision 4 [1], references the TN Americas consolidated application dated November 2012 [2], supplemented as follows:

Transnuclear, Inc. supplement dated: November 27 and December 18, 2012, (CoC R0, SAR R6)

Transnuclear, Inc supplements dated: January 27, 2014, (CoC R1 Certificate holder name change)

AREVA Inc. supplement dated: October 11, 2013, December 13, 2013, March 24, 2014, (CoC R2 SAR R7 and R8)

TN Americas LLC supplement dated: November 18, 2016, (CoC R3, Certificate holder name change)

TN Americas LLC supplement dated: November 8, 2017 (CoC R4 Renewal)

Two of these supplements provided changed pages for the Safety Analysis Report (SAR), and the remainder were administrative requests involving no changes to the SAR. A consolidated SAR has not been submitted since the original issue of the CoC. This request to revise the CoC includes the changed pages for review, and a consolidated-SAR will be submitted upon approval of changed pages. A performance based, risk-informed technical review of the SAR changes is needed to ship the allowed contents by the end of 2020.

Enclosures transmitted herein contain SUNSI. When separated from enclosures, this transmittal document is decontrolled.

Document Control Desk E-56457 April 23, 2020 Page 2 of 2 TN has a quality assurance program, approved by the NRC, which satisfies the provisions of Subpart H-Quality Assurance of 10 CFR Part 71.

Should the NRC staff require additional information to support review of this application, please contact Peter Vescovi at 336-420-8325, or by email at peter.vescovi@orano.group.

Sincerely, Don Shaw Digitally signed by Don Shaw Date: 2020.04.23 08:28:02

-04'00' Don Shaw Licensing Manager TN Americas LLC Electronic Information Exchange (EIE) Document Components:

001 NRC TN-LC Revision Application Transmittal Letter.pdf 002 Enclosure 1 - Summary of Proposed Changes 003 Enclosure 2 - Changed Pages for SAR Revision 9a (Proprietary and SUNSI) 004 Enclosure 3 - Changed Pages for SAR Revision 9a (Public) 005 Enclosure 4 - Affidavit Pursuant to 10 CFR 2.390 cc: Pierre Saverot, Senior Project Manager, U.S. Nuclear Regulatory Commission Peter Vescovi, Licensing Engineer, TN Americas LLC Don Shaw, Licensing Manager, TN Americas LLC

Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

This enclosure provides a summary of the proposed changes to assist in reviewing the affected sections of the safety analysis report (SAR). The changes affect the SAR as described in the following summary and more detailed description and justification of the changes that follows.

Contents Specification WE 16x16 was added to the list of fuel classes to be transported in the TN-LC-1FA Basket. Fuel designations were added for WE 17x17, WE 14x14, CE 16x16, WE 16x16 fuel classes.

Appendix 1.4.5, Table 1.4.5-1, Table 1.4.5-2, Table 1.4.5-4, and Figure 1.4.5-6, Appendix 6.10.4, Section 6.10.4.2, Section 6.10.4.4.1, Table 6.10.4-2, Table 6.10.4-26, were revised or added for this change.

Engineering Drawings The revisions to the SAR drawings include changes as recommended in NRC SFST ISG-20 (i.e., material specifications, dimensions, details, optional parts) that provide flexibility to make design changes without requiring prior NRC approval. Improvements are made to the design features based on experience with similar cask fabrication, operation, and maintenance. In addition, some non-conforming conditions that were dispositioned during fabrication require adding options to the SAR drawings to allow use as-fabricated.

Shielding Evaluation The TN-LC Unit 01 as-fabricated cask body has a reduction in its shielding capability due to localized areas where the lead thickness is less than the acceptance criterion, and may be as low as 3.10 inches. As a result of the reduced lead thickness condition for TN-LC Unit 01, Section 5.6.4.5 was added to SAR Appendix 5.6.4 for assessing the impacts of the non-conforming thickness conditions with respect to the limiting dose rates for TN-LC-1FA PWR fuel assembly and PWR fuel rods contents loaded in TN-LC transport cask. Chapter 5, Section 5.4.4, Appendix 5.6.4, Section 5.6.4.4.5 were revised or added for this change.

Containment Evaluation The NUHOMS-MP197HB transport cask is designed and tested for a leak rate of 1x10-7 ref cm3/s, defined as leak-tight per ANSI N14.5. Some contents do require the cask to be leak-tight for containment. A reference air leak rate that allows testing to less than leak-tight is established that satisfies containment requirements. A new Appendix 4.6.1, Containment Reference Leak Rate for 1FA, is added to determine an allowable leak rate for the 1FA contents.

Operating Instructions, and Acceptance Criteria and Maintenance The allowable leak rate provides flexibility for performing leak tests using methods less sensitive than the helium leak test that is required to demonstrate leak-tightness required for other contents.

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Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Affected sections of the SAR are revised to allow less sensitive leak rate measurement methods and take into account NRC IN-2016-04, ANSI N14.5-2014 Revision and Leakage Rate Testing Considerations.

Editorial Corrections Changes as noted in table of detailed description of changes.

Each item in the table below corresponds to a description of change that affects one or more sections in the SAR. Each Description of Change is itemized and may be associated with one or more affected sections in the SAR. Each affected section associated with an item may have a separate justification, or a single justification is provided at the end of the item that applies to all the affected sections. Affected sections that require changes may include an excerpt from the SAR, in addition to the change pages that are provided in Enclosure 2. A SAR page number is provided to assist in finding the affected section in the SAR or within the change pages provided after the table.

Design changes made to package design are subject to design control commensurate with original design as previously approved. An LR Number (e.g. LR719358-XXXX) is a reference to the review process implemented by the QA program that reviews design changes for impact on the SAR.

Chapter/Appendix/

Item Description and Justification Section Chapter 1 General Information 1.1 1.2.1.1 TN-LC Cask Body Description of Change Page 1-1 a) Add option to transport on flat rack, deleted description to transport single package, changed 21 pin can to pin can, added boat mode for transport, Page 1-2 changed carrier to carriers Page 1-3 b) Added word fuel before cladding, and added word that before the fuel Page 1-5, cladding Page 1-5a, c) Added word the before criticality analysis Page 1-6 d) Changes to inner cavity length, and shielding lead thickness due to nonconforming conditions during fabrication for TN-LC Unit 01.

e) Modified o-ring seals cross-section diameter.

f) Removed description of cask orientation for transportation.

SAR a) See changed page(s) b) See changed pages c) See changed pages d) See changed pages e) See changed pages 2 of 31

Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section f) Trunnions, attached to the cask body, are provided for lifting and handling operations, including rotation of the packaging between the horizontal and vertical orientations. The TN-LC transport package is transported in the horizontal orientation on a specially designed shipping frame. with the lid end facing the direction of travel.

Justification a) Editorial for clarification and add options for transport.

b) Editorial c) Editorial d) Evaluations done to accept the non-conforming conditions.

e) Editorial. Change to correct value.

f) Structural evaluation makes no assumption for cask orientation during routine and normal conditions for transport.

1.2 1.2.1.3 Impact Limiters Description of change:

Page 1-7 a) Added B24 material class for impact limiter bolts, and b) Specified number of fusible plugs.

SAR Each impact limiter is attached to the TN-LC transport cask by eight 1-8 UNC bolts made from SA-540 grade B23 or B24 class 1 material. The attachment bolts are designed to keep the impact limiters attached to the cask body during all NCT and HAC.

Each impact limiter is provided with four five fusible plugs that are designed to melt during a fire accident, thereby relieving excessive internal pressure. Each impact limiter has four hoist rings for handling and two support angles for supporting the impact limiter in a vertical position during storage. The hoist rings are threaded into the reinforcement blocks which are welded to the impact limiter gusset plates, while the support angles are welded to the outer shell. Prior to transport, the impact limiter hoist rings are removed and replaced with bolts.

Justification a) See Item 2.3 b) Conforms to package design specification and as-built.

1.3 1.2.2 Contents Description of Change Page 1-8 Refer to pin can instead of 25 pin can SAR See changed pages.

Justification Original design of pin can permitted loading of 25 individual fuel pins. A design change during fabrication modified the four corner fuel pin tubes to accommodate lid fastening bolts.

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Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section 1.4 Table 1-2 Basket Description of change:

Configurations in the TN-LC Packaging a) Changed Subbasket Type names to be consistent with other sections of SAR and drawings.

Page 1-12 b) Add max decay heat for pin can SAR See changed page Justification a) Editorial b) Adjusted decay heat for reduced number of fuel rods in pin can.

1.5 Appendix 1.4.1 TN-LC 1.5.1 Drawing 65200-71-01 Revision 7 and 8 TN-LC Cask Assembly (11 sheets)

Transport Package 1.5.2 Drawing 65200-71-20 Revision 5 TN-LC Impact Limiter Assembly Drawings Drawing (2 sheets) 1.5.3 Drawing 65200-71-90 Revision 4 and 5 TN-LC-1FA Basket (5 sheets) 1.5.4 Drawing 65200-71-96 Revision 5 TN-LC-1FA BWR Sleeve and Hold-Down Ring (2 sheets) 1.5.5 Drawing 65200-71-102 Revision 5 and 6 TN-LC-1FA 25 Pin Can Basket (4 sheets)

Updated page 1.4.1-1 to show most current revision.

Note: Drawings 65200-71-01, 65200-71-20, 65200-71-90 and 65200-71-102 were revised twice since last approved. Both revisions are provided.

1.5.1 Drawing 65200-71-01 LR719358-0008 Item(s) a)

Revision 7 and 8 TN-LC Cask Assembly (11 LR719358-0014 Item(s) b) through i) sheets) LR719358-0016 Item(s) j)

LR719358-0019 Item(s) k)

LR719358-0027 Item(s) l)

LR719358-0028 Item(s) m)

LR719358-0030 Item(s) n)

LR719358-0031 Item(s) o)

Description of change:

a) Items 3B, 7C, and 10A, code criteria changed to NB W/ EXCEPTIONS, NF W/ EXCEPTIONS, AND NB W/ EXCEPTIONS, respectively Add note 47 as follows, ASME CODE MATERIAL SPECIFICATION NOT REQUIRED FOR HELI- COIL THREAD INSERTS (Sheet 1 of 11) b) Revise Item material description for Item 3J and 3N to ASTM B36 or ASTM A193 B8. See Figure 1 on Page 2

.(Sheet 1 of 11) c) Revise Item description and pictorial representation for Items 3J and 3N to SHCS, Port Plug. See Figures 1 on Pages 2.

(Sheet 1 of 11) d) Revise Note 45 to allow an optional Vent/Drain/Test port plug configuration.

(Sheet 1 of 11) 4 of 31

Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section e) Create a new Item 3P with a quantity of 2 to replace two of Item 3J. Make 3P a NITS item. Reduce quantity of 3J to 1 (Sheet 1 of 11) f) Revise part of Note 16 to SST Vent/Drain/Test Port Plugs and add Brass Vent/Drain/Test Port Plugs - 130-168 in-lb Torque requirement per revised calculation TN-LC-0203.

(Sheet 2 of 11) g) Revise the edge of the two machined pocket trunnions on the bottom of the bottom flange to incorporate transport frame pivot pins lead-in chamfers.

(Sheet 8 of 11) h) Revise the 90° and 270° orientation (Sheet 1 of 11) i) Revise the torque values specified for the vent/drain/test port plugs. torque specified for the SST vent/drain/test port plug and optional brass vent/drain/test port plugs to 130-140 in-lb. (Sheet 1 of 11) j) Recess Item 24 (check valve) within neutron shielding using new Item (25)

(Sheet 1 of 11) k) Include ASME SA-193 B8 as an acceptable material for Items 3J (SHCS, PORT PLUG), 3N (SHCS, PORT PLUG, W/ EXTENSION), and 3P (SHCS, PORT PLUG) of SAR Drawing 65200-71-01 (TN-LC CASK ASSEMBLY).

(Sheet 1 of 11) l) As-fabricated TN-LC Ser. No. 01 shielding non-conformance resulted in a minimum lead thickness as verified by gamma scan that is less than the specified 3.38 minimum lead thickness.

(Sheet 1 of 11) m) As-fabricated TN-LC Ser. No. 01 non-conformance resulted in a minimum cask cavity length less than 182.50 inches. The as-fabricated measured cavity length is 182.12 inches.

(Sheet 2 of 11)

Add Note 46 - AS-FABRICATED CAVITY LENGTH OF 182.12 INCHES IS ALLOWED FOR TN-LC NO. 01 FOR USE ONLY WITH 1FA BASKET AND TRIGA BASKET WITH HEIGHT OF TOP SPACER REDUCED BY 0.50 INCHES.

(Sheet 1 of 11)

Call out note 46 at the minimum cavity length dimension on the drawing (Sheet 4 of 11) n) Add a groove between the TN-LC lid and bottom plugs O-ring grooves, (Sheet 7 and 11 of 11) o) Remove the bottom lead plug and its attachment hardware (Items 5, 6, 7A, 7B, 12, 13 and 14) from Drawing 65200-71-102 and Item 8J from Drawing 65200-71-01.

SAR See change page(s) 5 of 31

Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section Justification a) The material specified in the BOM is non-code material; however, the code criteria are specified as NB, NF, and NB for Items 3B, 7C, and 10A, respectively. This change adds clarification that although the design criteria are code-based, exception is taken to the material used. These Heli-Coil thread inserts are tested to confirm identified critical characteristics and dedicated to comply with the ASME code design and NAQ-1 program safety basis.

b) The addition of ASTM B36 is for clarification purposes only. The current revision of the drawing allows brass with no clarity on the applicable material specification. ASTM B36 has been reviewed and found to be an acceptable material for this application.

c) The primary reason for the vent/drain/test port plugs is for flooding, draining and leak testing of the confinement boundary and seals prior to transportation of the packaging. The only design function credited to the vent/drain port plug is the confinement design function with a leak-tight criterion of 1 x E-7 std atm cc/s.

The only purpose for the test port plug is to facilitate helium leak testing of the top Lid and Bottom Plug seals. Although, the test port plug is not considered part of the confinement boundary, the same material requirements and changes implemented for the vent/drain port plug (Item 3J) are also being implemented for the test port plug (Item 3N) to maintain consistency throughout. Refer to Item 1.5.1 e) for discussion on the quality classification change for the test port plugs.

The reason for the change to the vent/drain/test port plug fastener from a hex head to a socket head is for compatibility with the Helium Leak test tool that will be used during operations for cask cavity evacuation, helium filling and leak testing. The helium leak test tool will utilize a hex key (in lieu of a hex head socket) to perform various stages of tightening and loosening of the vent/drain/test port plugs during leak testing. The thread size, length engagement, and required torque during tightening are unaffected by this change d) The TN-LC Cask along with the port plug/washer/o-ring combination seal with a 1 x E-7 std atm cc/s Helium leak acceptance criterion was a first of a kind. During fabrication, testing of the original design using the Port Plug (Item 3J), washer, and Parker O-ring 2-015 configuration did not work during helium leak testing. It was therefore, concluded that the acceptance criterion as specified in the SAR was unachievable when the current washer/o- ring configuration was used. Consequently, Note 45 is revised to implement the new port plug design that was tested and proven successful.

e) The only purpose for the test port plug (Item 3J) is to facilitate helium leak testing of the top Lid and Bottom Plug inner O-ring seals. The test port plug is not considered part of the confinement boundary, as the confinement design function is performed by the inner o-ring (Items 3E and 8E) seals the test port plug is intended to test. Additionally, making the new Item 3P a NITS item, removes unnecessary restrictions and controls on the material simplifying the procurement process, should there ever be a need to procure material for a replacement plug during operations or maintenance.

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Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section f) Provides clarification on the applicable torque requirements when a Brass vent/drain/test port plug is used and when a SST vent/drain/test port plug is used. A 60-70 FT-LB torque requirement has been demonstrated to work well for a SST port plug but lower torque values specified on the drawing are sufficient. The vent/drain port plugs are part of the confinement boundary as they allow a direct path to the cask cavity when the plugs are removed. The calculated torque requirements shall be such that the bolt maintains its confinement integrity and leak-tightness acceptance criterion while installed and does not become loose during transportation of the cask. No other design function is credited to the vent/drain port plugs.

g) In most cases, the addition of lead chamfers is good practice as it provides the field operator with improved lead-in and alignment capabilities during operations. The worst case clearance between the pocket trunnion and the TN-LC Cask Transport Frame Pivot Pin is .020 inch (3.90 - 3.88). The lead-in chamfer will provide for better alignment control while lowering of the cask onto the Pivot Pins. The periphery of the bottom flange where the relatively small amount of material will be removed is not stressed during down-ending of the cask. The effective wall thickness is also not reduced; therefore, there is no adverse impact to the structural, thermal, shielding, or confinement design function of the TN-LC.

h) This change is a correction of an orientation oversight. The orientation shown at the bottom flange at the component level (Sheet 8, Section H-H) does not agree with the orientation of the bottom flange shown at the assembly level (Sheet 5, View D-D).

i) This proposed activity is essentially identical to Item 1.5.1 f) discussed above; except that clarification is provided in the Operations Sections 7.1.2.1 and 7.4.1 of the SAR in addition to the transportation drawings.

j) Avoid interference of check valve with impact limiter.

k) The proposed activity will add ASTM A193 B8 as an acceptable material for Items 3J 3N, and 3P on sheet 1 of design Drawing 65200-71-01. Although ASME SA-193 and ASTM A193 are identical specifications, ASTM A193 B8 material is not specifically identified as an acceptable material for Items 3J, 3N, and 3P on sheet 1 of SAR Drawing 65200-71-01.

l) Engineering drawings for the approval of the package design as provided in SAR Chapter 1.0 - General Information, specify a 3.50-inch nominal thickness for the lead component of the radial shielding, and a 3.38-inch minimum thickness for the lead. No change is made to design specification for the 3.50-inch nominal radial thickness, or 3.38-inch minimal thickness. A minimum thickness of 3.10-inch is specified for TN-LC No. 01 based on the evaluation of as-built gamma scan results. A minimum lead thickness of at least 3.12 inches was verified by gamma scan of a reworked calibration block.

m) Design specification for TN-LC minimum cavity length remains unchanged.

The non-conformance is dispositioned for TN-LC No. 01 by reconciling the shorter cavity length with the technical evaluations, operations, and maintenance and acceptance requirements.

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Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section Structural The impact of an under length cavity is interference with the basket (or basket stack) in the hot condition (insufficient hot gap). Each of the basket designs were evaluated. The NRUX, MTR, and TRIGA basket heights resulted in a hot gap interference with the inner cavity. A modification to the TRIGA top spacer eliminate the interference, and 1FA baskets did not have a hot gap interference. Neither the NRUX or MTR baskets was fabricated. A note is added to DWG 65200-71-01 to allow an exception to the minimum cavity length for TN-LC No. 01 and use of only the 1FA basket and TRIGA basket with modified top spacer. Appendix 2.13.10 was updated accordingly.

Thermal / Containment The heat flux is simulated in the thermal model over the active fuel length that is less than the cavity length. No axial gaps are assume between the cavity and lid or bottom of the inner shell. A reduction in inner cavity length of 0.38 inches has no effect on the thermal evaluation temperatures.

A reduction in length of 0.38 inches (182.50 inches minimum - 182.12 inches as-fabricated) results in approximately 0.2% reduction in the free volume. The MNOP and maximum pressure calculated using the reduced free volume would increase by approximately 0.2%. MNOP pressure is calculated as 16.9 pisg, but a pressure of 30 psig is used for structural evaluation. The result is an increase in the maximum pressure from 90.9 psig to 91.1 psig. The estimated increase in pressure is offset by the reduce quantity of helium fill gas that contributes about 30% total amount of gas in the TN-LC cask cavity.

Shielding A reduction in the cavity length below the specified minimum has no effect on radial or axial shielding. The TRIGA top spacer is not modeled; therefore, any change to the height of the spacer has no effect on the shielding evaluation.

Criticality Criticality control is provided primarily by the MTR, NRUX, TRIGA, and 1FA basket materials and dimensions. The cask body is modeled with nominal dimensions because the tolerances on the cask body dimensions have little effect on the system reactivity. A 0.38-inch reduction in cavity length below the minimum specified 182.50 inches has no significant effect on the criticality evaluations.

Operations Package Operations do not specify the cavity length or basket spacer lengths.

Acceptance Tests and Maintenance Acceptance Tests and Maintenance do not specify the cavity length or basket spacer lengths.

(LR 719358-0028) n) Minimizes the conductance time for a leak to reach the test port and to ensure that the time is repeatable. Without these grooves, a leak will need migrate up to 180° around the circumference of the sealing surface through a metal to metal interface (hence higher conductance time).

(LR 719358-0030) 8 of 31

Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section o) In the current design, the plug (at the bottom of the pin can) can move in one direction and it may not engage with the hole at the bottom of the cask.

Locating it is difficult because the plug diameter is very close to the diameter of the cask bottom opening. Also, the two #10 screws are not strong enough to withstand the pulling loads with the pin can, which can weigh approximately 1,000 lbs fully loaded. Overall it makes the design and operations simpler if there is no plug at the bottom of the pin can, and if the bottom cover with the lead plug is used on the cask to maintain the same amount of shielding. Therefore, this pin can bottom plug is removed from the design as well as the plug-less cover from the cask drawing (Item 8J).

(LR 719358-0031) 1.5.2 Drawing 65200-71-20 LR719358-0004 Item(s) a)

Revision 4 and 5 TN-LC LR719358-0023 Item(s) b)

Impact Limiter Assembly LR719358-0029 Item(s) c) through gg)

(2 sheets)

Description of change:

a) Change material of Item 8B to ASTM A240 TYPE 304.

b) Change lower limit Balsa density tfrom 10 to 7 lb/ft3, moisture content range from 6-10% to 6-12%.

c) Update note 2.

d) Update note 3 and delete note 15.

e) Update note 5 On sheet 2 area F4 change the (Ø30.50 I.D.) inner diameter of the impact limiter to Ø30.38 I.D.

f) Update note 8 and change to note 6.

g) Note 11: clarify that the length is the length of the bolt taper on the second sentence. On detail 1 make sure note points to actual taper.

h) Correct note 13, Item 22 is also plug-welded to Item 2.

i) Correct note 14, change Item 8 to Item 22.

j) Note 16 (welding code): remove edition and add ASME code section IX.

k) In parts list, add metric equivalent to the ASTM material specifications where applicable (see markups) and specify metric thicknesses for all the impact limiter plates.

l) Clarify STAINLESS STEEL material for Items 14, 15, 19, 20, 21A, 24 and 25.

m) Add material specification to Items 7, 8 and 8A like on 65200-30-20 latest revision.

n) Fusible plug (Item 16) material: replace equivalent with nylon.

o) Update part # and material for Item 17.

p) Add Gr. B24 as an authorized material for Item 21.

q) Make Item 10 in one piece and shorten gussets (Items 11) accordingly.

r) Correct the length of impact limiter bolt.

s) Update Item 8B like on latest 65200-30-20.

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Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section t) Add a note to clarify that the bolts Items 24 will be swapped for the hoist rings Items 18 during transportation.

u) Sheet 2 views B-B and C-C: remove tolerance on bolt circle diameter reference dimensions.

v) In detail 1 (sheet 2), remove weld callout of Item 8A to 10. Add a not to clarify that the bolts (item 24) will be swapped for the hoist rings (item 18) for transportation.

w) Sheet 2 section A-A: update tolerances per the markup and update 5.63 dimension. Add missing dimension of Item 23 and wood orientation + callout of one Item 22.

x) Sheet 2 section A-A: add depth of impact limiter.

y) Sheet 2, Detail 2B: update to clarify callouts of bolts tunnels welds (see markups), change to just detail 2 (no detail 2A).

z) Sheet 3, bolt tunnel weld assembly: change weld of Item 8 to 9 to a groove weld. Add allowance for weld face to be ground or machined flush at locations where there will be contact with Item 2 (at a minimum).

aa) Specify intermittent 1/8 fillet welds between each Item 11 and Items 1, 2 and

10. Same thing between each Item 12 and Items 2, 3 and 13. Add/remove details as required.

bb) Add a new diamond note to specify that the wood in the middle section of the impact limiter is split into four pie-shaped regions, with each having a grain orientation in the radial direction.

cc) Insert in new note 8 that Items 8 and 8A may be procured with a smaller inner diameter, and their inner diameter machined by the fabricator after full impact limiter assembly).

dd) Add another fusible plug on Item 6 and add holes on Item 2.

ee) Add in note 7 (sheet 1) the allowance to make each wood block from multiple pieces glued together.

ff) In detail 1 (sheet 2), change weld location of Item 2 to 3 to the other side of plate Item 3 (inside the impact limiter recess).

gg) Sheet 3 detail bolt tunnel assembly: change the length of Item 7 to 15.65.

SAR See change page(s)

Justification a) The current specification for Item 8B is ASTM A511, which is a tube specification. However, this Item is fabricated from a plate an cannot be fabricated from a tube.

b) See change previously approved for similar package design (Reference NRC Safety Evaluation Report , Docket No. 71-9302, Certificate of Compliance No.

9302, Revision No. 9, April 17, 2019, ADAMS Accession No.:

ML19112A168))

c) Lessons learned from MP197HB impact limiter fabrication: a more specific description of the gaps allowed (and between which components) is needed.

d) Clarification of the drilling instructions for the tamper-indicating device. Note 15 was an operational note with no package safety implications.

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Enclosure 1 to E-56457 Summary of Proposed Changes (Safety Analysis Report, TN-LC, Revision 9a)

Chapter/Appendix/

Item Description and Justification Section e) The minimum radial gap is not important to the safety of the package, only the maximum radial gap is important and therefore only the maximum radial gap is mentioned in the note.

The previous 30.50 (reference) nominal diameter was wrong and did not match the required bolt circle diameter of 32.50 considering material thicknesses and dimensions.

The maximum possible radial gap based on fabrication tolerances is .2875, which is bounded by the .38 value; also, this corrects the impact limiter inner diameter nominal value, which was wrong.

f) Other items than Item 1 (2, 3, 10 and 13) could benefit from the allowance to be built in several pieces. The requirement to use full penetration welds was missing. Inversely, fabricating the bolt tunnels from one piece of bar material could make the design easier to fabricate.

Clarifies how the items should be made and allows other items to be made from more than one piece, or from a single piece in the case of the bolts tunnels.

g) Clarification.

h) Correction.

i) Correction.

j) Some shops are ASME-certified instead of AWS, so this change makes fabrication easier. Editions to be specified in fabrication specification.

k) Makes fabrication easier in the event the impact limiter is fabricated in a country that follows the metric standard.

l) SST is not always understood (for example if the impact limiters are fabricated in a non-English speaking country). Specifying 300 series is not necessary for a NITS item, stainless steel is enough. This is an editorial change.

m) These specifications have been used for the MP197HB impact limiters and they are easier to procure for tubes.

n) Editorial and generic nylon has been shown to be an acceptable alternate to Rilsan.

o) The currently specified item, Parker 600-01 02-1/2, is a 1/2 Stat-O-Seal with a zinc-plated carbon steel washer and a nitrile rubber seal. The nitrile rubber seal would perform adequately but is not consistent with the materials used for the seals in the rest of the cask design, and there is a chance of rusting with the zinc-plated steel washer portion. The new part number has a stainless steel washer and a fluorocarbon seal. The seal material is consistent with the other elastomer O-rings/seals used elsewhere in the cask design, and changing the washer material to stainless steel will eliminate the possibility of rusting (also see DCR 1001190, change #35).

p) Adds flexibility for fabrication. Specifying either Gr. B23 or B24 allows suppliers more flexibility in meeting material requirements including fracture toughness tests. (See Item 2.3) 11 of 31

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Item Description and Justification Section q) Current design with multiple plates is difficult to fabricate: lots of welding, lots of distortion, it will be hard for the fabricator to meet the required dimensions. This change aligns this design with the MP197HB impact limiters design.

Simplifies the design and makes it easier to fabricate, reduces amount of welding required, which means less distortion, which allows the fabricator to better control the dimensions. Updated calculation TNLC-0242 accordingly.

Length of gussets is below the level of detail (not shown) r) 17 seems too long: based on 3D model, length should be 16.25, TBD after model check.

Improvement, bolt may be too long to perform its function and some important dimensions might be missing.

s) Part can be machined very accurately and will vastly improve tubes alignment.

t) Not clear on the drawings where bolts (Item 24) are supposed to go.

u) No reason to tolerance a reference dimension.

v) Its already shown on detail 2B.

w) Tolerances are updated per lessons learned on the MP197HB to specify tolerances that the fabricator can achieve. Dimension is updated to reflect actual impact limiter geometry. Corrected missing information.

x) Dimension is missing.

y) Add clarity to the details, prevent the creation of unnecessary paperwork for later.

z) Prevent interference of this fillet weld with Item 2 (see markup and sketches).

aa) These welds are not defined on the drawing.

bb) This note was missing (see MP197HB impact limiter Drawing MP197HB 8, note 14) and is required to ensure the impact limiter is built as intended in the design.

cc) Incorporation of lessons learned from fabrication of the MP197HB impact limiters. During their fabrication, the welding of the impact limiter plates resulted in alignment problems in the bolts tunnels on the cask side and the necessity to machine the IDs to a larger value. This change should eliminate this potential problem.

dd) This is so that the intermediate wood compartments communicate with the outer wood compartments. The center and intermediate compartments have no over-pressure relief system in case of a fire.

ee) Clarification, and simplifies the fabrication of the impact limiter.

ff) Welding from this side is easier, since Item 9 is not in the way like it was on the other side.

gg) Actual length from the design Drawing 65200-30-20. This dimension is not important to safety and may be adjusted to fit the overall length of bolt tunnel needed, therefore it is given as a reference dimension.

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Item Description and Justification Section 1.5.3 Drawing 65200-71-90 LR719358-0022 Item(s) a)

Revision 3 and 4 TN-LC-1FA Basket (5 sheets) LR719358-0032 Item(s) b), c), and d)

Description of change:

a) Remove note 11 and the mention flat in the BOM for Item 6.

b) Add an optional 3 long, 1/4 max deep recess at the top of the basket cavity.

c) Add 4 optional slots roughly 1/2 wide, up to 2 tall (enough to remove the threaded holes at the top of the walls and allow the insertion of a tool to lift the failed fuel can) through the thickness of the walls, see sketch below and markups (final dimensions per final Drawing 65200-30-90).

d) Add optional small drain holes at the bottom of the threaded holes in the top of the rails and basket plates.

e) Remove STK from the basket plates (Items 1 and 2) thicknesses on sheet 2 view D-D and sheet 5.

f) 1) slot width update to .55-.70 and 2) add missing dimensions and radii:

As a result of this change, the slot width on the SAR drawing changes from .55 to .70 max and optional radii are explicitly allowed in the cutout/recess.

g) Apply note 10 (diamond) to the weld of Item 8 to Items 1 or 2 on sheet 3.

h) Change 8.875 basket opening (sheet 2) and 8.875 and 10.875 basket plates widths (sheet 5) to reference dimensions (8.875) and (10.875).

SAR See change page(s)

Justification a) There is no formal analysis for these nuts and their torque; these studs/nuts are NITS and only serve to hold the rails on the basket during fabrication.

There is no reason for the washer to be flat, Belleville washers could be used instead. Washers, studs and nuts have no function during transportation and they are not needed for the basket to perform its structural, thermal, shielding and criticality functions.

b) Some room is needed at the top of the basket to fit the increased thickness at the top of the failed fuel can (FFC) so that the FFC can be lifted without deformations. The introduction of the use of an FFC for a PWR FA shall be the subject of a further application.

No impact to the design structural-wise: the recess is not tall enough to reach the first welded studs, and the short depth and length of the recess means it does not have any impact on the structural behavior of the basket. A sensitivity analysis will be performed to determine if a 1/4 recess depth is too much for shielding, in which case the depth may be reduced.

c) Some room is needed at the top of the basket for a tool to lift an FFC if one is used. The slots height is increased to 2 because the lifting slots are typically 1/2 tall and located 1 from the top of the failed fuel can.

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Item Description and Justification Section 1.5.3 No impact to the design structural-wise: the slots are not tall enough to reach the first welded studs, and the short width and height of the slots means they do not have any impact on the structural behavior of the basket. The sensitivity analysis mentioned in 1 will take the slots into account to ensure they have no effect on shielding.

d) Any water trapped in the threaded holes will make drying of the cavity very difficult during operations.

e) The mention STK implies a negative tolerance of -.01 to the nominal value, which is unnecessarily restrictive for fabrication. The usual practice is to not provide fabrication tolerances on the licensing drawings, and therefore this unnecessary restrictive fabrication tolerance is removed.

f) 1) Extra clearance might be needed for the FFC lifting tool. Only the max is needed as the min slot width would have less of an effect, and

2) Fabrication processes might require radii in these areas.

g) Editorial change for clarification.

h) The basket opening (and therefore basket plates widths) is controlled by note

2. Therefore, these dimensions can be provided as reference dimensions (for information only) on the drawing.

1.5.4 Drawing 65200-71-96 LR719358-0033 Revision 5 TN-LC-1FA BWR Sleeve and Hold- Description of change:

Down Ring (2 sheets) Add a new note to state Optional drain holes may be drilled at the base of all lifting holes on both BWR sleeve and hold-down ring (Items 1 through 4)

SAR See change page(s)

Justification Without a drain hole, water pools inside each hole during operations in the pool and this makes the cask cavity very difficult to dry.

1.5.5 Drawing 65200-71-102 LR719358-0020 Item(s) d)

Revision 5 and 6 TN-LC-1FA 25 Pin Can Basket (4 LR719358-0031 Item(s) a), b), and c) sheets) Description of change:

a) Replace small 6-32 UNX hex head screws with larger bolts and add alignment pins to facilitate underwater alignment pins are placed at the inside corners of the pin can tube.

b) Update the various design lengths (see markups) to reflect the as-built dimensions of the first fabricated pin can and TN- LC cask (Drawing 65200-71- 102). Update 2.13.10 accordingly.

c) Downgrade the quality categories of various item (Drawing 65200-71-102).

d) Remove mention of 25 pin can, replace with simply pin can, including in the title (Drawing 65200-71-102).

SAR See change page(s)

Justification a) Revise Pin Can Basket lid design to improve remote underwater operation 14 of 31

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Item Description and Justification Section b) The as-built cask cavity is shorter than it was supposed to be by roughly 3/8.

Even though the as-built pin can is also slightly shorter than it was supposed to be, with the current dimensions, the design gap of 1/4 is now down to 0.03.

These changes are meant to restore this 1/4 design gap, reflect the as-built final pin can dimensions (especially cavity length and overall length) and ensuring in the future will fit into the cask.

c) A lot of items that have been classified quality category A should be B or even NITS. Downgrading these items is a future cost-saving opportunity.

d) The quantity of rods that can be transported inside the pin can was decreased from 25 to 21 to make way for str is needs to be reflected throughout the drawing.

1.6 Appendix 1.4.5 Description of change:

TN-LC-1FA Basket Updated Table of Contents, List of Tables, and List of Figures.

Page 1.4.5-I and ii SAR See changed page(s)

Justification Editorial 1.7 Pages 1.4.5- Description of Change 1,2,3,8,9,14,16,18, 19 Changes to describe modifications to pin can.

Table 1.4.5-5 Specification for the MOX SAR Fuel Rods to be See changed page(s)

Transported in the TN-LC-1FA Basket Justification Table 1.4.5-6 Fuel Pin can design changed to improve underwater operation of lid.

Specification for the BWR Fuel to be Transported in the in the TN-LC-1FA Bask Table 1.4.5-10 Fuel Qualification Table for 21 PWR/EPR Fuel Rods (UO2)

Table 1.4.5-12 Fuel Qualification Table for 21 BWR Fuel Rods (UO2)

Table 1.4.5-14 Fuel Qualification Table for MOX PWR/BWR 21 Rods and MOX PWR/BWR 9 Rods 15 of 31

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Item Description and Justification Section 1.8 Table 1.4.5-1 PWR Fuel Description of Change Specification for the Fuel to be Transported in the Add WE 16x16 to Fuel Class, and revised Notes TN-LC-1FA Basket SAR Page 1.4.5-5 See changed page(s)

Justification See Change Item 6.3.

1.9 Table 1.4.5-2 PWR Fuel Description of Change Assembly Design Characteristics for Added fuel designations for WE 17x17, W 14x14, CE 16x16, and WE 16x16 Transportation in the TN- SAR LC-1FA Basket See changed page(s)

Page 1.4.5-6 Justification See Items 6.1 through 6.4.

1.10 Table 1.4.5-4 Summary of Description of Change PRA Requirements for PWR Fuel Assembly Added PRA requirements for WE 16x16.

Classes SAR Page 1.4.5-7 See changed page(s)

Justification See Change Item 6.1.

1.11 Figure 1.4.5-6 PRA Description of Change Insertion Locations for WE 16x16 Class Added PRA insertion locations for WE 16x16 Assemblies SAR Page 1.4.5-25 See changed page(s)

Justification See Change Item 6.4.

Chapter 2 Structural Evaluation 2.1 Appendix 2.13.8 TN-LC Description of change:

Basket Structural Evaluation Updated Table of Contents, List of Tables, and List of Figures.

Page 2.13.8-i thru iii SAR See changed page Justification Editorial 16 of 31

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Item Description and Justification Section 2.2 Chapter 2 Description of Change TN-LC Transport Package Refer to pin can instead of 25 pin can Structural Evaluation SAR Pages 2-3, 4, 5, and 49 See affected changed pages.

Appendix 2.13.8 Justification TN-LC Basket Structural Evaluation Original design of pin can permitted loading of 25 individual fuel pins. A design change during fabrication modified the four corner fuel pin tubes to accommodate Pages 2.13.8-1, 5, 8, 9, 11, lid fastening bolts.

12, 13, 48, 58, 59, 60, 65, 66, 67, 68 Appendix 2.13.10 TN-LC Transport Package Thermal Expansion Evaluation Pages 2.13.10-12, 15, and 17 2.3 Appendix 2.13.12 Description of Change TN-LC Transport Package Add flexibility to SAR to allow B23 or B24 for the grade of material allowed for Impact Limiter lid bolts, ram access cover bolts, and trunnion bolts for TN-LC. All references to SA-540 Gr. B23 Cl.1 are revised to add Gr. B24 Cl.1 Page 2-13.12-1, 8d, 14 SAR SA-540 GR. B23 or B24 CL. 1 Justification Design stress values for bolting materials for ASME BPVC Subsection NB (Section III, Class 1) are found in Section II, Part D, Table 4 for SA-540, B23 and B24, Class 1. Table 3 provides stress values for Subsection NF (Section III, Class 3). The design stress values are the same for Type/Grade B23 and B24, and are the same values in Table 3 and Table 4. Excerpts from Table 4 as follows show that the stress intensity values are the same for Grade B23 and B24.

The technical evaluations specify SA-540 Gr B24 Cl. 1 steel for bolts used on the trunnions, lid, and ram access cover. Specifying B23 or B24 allows suppliers more flexibility in meeting material requirements including tests for fracture toughness at -40 °F. The difference in composition may affect the fracture toughness.

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Item Description and Justification Section Chapter 3 Thermal Evaluation 3.1 Page 3-1 Description of change:

Reduced heat load value for PWR/BWR/EPR/MOX Fuel Pins.

SAR See changed page(s)

Justification Pin can capacity reduced from 21 to 25 fuel rods.

3.2 Page 3-4, 3-5, 3-15, 3-30, Description of Change 3-31, 3-32, 3-36, 3-43, 3-46, 3-51, 3-52, 3-53, 3-68, Pin can design change to reduce capacity from 25 to 21..

3-74, 3-76, 3-77, 3-79, 3- SAR 79a, 3-79b, 3-79c, 3-79e, 3-79f, 3-79g, 3-79h, 3-82, See changed page(s) 3-88 thru 91, , Justification Editorial and clarification that thermal evaluation for 25 fuel rods bounds 21 fuel rods Chapter 4 Containment Evaluation 4.1 Chapter 4 Containment Description of change:

TABLE OF CONTENTS Updated Table of Contents, List of Tables, and List of Figures.

Page 4-i SAR See changed page(s)

Justification Editorial 4.2 4.1.1 Description of Change Page 4-2 Corrected spelling of radiographic as an allowed weld inspection method.

SAR 4.1.1.3 Seals and Welds All containment boundary welds are full penetration bevel or groove welds to ensure structural and sealing integrity. These full penetration welds are designed per ASME III Subsection NB and are fully examined by radiographic or ultrasonic methods in accordance with Subsection NB.

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Item Description and Justification Section 4.3 4.1.1 Description of Change Page 4-3 See Item 2.3 SAR 4.1.1.4 Closure The containment vessel contains an integrally-welded bottom closure and a bolted and flanged top closure forging (lid). The lid forging is attached to the cask body with twenty (20), SA-540, Grade B23 or B24, Class 1, 1.0-inch diameter bolts and stainless steel washers. Closure of the bottom plug (with or without gamma shielding) is accomplished by eight (8), SA-540, Grade B23 or B24, Class 1, 0.5-inch diameter cap screws and stainless steel washers. The bolt torque required for the top lid and bottom plug are provided in Drawing 65200-71-01 in Chapter 1, Appendix 1.4.1. The closure bolt analysis is presented in Appendix 2.13.2.

Justification See Item 2.3 4.4 4.2 Containment under Description of Change Normal Conditions of Transport Add containment criteria for 1FA contents and a calculation to specify a leak rate criteria for the 1FA contents to allow test methods less sensitive than the helium Page 4-4 leak test.

4.3 Containment under SAR Hypothetical Accident Conditions See changed page(s)

Page 4-5 Justification 4.5 References Allows flexibility for performing leak tests using methods less sensitive than the helium leak test required to demonstrate leak-tightness required for other contents.

Page 4-7 Appendix 4.6.1 Containment Reference Leak Rate for 1FA Pages 4.6.1-i, and 4.6.1-1 thru 4.6.1-9 Chapter 5 Shielding Evaluation 5.1 Chapter 5 Shielding Description of Change Evaluation As-built lead shielding thickness for TN-LC Unit 01 is non-conforming to Page 5-1,5-2, and 5-3 specification.

SAR The TN-LC Unit 01 as-fabricated cask body has a reduction in its shielding capability due to localized areas where the radial lead thickness may be as low as 3.10 inches. See Appendices 1.4.1 and 5.6.4 for further details.

Justification See Item 5.8 20 of 31

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Item Description and Justification Section 5.2 5.2.1 Crud Evaluation for Description of Change Shielding Containment criteria allows leak rate for 1FA contents.

Page 5-3a SAR Since the TN-LC is designed to be leak-tight, no release calculations are performed except for the 1FA contents (see Appendix 4.6.1) in order to relax the pre-shipment leak- tightness criterion when shipping LWR fuel assemblies or rods.

Justification See Item 4.4 5.3 5.4.4 External Radiation Description of Change Levels Added word chapter and its Page 5-9 SAR The cask pay load with 1 LWR fuel assembly, referred to as 1 FA throughout this chapter and its appendices, results in 2 meters from side of the cask radial dose rates that are bounding for those due to other payloads of the cask.

Justification Editorial 5.4 5.4.4 External Radiation Description of Change Levels As-built lead shielding thickness for TN-LC Unit 01 is non-conforming to Page 5-10 specification.

SAR The TN-LC Unit 01 is the same as the TN-LC with the exception of localized reduced lead thickness on the side of the cask body. The shielding assessment of the TN-LC with a uniformly reduced lead thickness of 3.10 in. is performed in Appendix A.5.4.4.4.5.

Justification See Item 5.8 5.5 Appendix 5.6.4 TN-LC- Description of change:

1FA Basket Shielding Evaluation Updated Table of Contents, List of Tables, and List of Figures.

Pages 5.6.4-i through iv SAR See changed page(s)

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Item Description and Justification Section 5.6 5.6.4.4.4.1 Description of Change NCT, PWR Fuel Changed ratios for scaling factors.

Assembly SAR Package surface:

To estimate the dose rate at the shear key using the HAC source, the neutron Page 5.6.4-16 scaling factor is the ratio of the HAC to NCT neutron source magnitudes, or 2.256E+09/2.078E +09= 1.1. The gamma scaling factor is estimated by using the PWR fuel assembly response function to compute dose rates for the two in-core gamma sources and taking the ratio, or 2.12/2.58 = 0.82. Using the shear key results for the NCT source from Table 5.6.4-33, the dose rate using the HAC source is estimated as 0.82*23.2 + 1.1*(446+4.79) = 508 mrem/hr.

Justification Editorial to correct typos to existing analysis.

5.7 5.6.4.4.5 Reduced Lead Description of Change Thickness Assessment Add assessment for reduced lead thickness in TN-LC unit 01.

5.6.4-18a thru 18d SAR 5.6.4.4.5 Reduced Lead Thickness Assessment The gamma shield minimal lead thickness is 3.38 inches; during fabrication, and prior to the installation of the neutron shield, gamma scanning is used to verify the integrity of the poured lead shielding and a minimum thickness is confirmed by comparison of gamma scan results to a calibration block consisting of a known thickness of lead between steel plates of nominal thickness the same as used in the cask fabrication.

This section provides a comprehensive shielding evaluation for a reduced lead thickness, i.e. lead thickness below the minimal lead thickness of 3.38 inches, for the 1FA basket with 1 PWR fuel assembly and 25 PWR rods in 1 pin can. The assumed lead thickness is 3.10 inches.

The shielding evaluation is consistent with the shielding evaluation performed in Section 5.6.4.4.4 for a minimal lead thickness is 3.38 inches. The shielding evaluation for the reduced lead thickness consists of updated response functions, FQTs, bounding NCT and HAC sources for the 1 PWR fuel assembly and 25 PWR rods contents and shielding analysis of the 1 PWR fuel assembly and 25 PWR rods contents (more)

See changed page(s)

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Item Description and Justification Section 5.8 Table 5.6.4-59 thru 66 Description of Change Page 5.6.4-89a thru 89h Added response functions, Fuel qualification tables, Design Basis Gamma Source Terms, Design Basis Neutron Sources, and Dose Rate Summary tables for TN-LC No. 1 with reduced lead shielding thickness for 1FA contents.

SAR See changed page(s)

Justification The evaluation assesses the impact on shielding analysis for the 1-FA PWR fuel assembly and PWR fuel rods contents due to reduced lead thickness condition. The evaluation includes updated response functions, fuel qualification tables, design basis radiation gamma and neutron sources and dose rates for the contents above and an assumed uniformly distributed lead thickness of 3.10 in. The dose rates results demonstrate the shielding performance of the TN-LC, with a uniformly distributed lead thickness of 3.10 in., is not affected for the authorized 1-FA PWR fuel assembly and PWR fuel rods contents when loaded under the fuel qualification tables requirements developed in Table 5.6.4-61 and Table 5.6.4-62.

Chapter 6 Criticality Evaluation 6.1 6.10.4.2 Fissile Material Description of Change Contents Added WE 16x16 Class Assemblies Page 6.10.4-4 SAR The PWR fuel assemblies and their parameters are provided in Table 6.10.4-2. The KENO model fuel assemblies are constructed using these parameters. Note that WE 16x16 fuel class is not specifically analyzed as this fuel class is similar to WE 17x17 fuel class. WE 16x16 fuel class is a 235 fuel rods design (16x16 - 21 guide/instrument tubes) with fuel characteristics (pellet OD, clad thickness and clad OD) similar to those of WE 17x17 LOPAR. WE 17x17 fuel class is expected to bound WE 16x16 fuel class.

Justification WE 16x16 fuel class is bounded by WE 17x17 fuel class by virtue of similar fuel rods characteristics (pellet OD, clad thickness and clad OD), slightly smaller pitch (0.485 for WE 16x16 as opposed to 0.496 for WE 17x17) in a 16x16 lattice with 21 guide/instrument tubes (235 fuel rods for WE 16x6 fuel class as opposed to 264 fuel rods for WE 17x17 fuel class).

6.2 6.10.4.4 Single Package Description of Change Evaluation Added WE 16x16 Class Assemblies 6.10.4.4.1 Configuration SAR Page 6.10.4-12a 23 of 31

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Item Description and Justification Section 6.2 WE 17x17 Class Assemblies:

The most reactive WE 17x17 assembly evaluated is the WE 17x17 OFA fuel assembly, as shown in Table 6.10.4-8. These class of assemblies will remain subcritical and below the USL with the PRA configuration as shown in Figure 6.10.4-15. The number of PRAs required is 8, each at a minimum diameter of 0.88 cm. The maximum allowable U-235 enrichment is 5.00 weight percent. All rotationally symmetric configurations of the absorber rods are also acceptable.

Results for WE 17x17 class assembly are applicable to WE 16x16 class assembly.

Justification WE 16x16 fuel class is bounded by WE 17x17 fuel class by virtue of similar fuel rods characteristics (pellet OD, clad thickness and clad OD), slightly smaller pitch (0.485 for WE 16x16 as opposed to 0.496 for WE 17x17) in a 16x16 lattice with 21 guide/instrument tubes (235 fuel rods for WE 16x6 fuel class as opposed to 264 fuel rods for WE 17x17 fuel class).

6.3 Table 6.10.4-2 PWR Fuel Description of Change Assembly Parameters Add WE 16x16 fuel types to parameters Page 6.10.4-26 SAR See changed page(s).

Justification Design changes to contents, add WE 16x16 fuel class parameters in Table 6.10.4-2.

6.4 Table 6.10.4-26 Summary Description of Change of PRA Requirements Under all Conditions of Added WE 16x16 Assembly Class.

Transport for PWR Fuel SAR Assembly Classes See changed page(s).

Page 6.10.4-44 Justification WE 16x16 fuel class is bounded by WE 17x17 fuel class by virtue of similar fuel rods characteristics (pellet OD, clad thickness and clad OD), slightly smaller pitch (0.485 for WE 16x16 as opposed to 0.496 for WE 17x17) in a 16x16 lattice with 21 guide/instrument tubes (235 fuel rods for WE 16x6 fuel class as opposed to 264 fuel rods for WE 17x17 fuel class).

WE 17x17 PRAs requirements (PRAs number, locations, minimum B4C content for enrichment up to 5.00 wt% U-235) are applicable to the WE 16x16 fuel class.

Chapter 7 Package Operations 7.1 Chapter 7 Package Description of change:

Operations Updated Table of Contents, List of Tables, and List of Figures.

TABLE OF CONTENTS SAR Pages 7-i See changed page(s)

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Item Description and Justification Section 7.2 7.1.1 TN-LC Cask Description of Change Preparation for Loading Changes described below in excepts from SAR. Reference to Justification shown page 7-2 in brackets, for example [1 ] refers to Justification below for corresponding number in brackets.

SAR

5. Prior to removing the lid, sample the cask cavity atmosphere. If removing the lid at this stage, inspect the lid seals and sealing surfaces and verify that the O-ring seals have been replaced within the last 12 months.[1]

8. The lid, bottom plug and all drain/vent/test ports incorporate O-ring seals. O-ring seals may be reused. Prior to installation, the seals and sealing surfaces shall be inspected. Verify that the seals have been replaced within the last 12 months.[1]

10. Install the two lifting trunnions in place of the front trunnions plugs. Install the trunnion bolts and torque them to the torque specified on drawing 65200 01, Appendix 1.4.1, following the torquing sequence shown in Figure 7-1. [2]

14. Remove the bottom plug assembly, inspect the seals and sealing surfaces, verify that the O-ring seals have been replaced within the last 12 months, lubricate and reinstall the bottom plug assembly, torquing the bolts to the torque specified on drawing 65200-71-01, Appendix 1.4.1.[1]

15. Remove the two test ports, the drain port and the vent port, inspect the seals and sealing surfaces, verify that the O-ring seals have been replaced within the last 12 months, lubricate and reinstall each port (hand tight). The vent port on the lid may be left partially threaded to facilitate draining operations in step 14. The ports covers may be reinstalled over the two test ports at this time.[1]

page 7-3 19. If the cask lid has not already been removed, remove the bolts from the lid and lift the lid from the cask. Inspect the seals and sealing surfaces and verify that the O-ring seals have been replaced within the last 12 months,[1]

Notes:

If loading fuel rods in a TN-LC-1FA basket, place a pin can inside the BWR sleeve with a hold down ring in the basket as shown in drawing 65200-71-102. [3]

7.1.2 TN-LC Cask Wet 7.1.2.1 Preparing the TN-LC Cask for Downending Loading

1. Torque the drain port plug to the torque specified on drawing 65200-71-01, Page 7-4 Appendix 1.4.1. The drain port plug cover may be installed at this time.

2. Verify the lid O-ring seals are new. Not used.

3. Install the cask lid remaining bolts. Follow the torquing sequence shown in Figure 7-1, torque the lid bolts to 400-450 ft-lbs. the torque specified on drawing 65200-71-01, Appendix 1.4.1. [4] [2]

4. Discard cavity test port seal, and install new cavity test port seal. Not used.

6. Backfill with helium to 2.5 +/- 1.0 psig. [4]

Page 7-5 2. If the cask lid has not already been removed, remove the bolts from the lid and, using appropriate slings and/or the cask yoke with appropriate slings, lift the lid from the cask. Inspect the seals and sealing surfaces and verify that the O-ring seals have been replaced within the last 12 months, [1]

6. Install and torque the lid bolts to the torque specified on drawing 65200-71-01, Appendix 1.4.1, following the torquing sequence shown in Figure 7-1.[2]

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Item Description and Justification Section 7.2 Page 7-6 The cask is now ready to be ba k-filled with helium and down-ended as described starting from step 6 in section 7.1.2.1 above.[4]

7.1.4 TN-LC Cask 4. Install the impact limiters on the cask, torqueing the attachment bolts to 330-Preparation for Transport 375 ft-lbs the torque specified on drawing 65200-71-01, Appendix 1.4.1. [2].

7.1.4.1 Placing the TN-LC Cask onto the Conveyance Page 7-6 Page 7-7 10. Install the two trunnions in place of the trunnion plugs, torquing the trunnion bolts to 400-450 ft-lbs. the torque specified on drawing 65200-71-01, Appendix 1.4.1, in the sequence shown in Figure 7-1. [2]

7.2.2 Removal of Contents 3. Remove the drain port plug and install an appropriate fitting in the drain port.

from TN-LC Cask Alternatively, a cask port tool may be used to perform flooding and draining 7.2.2.1 activities. [4]

Unloading the TN-LC Cask in a Fuel Pool Page 7-8 7.2.2.2 Unloading the TN- NOTE: See Section 7.2.2.3 for dry unloading of a 25 pin can. [3]

LC Cask to a Hot Cell 7.2.2.3 Horizontal Unloading of a 25 Pin Can from the TN-LC Cask Page 7-9 This procedure is for handling a TN-LC cask with a 25 pin can [3]

7.4 Other Operations Make port plug cover removal conditional on having been previously installed.

2. Add a allowable leakage rate to permit flexibility for leak rate measurement method when replacing elastomeric seals, and require maintenance leak test when any o-ring is replaced.

3. Provide torque requirements when a Brass vent/drain/test port plug is used and when a SST vent/drain/test port plug is used. A 60-70 FT-LB torque requirement has demonstrated to work well for a SST port plug but may exceed material strength for a Brass vent/drain/test port plug.

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Item Description and Justification Section 7.2 7.4 Other Operations The acceptance criterion for pre-shipment leakage rate testing shall be either (a) a leakage rate of not more than the reference air leakage rate, or (b) no detected 7.4.1 Assembly leakage when tested to a sensitivity of at least 10-3 ref-cm3/s. [1]

Verification Leakage Testing of the The following steps present one method of performing the pre-shipment Containment Boundary verification leakage testing. Alternate methods and order of testing are acceptable as long as the above criteria is satisfied for the TN-LC containment boundary Page 7-11 seals.[4]

Vent Port Plug Seal Leakage Test

1. Remove the vent port plug cover if previously installed. Install the cask port tool in the vent port.[4]

6. Close the vent port plug, torquing it to 60-70 ft-lb. the torque specified on drawing 65200-71-01, Appendix 1.4.1. [2]

8. Connect a mass spectrometer leak detector to the cask port tool. [4]

10. Perform the leakage test pre-shipment leak test in accordance with Section 8.2.2. If either O-ring was replaced, the maintenance leak test in Section 8.2.2 shall be performed. If the leakage rate is greater than 1x10-7 ref cm3/s, repair or replace the vent port seal as required and retest. [1]

Page 7-12 Lid O-ring Leakage Test

13. Remove the lid test port plug cover if previously installed. [4]

17. Connect the leakage detector to the cask port tool. [4]

18. Evacuate the lid test port until the vacuum is sufficient to operate the leakage detection equipment per the manufacturer's recommendations. Perform a pres ure rise leakage test to confirm leakage rate past the outer seal is less than 7x10-7 ref cm3 [1]

19. Perform the pre-shipment leak test in accordance with Section 8.2.2. If either O-ring was replaced, the maintenance leak test in Section 8.2.2 shall be performed. Perform the helium leakage test. If the leakage rate is greater than 1x10-7 ref cm3/s, repair r replace the cask lid or the cask lid O-ring seals as required and retest. [1]

20. Remove the leakage detection equipment. [4]

21. Close lid test port plug and tighten it to the torque specified on drawing 65200-71-01, Appendix 1.4.1. Remove the cask port tool from the lid test port and replace the lid test port plug cover. [1]

Drain Port Plug Seal Leakage Test

22. Remove the cask drain port plug cover if previously installed. [4]

23. Verify that the cask drain port is closed and torqued to 60-70 ft-lbs. the torque specified on drawing 65200-71-01, Appendix 1.4.1.[2]

26. Connect the leakage detector to the cask port tool. [4]

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Item Description and Justification Section 7.2 Page 7-13 29. Remove the leakage detection equipment.[4]

Bottom Plug O-ring Leakage Test

31. Remove the bottom test port plug cover if previously installed.[4]

35. Connect the leakage detector to the cask port tool.[4]

36. Evacuate the bottom test port until the vacuum is sufficient to operate the leakage detection equipment per the manufacturer's recommendations.

Perform a pressure rise leakage test to confirm leakage rate past the outer seal is less than 7x10-7 ref cm3[1]

37. Perform the helium leakage test. If the leakage rate is greater than 1x10-7 ref cm3/s, repair or replace the bottom plug or the bottom plug O-ring seals as required and retest. pre-shipment leak test in accordance with Section 8.2.2. If either O-ring was replaced, the maintenance leak test in Section 8.2.2 shall be performed. [1]

38. Remove the leakage detection equipment.[4]

39. Close bottom test port plug and tighten it to the torque specified on drawing 65200-71-01, Appendix 1.4.1. Remove the cask port tool from the bottom test port and replace the bottom test port plug cover. [2]

Justification

1. Remove the requirement to replace the seal before every transport. Such a requirement is not necessary because elastomer O-rings can be reused, and doing so allows the user to perform a leak-test prior to shipment that does not need to be a helium leak test and greatly simplifies pre-shipment operations.

This is supported by the new release evaluation appendix which calculated allowable release rates. Also updated the applicable standard for leak-testing to the latest version to stay up-to-date with current, applicable standards.

2. In order to avoid listing the torque values in two places in the SAR (in this chapter and on the drawings), all the torque values are replaced with a reference to the drawing where these torques are specified instead.

3. Original design of pin can permitted loading of 25 inividual fuel pins. The capacity of the pin can was decreased from 25 to 21 in its new design, any mention of the actual capacity is removed. A design change during fabrication modified the four corner fuel pin tubes to accommodate lid fastening bolts

4. Optimization of the wording/sequence of operations/correction of eventual typos, etc.; general chapters improvements. Includes renumbering steps as required.

7.3 7.5 References Description of Change Page 7-14 Change ANSI-N14 edition from 1997 to 2014 SAR

1. ANSI N14.5-2014, American National Standard for Radioactive Materials -

Leakage Tests on Packages for Shipment, American National Standards Institute, Inc., New York, 2014.

Justification NRC Information Notice 2016-04, ANSI 14.5-2014 Revision and Leakage Rate Testing Considerations 28 of 31

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Item Description and Justification Section 7.4 7.6 Glossary Description of Change Page 7-15 Refer to pin can instead of 25 pin can Table 7-2 SAR Appendices Containing See change page(s).

Loading Procedures for Various TN-LC Baskets Justification Original design of pin can permitted loading of 25 individual fuel pins. A design Page 7-17 change during fabrication modified the four corner fuel pin tubes to accommodate Appendix 7.7.4 lid fastening bolts.

TN-LC-1FA Basket Wet and Dry Loading and Unloading 7.7.4.1 TN-LC-1FA Basket Wet Loading Page 7.7.4-1, 2 7.7.4.2 TN-LC-1FA Basket Dry Loading Page 7.7.4-3, 4 7.7.4.3 TN-LC-1FA Basket Wet Unloading Page 7.7.4-5, 6 7.5 Figure 7-1 Description of Change TN-LC Packaging Torquing Patterns Page numbers changed Page 7-18 SAR Figure 7-2 See change page(s).

Assembly Verification Justification Leakage Test Editorial Page 7-19 Chapter 8 Acceptance Tests and Maintenance 8.1 8.1.2 Weld Examinations Description of change:

Page 8-1a (LR 719358-0023)

Added weld specifications SAR The TN-LC impact limiter welds are designed, fabricated, and inspected in accordance with the AWS Structural Welding Code - Stainless Steel [17] or the ASME Code,Section III Subsection NF [1], and Section IX [5].

Justification Consistent with guidance in NUREG-3854, Fabrication Criteria for Shipping Containers.

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Item Description and Justification Section 8.2 8.1.3 Structural and Description of change:

Pressure Tests Eliminated options for bottom plug.

8.1.4.1 TN-LC Cask Leakage Tests SAR Page 8-2 See changed page(s) 8.1.4.1.2 Fabrication Justification Verification Leakage Option 2 of the bottom plug was removed from the design due to operational Tests considerations.

Page 8-3 8.2.2 Leakage Tests Page 8-14 8.3 8.1.6.1 Description of change:

Gamma Shield Test Added alternative way to inspect gamma shield, and specify the acceptance criteria for gamma scan as minimum specified lead thickness.

Page 8-4 SAR 8.1.6.1 Gamma Shield Test The TN-LC cask poured lead gamma shielding shall be inspected via gamma scanning at the intersections of a grid no larger than 6 x 6 inches on the outside of the shell prior to installation of the neutron shield. Lead gamma shielding that is packed into place shall be inspected for uncontrolled voids. This shall be done either by gamma scanning or by dimensional inspection and weighing to determine the average density of each part.

The acceptance criterion for the gamma scan is based on dose rate measurements of a test block constructed to replicate the layers of stainless steel, lead, and stainless steel in the TN-LC cask. The thickness of each stainless steel layer in the test block shall be no less than the minimum specified thickness of the corresponding cask shell, and the thickness of the lead layer in the test block shall be no less than the minimum thickness of lead specified for the cask. The dose rate measured using the test block shall be the maximum acceptable reading for the inspected cask.

Justification Fabrication specifications required use of alternate verification method.

8.4 8.1.9 Impact Limiter Description of change:

Wood Test Change lower limit Balsa density from 10 to 7 lb/ft3, moisture content range from Page 8-13 and 13a 6-10% to 6-12%, and redwood crush stress upper limit from 7500 to 7000 psi.

SAR See changed page(s)

Justification See change previously approved for similar package design (Reference NRC Safety Evaluation Report , Docket No. 71-9302, Certificate of Compliance No.

9302, Revision No. 9, April 17, 2019, ADAMS Accession No. ML19112A168))

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Item Description and Justification Section 8.5 8.2.2 Leakage Tests Description of change:

Page 8-14, 14a Added a table for leakage test requirements for 1FA contents.

SAR See changed page(s)

Justification See Item 4.4.

8.6 8.2.3.3 Description of change:

Valves, Rupture Discs, Changed criterial for replacing o-ring seals.

and Gaskets on Containment Vessel SAR Page 8-15 If the bottom plug or the lid is removed, the seals are replaced shall be inspected prior to transport of a loaded TN-LC package. The seals will be leakage tested after retorquing the bolts in accordance with Chapter 7, Section 7.4.

O-ring seals may be reused for transport of an empty TN-LC packaging. O-ring seals shall be replaced inspected prior to each shipment of a loaded TN-LC package and replaced at least every twelve months. For shipments with a loaded TN- LC-1FA basket the O-ring seals shall be replaced within six months of the shipping date or prior to the next shipment, whichever comes first.

Justification These changes because it is now acceptable to reuse the seals, which is OK because they are elastomer seals and meant to be reused.

8.7 8.3 References Description of change:

Page 8-17 a) Change ANSI-N14 edition from 1997 to 2014 b) Add ASME BPVC,Section IX c) Delete 2007 from AWS code specification SAR

4. ANSI N14.5-1997 2014, American National Standard for Leakage Tests on Packages for Shipment of Radioactive Materials.

5. ASME Boiler and Pressure Vessel Code,Section IX, 2004 Edition including 2006 addenda.

17. AWS D1.6/D1.6M-2007, Structural Welding Code - Stainless Steel.

Justification a) NRC Information Notice 2016-04, ANSI 14.5-2014 Revision and Leakage Rate Testing Considerations b) See Item 8.1 c) See Item 8.1 31 of 31

Application for Revision of Certificate of Compliance No. 9358 for the Model No. TN-LC, Revision 9a (Proprietary and SUNSI Version) Docket No. 71-9358 Withheld Pursuant to 10 CFR 2.390

Enclosure 3 to E-56457 Changed Pages for SAR Revision 9a (Public)

PUBLIC TN Americas LLC TN-LC TRANSPORTATION PACKAGE SAFETY ANALYSIS REPORT Docket Number 71-9358 Revision 9 April 2020 TN AMERICAS LLC 7135 Minstrel Way, Suite 300

• Columbia, MD 21045

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Revision History Rev. 0 May 2012 Initial Application for CoC 9358 Rev. 1 August 2011 Response to NRC Request for Supplemental Information (RSI-1)

Rev. 2 May 2012 Response to NRC Request for Additional Information (RAI-1)

Rev. 3 August 2012 Response to NRC Request for Additional Information (RAI-2)

Rev. 4 September 2012 Response to NRC Request for Supplemental Information (RSI-2) with respect to the RAI-2 request Rev. 5 October 2012 Incorporate drawing revisions with respect to RAI-2 Rev. 6 November 2012 Response to NRC Request (RSI-2 and RSI-3)

Rev. 7 October 2013 Initial Application for Revision 1 to CoC 9358 Rev. 8 March 2014 Response to NRC Request for Additional Information (RAI-1)

Rev. 9a April 2020 Revised specifications and technical evaluations to reflect as-built non-conformances for TN-LC Unit 01, revised containment evaluation to establish an allowable leak rate for 1FA contents TN-LC-0100 i

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 3.6.9 Thermal Evaluation of TN-LC Transport Cask with High Burnup Fuel Assemblies ....................................................................................................... 3-79h Chapter 4 Containment .......................................................................................................... 4-1 4.1 Description of Containment System ................................................................................ 4-1 4.1.1 Containment Boundary ........................................................................................ 4-1 4.2 Containment under Normal Conditions of Transport (Type B Packages) ................. 4-4 4.2.1 Containment of Radioactive Material .................................................................. 4-4 4.2.2 Pressurization of Containment Vessel ................................................................. 4-4 4.2.3 Containment Criteria ............................................................................................ 4-4 4.3 Containment under Hypothetical Accident Conditions (Type B Packages) ............... 4-5 4.3.1 Fission Gas Products ............................................................................................ 4-5 4.3.2 Containment of Radioactive Material .................................................................. 4-5 4.3.3 Containment Criterion .......................................................................................... 4-5 4.4 Special Requirements ....................................................................................................... 4-6 4.5 References .......................................................................................................................... 4-7 4.6 Appendices .......................................................................................................................... 4-7 4.6.1 Containment Reference Leak Rate for 1FA Contents .................................... 4.6.1-1 4.6.1.1 Criterion, Parameters and Assumptions ....................................... 4.6.1-1 4.6.1.2 Source Activities and Source Activity Densities ............................ 4.6.1-1 4.6.1.3 Determination of allowable leakage rates .................................... 4.6.1-3 4.6.1.4 Results ........................................................................................... 4.6.1-6 4.6.1.5 Conclusions ................................................................................... 4.6.1-9 4.6.1.6 References ..................................................................................... 4.6.1-9 Chapter 5 Shielding Evaluation ............................................................................................. 5-1 5.1 Description of the Shielding Design................................................................................. 5-1 5.1.1 Design Features .................................................................................................... 5-1 5.1.2 Summary Tables of Maximum Radiation Levels ................................................ 5-1 5.2 Source Specification .......................................................................................................... 5-3 5.2.1 Crud Evaluation for Shielding ............................................................................. 5-3 5.3 Shielding Model................................................................................................................. 5-5 5.3.1 Configuration of Source and Shielding ................................................................ 5-5 5.3.2 Material properties ............................................................................................... 5-6 5.3.3 Notes on VYAL-B and Resin-F Mixing and Installation .................................... 5-6 5.4 Shielding Evaluation ......................................................................................................... 5-8 5.4.1 Methods ................................................................................................................ 5-8 5.4.2 Input and Output Data .......................................................................................... 5-8 5.4.3 Flux-to-Dose-Rate Conversion ............................................................................ 5-8 5.4.4 External Radiation Levels .................................................................................... 5-8 5.5 References ........................................................................................................................ 5-11 5.6 Appendices ....................................................................................................................... 5-12 TN-LC-0100 viii

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 5.6.1 TN-LC-MTR Basket Shielding Evaluation ................................................... 5-6.1-1 5.6.1.1 Description of the Shielding Design ............................................. 5.6.1-1 5.6.1.2 Source Specification ...................................................................... 5.6.1-3 5.6.1.3 Shielding Model ............................................................................ 5.6.1-7 5.6.1.4 Shielding Evaluation ..................................................................... 5.6.1-9 TN-LC-0100 viiia

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 6.10.4.7 Fissile Material Packages for Air Transport.............................. 6.10.4-16 6.10.4.8 Benchmark Evaluations ............................................................. 6.10.4-17 6.10.4.9 Appendix ................................................................................... 6.10.4-19 Chapter 7 Package Operations .............................................................................................. 7-1 7.1 TN-LC Package Loading .................................................................................................. 7-1 7.1.1 TN-LC Cask Preparation for Loading.................................................................. 7-1 7.1.2 TN-LC Cask Wet Loading ................................................................................... 7-3 7.1.3 TN-LC Cask Dry Loading ................................................................................... 7-5 7.1.4 TN-LC Cask Preparation for Transport ............................................................... 7-6 7.2 TN-LC Package Unloading .............................................................................................. 7-7 7.2.1 Receipt of Loaded TN-LC Package from Carrier ................................................ 7-7 7.2.2 Removal of Contents from TN-LC Cask ............................................................. 7-8 7.3 Preparation of Empty Package for Transport ............................................................. 7-10 7.4 Other Operations ............................................................................................................ 7-11 7.4.1 Assembly Verification Leakage Testing of the Containment Boundary ........... 7-11 7.5 References ........................................................................................................................ 7-14 7.6 Glossary ........................................................................................................................... 7-15 7.7 Appendices ....................................................................................................................... 7-16 7.7.1 TN-LC-NRUX Basket Wet and Dry Loading and Unloading ....................... 7.7.1-1 7.7.1.1 TN-LC-NRUX Basket Wet Loading ............................................. 7.7.1-1 7.7.1.2 TN-LC-NRUX Basket Dry Loading ............................................. 7.7.1-3 7.7.1.3 TN-LC-NRUX Basket Wet Unloading ......................................... 7.7.1-4 7.7.1.4 TN-LC-NRUX Basket Dry Unloading.......................................... 7.7.1-5 7.7.2 TN-LC-MTR Basket Wet and Dry Loading and Unloading ......................... 7.7.2-1 7.7.2.1 TN-LC-MTR Basket Wet Loading ............................................... 7.7.2-1 7.7.2.2 TN-LC-MTR Basket Dry Loading ................................................ 7.7.2-3 7.7.2.3 TN-LC-MTR Basket Wet Unloading ............................................ 7.7.2-4 7.7.2.4 TN-LC-MTR Basket Dry Unloading ............................................ 7.7.2-5 7.7.3 TN-LC-TRIGA Basket Wet and Dry Loading and Unloading ...................... 7.7.3-1 7.7.3.1 TN-LC-TRIGA Basket Wet Loading ............................................ 7.7.3-1 7.7.3.2 TN-LC-TRIGA Basket Dry Loading ............................................ 7.7.3-3 7.7.3.3 TN-LC-TRIGA Basket Wet Unloading ........................................ 7.7.3-4 7.7.3.4 TN-LC-TRIGA Basket Dry Unloading ......................................... 7.7.3-5 7.7.4 TN-LC-1FA Basket Wet and Dry Loading and Unloading ........................... 7.7.4-1 7.7.4.1 TN-LC-1FA Basket Wet Loading ................................................. 7.7.4-1 7.7.4.2 TN-LC-1FA Basket Dry Loading ................................................. 7.7.4-3 7.7.4.3 TN-LC-1FA Basket Wet Unloading ............................................. 7.7.4-5 7.7.4.4 TN-LC-1FA Basket Dry Unloading .............................................. 7.7.4-6 TN-LC-0100 xi

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Chapter 1 General Information NOTE: References in this Chapter are shown as [1], [2], etc. and refer to the reference list in Section 1.3.

1.1 Introduction 1.1.1 General Description This chapter presents a general introduction and description of the Transnuclear TN-LC transport package. This application seeks authorization of the TN-LC transport package as a B(U)F-96, category I, spent fuel transport packaging container in accordance with the provisions of Title 10, Part 71 of the Code of Federal Regulations [1]. This Safety Analysis Report (SAR) describes the design features and presents the safety analyses, which demonstrate that the TN-LC transport package complies with applicable requirements of 10 CFR Part 71. The format and content of this document follow the guidelines of Regulatory Guide 7.9 [2].

The TN-LC packaging has been developed for exclusive-use transport of irradiated test, research, and commercial reactor fuel in a closed transport vehicle or an ISO container or flatrack fitted with a weather protection. The fuel is primarily of three basic types: highly enriched aluminum-uranium plate fuel, highly enriched aluminum-uranium pin fuel, and commercial light water reactor fuel assemblies and pins. Within the package, the fuel is contained in basket structures specifically designed for each fuel type that provide for suitable heat rejection and criticality control.

The packaging consists of a payload basket, a shielded cask body, a shielded closure lid, and top and bottom impact limiters. The packaging is of conventional design and utilizes ASME Type 304 and Type XM19 stainless steel as its primary structural materials. The packaging is designed to provide leak-tight containment of the radioactive contents under all normal conditions of transport (NCT) and hypothetical accident conditions (HAC).

The TN-LC packaging may be loaded or unloaded in either a spent fuel pool or a hot cell environment. The cask body is provided with a test/vent port and a drain port. The package is designed to be transported horizontally by highway truck, boat, or by rail in exclusive use. The TN-LC transport package is 230 inches long and 66 inches in diameter.

Based on the criticality assessment provided in Chapter 6, Criticality Evaluation, the Criticality Safety Index for the TN-LC transport package is 100 for the TN-LC-MTR, TN-LC-NRUX, and TN-LC-1FA (PWR fuel assembly payload), and 0 for the TN-LC-TRIGA and TN-LC-1FA (BWR fuel assembly and pin can payload), in accordance with 10 CFR 71.59 [1].

Based on the shielding assessment provided in Chapter 5, Shielding Evaluation, per 10 CFR 71.47(b)(4), the TN-LC transport package shall be transported by private carriers, and personnel in occupied locations shall wear dosimetry devices (except the for the transport of the TN-LC-NRUX basket).

Transnuclear, Inc. has an NRC approved quality assurance program (Docket Number 71-0250),

which satisfies the requirements of 10 CFR Part 71 Subpart H [1].

TN-LC-0100 1-1

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 1.1.2 Licensing Approach for the TN-LC Package

[

]

TN-LC-0100 1-2

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 1-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 payload cavity has a diameter of 18 inches and a minimum length of 182.5 inches. The TN-LC Unit 01 as-fabricated cask has a reduced cavity length of 182.12 inches. This is allowed for Unit 01 for use only with the 1FA and TRIGA baskets, with the height of the top spacer reduced by 0.50 inch (see Appendix 1.4.1). The end flanges are made from ASME SA-182, Type FXM19 stainless steel forgings. The bottom end of the cask has a drain to allow removal of water from the payload cavity and bottom plug for cask operations. A test port with a sealing washer is provided for testing the cask bottom access port cover seal.

The inner and outer shells are made from ASME SA-240, Type XM19 plate. Except for the closure bolts, trunnions and impact limiter attachments, the package is of primarily welded construction, using austenitic stainless steel. The inner and outer shells each may have two or more full penetration longitudinal seam welds and may have circumferential butt welds. The inner shell is 1 inch thick and is welded to each end structure using a full penetration weld. The outer shell is 1.5 inches thick and is connected to each end structure using a full penetration weld.

The cask is lifted using two removable martensitic stainless steel trunnions (SA-182 Type FXM19) which are bolted to the cask body using eight 1-8 UNC bolts. The threaded holes in the upper end cask structure have thread inserts for improved durability.

On the outside of the outer shell, in the region not covered by the impact limiters, is a neutron shield composed of an outer sheet (neutron shield shell) of 0.25 inch thick Type 304 stainless steel, separated from the outer shell by twenty aluminum shield boxes which are filled with neutron absorbing material. The outside diameter of the cask including neutron shield and neutron shield shell is 38.5 inches.

A set of eight 1-8 UNC bolts is used to attach each of two impact limiters. At the top end of the cask, two of the bolt holes are contained in the trunnion attachment blocks and the other six are within attachment blocks which are welded to the cask shell and extend through the neutron shield. At the bottom end of the cask, all eight attachment blocks are welded to the cask shell and extend through the neutron shield. The attachment is completed with the above-mentioned bolts which pass through the impact limiter and thread into the attachment blocks described above.

All lead shielding is made from ASTM B29 copper lead. The annular lead shield is cast-in-place through the upper end structure and is nominally 3.5 inches thick. The TN-LC Unit 01 as-fabricated cask body has a reduction in its shielding capability due to localized areas where the lead thickness may be as low as 3.10 inches. See Appendices 1.4.1 and 5.6.4 and Chapter 5 for further details. The shield at the bottom is made from lead sheet material that is packed firmly into place or poured and is also nominally 3.5 inches thick. The bottom lead cavity is closed using a 1.5 inch stainless steel plate.

The closure lid is machined from an ASME SA-182 F304 forging. It is attached to the cask using twenty 1-8 UNC ASME SA-540 Grade B23 Class 1 hex head bolts and stainless steel washers. The mating holes in the cask body are fitted with heavy duty thread inserts for improved durability. The mating surface of the lid features a step relief located at the bolt circle.

This relief prevents any contact between the lid and the body outside of the bolt circle, thus TN-LC-0100 1-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 preventing prying loads from being applied to the closure bolts. The closure lid includes two fluorocarbon O-ring seals of 3/16 inch cross-sectional diameter. The inner O-ring is the containment seal, and the outer is the test seal.

The gross weight of the loaded package is 51,000 lbs including a payload of 7,100 lbs. Table 1-1 summarizes the dimensions and weights of the TN-LC packaging components.

TN-LC-0100 1-5a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Trunnions, attached to the cask body, are provided for lifting and handling operations, including rotation of the packaging between the horizontal and vertical orientations. The TN-LC transport package is transported in the horizontal orientation on a specially designed shipping frame.

The spent fuel payload in the TN-LC transport cask is shipped dry in a helium atmosphere. The heat generated by the spent fuel is rejected to the environment by conduction, convection and radiation. No forced cooling is required. TN-LC packaging markings are specified on Appendix 1.4.1 drawing.

1.2.1.1.1 Containment Vessel The containment boundary for the TN-LC transport cask consists of the inner shell, the bottom flange, the bottom plug, the bottom plug inner O-ring, the top flange, the lid, the lid inner O-ring and vent and drain port plug bolts and seals. The containment system prevents leakage of radioactive material from the cask cavity and allows pre-shipment leakage testing of the assembled cask configuration.

The containment vessel prevents leakage of radioactive material from the cask cavity. It also maintains an inert atmosphere (helium) in the cask cavity. Helium within the cask assists in heat removal and provides an inert environment to protect the fuel assemblies/elements. To preclude air in-leakage, the cask cavity is pressurized with helium to above atmospheric pressure.

The TN-LC packaging containment system is designed, fabricated, examined and tested in accordance with the requirements of Subsection NB [3] of the ASME Code to the maximum practical extent. In addition, the design meets the requirements of Regulatory Guides 7.6 [4] and 7.8 [5]. Alternatives to the ASME Code are discussed in Chapter 2. The construction of the containment boundary is shown in the drawing provided in Appendix 1.4.1. The design of the containment boundary is discussed in Chapter 2, and the fabrication requirements (including examination and testing) of the containment boundary are discussed in Chapter 4.

1.2.1.1.2 Gamma and Neutron Shielding The lead and steel shells of the TN-LC transport cask provide shielding between the fuel and the exterior surface of the package for the attenuation of gamma radiation.

Neutron shielding is provided by a borated resin compound surrounding the outer shell. The resin compound is cast into long, slender aluminum containers. The containers are constructed from 6063 aluminum. The thickness of the resin is 3.75 inches. The array of resin-filled containers is enclosed within a 0.25 inches thick outer stainless steel shell. In addition to serving as resin containers, the aluminum containers provide a heat conduction path from the cask body to the neutron shield shell.

Non-containment welds are examined in accordance with the requirements of ASME B&PV Code Subsection NF [6].

The structural analysis of the TN-LC transport package is presented in Chapter 2.

TN-LC-0100 1-6

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 1.2.1.2 Tiedown and Lifting Devices There are two trunnion attachment blocks on the cask and two pocket trunnions in the bottom flange. The pocket trunnions are used to support the bottom of the cask during rotation. They may also be used for horizontal lifting. The trunnion attachment blocks accommodate removable trunnions for handling, lifting, and rotating the cask. These trunnion blocks are attached to the structural shell. The TN-LC transport cask trunnions have a single shoulder and are designed to be single failure proof. The trunnions are fabricated and tested in accordance with ANSI N14.6

[7]. During transport, four plugs, containing neutron shielding material, are bolted to the cask in lieu of the trunnions to prevent radiation streaming.

When the cask is in the horizontal position, a shear key slot on the bottom of the cask transfers the longitudinal tie-down loads. The shear key slot is welded to the structural shell and protrudes through the neutron shield. During transport, the receptacle interfaces with the shear key attached to the transport skid.

1.2.1.3 Impact Limiters The front and rear impact limiters, shown in Appendix 1.4.1 drawings, absorb energy during impact events by crushing balsa and redwood blocks. The two impact limiters are identical.

Each has an outside diameter of 66 inches and a height of 27.75 inches. The inner and outer shells are Type 304 stainless steel joined by radial gussets of the same material. The metal structure locates, supports, confines, and protects the wood energy absorption material.

Each impact limiter is attached to the TN-LC transport cask by eight 1-8 UNC bolts made from SA-540 grade B23 or B24 class 1 material. The attachment bolts are designed to keep the impact limiters attached to the cask body during all NCT and HAC.

Each impact limiter is provided with five fusible plugs that are designed to melt during a fire accident, thereby relieving excessive internal pressure. Each impact limiter has four hoist rings for handling and two support angles for supporting the impact limiter in a vertical position during storage. The hoist rings are threaded into the reinforcement blocks which are welded to the impact limiter gusset plates, while the support angles are welded to the outer shell. Prior to transport, the impact limiter hoist rings are removed and replaced with bolts.

The functional description, as well as the performance analysis of the impact limiters, is provided in Chapter 2.

1.2.2 Contents There are four basket designs provided for transport in the TN-LC packaging. Multiple fuel types are permitted in each basket. Details for each basket type are provided in Appendices 1.4.2 through 1.4.5. Table 1-2 lists each basket type provided for transport along with the required fuel spacers. In addition, the table lists the SAR appendix where the payload details can be found for each basket design. The TN-LC cask is designed and evaluated as shown in this SAR to transport the following contents:

• 1 PWR Assembly TN-LC-0100 1-7

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

• 1 BWR Assembly

• Up to 21 PWR (including MOX and EPR) or BWR fuel rods in a pin can basket

• Up to 26 NRU Fuel Assemblies

• Up to 26 NRX Fuel Assemblies

• Up to 54 MTR Fuel Assemblies

• Up to 180 TRIGA Fuel Elements Shipments in the TN-LC transport cask shall not exceed the following limits:

(a) The fuel and content limits specified in Appendices 1.4.2 through 1.4.5.

(b) The maximum decay heat for any payload is 3.0 kW.

(c) Radiation standards shall meet the requirements of 10 CFR 71.47 and 10 CFR 71.51.

(d) Surface contamination shall meet the requirements of 10 CFR 71.87.

(e) The maximum number of fuel rods (pins) in the pin can basket is specified in Appendix 1.4.5.

1.2.3 Special Requirements for Plutonium The TN-LC transport package may contain plutonium in excess of 0.74 Tbq (20 Ci) per package.

As such, the plutonium is in solid form within the fuel matrix and must remain in the solid form.

1.2.4 Operational Features The TN-LC transport package is not considered to be operationally complex and is designed to be compatible with reactor fuel pool loading/unloading at nuclear power plants, research reactors, Savannah River Site and Idaho National Laboratory. The package is also designed to be compatible with hot cells in Europe (as an example: French CEA Cadarache STAR facility, Belgian Nuclear Research Centre (SCK-SEN), German Institute for Transuranium Elements (ITU), Studsviks facility in Sweden, and Paul Scherrer Institute (PSI) in Switzerland) and the United States (as an example: Oak Ridge National Laboratory, Idaho National Laboratory, and General Electric Vallecitos Nuclear Center). All operational features are readily apparent from inspection of the General Arrangement Drawings provided in Appendix 1.4.1, Section 1.4.1.1.

The sequential steps to be followed for cask loading and unloading operations are provided in Chapter 7. The acceptance tests and maintenance program are specified in Chapter 8.

TN-LC-0100 1-8

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1-2 Basket Configurations in the TN-LC Packaging SAR Appendix with Maximum Spacer Basket Type Subbasket Type Detailed Contents Heat Load Required Description (kW)

--- Yes 1.4.2 0.39 TN-LC-NRUX

--- Yes 1.4.2 0.39 TN-LC-MTR --- Yes 1.4.3 1.5 TN-LC-TRIGA --- Yes 1.4.4 1.5 1FA-PWR Yes 1.4.5 3.0 TN-LC-1FA 1FA-BWR Yes 1.4.5 2.0 Pin Can No 1.4.5 2.5 TN-LC-0100 1-12

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Appendix 1.4.1 TN-LC Transport Package Drawings Drawing Number Title 65200-71-01 Revision 8 TN-LC Cask Assembly (11 sheets)

TN-LC Transport Cask 65200-71-02 Revision 0 Regulatory Plate (1 sheet)

TN-LC 65200-71-20 Revision 5 Impact Limiter Assembly (2 sheets)

TN-LC Transport Packaging 65200-71-21 Revision 1 Transport Configuration (1 sheet)

Drawing Number Title TN-LC-NRUX Basket 65200-71-40 Revision 4 Basket Assembly (5 sheets)

TN-LC-NRUX Basket 65200-71-50 Revision 4 Basket Tube Assembly (5 sheets)

Drawing Number Title TN-LC-MTR Basket 65200-71-60 Revision 4 General Assembly (4 sheets)

TN-LC-MTR Basket 65200-71-70 Revision 4 Fuel Bucket (2 sheets)

Drawing Number Title 65200-71-80 Revision 4 TN-LC-TRIGA Basket (5 sheets)

Drawing Number Title 65200-71-90 Revision 5 TN-LC-1FA Basket (5 sheets)

TN-LC-1FA BWR 65200-71-96 Revision 5 Sleeve and Hold-Down Ring (2 sheets)

TN-LC-1FA 65200-71-102 Revision 6 21 Pin Can Basket (4 sheets)

TN-LC-0100 1.4.1-1

Proprietary and Security Related Information for Drawing 65200-71-01, Rev. 7 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 65200-71-01, Rev. 8 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 65200-71-20, Rev. 5 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 65200-71-90, Rev. 5 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 65200-71-90, Rev. 5 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 65200-71-96, Rev. 5 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 65200-71-102, Rev. 5 Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing 65200-71-102, Rev. 6 Withheld Pursuant to 10 CFR 2.390

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Appendix 1.4.5 TN-LC-1FA Basket TABLE OF CONTENTS 1.4.5.1 TN-LC-1FA Basket Description............................................................................. 1.4.5-1 1.4.5.2 TN-LC-1FA Basket Contents ................................................................................. 1.4.5-2 1.4.5.2.1 PWR Fuel Assemblies .................................................................................... 1.4.5-2 1.4.5.2.2 BWR Fuel Assemblies .................................................................................... 1.4.5-2 1.4.5.2.3 Fuel Rods in the Pin Can ............................................................................... 1.4.5-2 1.4.5.3 References .............................................................................................................. 1.4.5-4 LIST OF TABLES Table 1.4.5-1 PWR Fuel Specification for the Fuel to be Transported in the TN-LC-1FA Basket ............................................................................................ 1.4.5-5 Table 1.4.5-2 PWR Fuel Assembly Design Characteristics for Transportation in the TN-LC-1FA Basket ................................................................................ 1.4.5-6 Table 1.4.5-3 Irradiated EPR Fuel Rod Parameters ....................................................... 1.4.5-6 Table 1.4.5-4 Summary of PRA Requirements for PWR Fuel Assembly Classes ............ 1.4.5-7 Table 1.4.5-5 Specification for the MOX Fuel Rods to be Transported in the TN-LC-1FA Basket ............................................................................................ 1.4.5-8 Table 1.4.5-6 Fuel Specification for the BWR Fuel to be Transported in the in the TN-LC-1FA Basket...................................................................................... 1.4.5-9 Table 1.4.5-7 BWR Fuel Assembly Design Characteristics (1) for Transportation in the TN-LC-1FA Basket .............................................................................. 1.4.5-10 Table 1.4.5-8 Fuel Qualification Table for a PWR Fuel Assembly ............................... 1.4.5-12 Table 1.4.5-8a Fuel Qualification Table for a PWR Fuel Assembly - 3.10 Lead Thickness ................................................................................................. 1.4.5-12a Table 1.4.5-9 Fuel Qualification Table for a BWR Fuel Assembly ............................... 1.4.5-13 Table 1.4.5-10 Fuel Qualification Table for 21 PWR/EPR Fuel Rods (UO2) ................. 1.4.5-14 Table 1.4.5-10a Fuel Qualification Table for 21 PWR Fuel Rods (UO2) - 3.10 Lead Thickness ................................................................................................. 1.4.5-14a Table 1.4.5-11 Fuel Qualification Table for 9 PWR/EPR Fuel Rods (UO2) ................... 1.4.5-15 Table 1.4.5-12 Fuel Qualification Table for 21 BWR Fuel Rods (UO2) .......................... 1.4.5-16 Table 1.4.5-13 Fuel Qualification Table for 9 BWR Fuel Rods (UO2) ............................ 1.4.5-17 Table 1.4.5-14 Fuel Qualification Table for MOX PWR/BWR 21 Rods and MOX PWR/BWR 9 Rods ..................................................................................... 1.4.5-18 TN-LC-0100 1.4.5-i

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 LIST OF FIGURES Figure 1.4.5-1 PRA Insertion Locations for WE 14x14 Class Assemblies .......................... 1.4.5-20 Figure 1.4.5-2 PRA Insertion Locations for WE 15x15 Class Assemblies .......................... 1.4.5-21 Figure 1.4.5-3 PRA Insertion Locations for BW 15x15 Class Assemblies .......................... 1.4.5-22 Figure 1.4.5-4 PRA Insertion Locations for BW 17x17 and WE 17x17 Class Assemblies ................................................................................................. 1.4.5-23 Figure 1.4.5-5 Poison Rod Assemblies (PRAs) ................................................................... 1.4.5-24 Figure 1.4.5-6 PRA Insertion Locations for WE 16x16 Class Assemblies .......................... 1.4.5-25 TN-LC-0100 1.4.5-ii

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Appendix 1.4.5 TN-LC-1FA Basket NOTE: References in this Appendix are shown as [1], [2], etc. and refer to the reference list in Section 1.4.5.3.

1.4.5.1 TN-LC-1FA Basket Description The TN-LC cask is designed to contain the TN-LC-1FA basket assembly, with either (a) one fuel assembly (PWR or BWR) or (b) one pin can with up to 21 fuel rods (and spacers) while remaining completely supported by the transport cask.

The basket structure is designed, fabricated and inspected in accordance with ASME B&PV Code Subsection NG [1]. Alternatives to the code are provided in Chapter 2. The overall length of the basket is 181.5 in. and has a diameter of 17.5 in. The details of the TN-LC-1FA basket are shown on drawing 65200-71-90, 65200-71-96 and 65200-71-102 in Chapter 1, Appendix 1.4.1.

The PWR basket structure consists of a thick square-shaped welded or bolted tube assembly which is attached to the solid aluminum support rails. The poison plate is sandwiched between each rail and frame on all four sides of the compartment. The BWR compartment, which slides inside the PWR compartment (to accommodate the smaller cross section of a BWR assembly), is comprised of a 17.5 inch long hold-down ring and a 164 inch long BWR sleeve. The hold-down ring is designed for BWR fuel assembly loading to provide lateral clearance for a fuel grapple, if necessary. After fuel loading, the hold-down ring is installed to provide continuous transfer of basket loads to the cask.

The minimum B-10 areal density of the poison plate is 16.7 mg/cm2 if boron aluminum alloy or metal matrix composite (MMC) is used. The minimum B-10 areal density of the poison plate is 20.0 mg/cm2 if Boral is used.

The basket structure is open at each end. Therefore, longitudinal fuel assembly or fuel pin can loads are applied directly to the cask body and not the fuel basket structure. The fuel assembly or fuel pin can is supported laterally by the stainless steel tube assembly. The basket is supported laterally by the basket rails and the cask shell. The solid aluminum basket rails are oriented parallel to the axis of the cask and are attached to the periphery of the basket to provide support and to establish and maintain basket orientation.

The pin can is a welded 5x5 square array of stainless steel 1 in. tubes which are wrapped in a stainless steel plate and slides inside the BWR basket. The top of the pin can has a bolted closure lid, and the four corner tubes are shortened at the top to install four solid rods: two with threads for the lid bolts and two with locating pins to facilitate installation of the lid underwater.

The closure lid has threads to attach a lifting handle to allow handling of the pin can.

A shear key, welded to the inner wall of the cask, mates with a notch in a basket support rail to prevent the basket from rotating during normal operations.

The maximum allowable heat load for the TN-LC cask with TN-LC-1FA basket is 3.0 kW.

TN-LC-0100 1.4.5-1

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 1.4.5.2 TN-LC-1FA Basket Contents The TN-LC-1FA basket has three different types of intact payload: PWR fuel assemblies, BWR fuel assemblies, and fuel rods from PWR, BWR, MOX, and EPR fuel assemblies. Intact payloads are fuel assemblies or fuel rods with no known or suspected cladding defects greater than hairline cracks or pinhole leaks.

1.4.5.2.1 PWR Fuel Assemblies The TN-LC-1FA basket is designed to transport one intact PWR fuel assembly, as specified in Table 1.4.5-1. The PWR fuel qualification table (FQT) is provided in Table 1.4.5-8. The fuel to be transported is limited to a maximum assembly average initial enrichment of 5.0 wt. % 235U except for CE 15x15 class assemblies (maximal assembly initial enrichment of 3.7 wt. % 235U).

The maximum assembly average burnup is limited to 62 GWd/MTU. The maximum allowable heat load for the TN-LC-1FA basket loaded with a PWR fuel assembly is 3.0 kW.

In addition to the poison plates provided in the basket, Poison Rod Assemblies (PRAs) are required while transporting PWR fuel assemblies in order to ensure that the maximum reactivity is subcritical and below the Upper Subcritical Limit (USL). The PRAs consist of a cluster of absorber rods containing B4C pellets inserted into the guide tubes of the fuel assembly. A typical PRA is illustrated in Figure 1.4.5-5. The minimum required B4C content of the absorber rods in the PRA is 40 percent Theoretical Density (TD) (75 percent credit is taken in the criticality analysis, or 30 percent TD). A summary of the number of absorber rods required in the PRA for each PWR fuel class is shown in Table 1.4.5-4. PRA loading configurations are also illustrated in Figure 1.4.5-1 through Figure 1.4.5-4.

1.4.5.2.2 BWR Fuel Assemblies The TN-LC-1FA basket is designed to transport one intact BWR fuel assembly as specified in Table 1.4.5-6. Basket cell sleeves are used to reduce the area within the 1FA basket for BWR fuel. The BWR FQT is provided in Table 1.4.5-9. The fuel to be transported is limited to a maximum assembly average initial enrichment of 5.0 wt. % 235U. The maximum allowable assembly average burnup is limited to 62 GWd/MTU. The maximum allowable heat load for the TN-LC-1FA basket loaded with a BWR fuel assembly is 2.0 kW.

1.4.5.2.3 Fuel Rods in the Pin Can The TN-LC-1FA basket is designed to transport up to 21 intact light water reactor fuel rods in the pin can. This includes irradiated PWR, BWR, MOX, and EPR fuel rods. The maximum peak burnup for fuel rods is 90 GWd/MTU. Two designs are available, with cavity lengths of 180.24 in. or 169.55 in. The pin can with the shorter cavity length is heavily shielded with lead at the ends, while the pin can with the longer cavity length does not feature axial lead shielding.

The longer cavity pin can is used only for EPR pins, which are much longer than a standard fuel rod (an EPR rod is approximately 179.24 in. long). All other rods are transported in the shorter cavity pin can with heavy axial shielding.

PWR and BWR intact fuel rods may be from any of the fuel assemblies listed in Table 1.4.5-1 or Table 1.4.5-6, respectively. The pin can may transport up to 21 fuel rods, although the cooling times are reduced if 9 or fewer rods are transported. When transporting 9 or fewer rods, the rods TN-LC-0100 1.4.5-2

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 shall be placed in the center 3x3 region of the pin can. PWR rod FQTs are shown in Table 1.4.5-10 and Table 1.4.5-11 for the 21 and 9 rod configurations, respectively. BWR rod FQTs are shown in Table 1.4.5-12 and Table 1.4.5-13 for the 21 and 9 rod configurations, respectively.

MOX rods have the same geometry as PWR or BWR rods, as defined in Table 1.4.5-1 and Table 1.4.5-5, although with a different fuel composition. The composition of MOX fuel is specified in Table 1.4.5-5.

The MOX rod FQT is provided in Table 1.4.5-14 for both 21 and 9 rods. The MOX rod FQT is applicable to both BWR and PWR MOX rods.

EPR rods may be either standard (UO2) or MOX. UO2 EPR rods have a uranium loading of 0.0020 MTU/rod, which is bounded by the B&W 15x15 Mark B10 rod listed in Table 1.4.5-1.

Therefore, EPR rods are governed by the PWR rod FQTs (Table 1.4.5-10 and Table 1.4.5-11),

while MOX EPR rods are governed by the MOX rod FQTs (Table 1.4.5-14).

Solid stainless steel spacers are inserted into the tubes prior to fuel rod loading to leave approximately 2 in. of each fuel rod protruding above the top of the base 5x5 array of tubes in the pin can assembly to allow handling.

The maximum allowable heat load for TN-LC cask with TN-LC-1FA basket loaded with fuel rods in the pin can is 2.5 kW (120 watts per rod) for the 21 rod option and 1.8 kW (220 watts per rod) for the 9 rod option.

TN-LC-0100 1.4.5-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-1 PWR Fuel Specification for the Fuel to be Transported in the TN-LC-1FA Basket PHYSICAL PARAMETERS:

(1)(2) Intact unconsolidated B&W 17x17, WE 17x17, Fuel Class CE 16x16, B&W 15x15, WE 15x15, CE 15x15, WE 14x14, WE 16x16, and CE 14x14 class PWR assemblies (without control components) that are enveloped by the fuel assembly design characteristics listed in Table 1.4.5-2. Reload fuel manufactured by the same or other vendors but enveloped by the design characteristics listed in Table 1.4.5-2 is also acceptable.

Maximum Assembly + PRA Weight 1850 lbs Fissile Material UO2 Maximum Initial Uranium Content(4) 490 kg/assembly Maximum Unirradiated Assembly Length 178.3 inches THERMAL/RADIOLOGICAL PARAMETERS:

Fuel Assembly Average Burnup, Enrichment and Per Table 1.4.5-8 Minimum Cooling Time Maximum Planar Average Initial Enrichment 5.0(3) wt.% 235U Maximum Decay Heat(5) 3.0 kW per Assembly

• 16.7 mg/cm2 (Natural or Enriched Boron Aluminum Alloy / Metal Matrix Composite (MMC))

Minimum B-10 content in poison plates loading

• 20.0 mg/cm2 (Boral)

Minimum number of absorber rods per PRA as a Per Table 1.4.5-4 function of assembly class Notes:

1. Up to 21 PWR fuel rods from any of the PWR fuel assemblies listed in Table 1.4.5-2 may also be transported in the TN-LC-1FA basket in a pin can. The fuel rods are loaded in a pin can with a cavity length of 169.55 inches (Option 3) which is placed within the TN-LC-1FA basket. The maximum peak burnup for the fuel rods is 90 GWd/MTU. The required cooling time as a function of a PWR fuel rod burnup and enrichment are provided in Table 1.4.5-10 for 21 rods and Table 1.4.5-11 for 9 rods, respectively.

2. Up to 21 EPR fuel rods from any of the fuel class listed in Table 1.4.5-2 and meeting EPR rod parameters specified in Table 1.4.5-3 may also be loaded in the TN-LC-1FA basket. The fuel rods are loaded in a pin can with a cavity length of 180.24 inches (Option 1 and Option 2) which is placed within the TN-LC-1FA basket.

The maximum peak burnup for the fuel rods is 90 GWd/MTU. The required cooling time as a function of an EPR fuel rod burnup and enrichment are provided in Table 1.4.5-10 for 21 rods and Table 1.4.5-11 for 9 rods, respectively.

3. For CE 15x15, the maximum planar average initial enrichment is 3.70 wt. % 235U.

4. The maximum initial uranium content is based on the shielding analysis. The listed value is higher than the actual.

5. The maximum decay heat per rod is 220 watts when loading up to 9 rods. The maximum decay heat per rod is 120 watts when loading 10 or more (up to 21) rods.

TN-LC-0100 1.4.5-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-2 PWR Fuel Assembly Design Characteristics for Transportation in the TN-LC-1FA Basket B&W B&W WE CE WE CE WE CE WE Assembly Class 15x15 17x17 17x17 15x15 15x15 14x14 14x14 16x16 16x16 OFA, Fuel Designations KOFA, KSFA, OFA, STD, (for information STD, V5H, GUARDIAN,

- - - - - KOFA, KOFA, only, including but RFA, PLUS7, STD 16ACE7 not limited to) 17ACE7, HIPER16 HIPER17 Number of Fuel 208 264 264 216 204 176 179 236 235 Rods Number of Guide /

17 25 25 9 21 5 17 5 21 Instrument Tubes Rod Pitch(1), in 0.496 0.556 0.506 0.496 0.568 0.502 0.550 0.563 0.580 (mm) (12.60) (14.12) (12.85) (12.60)

Pellet Diameter(1), 0.323 0.368 0.326 0.323 0.374 0.323 0.360 0.367 0.382 in (mm) (8.20) (9.35) (8.28) (8.20)

Clad Outer 0.360 0.400 0.374 0.360 Diameter(1), in 0.416 0.379 0.417 0.422 0.440 (9.14) (10.16) (9.50) (9.14)

(mm)

Clad Thickness(1), 0.022 0.022 0.023 0.022 0.024 0.024 0.026 0.024 0.026 in (mm) (0.57) (0.57) (0.57) (0.57)

Notes:

1. The fuel assembly fabrication documentation may be used to demonstrate compliance with these fuel assembly parameters. The fuel assembly parameters are design nominal values. The maximum and minimum dimensions are specified to bound variations in design nominal values among fuel assemblies within a fuel assembly class (or an array type).

Table 1.4.5-3 Irradiated EPR Fuel Rod Parameters Parameter Value Maximum Unirradiated Length 179.5 inches Cladding Thickness Nominal 0.022 inch Maximum Initial Uranium Content 2 kgU/rod TN-LC-0100 1.4.5-6

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-4 Summary of PRA Requirements for PWR Fuel Assembly Classes Diameter of Minimum Number of Absorber B4 C B4 C Assembly Rods in PRAs and Absorber Content Class Locations (cm) (g/cm)

WE 17x17 8, Per Figure 1.4.5-4 0.88 0.613 WE 16x16 8, Per Figure 1.4.5-6 0.88 0.613 CE 16x16 5, All Guide Tubes 1.02 0.824 BW 15x15 8, Per Figure 1.4.5-3 0.88 0.613 CE 15x15 1, Center Guide Tube 0.76 0.475 WE 15x15 8, Per Figure 1.4.5-2 0.88 0.613 CE 14x14 5, All Guide Tubes 1.02 0.824 WE 14x14 8, Per Figure 1.4.5-1 0.88 0.613 BW 17x17 8, Per Figure 1.4.5-4 0.76 0.475 TN-LC-0100 1.4.5-7

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-5 Specification for the MOX Fuel Rods to be Transported in the TN-LC-1FA Basket PHYSICAL PARAMETERS:

• Up to 21 PWR MOX fuel rods with physical parameters as those listed in Table 1.4.5-1.

• Up to 21 BWR MOX fuel rods with physical parameters as those listed in Table 1.4.5-6.

• Up to 21 EPR MOX fuel rods with physical parameters as those listed in Table 1.4.5-3.

Fissile Material UO2, PuO2 (Mixed Oxide or MOX)

Heavy Metal (HM) Content 2.5 kgU/rod CRITICALITY PARAMETERS 235

• U Content in UO2 : 0.5 235U 0.7 wt. %

• Plutonium Content: Pu / (U + Pu) 7.0 wt. %

Initial MOX composition:

• Initial Pu-239 Content in PuO2 60.0 wt. %

• Initial Pu-241 Content in PuO2 7.5 wt. %

THERMAL/RADIOLOGICAL PARAMETERS:

• Pu-238 / Pu-239 4.0 wt. %

Initial MOX Composition for Fuel Qualification

• Pu-239/ PuO2 50 wt. %

• Am-241 / PuO2 70 ppm

• 235U /U 0.5 wt. %

Burnup and Minimum cooling time for MOX rods Per Table 1.4.5-14

• 2.5 kW for the pin can with up to 21 rods Maximum Decay heat per pin can

• 1.8 kW for the pin can with up to 9 rods

• 16.7 mg/cm2 (Boron Aluminum Alloy / Metal Matrix Minimum B10 content in poison plates loading Composite (MMC))

• 20.0 mg/cm2 (Boral)

TN-LC-0100 1.4.5-8

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-6 Fuel Specification for the BWR Fuel to be Transported in the in the TN-LC-1FA Basket PHYSICAL PARAMETERS:

Intact 7x7, 8x8, 9x9 or 10x10 BWR assemblies Fuel Class(1) manufactured by General Electric or Exxon/ANF or FANP or ABB or reload fuel manufactured by same or other vendors that are enveloped by the fuel assembly design characteristics listed in Table 1.4.5-7.

Fuel may be transported with or without channels, Channels channel fasteners, or finger springs.

Fissile Material UO2 Maximum Assembly Weight with Channels 790 lbs Maximum Unirradiated Assembly Length 176.6 inches THERMAL/RADIOLOGICAL PARAMETERS:

Maximum Lattice Average Initial Enrichment 5.0 wt. % 235U Fuel Assembly Average Burnup, Enrichment and Minimum Per Table 1.4.5-9 Cooling Time Maximum Decay Heat(2) 2.0 kW per Assembly

• 16.7 mg/cm2 (Boron Aluminum Alloy / Metal Minimum B10 Content in Poison Plates Matrix Composite (MMC))

• 20.0 mg/cm2 (Boral)

Notes:

1. Up to 21 fuel rods from any of the BWR fuel assemblies listed in Table 1.4.5-7 may also be transported in the TN-LC-1FA basket in the pin can. The fuel rods are loaded in a pin can with a cavity length of 169.55 inches (Option 3) which is placed within the TN-LC-1FA basket. The required cooling time as a function of BWR fuel rod burnup and enrichment are provided in Table 1.4.5-12 for 21 rods and Table 1.4.5-13 for 9 rods respectively.

2. The maximum decay heat per rod is 220 watts when loading up to 9 rods. The maximum decay heat per rod is 120 watts when loading 10 or more (up to 21) rods.

TN-LC-0100 1.4.5-9

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-8a Fuel Qualification Table for a PWR Fuel Assembly - 3.10 Lead Thickness (Minimum required years of cooling time after reactor core discharge)

Burn-up, Initial Assembly Averaged 235-U Enrichment, wt.%.

GWd/

MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 3.0 3.0 2.9 2.9 2.8 2.8 2.8 2.8 2.7 2.7 2.7 2.7 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.4 2.4 2.4 2.4 2.4 20 4.7 4.6 4.5 4.4 4.3 4.2 4.2 4.1 4.0 4.0 3.9 3.9 3.8 3.8 3.8 3.7 3.7 3.6 3.6 3.6 3.6 3.5 3.5 3.5 3.5 3.4 3.4 3.4 3.4 3.4 3.4 3.3 3.3 3.3 3.3 3.3 30 6.7 6.5 6.3 6.2 6.0 5.9 5.7 5.6 5.5 5.3 5.2 5.1 5.1 5.0 4.9 4.8 4.7 4.7 4.6 4.6 4.5 4.5 4.4 4.4 4.3 4.3 4.2 4.2 4.2 4.1 4.1 4.1 4.1 4.0 39 7.1 6.9 6.7 6.6 6.4 6.3 6.2 6.0 5.9 5.8 5.7 5.6 5.5 5.4 5.4 5.3 5.2 5.2 5.1 5.0 5.0 4.9 4.9 4.8 4.8 4.8 4.7 4.7 4.6 4.6 4.6 40 6.4 6.2 6.1 6.0 5.9 5.8 5.7 5.6 5.5 5.4 5.4 5.3 5.2 5.2 5.1 5.1 5.0 5.0 4.9 4.9 4.8 4.8 4.7 4.7 4.7 50 9.6 9.4 9.2 8.9 8.7 8.5 8.3 8.1 7.9 7.8 7.6 7.5 7.3 7.2 7.1 6.9 6.8 6.7 6.6 6.5 6.4 6.3 6.2 6.2 6.1 55 12.0 11.7 11.4 11.1 10.8 10.5 10.3 10.0 9.8 9.6 9.4 9.1 8.9 8.8 8.6 8.4 8.2 8.1 7.9 7.8 7.7 7.5 7.4 7.3 7.2 60 14.8 14.4 14.1 13.7 13.4 13.0 12.7 12.4 12.1 11.8 11.5 11.3 11.0 10.7 10.5 10.3 10.1 9.8 9.6 9.4 9.3 9.1 8.9 8.8 8.6 61 15.4 15.0 14.7 14.3 13.9 13.6 13.3 12.9 12.6 12.3 12.0 11.7 11.5 11.2 10.9 10.7 10.5 10.2 10.0 9.8 9.6 9.4 9.3 9.1 8.9 62 16.1 15.7 15.3 14.9 14.5 14.2 13.8 13.5 13.1 12.8 12.5 12.2 11.9 11.7 11.4 11.1 10.9 10.7 10.4 10.2 10.0 9.8 9.6 9.4 9.3 Enr. wt.% 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 Note:

1. Explanatory notes and limitations regarding the use of this table follow Table 1.4.5-14.

TN-LC-0100 1.4.5-12a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-10 Fuel Qualification Table for 21 PWR/EPR Fuel Rods (UO2)

(Minimum required years of cooling time after reactor core discharge)

Burnup, Enrichment, wt. %.235U GWd/

MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 20 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 39 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 40 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 45 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 50 0.30 0.30 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 55 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 60 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 61 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.30 0.30 62 0.40 0.40 0.40 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 65 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 70 0.50 0.50 0.50 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 0.45 75 0.65 0.65 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.60 0.50 0.50 0.50 0.50 80 0.85 0.85 0.75 0.75 0.75 0.75 0.75 0.70 0.70 0.70 0.70 0.70 0.70 0.70 85 1.05 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 90 1.25 1.25 1.25 1.15 1.15 1.15 1.10 1.10 1.10 1.00 1.00 1.00 1.00 0.95 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 Notes:

1. Explanatory notes and limitations regarding the use of this table follow Table 1.4.5-14.

TN-LC-0100 1.4.5-14

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-10a Fuel Qualification Table for 21 PWR Fuel Rods (UO2) - 3.10 Lead Thickness (Minimum required years of cooling time after reactor core discharge)

Burn-up, Initial Assembly Averaged 235-U Enrichment, wt.%.

GWd/

MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 20 0.22 0.22 0.22 0.22 0.22 0.22 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 30 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 39 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 40 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 45 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 50 0.31 0.30 0.30 0.30 0.30 0.30 0.30 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.28 0.28 0.28 0.28 0.28 0.28 0.28 0.28 0.28 0.28 55 0.37 0.37 0.37 0.36 0.36 0.35 0.35 0.35 0.35 0.34 0.34 0.34 0.34 0.33 0.33 0.33 0.33 0.33 0.33 0.32 0.32 0.32 0.32 0.32 0.32 60 0.50 0.49 0.48 0.47 0.47 0.46 0.45 0.45 0.44 0.44 0.43 0.43 0.42 0.42 0.41 0.41 0.41 0.40 0.40 0.40 0.39 0.39 0.39 0.38 0.38 61 0.53 0.52 0.51 0.50 0.50 0.49 0.48 0.47 0.47 0.46 0.46 0.45 0.44 0.44 0.44 0.43 0.43 0.42 0.42 0.42 0.41 0.41 0.41 0.40 0.40 62 0.56 0.55 0.54 0.53 0.53 0.52 0.51 0.50 0.49 0.49 0.48 0.48 0.47 0.46 0.46 0.45 0.45 0.44 0.44 0.44 0.43 0.43 0.43 0.42 0.42 65 0.56 0.55 0.55 0.54 0.53 0.53 0.52 0.51 0.51 0.50 0.50 0.49 0.49 0.49 70 0.71 0.70 0.69 0.68 0.67 0.67 0.66 0.65 0.65 0.64 0.63 0.62 0.62 0.61 75 0.87 0.86 0.85 0.84 0.83 0.82 0.80 0.79 0.79 0.78 0.77 0.76 0.75 0.75 80 1.04 1.03 1.01 1.00 0.99 0.97 0.96 0.95 0.94 0.93 0.92 0.91 0.90 0.89 85 1.24 1.22 1.20 1.18 1.16 1.15 1.13 1.12 1.10 1.09 1.08 1.06 1.05 1.04 90 1.47 1.44 1.41 1.39 1.37 1.34 1.32 1.30 1.29 1.27 1.25 1.23 1.22 1.20 Enr. wt.% 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 Note:

1. Explanatory notes and limitations regarding the use of this table follow Table 1.4.5-14.

TN-LC-0100 1.4.5-14a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-12 Fuel Qualification Table for 21 BWR Fuel Rods (UO2)

(Minimum required years of cooling time after reactor core discharge)

Burnup, Enrichment, wt. %.235U GWd/

MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 20 0.30 0.30 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 39 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 40 0.40 0.40 0.40 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 45 0.45 0.45 0.45 0.45 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 0.40 50 0.60 0.60 0.60 0.60 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.55 0.50 0.50 0.50 0.50 0.50 0.50 0.50 55 0.75 0.75 0.75 0.75 0.75 0.70 0.70 0.70 0.70 0.70 0.70 0.70 0.70 0.70 0.65 0.65 0.65 0.65 0.65 0.65 0.65 0.65 0.65 0.65 0.65 60 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.75 0.75 0.75 0.75 0.75 0.75 0.75 0.75 61 1.05 1.05 1.00 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.75 0.75 0.75 0.75 62 1.10 1.05 1.05 1.05 1.05 1.00 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.90 0.90 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 0.85 65 1.05 1.05 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.90 0.90 0.90 0.90 0.90 70 1.20 1.20 1.20 1.15 1.15 1.15 1.15 1.15 1.15 1.10 1.10 1.10 1.10 1.10 75 1.45 1.45 1.45 1.40 1.40 1.40 1.30 1.30 1.30 1.30 1.25 1.25 1.25 1.25 80 1.70 1.70 1.65 1.65 1.60 1.60 1.60 1.50 1.50 1.50 1.45 1.45 1.45 1.45 85 2.15 2.05 2.00 2.00 1.95 1.85 1.85 1.80 1.80 1.70 1.70 1.65 1.65 1.65 90 2.60 2.55 2.50 2.40 2.35 2.30 2.20 2.15 2.15 2.10 2.00 2.00 1.95 1.95 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 Notes:

1. Explanatory notes and limitations regarding the use of this table follow Table 1.4.5-14.

TN-LC-0100 1.4.5-16

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 1.4.5-14 Fuel Qualification Table for MOX PWR/BWR 21 Rods and MOX PWR/BWR 9 Rods 9 Rods 21 Rods Burnup, GWd/MTHM 0.5 wt.% of 0.7 wt.% of 0.5 wt.% of 0.7 wt.% of 235 235 235 235 U U U U 10 0.25 0.25 0.25 0.25 20 0.25 0.25 0.30 0.30 30 0.25 0.25 0.50 0.50 40 0.25 0.25 0.95 0.95 45 0.25 0.25 1.25 1.25 50 0.35 0.35 1.70 1.70 55 0.40 0.40 2.20 2.10 60 0.45 0.45 2.80 2.70 62 0.55 0.55 3.75 3.65 Notes:

1. Explanatory notes and limitation regarding the use of this table follow Table 1.4.5-14.

TN-LC-0100 1.4.5-18

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Notes:

General

• Use burnup and enrichment to lookup minimum cooling time in years. Licensee is responsible for ensuring that uncertainties in fuel enrichment and burnup are correctly accounted for during fuel qualification.

• For values not explicitly listed in the tables, round burnups up to the first value shown, round enrichments down, and select the cooling time listed.

• UO2 Fuel with an initial enrichment less than 0.7 (or less than the minimum provided above for each burnup) or greater than 5.0 wt. % 235U is unacceptable for transportation.

• Shaded areas in these Tables indicate fuel is not analyzed for loading.

For Fuel Assemblies

• Burnup = Assembly Average burnup.

• Enrichment = Assembly Average Enrichment.

• Fuel assembly with a burnup greater than 62 GWd/MTU is unacceptable for transportation.

For Fuel Rods

• Burnup = Maximum burnup.

• Enrichment = Rod Average Enrichment.

• When transporting 21 or less fuel rods, the rods shall be placed in a specially designed pin can.

• When transporting 9 or less fuel rods, the rods shall be placed in the 3x3 region of the pin can.

• Fuel rods with a burnup greater than 90 GWd/MTU are unacceptable for transportation.

Example: Per Table 1.4.5-8, a PWR assembly with an initial enrichment of 4.85 wt. % 235U and a burnup of 41.5 GWd/MTU is acceptable for transport after a 3.95-year cooling time as defined by 4.8 wt. % 235U (rounding down) and 45 GWd/MTU (rounding up) on the qualification table (other considerations not withstanding).

TN-LC-0100 1.4.5-19

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Note: This configuration indicates the relative location of the poison rods within the guide tubes and does not provide any other fuel class specific information. Any other configuration of poison rods that is rotationally symmetric is also acceptable.

Figure 1.4.5-6 PRA Insertion Locations for WE 16x16 Class Assemblies TN-LC-0100 1.4.5-25

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The impact limiter stainless steel cylinders, gussets, and end plates are designed to position and confine the balsa and redwood blocks to minimize the impact forces and to prevent excessive deformation of the limiters. The stainless steel shell is also designed to support and isolate the wood blocks from ambient moisture and pressure during normal operation.

The impact limiter and attachments are designed to withstand the impact loads and to prevent separation of the limiters from the cask during an impact. The design of the impact limiters and attachments is specified in Appendix 2.13.12.

2.1.1.3 TN-LC Cask Basket Assemblies The TN-LC cask will accommodate several different basket assemblies each containing a unique payload. The TN-LC-NRUX and TN-LC-MTR baskets consist of a base basket assembly that holds fuel buckets which, in turn, support the spent fuel assemblies. The TN-LC-TRIGA basket is comprised of a stack of basket segments that support the fuel assemblies/elements. The TN-LC-1FA basket has three modes of use. The base TN-LC-1FA basket may be loaded with a single PWR fuel assembly or, with the addition of an internal sleeve, a single BWR fuel assembly. LWR fuel pins may be loaded into the pin can which is placed within the base TN-LC-1FA basket. A detailed description of each basket configuration is provided in Chapter 1, Appendices 1.4.2 through 1.4.5.

The details of each basket are shown on drawings provided in Appendix 1.4.1.

2.1.1.4 TN-LC-NRUX Basket Assembly The TN-LC-NRUX basket is designed to accommodate up to 26 fuel assemblies. The TN-LC-NRUX basket structure consists of two removable subassemblies. Those subassemblies are comprised of 13 stainless steel tubes welded together, wrapped in a stainless steel plate and capped on the bottom with a perforated stainless steel plate. The two basket tube subassemblies are centered in the basket assembly with a series of guide plates and guide plate supports. Two aluminum basket assembly tube caps are used to confine the contents of each tube and to limit axial motion of the fuel and tube subassemblies with respect to the cask.

The TN-LC-NRUX basket tube assembly bottom covers and tube caps, combined with top and bottom cask spacers, are designed to transmit longitudinal fuel assembly loads to the cask body.

The fuel assemblies are supported laterally by the stainless steel tubes, tube wraps, guide plates, guide plate supports, and basket assembly tube. The complete basket assembly is supported laterally by the cask shell. The basket structure is oriented parallel to the axis of the cask and establishes and maintains fuel assembly orientation.

2.1.1.5 TN-LC-MTR Basket Assembly The TN-LC-MTR basket is designed to accommodate up to 54 MTR fuel elements. The TN-LC-MTR basket is designed to accommodate eighteen fuel buckets. The fuel buckets are fabricated from stainless steel plates that are welded together to form three compartments and a perforated bottom plate. Three stacks of six fuel buckets each are centered in the basket assembly with two stainless steel divider plates and four thick outer plates. The thick outer plates have aluminum rails bolted onto the outside which provide the transition to the cask inside shell.

TN-LC-0100 2-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The basket structure combined with the fuel buckets and top caps is designed to transfer longitudinal fuel assembly and bucket loads to the cask body. The fuel assemblies and buckets are supported laterally by the stainless steel outer plates and aluminum rails. The basket assembly is supported laterally by the cask shell. The basket structure and fuel buckets are oriented parallel to the axis of the cask and establish and maintain fuel orientation.

2.1.1.6 TN-LC-TRIGA Basket Assembly The TN-LC-TRIGA basket structure is comprised of five stacked TN-LC-TRIGA basket segments and a cask spacer to limit axial cask to basket gaps.

The TN-LC cask is designed to contain and support the TN-LC-TRIGA basket assembly, cask spacer, and up to 180 TRIGA fuel assemblies/elements.

Each of the basket structures consist of a welded assembly of stainless steel tubes (fuel compartments) separated by poison plates and surrounded by a larger stainless steel box or wrap and aluminum support rails.

Longitudinal fuel element loads are applied through the stack of baskets and then to the cask body. The fuel elements are supported laterally by the stainless steel structural boxes. The baskets are supported laterally by the aluminum basket rails. The basket rails are oriented parallel to the axis of the cask and are attached to the periphery of the basket to provide support as well as to establish and maintain basket orientation.

2.1.1.7 TN-LC-1FA Basket Assembly The TN-LC-1FA basket is comprised of the basket assembly, a fuel pin tube can (when loaded with loose fuel pins), basket cell spacers, and cask spacers to limit axial cask-to-basket gaps.

The TN-LC cask is designed to contain the TN-LC-1FA basket assembly, end caps, one fuel assembly or fuel pin tube can with up to 21 fuel pins, and spacers while remaining completely supported by the transport cask.

The basket structure consists of a thick square-shaped welded or bolted tube assembly surrounded by poison plates and solid aluminum support rails.

The basket structure is open at each end. Therefore, longitudinal fuel assembly or fuel pin can loads are applied directly to the cask body and not the fuel basket structure. The fuel assembly or fuel pin can is supported laterally by the stainless steel tube assembly. The basket is supported laterally by the basket rails and the cask shell. The solid aluminum basket rails are oriented parallel to the axis of the cask and are attached to the periphery of the basket to provide support and to establish and maintain basket orientation.

A hold down ring is designed for BWR fuel assembly loading to provide lateral clearance for a fuel grapple, if necessary. After fuel loading, the hold down ring is installed to provide continuous transfer of basket loads to the cask.

The pin can is a welded 5x5 square array of stainless steel 1 in. tubes which are wrapped in a stainless steel plate. The top of the pin can has a bolted closure lid. The closure lid has threads to attach a lifting interface to allow handling of the pin can.

TN-LC-0100 2-4

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 2.1.2 Design Criteria The packaging consists of the following major components:

Cask Body Impact Limiters The structural design criteria for these components are described below.

2.1.2.1 Basic Design Criteria Cask Containment Vessel The containment vessel consists of the inner shell including the flange inside of the lid inner O-ring, the bottom flange, the bottom plug, and the lid. The lid and bottom plug bolts and inner O-ring are also part of the containment vessel as are the drain and vent port plug bolts and seals.

The containment vessel is designed to the maximum practical extent as an ASME Class I component in accordance with the rules of the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB [3]. The Subsection NB rules for materials, design, fabrication and examination are applied to all of the above components to the maximum practical extent. In addition, the design meets the requirements of Regulatory Guides 7.6 [5] and 7.8 [6]. Alternatives to the ASME Code are discussed in Section 2.1.4 and Appendix 2.13.13 of this Chapter.

The acceptability of the containment vessel under the applied loads is based on the following criteria:

Title 10, Code of Federal Regulations, Part 71 Regulatory Guide 7.6 Design Criteria ASME Code Design Stress Allowables Preclusion of Fatigue Failure Preclusion of Brittle Fracture The stresses due to each load are categorized as to the type of stress induced, such as membrane or bending, and the classification of stress, such as primary or secondary. Stress limits for containment vessel components, other than bolts, for NCT (ASME Level A) and HAC (ASME Level D) are given in Table 2-1.

The primary membrane stress and primary membrane plus bending stress are limited to Sm (Sm is the Code allowable stress intensity) and 1.5 Sm, respectively, at any location in the cask for NCT (ASME Level A).

The HAC events are evaluated as short duration, Level D conditions. The stress criteria are taken from Section III, Appendix F of the ASME Code [3]. For elastic quasi-static analysis, the TN-LC-0100 2-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 2-10 Weights and Center of Gravity Locations for TN-LC Cask and Its Major Components Basket Basket Total Total Assembly C.G. Assembly Assembly Cask & Basket Configuration(1)

Weight Weight C.G.

(lbs) (inch) (lbs) (inch)

TN-LC Cask with 1FA Basket Assembly PWR Fuel 6,061 90.75 50,002 98.48 BWR Fuel 6,840 88.90 50,781 98.12 Pin Can (Options 1 and 2) 6,990 96.34 50,931 99.11 Pin Can (Option 3) 7,075 96.03 51,016 99.06 TN-LC Cask with MTR Basket Assembly 30-70-S 6,269 92.26 50,210 98.64 30-70-M 5,931 92.17 49,872 98.67 30-70-L 5,593 92.07 49,534 98.71 TN-LC Cask with NRUX Basket Assembly NRUX 3,566 99.70 47,507 99.56 TN-LC Cask with TRIGA Basket Assembly Option 1 5,150 95.95 49,091 99.17 Option 2 5,342 109.01 49,283 100.58 MIN 47,507 98.12 MAX 51,016 100.58 Notes:

1. Maximum Cask Weight 43,941 lbs, Cask C.G. location 99.55 inch.

TN-LC-0100 2-49

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on Pages 2.13.8-i through 2.13.8-iii, and 2.13.8-1 Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 2.13.8-i

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 2.13.8-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on Pages 2.13.8-8 and 2.13.8-9 Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 2.13.8-8

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on Pages 2.13.8-11 through 2.13.8-13 Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 2.13.8-11

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 2.13.8-48

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on Pages 2.13.8-58 through 2.13.8-60 Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 2.13.8-58

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on Pages 2.13.8-65 through 2.13.8-68 Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 2.13.8-65

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 2.13.10-1 Dimensions and Average Temperatures Used in Calculating Fuel Assembly Axial Thermal Expansion in the TN-LC Transport Cask TN-LC Basket in LIC,FCav,Cold, Tavg,FA, Cladding Operation Condition Transport Cask inches F Materials NRUX 115.70 196 MTR-S 27.42 245 MTR-M 33.25 245 Aluminum (1)

MTR-L 42.00 245 Hot NCT TRIGA (Option 1) 31.00 250 (100°F Ambient) TRIGA (Option 2) 48.05 250 1FA (BWR) 182.12 (2) 396 1FA (PWR) 182.12 (2) 401 1FA (Pin Can Options 1 Zircaloy-4 and 2) 180.24 456 1FA (Pin Can Option 3) 169.55 456 1FA (Pin Can Options 1 Cold NCT and 2) 180.24 335

(-40°F Ambient) 1FA (Pin Can Option 3) 169.55 335 Notes:

1. [ ]

2. Conservatively considering TN-LC Unit #1 as-built cavity length.

Table 2.13.10-2 Dimensions and Average Temperatures Used in Calculating Basket Assembly Axial Thermal Expansion in the TN-LC Transport Cask TN-LC Basket in LBasket,Cold, LIC,TC,Cold, Tavg,Basket, Tavg,TC,Shell, Operation Condition Transport Cask in. in. F F NRUX 181.5 181.75 199 MTR 182.0 182.5 256 Hot NCT TRIGA 179.5 182.5 255 n/a 1FA (BWR) 181.5 182.12 (3) 298 1FA (PWR) 181.5 182. 12 (3) 278 1FA 181.87 182. 12 (3) 283 (1) 210.5 (2)

Cold NCT 1FA 181.87 182. 12 (3) 185 (1) n/a Notes:

1. Based on the average value among pin can, cask top and bottom maximum temperatures from Chapter 3.

2. Average temperature based on TN-LC transport cask temperature plot in Chapter 3. Thermal expansion for the transport cask is considered for this basket considering the higher heat load and the lower initial clearance.

3. Conservatively considering TN-LC Unit #1 as-built cavity length.

TN-LC-0100 2.13.10-12

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 2.13.10-3 Dimensions and Average Temperatures Used in Calculating Basket Rail Axial Thermal Expansion in the TN-LC Transport Cask TN-LC Basket in Operation Condition LRail,Cold, LIC,Rail,Cold, Tavg,Rail, Transport Cask In. In. °F MTR 176.5 177 220 TRIGA 47.3 1 48.3 231 Hot NCT 1FA (BWR) 180.5 182. 12 (3) 261 1FA (PWR) 180.5 182. 12 (3) 268 1FA (Pin Can) 180.5 182. 12 (3) 268 Cold NCT 1FA (Pin Can) 180.5 182. 12 (3) 250 2 Notes:

1. At least 1 in. space among basket rail segments for thermal growth based on Chapter 1 drawings.

2. Based on the bounding maximum basket temperature for cold NCT.

3. Conservatively considering TN-LC Unit #1 as-built cavity length.

Table 2.13.10-4 Dimensions and Average Temperatures Used in Calculating Basket Assembly Radial Thermal Expansion in the TN-LC Transport Cask Operation TN-LC Basket IDTC,Cold ODBasket,Cold WSS,Basket,Cold WRail,Cold Tavg,Basket Tavg,Rail Condition in Transport Cask In. In. In. In. o F o F

NRUX 18 17.55 n/a n/a 191 n/a MTR 18 17.55 12.94 2.305 256 220 TRIGA 18 17.55 12.12 2.715 255 231 Hot NCT 1FA (BWR) 18 17.55 11.375 3.0875 267 261 1FA (PWR) 18 17.55 11.375 3.0875 278 268 1FA (Pin Can) 18 17.55 11.375 3.0875 277 268 Cold NCT 1FA (Pin Can) 18 17.55 11.375 3.0875 250 1 250 1 Notes:

1. Based on the bounding maximum basket temperature for cold NCT.

TN-LC-0100 2.13.10-13

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 2.13.10-7 Maximum Allowable Irradiated FA Length in a Fuel Basket Operating Tavg,FA LIC,FCan,Cold LFA,Cold,Max TN-LC Basket Type Condition °F In in NRUX 196 (1) 115.70 115.51 MTR-S 245 (1) 27.42 27.36 MTR-M 245 (1) 33.25 33.17 MTR-L 245 (1) 42.00 41.90 TRIGA (Option 1) 250 (1) 31.00 30.93 TRIGA (Option 2) 250 (1) 48.05 47.94 Hot NCT 1FA (BWR) 396 (2) 182.12 181.69 (100ºF Ambient) 1FA (PWR) 401 (2) 182.12 181.68 1FA (Pin Can Options 1 and 2) 456 (2) 180.24 179.73 1FA (Pin Can Option 3) 456 (2) 169.55 169.08 1FA (Pin Can Options 1 Cold NCT and 2) 335 (2) 180.24 179.89

(-40ºF Ambient) 1FA (Pin Can Option 3) 335 (2) 169.55 169.23 Notes:

1. Fuel cladding material based on aluminum cladding.

2. Zircaloy assumed for 144 active fuel region and SA-240 type 304 for fuel assembly components excluding active fuel region.

Table 2.13.10-8 Minimum Axial Clearance for the Basket in Transport Cask Operating SS,Basket TC LBasket,Hot LIC,TC,Hot Bask_Axi_Hot_Gap TN-LC Basket Type Condition °F-1 °F-1 In. in. in.

NRUX 8.90E-06 n/a 181.71 181.75 0.04 MTR 9.11E-06 n/a 182.31 182.50 0.19 TRIGA 9.11E-06 n/a 182.31 182.50 0.19 Hot NCT 1FA (BWR) 9.20E-06 n/a 181.88 182.12 0.24 1FA (PWR) 9.16E-06 n/a 181.85 182.12 0.27 1FA (Pin Can) 9.17E-06 8.52E-06 182.23 182.34 0.11 Cold NCT 1FA (Pin Can) 8.97E-06 n/a 182.06 182.12 0.06 TN-LC-0100 2.13.10-15

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 2.13.10-9 Minimum Axial Clearance for the Basket Rail in TN-LC Transport Cask Operating Al,Rail LRail,Hot LIC,Rail,Hot Rail_Axi_Hot_Gap TN-LC Basket Type Condition °F-1 in. in. in.

MTR 1.30E-05 176.85 177.00 0.15 TRIGA 1.31E-05 47.40 48.30 0.90 NCT 1FA (BWR) 1.31E-05 180.95 182.12 1.17 1FA (PWR) 1.32E-05 180.97 182.12 1.15 1FA (Pin Can) 1.32E-05 180.97 182.12 1.15 Cold NCT 1FA (Pin Can) 1.24E-05 180.93 182.12 1.19 Table 2.13.10-10 Minimum Radial Clearance for the Basket in the TN-LC Transport Cask Operating TN-LC Basket SS,Basket Al,Rail ODBasket,Hot IDTC,Hot Bask_Rad_Hot_Gap Condition Type 1/°F 1/°F in. in. in.

NRUX 8.88E-06 1.28E-05 17.569 18.000 0.216 MTR 9.11E-06 1.30E-05 17.581 18.000 0.210 TRIGA 9.11E-06 1.31E-05 17.582 18.000 0.209 Hot NCT 1FA (BWR) 9.13E-06 1.31E-05 17.586 18.000 0.207 1FA (PWR) 9.16E-06 1.32E-05 17.588 18.000 0.206 1FA (Pin Can) 9.15E-06 1.32E-05 17.588 18.000 0.206 Cold NCT 1FA (Pin Can) 9.10E-06 1.31E-05 17.583 18.000 0.208 Table 2.13.10-11 Minimum Clearance for the Poison Plate Thermal Expansion in a Fuel Basket Hot NCT Cold NCT WWrap,Cold, in. 11.875 11.875 WPoison,Cold, in. 11.845 11.845 Tavg,Wrap, °F n/a n/a Tavg,Poison, °F 255 250 SS,Wrap, 1/°F n/a n/a Al,Poison, 1/°F 1.31E-05 1.31E-05 WWrap,Hot, in. 11.875 11.875 WPoison,Hot, in. 11.874 11.873 Poison_Hot_Gap, in. 0.001 0.002 TN-LC-0100 2.13.10-16

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 2.13.10-12 Minimum Axial Clearance for the Poison Plate Thermal Expansion in a Fuel Basket Basket Type TRIGA (Option 1) TRIGA (Option 2) 1FA Insert_End,Cold, in. 28.76 45.81 181 LPoison,Cold, in. 28.31 45.36 180.5 Tavg,Poison, °F 255 255 275 Poison, 1/°F 1.31E-05 1.31E-05 1.32E-05 LInsert_End,Hot, in. 28.760 45.810 181.000 LPoison,Hot, in. 28.379 45.470 180.988 Poison_Axi_Hot_Gap, in. 0.381 0.340 0.012 Table 2.13.10-13 Maximum Length of Fuel Assemblies, Rods and Fuel Elements for TN-LC Transport Cask TN-LC Basket Type LFA,Cold,Max Operating LIC,FCan,Cold Condition in. in.

NRUX 115.70 115.51 MTR-S 27.42 27.36 MTR-M 33.25 33.17 MTR-L 42.00 41.90 TRIGA (Option 1) 31.00 30.93 TRIGA (Option 2) 48.05 47.94 Hot NCT 1FA (BWR) 182.50 182.06 (100°F Ambient) 1FA (PWR) 182.50 182.06 1FA (Pin Can Options 1 and 2) 180.24 179.73 1FA (Pin Can Option 3) 169.55 169.08 1FA (Pin Can Options 1 Cold NCT and 2) 180.24 179.89

(-40°F Ambient) 1FA (Pin Can Option 3) 169.55 169.23 TN-LC-0100 2.13.10-17

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Appendix 2.13.12 TN-LC Transport Package Impact Limiter Analysis Using LS-DYNA NOTE: References in this Appendix are shown as [1], [2], etc. and refer to the reference list in Section 2.13.12.7.

2.13.12.1 Description of Impact Limiters and Load Path Each of the impact limiters consists of 2 energy absorbing materials: (1) redwood and (2) balsa wood. The front and rear impact limiters are identical. Each has an outside diameter of 66.0 inches and a height of 27.8 inches. The inner and outer shells are Type 304 stainless steel joined by radial gussets of the same material.

The materials and grain orientations are selected to reduce the deceleration to prevent excessively high stresses in the cask during the impact. A 1.5 inch layer of balsa wood with grain parallel to the end of the cylindrical cask is provided on the top face of the impact limiter to minimize decelerations during a one foot end drop.

During the end drop, all of the wood in the central part of the impact limiter that is directly backed-up by the cask body will crush. The wood in the corner and side of the limiter will tend to slide around the side of the cask since it is not supported or backed-up by the body. The central part of the impact limiter consists of balsa wood with grain perpendicular to the axial direction, redwood with grain oriented axially, and balsa wood with grain oriented axially going radially outwards from the center.

A ring of balsa wood (consisting of 8 segments) is located in the sides of the pie-shape compartments which surround the end of the cylindrical surface of the cask with grain direction oriented radially. This ring of wood absorbs most of the kinetic energy during side drop and slap down events.

The corners of the pie-shaped compartments are filled with balsa wood with grain oriented axially. The primary function of this region is energy absorption during a 30 foot corner drop.

Each impact limiter is attached to the cask by eight attachment bolts. The attachment bolts have been sized to withstand the loads transmitted during a low angle slap down. This analysis is described in Section 2.13.12.6.

TN-LC-0100 2.13.12-1

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

[

]

The crush strength data for redwood is presented in Table 2.13.12-6. The stress-strain relationship for redwood is also shown in Figure 2.13.12-22, where they are also compared to the material properties used in the benchmark analysis.

Bolt and Alignment Tubes There are 8 bolts that attach each impact limiter to the cask model. The following elastic, linearly plastic material properties are used for the bolts.

SA-540 GR. B23 or B24 CL. 1 (at room temperature and -40 °F)

= 0.29 / 386.4 = 7.505 x 10-4 lb-s2/in.4 E = 27.8 x 106 psi [8]

= 0.3 [9]

Sy = 150.0 ksi [8]

Tangent Modulus, ET = 2% E = 5.54 x 105 psi Bolts are modeled as circular cross section beams with bolts located at the plane of symmetry having half the moment of inertia and volume of a full bolt. The springs representing the sections of bolts within the impact limiter were given conservatively high spring rates in tension TN-LC-0100 2.13.12-8d

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 bolt is obtained by compiling the strain value of the highest strained integration point for each beam element and then taking the maximum of each beam element within the bolt.

The peak bolt axial and shear stresses for each bolt for the 5º Slap Down with Soft wood properties are shown in Table 2.13.12-18.

Tables 2.13.12-17 and 2.13.12-18 indicate that none of the impact limiter bolts have combined axial and shear stress ratios that exceed 1.0. ASTM [11] indicates that for material SA-540 GR.

B23 or B24 CL. 1, the minimum elongation is 0.1. Based on the average cross-sectional strain, only a single bolt on each impact limiter is predicted to fail. Therefore, it is concluded that the impact limiters will remain connected to the cask during and after the postulated 30 drop.

Accelerations for the detailed bolt model runs are shown in Tables 2.13.12-19 and 2.13.12-20 and closely match the original results.

2.13.12.6.2 Minimum Engagement Length for Attachment Bolt and Block The minimum engagement length Le for the bolt and block is ([10], p. 1490):

2A t Le 1

3.1416 K n max .57735n E s min K n max 2

For a 1 -8UNC 2A bolt:

At = tensile stress area = 0.606 in.2; n = number of threads per inch = 8; Knmax = maximum minor diameter of internal threads = 0.890 in. ([10], p. 1728);

Esmin = minimum pitch diameter of external threads = 0.9067 in. ([10], p. 1728).

Substituting the values given above:

2 0.606 Le 0.751in .

1 3.1416 0.890 .57735 8 0.9067 0.890 2

The required engagement length Q is ([10], p. 1491):

Q J Le Where J is equal to ([10], p. 1490):

As Su bolt J

An Su block TN-LC-0100 2.13.12-14

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Chapter 3 Thermal Evaluation NOTE: References in this Chapter are shown as [1], [2], etc. and refer to the reference list in Section 3.5.

This chapter presents the thermal evaluations which demonstrate that the TN-LC transport cask meets thermal requirements of 10CFR71 [1] for transportation of commercial or research reactor spent fuel as described in Chapter 1, Section 1.2.2. The thermal analysis of the TN-LC package considers transportation within and without an ISO container to evalaue the worst conditions during NCT and HAC.

The maximum heat load per shipment allowed for transportation in TN-LC transport cask varies for the different basket types from 0.39 kW to 3.0 kW. The table below summarizes the maximum heat load per shipment for transportation.

FA Types and Decay Heat Loads for NCT FA Type Heat Load Heat Load (kW) (Btu/hr)

PWR 3.00 10237 BWR 2.00 6825 MTR 1.50 5118 TRIGA 1.50 5118 NRU/NRX 0.39 1331 PWR/BWR/EPR/MOX Fuel Pins 2.50 8531 Thermal performance of the TN-LC transport cask is evaluated based on finite element analyses using ANSYS computer code [9].

This evaluation demonstrates that packaging component temperatures are within material temperature limits and fuel cladding temperatures meet the thermal requirements of ISG-11 [2],

where applicable.

3.1 Description of Thermal Design Criteria The TN-LC transport cask is designed to passively reject decay heat under Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) while maintaining packaging temperatures and pressures within specified limits. Objectives of the thermal analyses performed for this evaluation include:

(a) Determination of maximum component temperatures with respect to cask materials limits to ensure components perform their intended safety functions, (b) Determination of temperature distributions to support the calculation of thermal stresses, (c) Determination of the cask cavity gas temperature to support containment pressure calculations, and TN-LC-0100 3-1

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The main design features of the fuel baskets are described in the sections that follow.

3.1.1.2 TN-LC-NRUX Basket The structure of the TN-LC-NRUX basket is described in Appendix 1.4.2. The stainless steel tubes provide the necessary heat conduction path from the fuel assemblies to the perimeter of tube subassemblies. The guid plate supports provide additional conduction path between the guide plates and the basket shell. For conservatism, the guide plate supports are not included in the 2D TN-LC-NRUX basket model. Radiation heat transfer between spaces among tubes, wrap plates, guide plates, basket shell and the cask inner shell are considered in the TN-LC-NRUX basket model. The bounding NRX effective fuel conductivity is used in the model.

3.1.1.3 TN-LC-MTR Basket The structure of the TN-LC-MTR basket is described in Appendix 1.4.3. The bucket assemblies, in combination with the basket plates and solid aluminum rails, provide the necessary heat conduction path from the fuel assemblies to the perimeter of the TN-LC cask. Radiation between basket rail and the cask inner shell gap is considered in the TN-LC-MTR basket model.

3.1.1.4 TN-LC-TRIGA Basket The structure of the TN-LC-TRIGA basket is described in Appendix 1.4.4. The heat conduction path from the fuel assemblies to the basket rails is provided by the fuel compartment assemblies and the poison plates, which are sandwiched between the fuel compartments. The heat conduction path from the fuel compartment assemblies to the perimeter of the cask inner shell is provided by aluminum rails. Radiation between basket rail and the cask inner shell is considered in the TN-LC-TRIGA basket model.

3.1.1.5 TN-LC-1FA Basket The structure of the 1FA basket is described in Appendix 1.4.5. The heat conduction path from the fuel assembly to the perimeter of the TN-LC cask inner shell is provided by the fuel compartment in combination with poison plates around the fuel compartment, the BWR sleeve (only for BWR fuel assemblies), and the aluminum basket rails assembly. The heat conduction path from the fuel rods to the perimeter of the pin can assembly is provided by the tubes containing the fuel rods and the pin can side wall. Radiation heat transfer between the basket rails and the cask inner shell space is considered in the TN-LC-1FA basket model with one PWR fuel assembly. Radiation heat transfer between the sleeve and basket frame is also considered in the TN-LC-1FA basket model with one BWR fuel assembly. Radiation heat transfer between the sleeve and the pin can is added in the TN-LC-1FA basket model with the pin can assembly.

The bounding effective fuel conductivity for the PWR, BWR, and pin can assembly are used in the model.

TN-LC-0100 3-4

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 3.1.2 Contents Decay Heat The design basis decay heat loading for the irradiated fuel to be transported within the TN-LC cask is a function of the irradiation history and the cooling time since discharge. Chapter 1, Section 1.2.2, provides details of the fuel elements to be transported. For the purposes of this evaluation, the design basis decay heat loadings are as shown the in the following table.

Heat Load per Maximum No. of Elements/

Basket Type Element/Assembly Allowable Heat Assemblies in Basket (W) Load (kW)

TN-LC-NRUX 26 NRU/NRX Assemblies 15/Assembly 0.39 TN-LC-MTR 54 Elements 30/Element 1.50 TN-LC-TRIGA 180 Elements 8.33/Elements 1.50 TN-LC-1FA 1 BWR Assembly 2000/Assembly 2.00 (BWR)

TN-LC-1FA 1 PWR Assembly 3000/Assembly 3.00 (PWR)

TN-LC-1FA 9 Fuel Pins (1) 200/Pin 1.80 (2)

TN-LC-1FA 21 Fuel Pins 120/Pin 2.50 Notes

1. The thermal evaluation considers 13 fuel pins with 220 W per pin, which conservatively bounds the above 9 fuel pins configuration.

2. The thermal evaluation considers 25 fuel pins with 120 W per pin, which conservatively bounds the above 21 fuel pins configuration.

The TN-LC transport cask is designed to transport low burnup fuel assemblies/fuel pins with burnup 45 GWd/MTU and high burnup fuel assemblies/fuel pins with burnup > 45 GWd/MTU.

Since each of the fuel pins loaded are encapsulated in individual tubes within the TN-LC-1FA basket, the physical configuration of the fuel pins is assured during transportation and the thermal evaluations presented for NCT in Section 3.3 and for HAC in Section 3.4 remain valid along with the confirmatory evaluations presented in Section 3.6.8 during transportation for NCT and HAC.

The physical configuration of the PWR/BWR fuel assemblies in the TN-LC-1FA basket can potentially alter during transportation thereby altering the heat transfer distribution within the cask cavity. To bound the uncertainties in the physical configuration of the fuel assemblies due to the paucity of the structural properties of the high burnup fuel assemblies for NCT and HAC, thermal evaluations are performed to bound both possible configurations. For the case that the physical configuration of the fuel assemblies is not altered (for both low burnup and high burnup fuel assemblies), the thermal evaluations presented for NCT in Section 3.3 and for HAC in Section 3.4 remain valid along with the confirmatory evaluations presented in Section 3.6.8. For the case that the physical configuration of the high burnup fuel assemblies may not be guaranteed, evaluations are presented in Section 3.6.9 to provide assurance that the containment of the TN-LC TC is maintained during transportation for NCT and HAC.

TN-LC-0100 3-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Fuel Effective Axial Thermal Conductivities, Effective Fuel Density and Specific Heat used for Transient Analysis of TN-LC Cask with TN-LC-1FA Pin Can Basket Tavg keff_axial Cp

(°F) (Btu/hr-in-°F) (lbm/in3) (Btu/lbm-oF) 200 0.1636 300 0.1725 400 0.1829 0.0895 0.093 500 0.1914 600 0.1986 800 0.2160

16. Effective Conductivity for Top and Bottom Gamma Shielding (See Section 3.3.1.3 for calculation of effective properties)

Temperature kplate kair keff_axial

(°F) (K) (Btu/hr-in.-°F) (W/m-K) (Btu/hr-in.-°F) (Btu/hr-in.-°F)

-100 200.0 1.767 0.0182 0.0009 0.028

-10 250.0 1.733 0.0223 0.0011 0.035 80 300.0 1.700 0.0261 0.0013 0.040 260 400.0 1.637 0.0330 0.0016 0.051 440 500.0 1.579 0.0395 0.0019 0.060 620 600.0 1.512 0.0456 0.0022 0.069 Temperature kplate kair keff_radial

(°F) (K) (Btu/hr-in.-°F) (W/m-K) (Btu/hr-in.-°F) (Btu/hr-in.-°F)

-100 200.0 1.767 0.0182 0.0009 1.713

-10 250.0 1.733 0.0223 0.0011 1.681 80 300.0 1.700 0.0261 0.0013 1.649 260 400.0 1.637 0.0330 0.0016 1.587 440 500.0 1.579 0.0395 0.0019 1.531 620 600.0 1.512 0.0456 0.0022 1.466

17. Effective Conductivity for Air in ISO Container Region 3 (See Section 3.3.1.3 for calculation of effective properties)

Ti To T avg T avg k µ Cp Pr Ra b Nu k_eff k_eff o o o 3 o

( F) ( F) ( F) (K) (W/m-K) (1/K) (kg/m-s) (kg/m ) (J/kg-K) (---) (---) (---) (W/m-K) (Btu/hr-in- F) 130 100 115 319 0.0275 3.13E-03 1.94E-05 1.11E+00 1008 0.71 4.97E+08 29.99 0.82 0.0397 140 110 125 325 0.0279 3.08E-03 1.97E-05 1.09E+00 1008 0.71 4.60E+08 29.41 0.82 0.0395 150 120 135 330 0.0283 3.03E-03 1.99E-05 1.07E+00 1009 0.71 4.26E+08 28.84 0.81 0.0392 160 130 145 336 0.0287 2.98E-03 2.02E-05 1.05E+00 1009 0.71 3.94E+08 28.29 0.81 0.0390 170 140 155 341 0.0290 2.93E-03 2.04E-05 1.03E+00 1010 0.71 3.66E+08 27.77 0.81 0.0388 180 150 165 347 0.0294 2.88E-03 2.07E-05 1.02E+00 1010 0.71 3.40E+08 27.26 0.80 0.0386 190 160 175 353 0.0298 2.84E-03 2.09E-05 1.00E+00 1011 0.71 3.17E+08 26.77 0.80 0.0384 200 170 185 358 0.0302 2.79E-03 2.12E-05 9.86E-01 1011 0.71 2.95E+08 26.30 0.79 0.0383 TN-LC-0100 3-15

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The planar thermal conductivity of the gap, which is perpendicular to the gap height, is assumed to be that of air. Heat capacity for this region is conservatively ignored. Further, radiation between the inner surface of the shield plate and the outer surface of the cask is modeled using the AUX12 processor. The effective conductivity values across the gap between the thermal shield plate and cask outer shell are listed in Section 3.2.1, material 21.

3.3.1.4 TN-LC Fuel Basket Model The following assumptions and conservatism are considered for the fuel basket model:

Commercial fuel assemblies (PWR, BWR or Fuel Pins) loaded in the TN-LC-1FA basket shall have a calculated maximum fuel cladding temperature in accordance with the guidance in ISG-11, Rev. 3, [2]. For NCT, the cladding temperature shall not exceed 400°C (752°F).

Research reactor fuel assemblies with aluminum cladding loaded in the TN-LC-NRUX/MTR/TRIGA basket shall have a maximum calculated cladding temperature less than 204°C (400°F) during NCT as specified in Section 3.1. This criterion is conservatively established to ensure the integrity of the aluminum cladding for NCT.

No convection is considered within the basket models.

The maximum inner shell temperatures resulting from the TN-LC transport cask model within an ISO container described in Section 3.3.1.1 are conservatively applied as uniform temperature boundary conditions in the fuel basket models.

The nominal cold radial gap of 0.25 in. between basket rail/shell and the cask inner shell is assumed conservatively for the hot conditions in the fuel basket models.

For the TN-LC-MTR basket model, a helium gap of 0.01 in. is considered between the basket rail and the outer plate to calculate effective conductivities of the outer plate in the cross section of the basket.

For the TN-LC-TRIGA basket model, 0.01 in. gaps on either side of the poison plates and wrap plates, and between any two adjacent plates are considered to calculate effective conductivities for these components in the cross section of the basket.

For the TN-LC-1FA basket models (including 1 PWR, 1 BWR and Pin Can Types), 0.01 in. gaps on either side of the poison plates and between the basket rail and the frame plate are considered to calculate effective conductivities for these components in the cross-section of the basket.

The gaps between adjacent components are related only to the flatness and roughness tolerances of the plates. The micro gaps related to these tolerances are non-uniform and provide interference contact at some areas and gaps on the other areas as shown schematically in Figure 3-10.

For the purpose of thermal evaluation, surfaces of intermittent contact between adjacent components are conservatively modeled as a uniform gap of 0.01 in. Based on the MP197 SAR, Appendix A, Section A.3.6.7.4 [21], the assumed gap size of 0.01 in. is approximately two times larger than the contact resistances between the adjacent components and is therefore conservative.

TN-LC-0100 3-30

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Heat loads considered for the thermal analysis of the TN-LC fuel baskets, i.e., TN-LC-NRUX, TN-LC-MTR, TN-LC-1FA (Pin Can), as listed in Table 3-14 are higher than the maximum allowable heat loads for the these baskets listed in Section 3.1.2. No axial heat transfer is considered through the basket to provide additional conservatism in calculation of the maximum fuel cladding and maximum component temperatures. This assumption includes additional conservatism in evaluation of the maximum temperature gradients through the basket.

For the TN-LC-1FA basket model, the fuel effective conductivity is selected based on the irradiated UO2 fuel conductivity. The small differences between the irradiated UO2, MOX, and EPR fuel conductivities have an insignificant impact on thermal evaluation of the TN-LC-1FA basket as shown in Appendix 3.6.7.

All other dimensions are based on nominal dimensions of TN-LC fuel baskets.

The thermal evaluation of the TN-LC transport cask determined that the maximum temperature of the cask inner shell during transportation with an ISO container is higher than that without an ISO container. The bounding maximum cask inner shell temperatures are rounded up to provide additional margin for conservatism. The resulting values are applied as uniform temperature boundary conditions to the fuel basket models. The uniform temperature boundary conditions used for the basket models are summarized in Table 3-15. The maximum cask inner shell temperature resulting for the TN-LC-MTR basket is used to evaluate TN-LC-TRIGA basket. Similarly, the maximum cask inner shell temperature resulting for TN-LC-IFA with PWR fuel assemblies is used to evaluate TN-LC-IFA with BWR or fuel pins. This approach is conservative as discussed in Section 3.3.1.1. The TN-LC-1FA pin can basket with 3 kW heat load shows the largest maximum basket temperature gradient among all fuel baskets during hot NCT. Therefore, to bound the maximum basket temperature gradients for the TN-LC transport cask during cold NCT, the TN-LC-1FA pin can basket for cold NCT with -40°F ambient and no insolation for 3 kW heat load is evaluated.

Decay heat load is applied as a uniform heat generation within a homogenized fuel region. The various heat loads used in this analysis are shown in the table below and are computed as follows:

Q q = x PF A x L Fuel Where:

q = uniform heat generation rate (Btu/hr-in.3),

Q = decay heat load per fuel element/assembly (Btu/hr),

A = area per fuel compartment (in.2),

LFuel = active fuel length per fuel element/assembly (in.),

PF = peaking factor.

TN-LC-0100 3-31

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Heat Generation in Fuel Basket Model Total Heat Active Fuel Length Heat Heat Load per Load Peaking per Generation Basket Type Element/Assembly Per Basket Factor Element/Assembly (Btu/hr-(W)

(kW) (in.) in.3)

TN-LC-NRUX 0.39 15/Assembly 1.0 96.0 0.135 TN-LC-MTR 1.62 30/Element 1.0 22.0 (1) 0.384 TN-LC-TRIGA 1.5 4x8.33/Element 1.0 14.0 0.671 TN-LC-1FA (BWR) 2.0 2000/Assembly 1.2 144.0 1.580 TN-LC-1FA (PWR) 3.0 3000/Assembly 1.1 144.0 0.993 (2)

TN-LC-1FA (Pin Can) 3.0 3000/Can 1.1 144.0 3.128 Notes:

1. The shortest active fuel length is considered to bound all design basis MTR fuels listed in Chapter 1.

2. The active fuel length of 144 in. is typical for PWR and BWR fule rods and bounds the active fuel length for MOX and EPR fuel rods.

The material properties used in the basket models are listed in Section 3.2.1. Effective conductivity for basket plates is calculated in Section 3.3.1.5.

Except for the contact gaps described above in this section, all the other gaps considered in the fuel basket model are based on cold nominal gaps. The geometry of the models and their mesh densities are shown in Figure 3-11 and Figure 3-12. Mesh sensitivities of the basket models are discussed in Appendix 3.6.4.

Typical boundary conditions for the fuel basket model are shown in Figure 3-13.

3.3.1.5 Effective Thermal Properties for Basket Components

1. Effective Conductivity for Basket Poison, Wrap and Outer Plates A helium gap of 0.01 in. is assumed between any two adjacent plates to account for contact resistance and fabrication imperfections between the plates.

The gaps in the plates build up serial thermal resistances through the thickness of the plates and parallel thermal resistances perpendicular to the thickness of the plates.

For conservatism in the calculation of the effective conductivity for the poison plates, the conductivity value of helium is based on the bounding value at 70°F (= 0.0072 Btu/hr-in.-°F). The conductivity value of the poison plate is based on 95 percent of the Boral core thermal conductivity at 500°F (=3.513 Btu/hr-in.-°F). No credit is taken for the conductivity through the aluminum cladding of the Boral plate.

The conductivities for the basket wrap (TN-LC-MTR basket) and outer plates (TN-LC-TRIGA basket) are taken from Section 3.2.1 based on stainless steel, SA-240 Type 304.

TN-LC-0100 3-32

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The resulting maximum temperatures for the fuel basket and fuel cladding, and maximum basket temperature gradient (Tbasket) based on TN-LC-1FA pin can basket for cold conditions (-40°F ambient and no insolation), are listed in Table 3-18.

Table 3-17 shows that the maximum basket temperature gradient (Tbasket) for hot NCT is 139°F for the TN-LC-1FA pin can basket and 56°F for other fuel baskets. However, the worst case condition for the bounding maximum basket temperature gradient is cold NCT with -40°F ambient. As can be seen from Table 3-18, the bounding maximum basket temperature gradient (Tbasket) during the worst case condition (cold NCT, -40°F ambient) is 181°F for the TN-LC-1FA pin can basket. The maximum basket temperature gradient difference between hot and cold NCT (181°F-139°F=42°F) for the TN-LC-1FA pin can basket combined with the bounding maximum basket temperature gradient (Tbasket) for other fuel baskets during the worst case condition (cold NCT, -40°F ambient) is calculated as 98°F (56°F+42°F). The conservative basket temperature gradient limit with additional margins (Tbasket, limit) used for fuel basket maximum stress evaluation is summarized in Table 3-5.

The minimum temperatures for fuel cladding and fuel basket components during NCT are bounded by a daily average ambient temperature of -40°F based on assuming no credit for decay heat for fuel contents in fuel baskets.

The maximum fuel cladding temperatures calculated for the TN-LC-NRUX/MTR/TRIGA basket in TN-LC transport cask for NCT and shown in Table 3-2, are lower than the allowable limit of 400°F. The maximum fuel cladding temperatures calculated for the TN-LC-1FA basket loaded with a PWR/BWR/pin can in the TN-LC transport cask for NCT are lower than the allowable limit of 752°F.

The minimum temperatures for fuel cladding and basket components are based on assuming no credit for decay heat for cold NCT and are summarized Table 3-3.

The maximum fuel cladding and basket component temperatures for cold NCT at -20°F ambient are bounded by those for hot NCT with 100°F shown in Table 3-2.

The maximum temperatures for fuel cladding and basket components for cold conditions at

-40°F ambient are summarized in Table 3-5. The maximum basket temperature gradient for cold NCT at -40°F ambient conditions shown in Table 3-5 bounds that for cold NCT at -20°F ambient conditions.

All materials can be subjected to a minimum environment temperature of -40°F (-40°C) without any adverse effects.

3.3.3 Maximum Normal Operating Pressure The maximum internal pressure for the TN-LC cask for NCT and HAC is determined based on the maximum allowable heat load of 3 kW and a maximum burnup of 70,000 MWD/MTU. The limiting fuel assembly type considered in this evaluation is the B&W 15x15 assembly.

TN-LC-0100 3-36

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 3.3.4.2 Dry Loading/Unloading During dry loading/unloading, the impact limiters are detached from the TN-LC cask and the cask body upper and lower segments beyond the neutron shield are open to environment for heat dissipation. In addition, the loading/unloading operations occur within a building which protects the cask from direct solar impact. Therefore, the ambient boundary conditions specified for NCT in Section 3.3 remain bounding for the dry loading/unloading operations.

The TN-LC transport cask model for NCT described in Section 3.3.1.1 does not include the baskets and applies the heat load as a uniform heat flux on the inner surface of the inner shell of the cask. Since this model does not include the basket and the helium backfill, the boundary conditions considered for the TN-LC transport cask model are bounding for dry loading/unloading operations and the resulting cask temperatures for NCT in Section 3.3.2 bound those for dry loading/unloading conditions.

Based on the above discussion, the cask inner shell temperatures resulting from the TN-LC transport cask model without the ISO container can be used conservatively to determine the maximum basket component and fuel cladding temperatures for dry loading/unloading conditions. The maximum cask inner shell temperatures for NCT conditions without the ISO container are 204°F and 166°F for 3 kW and 1.85 kW heat loads, respectively, as reported in Table 3-9. For conservatism, cask inner shell temperatures of 210°F and 170°F are used in this calculation for evaluation of the TN-LC-1FA pin can and TN-LC-MTR baskets under dry loading/unloading conditions.

The TN-LC-1FA fuel pin basket and TN-LC-MTR basket models described in Section 3.3.1.4 are used in this calculation to determine the bounding maximum basket component and fuel cladding temperatures for fuel pins and research reactor fuels under dry loading/unloading conditions. The properties of backfill gas in these models are changed from helium to air in order to simulate the dry loading/unloading conditions. The effective fuel conductivities in these basket models were calculated in Appendix 3.6.6 considering helium as backfill gas. For evaluation of the dry loading/unloading conditions, the effective conductivities of fuel pins and MTR fuel elements are recalculated considering air as backfill gas. The fuel assembly models described in Appendix 3.6.6 are used for recalculation of the effective fuel conductivities for dry loading/unloading conditions. The same methodology as described in Appendix 3.6.6 is used in this section to determine the effective fuel conductivities.

Steady state conditions are considered for dry loading/unloading conditions. Therefore, no time limits are applicable for these operations.

The following temperature limits are considered as design criteria for the evaluation of loading/unloading conditions:

• For commercial fuel assemblies (PWR, BWR or Fuel Pins) loaded in the TN-LC-1FA basket, the fuel cladding temperature is limited to 400°C (752°F). For short-term operations, such as vacuum drying, temperature differences greater than 65°C (117°F) are not permitted for repeated cycling of fuel cladding temperature during drying and backfilling operations in accordance with guidance provided by ISG-11, Rev 3 [2].

TN-LC-0100 3-43

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 this change in temperature occurs only once during loading/unloading steps, this change is not considered as repeated cycling of fuel cladding temperature and, therefore, the thermal cycling limit of 65°C (117°F) for short-term operations set by ISG-11 [2] is satisfied.

Typical temperature distributions for the TN-LC-MTR and TN-LC-1FA (pin can) baskets for dry loading/unloading conditions are shown in Figure 3-16.

As discussed above, the maximum cask component, basket component, and fuel cladding temperatures for wet loading operations are bounded by those calculated for NCT in Section 3.3.

Therefore, no additional thermal evaluation is needed for wet loading operations. The bounding wet unloading operation is the reflood of the cask cavity with water. For this operation, procedural controls assure that the cask will not be over pressurized. The maximum fuel cladding temperature during wet unloading operation remains bounded by the maximum fuel cladding temperature for NCT.

The maximum cask component temperatures for dry loading/unloading conditions are bounded by the NCT as discussed above.

The maximum fuel cladding and basket component temperatures resulting from the bounding cases discussed for dry loading/unloading conditions are listed in Table 3-19.

TN-LC-0100 3-46

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 maximum cask component temperatures at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the fire accident bound the maximum temperatures for the steady state conditions.

Transient runs for HAC were made with the TN-LC cask basket with a PWR fuel assembly and the basket-to-cask inner shell gap homogenized. The effective thermal properties calculation and results for the homogenized basket are shown in Appendix 3.6.5.

SOLID70 elements are used to model the TN-LC homogenized basket and gap region. The elements for other components are the same as those described in Section 3.3.1 for NCT.

Ambient conditions for HAC are based on 10CFR71 [1] requirements and are applied on the boundaries of the transport cask model. These conditions are listed in the table below.

Design Load Cases for HAC Ambient Duration Period Insolation temperature (°F) (hr)

Initial Conditions 100 Yes N/A Fire 1475 No 0.5 Wood Smoldering 100 Yes 0.5 Cool-Down 100 Yes N/A Insolation is applied as a heat flux over the transport cask outer surfaces using average insolation values from 10CFR71 [1]. The insolation values are averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and multiplied by the surface absorptivity factor to calculate the solar heat flux. The solar heat flux values used in the TN-LC transport cask model for HAC are summarized in the table below.

Solar Heat Flux for Cool-Down Period Insolation Total solar heat flux Surface Solar Shape over 12 hrs [1] averaged over 24 hrs Material Absorptivity (gcal/cm2) (Btu/hr-in.2)

All Curved 400 1.0 0.4267 materials Flat vertical 200 1.0 0.2133 Convection and radiation heat transfer from the transport cask outer surfaces are combined together as total heat transfer coefficients using the same methodology described in Section 3.3.1.

The highest peak temperature of the cask inner shell resulting from the transient TN-LC transport cask model was used as a steady-state, uniform boundary condition for a two-dimensional model of the TN-LC-1FA pin can basket. The 2D model of the TN-LC-1FA pin can basket is the same model described in Section 3.3.1.4 for evaluating NCT. A heat load of 3.0 kW is used in the model and the cask inner shell temperature and the heat generating rates are applied on the TN-LC-1FA pin can basket model for HAC using the same methodology as was used for NCT.

TN-LC-0100 3-51

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The transient thermal evaluation of the TN-LC transport cask showed that the highest peak temperature of the cask inner shell during HAC is 445°F, as shown in Table 3-6. For conservatism, a steady-state, uniform cask inner shell temperature of 450°F is applied as the boundary conditions on the nodes representing the inner surface of the cask inner shell in the TN-LC-1FA pin can basket model for HAC evaluation.

The TN-LC-1FA pin can basket with 3.0 kW heat load provides the highest maximum temperatures for the fuel cladding among all fuel baskets for NCT. Since a steady-state model of the basket is used to determine the maximum fuel cladding temperature, the same behavior is assumed for HAC. Therefore, the maximum fuel cladding temperature determined using the 2D model of the TN-LC-1FA pin can basket is the bounding cladding temperature for other fuel baskets loaded into the TN-LC cask.

3.4.3 Maximum Temperatures and Pressure Temperature distributions for the TN-LC transport cask under HAC are shown in Figure 3-20 and the time temperature histories are shown in Figure 3-21 through Figure 3-23. Typical temperature distribution of the fuel basket components for HAC is shown in Figure 3-24.

The maximum component temperatures for transient runs are listed in Table 3-6. The calculated maximum temperatures of the TN-LC cask components for HAC are lower than the allowable limits.

The seal O-rings are not explicitly considered in the models. The maximum seal temperatures are retrieved from the models by selecting the nodes at the locations of the corresponding seal O-rings.

The maximum long-term temperature of 275°F (135°C) calculated for the Viton fluorocarbon seals is for the top cavity port seal during steady state cool-down conditions.

As seen in Table 3-6, the maximum Viton fluorocarbon seal temperatures are below the long-term limit of 400°F (204°C), except for the top lid seal.

The length of the time intervals, in which the seal temperatures are above the long-term limit of 400°F, are extracted from the data based on the transient runs and listed in Table 3-22. The maximum short-term seal temperature is 449°F (232°C) for the top lid seal, which remains at most for one hour at an elevated temperature. This short-term temperature is well below the specified short-term limit of 482°F (250°C) for the Viton fluorocarbon seals.

The maximum temperature of gamma shielding (lead) is 558°F (292°C), which is well below the lead melting point of 621°F (327°C).

The resins and wood are assumed to be decomposed or charred after fire accident. Therefore, the maximum temperatures for these components are irrelevant for HAC.

The maximum temperatures for fuel cladding and fuel basket components for HAC are listed in Table 3-7. If the physical configuration of the fuel assemblies is not altered, the maximum fuel cladding temperature of 694°F (368°C) calculated for fuel pins in the TN-LC-1FA pin can TN-LC-0100 3-52

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 basket with 3.0 kW heat load is the bounding fuel cladding temperature for all other fuel types in the TN-LC-NRUX, MTR, TRIGA and 1FA baskets.

The maximum fuel cladding temperatures calculated for the TN-LC-1FA pin can basket in the TN-LC transport cask for HAC is lower than the allowable limit of 1,058°F (570°C) for LWR fuel. This bounding maximum calculated fuel cladding temperature of 694°F is lower than the lowest melting point of aluminum alloys 1100 and 6063 (1140°F/616°C [5]).

The maximum pressure in the cask cavity for HAC is calculated using the same methodology described in Section 3.3.3.

1. TN-LC Cask Cavity HAC Pressure The maximum cask cavity pressure in the TN-LC transport cask for HAC is calculated using the same methodology and assumptions as described for NCT in Section 3.3.3. The methodology and assumptions used are defined in Section 3.3.3.

The total amount of gas in the TN-LC cask cavity for HAC, nhe,HAC, is calculated as follows:

n he ,HAC = n he ,initial + n Free FA ,HAC + n CC HAC

= 9.6 + 16.52 + 2.24

= 28.36 g moles.

Where:

n he,initial = number of moles of helium fill gas in cask cavity = 9.6 g-moles (Section 3.3.3),

n Free FA ,HAC = number of moles of fuel rod fill/fission gas released for HAC = 16.52 g-moles (Table 3-11),

n CC HAC = number of moles of fill gas released from CCs for HAC = 2.24 g-moles (Table 3-12),

The maximum pressure in the cask cavity for HAC, PHAC, is calculated as:

4 psia o 1.4504

• 10 * (n he , HAC )

• R

• Tavg, he , HAC * (5 / 9 K / R )

Pa PHAC =

Vfree,cavity * (1.6387

• 10 5 m 3 / in.3 )

4 psia o o 1.4504

• 10 * (28.36 g moles) * (8.314 J / mol K ) * (1041 R ) * (5 / 9 K / R )

Pa

=

(11,429 in 3 ) * (1.6387

• 10 5 m 3 / in.3 )

= 105.6 psia = 90.9 psig.

Where:

TN-LC-0100 3-53

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The resulting maximum temperatures for hot NCT with insolation using design basis and fine meshed models are listed in the table below.

Maximum/Average Temperatures of Fuel Basket Component for Hot NCT Heat Load (kW) 0.390 1.62 1.5 Basket Type TN-LC-NRUX TN-LC-MTR TN-LC-TRIGA Mesh Type Coarse Fine Coarse Fine Coarse Fine Maximum Tmax Tmax T Tmax Tmax T Tmax Tmax T Temperature (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F)

Basket Shell/Rail 178 178 +0 220 220 +0 231 231 +0 Guide Plate (Side, Center and A 191 192 +1 -- -- -- -- -- --

Plates)

Tube Wrap/Outer Plate 193 193 +0 250 250 +0 239 239 +0 (Compartment)

Poison Plate -- -- -- -- -- -- 255 257 +2 Tube/Bucket 199 200 +1 256 256 +0 255 257 +2 Fuel Cladding 205 205 +0 262 262 +0 266 268 +2 Average Tavg Tavg T Tavg Tavg T Tavg Tavg T Temperature (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F)

Helium 182 182 +0 220 220 +0 217 217 +0 Fuel Cladding 196 196 +0 245 245 +0 250 251 +1 Heat Load (kW) 2.0 3.0 3.0 TN-LC-1FA TN-LC-1FA TN-LC-1FA Basket Type (BWR) (PWR) (Pin Can)

Mesh Type Coarse Fine Coarse Fine Coarse Fine Maximum Tmax Tmax T Tmax Tmax T Tmax Tmax T Temperature (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F)

Basket Rail 261 261 +0 268 268 +0 268 268 +0 Poison Plate 265 265 +0 275 275 +0 274 274 +0 Frame 267 267 +0 278 279 +1 277 277 +0 Sleeve Wall 298 298 +0 -- -- -- 318 318 +0 Pin Can Wall -- -- -- -- -- -- 379 380 +1 Fuel Cladding 497 496 -1 520 519 +1 543 542 +1 Average Tavg Tavg T Tavg Tavg T Tavg Tavg T Temperature (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F) (°F)

Helium 260 260 +0 256 256 +0 282 282 +0 Fuel Cladding 396 396 +0 401 401 +0 455 456 +1 Based on the results presented in the table above, there is an insignificant effect on maximum temperatures for fuel cladding and basket components (within 2°F) by refining mesh size from TN-LC-0100 3-68

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 3.6.6 Effective Thermal Properties of the PWR and BWR Fuel Assemblies The purpose of this appendix is to determine the effective thermal conductivity, specific heat and density for the fuel assemblies within the fuel basket assemblies for use in the analysis of the thermal performance of the TN-LC transport package. The TN-LC transport package is designed to accommodate the contents described in Chapter 1. The general fuel descriptions are:

• Fuel pins: UO2, MOX, and EPR,

• PWR or BWR fuel assemblies,

• NRU and NRX fuel assemblies,

• MTR fuel elements,

• TRIGA fuel elements.

The characteristics of the design basis fuel assemblies/elements selected for calculation of effective fuel properties are listed in table below:

Characteristics of Design Basis Contents in TN-LC Transport Package 1FA 1FA MTR TRIGA Fuel Basket Type NRUX (21-Pin Can) (9-Pin Can) (ORR#1) (Al Clad)

Decay Heat, W 120/Pin 220/Pin 10/Element 33/Element 15/Element Fuel Compartment Open 5.0 x 5.0 2.245 3.48 x 3.48 3.48 x 3.48 Size, in.

107.9 (NRU) /

Active Fuel Length, in. 144 22/Element 14/Element 96 (NRX)

Active Fuel Cross 0.216 (NRU)/

0.3088 2.5 x 0.02 1.41 Section Size, in. 0.250 (NRX) 0.03 (NRU)

Cladding Thickness, in. 0.0225 0.014 0.03 0.045 (NRX)

The fuel pin sizes are based on the minimum dimensions for the allowable fuel pins to maximize the gaps and bound the lowest effective conductivity.

The following assumptions were made:

(a) Fuel elements or fuel rods are centered in the fuel compartment.

(b) Heat generation is uniformly distributed along active fuel region. Decay heat load has negligible effect on the transverse effective thermal conductivities.

(c) The fuel compartment opening size for the PWR FA in the TN-LC-1FA basket is identical to the compartment opening size of the 24PTH basket. The bounding effective fuel properties for PWR fuel assemblies in the 24PTH basket are evaluated in Appendix P, Section P.4.8 of [25]. Evaluations in Appendix P, Section P.4.8 of [25] are based on unirradiated UO2 properties. The evaluation in Chapter 4, Section 4.8.6 of [15] shows that TN-LC-0100 3-74

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 surfaces to create the radiation super-element. The LINK32 elements were unselected prior to the solution of the model. The model was run with a series of isothermal boundary conditions applied to the outermost nodes representing fuel compartment walls. The conductivity of helium is used for the back fill gas.

The finite element models of TN-LC contents are shown in Figure 3-25.

The isotropic effective thermal conductivity of a heat generating square, such as the fuel pins in the TN-LC-1FA basket, can be calculated as described in [29]:

Q Q react k eff = 0.29469 x = 0.29469 x 4 x L x (Tc Ta ) 4 x (Tc Ta )

Where:

Q = decay heat load per fuel assembly (Btu/hr)

Qreact = reaction solution in the model =Q/L (Btu/hr-in.)

L = active fuel length (in.)

Tc = maximum temperature of fuel assembly (°F)

Ta = compartment wall temperature (°F)

The isotropic effective thermal conductivity of a heat generating cylinder such as the fuel assembly in TN-LC-NRUX basket tubes can be calculated as described in [4]:

Q Q react k eff = =

4 x x L x (Tc Ta ) 4 x x (Tc Ta )

In determining the temperature dependent effective fuel conductivities an average temperature, equal to (Tc + To)/2, is used for the fuel temperature.

Section 3.2.1, Material 15 provides the calculated fuel effective thermal conductivities. Effective fuel density, specific heat and axial fuel thermal conductivity used for transient thermal analysis of the TN-LC cask transporting a PWR fuel assembly are based on the bounding values from the 24PTH analysis in Appendix P.4, Section P.4 of [25].

Effective fuel density, specific heat and axial thermal conductivity used for transient thermal analysis of TN-LC cask transporting PWR fuel pins are computed in Section 3.6.8.2.

Figure 3-26 shows typical temperature plots for the TN-LC payload fuel assemblies with 800°F applied to the compartment walls.

Figure 3-27 shows a comparison plot of fuel transverse effective thermal conductivities for the TN-LC contents.

TN-LC-0100 3-76

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 3.6.7 Bounding Transverse Fuel Effective Thermal Conductivity for UO2, MOX, and EPR Irradiated Fuels 3.6.7.1 UO2 and MOX Irradiated Fuel Assembly Thermal Conductivity Evaluation The effect of using MOX fuel instead of UO2 fuel on the transverse fuel effective thermal conductivity is evaluated based on a sensitivity analysis of a WE 14x14 FA. As shown in Appendix M, Section M.4.8.1 of [25], the WE 14x14 FA model provides the bounding (lowest) transverse effective conductivity among PWR FAs. As listed in Table 1.4.5-2, WE 16x16 fuel assembly has the similar fuel assembly geometry with the bounding WE 14x14 fuel assembly.

Since WE 16x16 has more fuel rods (235 fuel rods for WE 16x16 fuel class as opposed to 179 fuel rods for WE 14x14 fuel class) and smaller rod pitch (0.496 rod pitch for WE 16x16 fuel class as opposed to 0.556 rod pitch for WE 14x14 fuel class), WE 16x16 fuel assembly is expected to have better thermal performance than WE 14x14 for the same thermal condition.

Therefore, it is concluded that WE 16x16 fuel assembly remains bounded by WE 14x14 fuel assembly. The WE 14x14 FA model developed in Appendix M, Section M.4.8.1 of [25] is used for the sensitivity analysis in this calculation.

The characteristics of the WE 14x14 model used for the sensitivity analysis are listed in the table below.

Characteristics of Design Basis FAs and Surface Properties Design Basis FA Type WE 14x14 Decay Heat, kW 0.75 Fuel Compartment Open Size, in. 8.7 x 8.7 Active Fuel Length, in. 144 Emissivity of Zircaloy 0.8 Emissivity of Stainless Steel Wall 0.46 Emissivity of Al/Poison Wall 0.85 Emissivity of Symmetry Plane 0.001 In the sensitivity analysis, the transverse fuel effective conductivities are calculated once using the irradiated conductivity of UO2 for the fuel pellets. The conductivity of the fuel pellet is then replaced by the conductivity of irradiated MOX.

Effect of irradiation on thermal conductivity for UO2 fuel is calculated as k UO2 irr / k UO2 un-irr It is assumed that this effect can also be used to calculate the irradiated MOX fuel thermal conductivity:

k MOX irr = k MOX un-irr

• k UO2 irr / k UO2 un-irr.

The results of irradiated MOX fuel thermal conductivity are shown in the table below.

TN-LC-0100 3-77

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 3.6.7.2 UO2 and EPR Irradiated Fuels Evaluation The parameters of EPR and UO2 fuel rods are shown in the table below.

Irradiated PWR and EPR Fuel Rod Parameters Parameter PWR UO2 rod EPR UO2 rod 4,300 mm for Active length 1,300 MWe 4,551 mm 3,700 mm for 900 MWe Min.: 8.04 mm Pellets Diameter: 8.192 +/-0.012 mm Max.: 8.30 mm Alloy containing at least Cladding M5 95% Zirconium Max. 9.70 mm External diameter of cladding 9.500 +/-0.040 mm Min. 9.36 mm Max.: 0.68 mm 0.570 mm nominal Cladding thickness Min.: 0.52 mm 0.535 mm minimal The design of the EPR and UO2 fuel assembly is similar as seen from the parameters shown in the table above.

As seen in the table above, fuel rods for PWR and EPR FAs have the same UO2 fuel. They have similar geometry with close pellet and cladding dimension. Further M5 has a higher thermal conductivity than zircaloy [30]. Therefore, FA effective thermal conductivity calculated for a PWR fuel pin can be used for an EPR fuel pin.

3.6.8 Confirmatory Analysis for NCT and HAC using Coupled Model The thermal evaluations presented in Section 3.3 for NCT and Section 3.4 for HAC are performed using separate thermal models representing the TN-LC TC and the various basket types allowed within the TN-LC TC.

The TN-LC TC thermal model used in Section 3.3 for NCT is a 3D model that includes the transport cask components but does not include the basket contents whereas the TN-LC TC thermal model used for the HAC evaluations as described in Section 3.4 includes a homogenized basket to account for the thermal capacity during transient evaluations. The 3D thermal model of the TN-LC determines the maximum component temperatures for the cask components and provides the temperatures distribution for the cask inner shell for use in the thermal evaluation of the various basket types. For the various basket types allowed within the TN-LC TC using the maximum inner shell temperature from the 3D TN-LC TC models, 2D thermal models as described in Section 3.3.1.4 are used for NCT and HAC to evaluate the thermal performance during transport.

Although use of a 2D thermal model to predict the maximum fuel cladding and basket component temperatures is conservative, 90o 3D coupled models of the TN-LC TC and the TN-LC-1FA basket with the maximum allowed heat load of 3.0 kW are developed to determine the bounding TN-LC-0100 3-79

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 configuration and to verify that the maximum temperatures determined in Section 3.3 for NCT and Section 3.4 for HAC remain bounding.

These evaluations also provide a measure of the variations in the thermal performance due to the various types of content allowed within the TN-LC-1FA basket (PWR fuel assembly or PWR fuel pins). They also quantify the impact of the environment (With ISO or Without ISO Container) on the thermal performance of the TN-LC TC during NCT and HAC.

The coupled models are created by extruding the 2D model of the TN-LC-1FA basket developed in Section 3.3.1.4 and introducing the elements and nodes from the extruded basket model into the TC model with ISO container. The 2D model of the TN-LC-1FA basket is extruded such that the locations of the nodes in the radial and axial direction correspond to the location of the nodes in the 3D TN-LC TC model to ensure mesh continuity. For the thermal evaluations without the ISO container, the elements representing the ISO container and the air within the ISO container cavity are deleted.

The coupled model of the TN-LC TC and TN-LC-1FA basket explicitly models the cask components as well as the basket components (including poison plates, back-filled gas and aluminum transition rails) and considers a homogenized fuel assembly within the fuel compartment. For the TN-LC-1FA basket with fuel pins, the BWR sleeve along with the pin can side wall is also explicitly modeled. The 25 fuel pins1 along with the stainless tubes are represented as a homogenized region with effective properties. The effective thermal properties are presented in Appendix 3.6.6 for the PWR fuel assembly. For the pin can, the effective properties are presented in Appendix 3.6.8.2. The geometry of the coupled model is shown in Figure 3-28 and Figure 3-29.

Decay heat load is applied as heat generation boundary conditions over the elements representing homogenized fuel assembly or the homogenized pin can. The base heat generation rate is multiplied by peaking factors along the axial fuel length to represent the axial decay heat profile consistent with the approach described in Appendix 3.6.8.1.

The ambient boundary conditions for the coupled model are identical to those described in Section 3.3 and Section 3.4 for the TC model under NCT and HAC, respectively. For the thermal evaluations of the TN-LC TC within an ISO container under HAC, it is assumed that the fire engulfs the entire ISO container and heats up the air within the ISO container providing heat input into the cask. A uniform temperature of 850°F is considered within the ISO container cavity for the entire 30 minute duration to provide heat input from the fire through the ISO container wall into the TN-LC transport cask. The assumed uniform temperature of 850°F is the highest average temperature between the fire temperature and the cask outer shell under NCT.

Further the impact limiters are assumed to be intact for the HAC evaluation with the TN-LC TC in the ISO container.

The configurations listed in the table below are considered to perform the evaluations. The maximum fuel cladding temperature is used as the criterion to determine the limiting configuration of the TN-LC package for NCT and HAC.

1 The thermal analysis conservatively considers 25 fuel pins with heat load of 120 W per pin, which is bounding for the 21 fuel pins configuration.

TN-LC-0100 3-79a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 List of Evaluated Cases to Determine the Bounding Configuration Heat Operating Impact ISO Run # Load Fuel Type Smoldering(1)

Condition Limiter Container (kW)

PWR Fuel N1 No Yes Assembly N2 Pin Can No Yes NCT Intact PWR Fuel N3 No No Assembly N4 Pin Can No No 3.0 PWR Fuel H1 No Yes Assembly Intact H2 Pin Can No Yes HAC PWR Fuel H3 Yes No Assembly Damaged H4 Pin Can Yes No Notes:

1. Immediately after the end of the fire, a 0.5-hour wood smoldering process is considered with a char wood temperature of 900°F, which is directly applied over the damaged impact limiter inner skin as described in Section 3.4.2.

The maximum fuel cladding and seal temperatures resulting from the above evaluations for NCT (configurations N1 through N4) are summarized in the table below.

Maximum Temperatures for Fuel Cladding and Seals in TN-LC for NCT NCT Coupled Model Decoupled Model N1 N3 N2 N4 Loading (Table 3-1, Table 3-10 and (PWR Fuel (PWR Fuel (Pin Can) (Pin Can)

Condition Table 3-16) Assembly) Assembly )

with ISO without ISO with ISO Container without ISO Container Container Container Tmax Tmax Tmax Tmax

(°F) (°F) (°F) (°F)

Fuel 543 (PIN)/

N/A 522 512 508 495 Cladding 520 (PWR)

Seals 205 173 209 210 188 188 As seen from the above table, the limiting configuration for NCT is the TN-LC package with PWR fuel assemblies inside the ISO container. The maximum fuel cladding temperature for the limiting configuration resulting from the coupled model remains bounded by the maximum value reported in Table 3-2 determined using the decoupled model. The maximum seal temperature of 210°F determined for load case N2 exceeds the maximum seal temperature of 205°F (See Table 3-1) determined using the decoupled model. However, considering the large margin to the long TN-LC-0100 3-79b

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 term seal temperature limit of 400oF specified in Section 3.1, there is no effect on the performance of the seal and containment of the TN-LC TC is assured during NCT.

Based on NCT evaluations summarized in the above table, the maximum fuel cladding temperature for fuel pins is bounded by the maximum fuel cladding temperature for PWR fuel assemblies. Configuration H2 is evaluated to provide further evidence for this conclusion. The maximum fuel cladding and seal temperatures resulting from the evaluations H1 through H4 described above are summarized in the table below.

Maximum Fuel Cladding and Peak Seal Temperatures in TN-LC for HAC HAC Coupled Model (Tmax/T)

Decoupled Model H1 H3 (Table 3-6 and Table 3-7 ) H2 H4 Loading (PWR Fuel (PWR Fuel (Pin Can) (Pin Can)

Condition Assembly ) Assembly )

with ISO without ISO with ISO Container without ISO Container Container Container Tmax Tmax Time T Tmax Time T Tmax Time T Tmax Time T

(°F) (°F) (hr) (°F) (°F) (hr) (°F) (°F) (hr) (°F) (°F) (hr) (°F)

Fuel 694 --- --- 575 8.43 556 570 8.47 553 583 8.62 536 (1)

Cladding ---

Seals --- 449 1.11 275 285 5.43 267 284 5.47 266 451 1.11 261 Notes:

1. Bounded by Configuration H3.

As shown in the above table, the limiting configuration for HAC is the TN-LC package with PWR fuel assemblies inside the ISO container. The maximum fuel cladding temperature for the limiting configuration resulting from the coupled model remains bounded by the maximum value reported in Table 3-7.

As seen in the above table, the short-term peak seal temperature of the top lid seal for the TN-LC TC without the ISO container remains below the short term limit of 482°F specified for Viton fluorocarbon O-rings in Section 3.1. The length of the time intervals, in which the seal temperatures are above the long-term limit of 400°F, are extracted from the data based on the transient evaluation H3 and presented in the table below.

TN-LC-0100 3-79c

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 1.2 1.0 Peaking Factor 0.8 0.6 DOE/RW-0472 0.4 TN-LC Coupled Model 0.2 0.0 0 24 48 72 96 120 144 Active Fuel Length (in)

Peaking Factor Curve for PWR Fuel As seen in Table 3-23, the normalized area under the peaking factor curve is smaller than 1.0. To avoid any degradation of decay heat load, a correction factor of 1.0154 calculated as follows is used when applying the peaking factors.

Area under Axial Heat Profile Nomalized Area under Curve = = 0.9848 Active Fuel Length Active fuel length = 144 1

Correction Factor = = 1.0154 Normalized Area under Curve 3.6.8.2 Effective Thermal Properties for the TN-LC-1FA 25 Pin Can Basket Effective transverse thermal conductivity for the fuel pins and the enclosing tubes in TN-LC-1FA pin can basket is determined in Appendix 3.6.6. It should be noted that the pin can side wall is explicitly modeled and is not included in the calculation of the effective thermal properties.

This section computes the effective density, specific heat and axial thermal conductivity for the TN-LC-1FA pin can basket with fuel pins for use in the transient thermal analyses. The effective properties computed in this section are summarized in Section 3.2.1, Material 15.

In computing the effective thermal properties, specific heat and density based on the bounding lowest values are considered for the fuel pellet, fuel cladding and the stainless steel pin tubes.

The bounding properties used are presented in the table below:

TN-LC-0100 3-79e

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Specific Heat and Density of Fuel Pellet, Fuel Cladding and Pin Tubes Density of Fuel Pellet, (lbm/in3) UO2 0.396 Specific Heat of Fuel Pellet, (Btu/lbm-°F) CP_UO2 0.056 3

Section 4.8 of Refeence [15]

Density of Fuel Cladding, (lbm/in ) Z 0.237 Specific Heat of Fuel Cladding, (Btu/lbm-°F) CP_Z 0.067 3

Density of Stainless Steel, (lbm/in ) SS 0.284 Section 3.2.1, Mateial 5 Specific Heat of Stainless Steel, (Btu/lbm-°F) CP_SS 0.116 The characteristics of the design basis fuel pins and the pin can selected for calculation of effective fuel properties in this section are based on the following dimensions and are summarized below:

Characteristics of Design Basis Fuel Pins in TN-LC Transport Package Number of Fuel Pins 25 Active Fuel Length, La (in) 144 Fuel Pin OD (in) 0.36 Cladding Thickness (in) 0.0225 Fuel Pellet OD (in.) 0.3088 Total Volume of Fuel Pellets, VUO2 (in3) 269.6 Total Mass of Fuel Pellet, MUO2 (lbm) 106.8 2

Total Area of Cladding, AZr (in ) 0.596 3

Total Volume of Cladding, VZr (in ) 85.9 Total Mass of Cladding, MZr (lbm) 20.4 Number of Pin Tubes 25 Pin Can Compartment Width, a (in) 5.0 2

Area of Pin Can Compartment, Aassembly (in ) 25.0 Pin Tube OD (in) 1.0 Pin Tube Thickness (in) 0.065 2

Total Area of Pin Tubes, Atube (in ) 4.77 3

Total Volume of Pin Tubes, Vtube (in ) 687.3 Total Mass of Pin Tubes, Mtube (lbm) 195.2 Basket Effective Density and Specific Heat of TN-LC-1FA Pin Can basket with fuel pins are calculated, respectively using equations listed below.

r fuel = M i

=

M UO 2 + M Zr + M tube Vassembly a 2 La TN-LC-0100 3-79f

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 C P _ fuel =

M C i Pi

=

M UO 2 C P _ UO 2 + M Zr C P _ Zr + M tube C P _ tube M i M UO 2 + M Zr + M tube Where:

Mi = Mass of fuel pellets, fuel cladding and stainless steel pin tubes, respectively lbm CPi = Specific heat of fuel pellets, fuel cladding and stainless steel pin tubes, respectively, Btu/lbm-°F Vassembly = Total volume of pin can compartment =a2xLa=3600 in3 La = Active fuel length = 144 in a = Pin can compartment width =5.0 in.

Using the above formulas and the physical properties determined in the previous tables, effective density and specific heat are computed in the following tables for the TN-LC-1FA pin can basket with fuel pins.

Effective Density of TN-LC-1FA Pin Can Basket with Fuel Pins Components Material Total Mass (lbm)

Fuel Pellets UO2 106.8 Fuel Cladding Zircaloy-4 20.4 Pin Tube Stainless Steel 195.2 Total 322.4 Homogenized Fuel in Pin Can Basket Pin Can Compartment Width (in) 5.0 Active Fuel Length, La (in) 144 3

Vassembly (in ) 3600.0 fuel (lbm/in3) 0.0895 Effective Specific Heat of TN-LC-1FA Pin Can Basket with Fuel Pins Mass Specific Heat Component (lbm) (Btu/lbm-°F)

Fuel Pellets 106.8 0.056 Fuel Cladding 20.4 0.067 Pin Tube 101.0 0.116 Homogenized Fuel in Pin Can Basket CP_fuel (Btu/lbm-°F) 0.0930 TN-LC-0100 3-79g

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Effective Axial Thermal Conductivity of TN-LC-1FA Pin Can Basket with Fuel Pins The effective axial thermal conductivity for the TN-LC-1FA pin can basket with fuel pins is calculated as area average as shown below:

k Zr

• AZr + k SS

• Atube k eff _ axial =

A assembly Where:

kZr, kSS = Thermal conductivity of fuel cladding and stainless steel tubes, respectively, Btu/hr-in-°F AZr, Atube = Area of fuel cladding and stainless steel tubes, respectively, in2 keff_axial = Effective axial thermal conductivity for the TN-LC-1FA pin can basket, Btu/hr-in-°F Aassembly = Area of pin can compartment = 25 in2.

Effective Axial Conductivity of TN-LC-1FA Pin Can Basket with Fuel Pins kZr kSS Temperature (Section 3.2.1, (Section 3.2.1, keff_axial Material 2) Material 5)

(°F) (Btu/hr-in-°F) (Btu/hr-in-°F) (Btu/hr-in-F) 200 0.654 0.775 0.1636 300 0.69 0.817 0.1725 400 0.726 0.867 0.1829 500 0.756 0.908 0.1914 600 0.786 0.942 0.1986 800 0.852 1.025 0.2160 3.6.9 Thermal Evaluation of TN-LC Transport Cask with High Burnup Fuel Assemblies

[

]

TN-LC-0100 3-79h

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 3-4 Maximum Temperatures of TN-LC Cask Components for Cold NCT Ambient Temperature -20°F -40°F Temperature Temperature

(°F) (°F)

Inner Shell 88 70 Gamma Shield 87 69 Gamma Shield (Top and Bottom) 52 34 Outer Shell 73 55 Neutron Shield Boxes 68 50 Neutron Shield Resin(1) 63 44 Neutron Shield Shell 63 45 Cask Lid 52 34 Cask Bottom Flange 36 18 Cask Top Flange 55 37 Wood in Impact Limiter 51 33 Notes:

1. The resin temperature is the volumetric average temperature at the hottest cross-section.

Table 3-5 Maximum Fuel Basket/Fuel Cladding Temperature and Maximum Fuel Basket Temperature Gradient for Cold NCT Fuel Basket Type 1FA (Pin Can) Others Except 1FA (Pin Can)

Tmax, Fuel (°F) 432 Tmax, basket (°F) 250 Tbasket (°F) 190 100 TN-LC-0100 3-82

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 3-14 Basket Types and Heat Loads Used in Thermal Calculations for TN-LC Transport Cask Total Heat Heat Load No. of Load Considered Heat Load per Elements/ Peaking Considered in Cask Basket Type Element/Assembly Assemblies in Factor in Basket Model (W)

Basket Model (kW)

(kW)

TN-LC- 26 NRU/NRX 15/Assembly 1.0 0.390 0.5 NRUX Assemblies TN-LC-MTR 54 Elements 30/Element 1.0 1.62 1.85 TN-LC-180 Elements 8.33/Elements 1.0 1.5 1.5 TRIGA TN-LC-1FA 1 BWR 2000/Assembly 1.2 2.0 2.0 (BWR) Assembly TN-LC-1FA 1 PWR 3000/Assembly 1.1 3.0 3.0 (PWR) Assembly TN-LC-1FA 9 Fuel Pins (1) 220/Pin 1.1 2.86 3.0 (Pin Can) 21 Fuel Pins (2) 120/Pin 1.1 3.0 3.0 Notes:

1. The thermal model considers 13 fuel pins with heat load of 220 W per pin, which is bounding for the 9 fuel pins configuration.

2. The thermal model considers 25 fuel pins with heat load of 120 W per pin, which is bounding for the 21 fuel pins configuration.

Table 3-15 Temperature Boundary Conditions for Fuel Basket Models Uniform Cask Inner Shell Ambient Temperature in Basket Type Temperature Fuel Basket Model

(°F)

(°F)

TN-LC-NRUX 100 170 TN-LC-MTR 100 200 TN-LC-TRIGA 100 200 TN-LC-1FA 100 240 (PWR)

TN-LC-1FA 100 240 (BWR)

TN-LC-1FA 100 240 (Pin Can) -40 69 TN-LC-0100 3-88

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 3-16 Maximum/Average Temperatures of Fuel Basket Component for Hot NCT Basket Type TN-LC-NRUX TN-LC-MTR TN-LC-TRIGA Maximum Tmax Tmax Tmax Temperature (°F) (°F) (°F)

Basket Shell/Rail 178 220 231 Guide Plate (Side, 191 -- --

Center and A Plates)

Tube Wrap/Outer 193 250 239 Plate (Compartment)

Poison Plate -- -- 255 Tube/Bucket 199 256 255 Fuel Cladding 205 262 266 Average Tavg Tavg Tavg Temperature (°F) (°F) (°F)

Helium 182 220 217 Fuel Cladding 196 245 250 TN-LC-1FA TN-LC-1FA Basket Type TN-LC-1FA (BWR)

(PWR) (Pin Can)

Maximum Tmax Tmax Tmax Temperature (°F) (°F) (°F)

Basket Rail 261 268 268 Poison Plate 265 275 274 Frame 267 278 277 Sleeve Wall 298 -- 318 Pin Can Wall -- -- 379 Fuel Cladding 497 520 543 Average Tavg Tavg Tavg Temperature (°F) (°F) (°F)

Helium 260 256 282 Fuel Cladding 396 401 455 TN-LC-0100 3-89

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 3-17 Maximum Basket Temperature Gradients Calculation for Hot NCT 1FA 1FA 1FA Fuel Basket Type TRIGA MTR NRUX (Pin Can) (PWR) (BWR)

Tamb (oF) 100 100 100 100 100 100 Tmax, basket (oF) 379 278 298 255 256 199 TInner Shell (oF) 240 240 240 200 200 178 Tbasket (oF) 139 38 58 55 56 21 Table 3-18 Maximum Basket/Fuel Cladding Temperature and Maximum Basket Temperature Gradient for Cold NCT (-40°F Ambient)

Fuel Basket Type 1FA (Pin Can)

Tamb (°F) -40 Tmax, Fuel (°F) 432 Tmax, basket (°F) 250 TInner Shell (°F) 69 Tbasket (°F) 181 Tavg, Fuel (°oF) 335 Tavg, Helium (°F) 125 TN-LC-0100 3-90

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 3-19 Maximum Fuel Cladding and Basket Component Temperatures for Dry Loading/Unloading Conditions TN-LC with NRU/NRX/MTR/TRIGA Fuels Maximum Temperature Temperature Limit Component

(°F) (°F)

Basket Shell/Rail 225 ---

Basket Plate 281 ---

Tube/Bucket 309 ---

Fuel Cladding 320 400 TN-LC with PWR/BWR Fuel Assemblies / Fuel Pins Maximum Temperature Temperature Limit Component

(°F) (°F)

Basket Rail 274 ---

Poison Plate 306 ---

Frame 309 ---

Sleeve Wall 383 ---

Pin Can Wall 476 ---

Fuel Cladding 720 752 Table 3-20 Comparison of Maximum Fuel Cladding Temperatures TN-LC-MTR Basket Dry Loading/Unloading NCT (Table 3-2)

T Tmax Tmax

(°F)

Component (°F) (°F)

Fuel Cladding 320 262 +58 TN-LC-1FA (Pin Can) Basket Dry Loading/Unloading NCT (Table 3-2)

T Tmax Tmax

(°F)

Component ( °F) (°F)

Fuel Cladding 720 543 +177 TN-LC-0100 3-91

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Chapter 4 Containment TABLE OF CONTENTS 4.1 Description of Containment System .................................................................................... 4-1 4.1.1 Containment Boundary................................................................................................ 4-1 4.2 Containment under Normal Conditions of Transport (Type B Packages) ......................... 4-4 4.2.1 Containment of Radioactive Material ......................................................................... 4-4 4.2.2 Pressurization of Containment Vessel ......................................................................... 4-4 4.2.3 Containment Criterion ................................................................................................ 4-4 4.3 Containment under Hypothetical Accident Conditions (Type B Packages) ....................... 4-5 4.3.1 Fission Gas Products .................................................................................................. 4-5 4.3.2 Containment of Radioactive Material ......................................................................... 4-5 4.3.3 Containment Criterion ................................................................................................ 4-5 4.4 Special Requirements .......................................................................................................... 4-6 4.5 References ........................................................................................................................... 4-7 4.6 Appendices .......................................................................................................................... 4-7 4.6.1 Containment Reference Leak Rate for 1FA Contents ................................................. 4-7 LIST OF FIGURES Figure 4-1 TN-LC Cask Containment ....................................................................................... 4-8 TN-LC-0100 4-i

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The containment vessel is hydrostatically tested in accordance with the requirements of the ASME B&PV Code,Section III, Article NB-6200.

Even though the Code is not strictly applicable to transport casks, it is the intent to follow Section III, Subsection NB of the Code as closely as possible for design and construction of the containment vessel. The casks may, however, be fabricated by other than N-stamp holders and materials may be supplied by other than ASME Certificate Holders. Thus the requirements of NCA are not imposed. TN's quality assurance requirements, which are based on 10CFR71 Subpart H and NQA-1, are imposed in lieu of the requirements of NCA-3850. This SAR is prepared in place of the ASME design and stress reports. Surveillances are performed by TN and other personnel rather than by an Authorized Nuclear Inspector (ANI).

The materials of the TN-LC packaging will not result in any significant chemical, galvanic or other reaction as discussed in Chapter 2.

4.1.1.2 Containment Penetrations The only penetrations through the containment boundary are the drain and vent ports, bottom plug plate (with or without gamma shielding) and the top closure plate (lid). Each penetration is designed to maintain a leak rate not to exceed 1.0 x 10-7 ref cm3/s, defined as leak tight per ANSI N14.5 [4]. To obtain these seal requirements, each penetration has an O-ring face seal type closure. Additionally, the lid and bottom plug penetrations have double O-ring configurations.

4.1.1.3 Seals and Welds All containment boundary welds are full penetration bevel or groove welds to ensure structural and sealing integrity. These full penetration welds are designed per ASME III Subsection NB and are fully examined by radiographic or ultrasonic methods in accordance with Subsection NB.

Additionally, a liquid penetrant examination is performed on these welds.

Containment seals are located at the bottom plug plate, lid, the drain plug and the vent plug. The inner seal, when two seals are provided, is the primary containment seal. The outer, secondary seals, facilitate leak testing of the inner containment seal of the bottom plug and the lid. There are also test ports provided for these two closures. The test ports are not part of the containment boundary.

All the seals used in the TN-LC cask containment boundary are static face seals. The seal areas are designed such that no significant plastic deformation occurs under normal and accident loads as shown in Chapter 2. The bolts are torqued to maintain seal compression during all load conditions as shown in Appendix 2.13.2. The seals used for all of the penetrations are fluorocarbon elastomer O-rings. All seal contact surfaces are stainless steel and are machined to a 32 RMS or finer surface finish. The dovetail grooves in the cask lid and the bottom end plug cover plate are intended to retain the seals during installation. The volume of the grooves is controlled to allow the mating metal surfaces to contact under bolt loads, thereby providing uniform seal deformation in the final installation condition.

TN-LC-0100 4-2

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 A fluorocarbon elastomeric seal was chosen for use on the TN-LC package because it has acceptable characteristics over a wide range of parameters. The fluorocarbon compound specified is V1289-75 or equivalent. The V1289-75 compound as described by the Parker Technical Bulletin ORD 5743 [8] is specially formulated for use at temperatures as low as -55°F while maintaining the upper temperature limit of 400°F. The selected seals remain leak tight (leak rate not exceeding 1.0 x 10-7 ref cm3/s) at 482°F for accident conditions as shown in Parker O-Ring Handbook [7].

4.1.1.4 Closure The containment vessel contains an integrally-welded bottom closure and a bolted and flanged top closure forging (lid). The lid forging is attached to the cask body with twenty (20), SA-540, Grade B23 or B24, Class 1, 1.0 inch diameter bolts and stainless steel washers. Closure of the bottom plug (with or without gamma shielding) is accomplished by eight (8), SA-540, Grade B23 or B24, Class 1, 0.5 inch diameter cap screws and stainless steel washers. The bolt torque required for the top lid and bottom plug are provided in Drawing 65200-71-01 in Chapter 1, Appendix 1.4.1. The closure bolt analysis is presented in Appendix 2.13.2.

Closure of each of the vent and drain ports is accomplished by a single 0.5 inch brass or ASTM A193, Grade B8 bolt with an elastomer seal under the head of the bolt.

TN-LC-0100 4-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 4.2 Containment under Normal Conditions of Transport (Type B Packages) 4.2.1 Containment of Radioactive Material As described earlier, the TN-LC cask is designed and tested for a leak rate of 1.0 x 10-7 ref cm3/s, defined as leak tight per ANSI N14.5 [4]. Additionally, the structural and thermal analyses presented in Chapters 2 and 3, respectively, verify that there is no release of radioactive materials under any of the normal or accident conditions of transport.

4.2.2 Pressurization of Containment Vessel The TN-LC cask contains one of four basket designs holding dry irradiated fuel and helium gas which is used to backfill the cask after drying. Therefore, the pressure in the TN-LC cask when loaded with fuel is from helium that has been backfilled into an evacuated cask cavity to a pressure of 2.5 +/- 1 psig at the end of loading. If the TN-LC cask contains design basis fuel at thermal equilibrium, the cask cavity helium temperature with 100°F ambient air and maximum insolation is 282°F. The maximum normal operating pressure (MNOP) is calculated in Chapter 3 to be 16.9 psig. The analyses in Chapters 2 and 3 demonstrate that the TN-LC cask effectively maintains containment integrity with a cavity pressure of 30 psig.

4.2.3 Containment Criterion The TN-LC cask is designed to be leak tight. The acceptance criterion for fabrication verification and periodic verification leak tests of the TN-LC cask containment boundary shall be 1.0 x 10-7 ref cm3/s. The test must have a sensitivity of at least one half the acceptance criterion, or 5.0 x 10-8 ref cm3/s. The testing of the containment boundary is described in Chapter 8.

For shipment of 1FA contents (PWR, BWR or pin can) in the TN-LC, a leak rate test criterion that is less restrictive than the leak-tight criterion may be used. The leak rate criterion for 1FA contents is established in Appendix 4.6.1.

TN-LC-0100 4-4

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 4.3 Containment under Hypothetical Accident Conditions (Type B Packages) 4.3.1 Fission Gas Products There is no need to explicitly determine the source term available for release. As described earlier, the TN-LC cask is designed and tested for a leakage rate of 1.0 x 10-7 ref cm3/s, defined as leak tight per ANSI N14.5 [4].

4.3.2 Containment of Radioactive Material The TN-LC cask is designed and tested to be leak tight. The results of the structural and thermal analyses presented in Chapters 2 and 3, respectively, verify the package will meet the leakage criteria of 10CFR71.51 for the hypothetical accident scenario.

4.3.3 Containment Criterion This package has been designed and is verified by leakage testing to meet the leak-tight criteria of ANSI N14.5 [4]. The results of the structural and thermal analyses presented in Chapters 2 and 3, respectively, verify the package will meet the leakage criteria of 10CFR71.51 for all the hypothetical accident conditions.

For shipment of 1FA contents (PWR, BWR or pin can) in the TN-LC, a leak rate test criterion that is less restrictive than the leak-tight criterion may be used.The leak rate criterion for 1FA contents is established in Appendix 4.6.1.

TN-LC-0100 4-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 4.5 References

1. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Subsection NB, 2004 edition including 2006 Addenda.

2. USNRC Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessel, Rev. 1, March 1978.

3. USNRC Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Cask, Rev. 1, March 1989.

4. ANSI N14.5-2014, American National Standard for Radioactive Material - Leakage Tests on Packages for Shipment, June 2014.

5. United States Air Force Military Specification MIL-R-83485, Rubber, Fluorocarbon Elastomer, Improved Performance at Low Temperatures, December 8, 1976.

6. Society of Automotive Engineers (SAE) Aerospace Material Specification (AMS) AMS-R-83485, Rubber, Fluorocarbon Elastomer, Improved Performance at Low Temperatures, May 1, 1998.

7. Parker O-Ring Handbook, Publication No. ORD-5700, 2007 Edition, Parker Seals www.parkerorings.com.

8. Low Temperature FKM V1289-75, Parker Technical Bulletin ORD5743, Parker Seals www.parkerorings.com.

4.6 Appendices 4.6.1 Containment Reference Leak Rate for 1FA Contents TN-LC-0100 4-7

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Appendix 4.6.1 Containment Reference Leak Rate for 1FA Contents TABLE OF CONTENTS 4.6.1 Containment Reference Leak Rate for 1FA Contents ............................................... 4.6.1-1 4.6.1.1 Criterion, Parameters and Assumptions...................................................... 4.6.1-1 4.6.1.2 Source Activities and Source Activity Densities ........................................ 4.6.1-3 4.6.1.3 Determination of allowable leakage rates ................................................... 4.6.1-6 4.6.1.4 Results ......................................................................................................... 4.6.1-9 4.6.1.5 Conclusions ................................................................................................. 4.6.1-9 4.6.1.6 References ................................................................................................... 4.6.1-9 LIST OF TABLES Table 4.6.1-1 10 CFR 71 Containment Criterion for Type B Transportation Packages ...... 4.6.1-1 Table 4.6.1-2 TN-LC Cavity Gas Temperatures, Pressures and Properties ......................... 4.6.1-2 Table 4.6.1-3 Gas Fission Products Volume ........................................................................ 4.6.1-4 Table 4.6.1-4 Gas Fission Products Activities ..................................................................... 4.6.1-5 Table 4.6.1-5 Gas Fission Products Released Activities - NCT (A2) .................................. 4.6.1-5 Table 4.6.1-6 Gas Fission Products Released Activies - HAC (A2 ) ................................... 4.6.1-6 Table 4.6.1-7 Activities and Activity Densities per Release Type ...................................... 4.6.1-6 TN-LC-0100 4.6.1-i Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 4.6.1 Containment Reference Leak Rate for 1FA Contents 4.6.1.1 Criterion, Parameters and Assumptions 4.6.1.1.1 Criterion There are two release rates calculated corresponding to the requirements for NCT and HAC in 10 CFR 71.51 [1]. The smallest value of these is determined to be the bounding release rate.,

although NCT and HAC values are both reported. Table 4.6.1-1 provides the containment criteria for the TN-LC transportation cask.

Table 4.6.1-1 10 CFR 71 Containment Criterion for Type B Transportation Packages Radioactive Release Rate Transport Condition Value Normal Conditions of Transport Less than A2 x10-6 per hour (NCT)

Hypothetical Accident Less than A2 per week (excluding 85Kr)

Conditions (HAC) Less than 10A2 per week (85Kr) 4.6.1.1.2 Parameters The TN-LC equipped with the 1FA basket is designed to transport the following:

• PWR irradiated fuel assemblies;

• BWR irradiated fuel assemblies loaded in the BWR sleeve fitted in the 1FA basket;

• Irradiated pins (UO2, MOX, and EPR) loaded in the 1FA pin can, which is itself loaded in the BWR sleeve fitted in the 1FA basket.

The following contributions are considered in determining the releasable source term for packages designed to transport irradiated fuel assemblies or rods: (1) the radionuclides comprising the fuel, (2) the radionuclides on the surface of the fuel rods/assemblies, and (3) the residual contamination on the inside surfaces of the containment vessel. However, studies have indicated that the contamination due to residual activity on the cask interior surfaces is negligible as compared to crud deposits on the fuel rods or assemblies [5]. This is helped by the fact that the TN-LC cask interior surfaces are cleaned before each loading operations. Therefore, this residual contamination on the interior surfaces of the cask is neglected in the following analysis and the following source terms are considered.

Based on Chapter 3 section 3.3.3, the bounding fuel assembly for this analysis is the B&W 15x15, which has 208 fuel rods per assembly. However, because the crud areal density for BWR is much higher than for PWR (by almost 10x), for the crud evaluation, this analysis considers 49 fuel rods per assembly with a Ø0.57 in rod O.D. (per Table 1.4.5-7), and the rod length is conservatively taken equal to 168 in.

The cask cavity free volume is equal to 11,429 in3 = 187,288 cm3 (see Chapter 3 section 3.3.3).

TN-LC-0100 4.6.1-1 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The calculation methodology is based on NUREG/CR-6487 [5]. ANSI N14.5 [3] provides the basis and methods for determining the maximum allowable reference air leakage rate for leak testing purposes.

Table 4.6.1-2 shows cavity pressures and gas temperatures relevant to TN-LC transport during NCT and HAC from Chapter 3 Tables 3-8 & 3-11. The temperatures used are the average cavity gas temperatures for the bounding FA with spent fuel contents. The table also lists the gas properties relevant to the analysis.

Table 4.6.1-2 TN-LC Cavity Gas Temperatures, Pressures and Properties Ref. Air In-Parameter NCT HAC Leakage at STP Pressure (equal to fluid 2.48 atm abs 13.35 atm abs 1 atm abs upstream pressure, Pu) (16.9 psig) (90.9 psig) 483.1K 578K Fluid Temperature (T) 298K (410°F) (581°F)

Fluid type Helium Helium Air Fluid molecular weight 4 g/mol 4 g/mol 29 g/mol Fluid downstream pressure1, 0.25 atm abs 0.25 atm abs 0.01 atm abs Pd Fluid viscosity µ(T) 2 0.0285 [6] 0.0322 [6] 0.0185 cP [4]

Average stream pressure 1.37 6.80 0.51 Pa =(Pu + Pd)/2 The calculation is performed in several steps that are described below:

• Step 1: Determine the source terms for the containment evaluation. Source terms are derived for normal conditions of transport (NCT) and hypothetical accident conditions (HAC).

• Step 2: Calculate effective A2 values for each transportation condition (NCT and HAC).

• Step 3: Using the effective A2 values and the source concentrations (source activity available for release in the cask), calculate the permissible leakage rate to satisfy 10 CFR 71 [1].

• Step 4: Using ANSI N14.5 [3], calculate the reference air leakage rates.

4.6.1.1.3 Assumptions Although the possibility of forming and releasing a crud-aerosol is minimal, a leakage rate can be calculated for this phenomenon. It is most probable that the crud particles will plug the small holes through which gas leakage usually occurs. However, it is conservatively assumed that the particulate leakage rate is the same as the gas leakage rate.

1 See Assumptions, Section 4.6.1.1.3.

2 Viscosities for helium at NCT and HAC temperatures were extrapolated from the outputs of [6] at various lower temperatures using a second order polynomial.

TN-LC-0100 4.6.1-2 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 As a conservative approach and to simplify calculations, it is assumed that crud spallation and cladding breaches occur instantaneously after fuel loading and container closure operations.

Therefore, the source term becomes time-independent, and all the radioactivity in the fill gas that is available for release from the containment vessel (should a leak occur) is available initially.

This assumption ensures the maximum amount of radioactive inventory is available should a leak occur just after closure.

The length of the leakage hole for calculation of the permissible leak rates is assumed to be 0.275 in (0.7 cm), which is the diameter of the seal cross-section.

The downstream pressure is conservatively taken as 0.25 atm abs for both the NCT and the HAC analyses.

It is conservatively assumed that the fuel pins experience cladding failure during NCT and HAC at rates of 5% and 100%, respectively.

4.6.1.2 Source Activities and Source Activity Densities 4.6.1.2.1 Source Activity Density Due to Crud Spallation From Fuel Rods When fuel rods are subject to the radioactive and corrosive environment of a PWR or BWR reactor, radioactive flaky material is formed on the outside surface of the fuel rods. Some of this material is loosely bound to the fuel rod surface and can be dislodged in some circumstances and possibly form an aerosol in the cavity atmosphere. Vibration and flowing gases for example have been shown to dislodge some of the particles, forming a powder aerosol in the surrounding gas, which could present a potential dispersion situation, although studies [2] show the crud to be very stable and adherent.

The typical chemical composition of materials in this aerosol would be primarily oxides of the constituents of reactor hardware metals including radionuclides such as Co-60. If there was any failed fuel in the reactor or storage pool along with the hardware, there might also be minute traces of fission products.

Measurements [5] have shown that the bounding value for crud surface activity for PWR and BWR rods is 1254 µCi/cm2 = 4.64 x 107 Bq/cm2 at discharge. Since 60Co dominates the crud activity [5], the A2 of 60Co (0.4 TBq) is used for the crud.

The surface area per rod is calculated based on the rod dimensions and the surface area associated with the assembly hardware is neglected [5].

The total surface area per rod is:

SAR = 168 x x 0.57 = 300.8 in2 = 1,940.9 cm2.

Measurements have also shown that 15% is a reasonable value for crud spallation for both PWR and BWR fuel rods under normal transport conditions. For hypothetical accident conditions, 100% crud spallation is assumed.

TN-LC-0100 4.6.1-3 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Therefore, the total activity released from crud spallation under NCT is equal to:

4.64 x 107 x 49 x 1,940.9 x 0.15 = 6.62 x 1011 Bq = 1.66 A2.

The total activity released from crud spallation under HAC is equal to:

4.64 x 107 x 49 x 1,940.9 x 1 = 4.41 x 1012 Bq = 11.03 A2.

Expressed in A2/cm3, the activity density inside the containment vessel is therefore:

1.66 NCT: 187,288 = 8.83 x 106 A2/cm3.

11.03 HAC: 187,288 = 5.89 x 105 A2/cm3.

4.6.1.2.2 Source Activity Density Due to Release of Fines From Cladding Breaches A breach in the cladding of a fuel rod may allow radionuclides to be released from the resulting cladding defect into the interior of the TN-LC. If there is a leak in the containment vessel, then the radioisotopes emitted from a cladding breach that were aerosolized can be entrained in the gases escaping from the TN-LC and result in a radioactive release to the environment.

Based on [5], the contribution of fines released to the total source term is less than 1% of the total. Therefore, the release of fines is neglected in this analysis.

4.6.1.2.3 Source Activity Density Due to Release of Gaseous Fission Products and Volatiles from Cladding Breaches Based on [5], the contribution of volatiles released to the total source term is approximately 1%

of the total. Therefore, the release of volatiles is neglected in this analysis.

The fission gases and their volumes are listed in Table 4.6.1-3 below. The maximum inventory over the first 10 years of cooling is conservatively considered.

Table 4.6.1-3 Gas Fission Products Volume Mol per Molar Assembly Element Mol/tU assembly1 volume (I) volume (I) 3 H 0.0092 0.0188 11.2 0.103 Br 0.1074 0.2191 11.2 1.203 Kr2 3.2506 6.6339 22.4 72.814 85 Kr 0.2313 0.4720 22.4 5.181 I 0.7254 1.4804 11.2 8.124 Xe 37.6468 76.8302 22.4 843.288 Total 41.971 - - 930.713 1

The B&W 15x15fuel assembly has 0.490 tU.

2 Except 85Kr.

TN-LC-0100 4.6.1-4 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Based on [5], it is conservatively assumed that 3% of the fuel rods develop a breach during normal conditions of transport, and 100% during hypothetical conditions of transport. It is also conservatively considered that 30% of the fission gases escape from the fuel pellet matrix.

The total volume of fission products gases per fuel assembly is 930.713 liters and the total quantity of fission product gases per fuel assembly is 41.971 mol.

4.6.1.2.4 Activities Among the various gaseous isotopes contained in the fuel assemblies, only 3H and 85Kr are important [5], and the other isotopes are assumed to exist in very small quantities due to their short half-lives.

The activities of 3H and 85Kr are given in Table 4.6.1-4 below.

Table 4.6.1-4 Gas Fission Products Activities 3 85 H Kr Specific Activity (Bq/g) 3.626 x 1014 1.480 x 1013 Activity Bq/tU 4.10 x 1013 5.93 x 1014 Activity / assembly1 (Bq) 2.011 x 1013 2.907 x 1014 4.6.1.2.5 Total Activity and Volume Released in Containment Cavity Due to Cladding Failure (NCT)

According to Section 4.6.1.2.3, 30% of the 3H and 85Kr is released from the fuel, and 5% of the fuel cladding fails in NCT. Therefore, the following activity and volume of fission products gases are released in the containment cavity following a cladding failure.

Total volume of fission gases released in the containment cavity in NCT:

VRel_fg-NCT = 930.713 x 0.30 x 0.05 14 liters = 14,000 cm3.

Table 4.6.1-5 Gas Fission Products Released Activities - NCT (A2) 3 85 H Kr Total activity released into the package 3.017 x 1011 4.360 x 1012 containment cavity (Bq)

A2 (Bq) 4.0 x 1013 1.00 x 1014 Activity released into the package 0.0075 0.0436 containment cavity (A2) 1 This value is the same as the activity / package since there is only one assembly per package.

TN-LC-0100 4.6.1-5 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The total activity ANCT of gaseous fission products released into the package containment cavity during NCT is 0.051 A2. The activity density is 0.051 / 187,288 = 2.73 x 10-7 A2/cm3.

4.6.1.2.6 Total Activity and Volume Released in Containment Cavity Due to Cladding Failure (HAC) 30% of 85Kr is released from the fuel, and 100% of the fuel cladding fails in HAC. Therefore, the following activity and volume of fission products gases are released in the containment cavity following a cladding failure.

Total volume of fission gases released in the containment cavity in HAC:

VRel_fg-HAC= 930.713 x 0.30 x 1 280 liters = 280,000 cm3.

Table 4.6.1-6 Gas Fission Products Released Activies - HAC (A2) 3 85 H Kr Total activity released into the package 12 6.034 x 10 8.720 x 1013 containment cavity (Bq)

A2 (Bq) 4.0 x 1013 1.00 x 1014 Activity released into the package 0.1509 0.8720 containment cavity (A2)

The total activity AHAC of gaseous fission products released into the package containment cavity during HAC is 1.023 A2. The activity density is 1.023 / 187,288 = 5.46 x 10-6 A2/cm3.

4.6.1.2.7 Summary of Activities and Activity Densities Table 4.6.1-7 Activities and Activity Densities per Release Type NCT HAC Release Type A2 A2/cm3 A2 A2/cm3 Crud 1.66 8.83 x 10-6 11.03 5.89 x 10-5 Fission Gases 0.05 2.73 x 10-7 1.02 5.46 x 10-6 Total 1.71 CN = 9.11 x 10-6 12.05 CA = 6.44 x 10-5 4.6.1.3 Determination of allowable leakage rates 4.6.1.3.1 Determine the allowable release rate for NCT and HAC The allowable release rate for NCT, RN, is calculated as:

= 106 2 = 2.78 x 1010 2 / .

TN-LC-0100 4.6.1-6 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 The allowable release rate for HAC, RA, is calculated as:

= 2 = 1.65 x 106 2 /.

4.6.1.3.1.1 NCT Allowable Leak Rate, LN The following equation from ANSI N14.5 [3] is used to calculate the permissible volumetric gas release rate at cask conditions:

= .

Where CN is the average activity concentration of the gas available for release inside the cask previously calculated (in A2/cm3) and RN is the allowable release rate set above.

The maximum allowable leakage rate during normal condition of transport is:

2.78x1010

= = = 3.1 x 105 3 /.

9.11x106 4.6.1.3.1.2 HAC Allowable Leak Rate, LA The maximum allowable leakage rate during HAC is:

1.65x105

= = 6.44x105 = 2.6 x 102 3 /.

4.6.1.3.2 Determine the reference air leakage rates 4.6.1.3.2.1 NCT Reference Air Leakage Rate, LRN The air leakage rate at standard conditions that is equivalent to LN is determined following the example 21 in ANSI N14.5 [3]. The following formula is used for calculating the reference air leakage rate, where Lu is the upstream volumetric leakage rate in cm3/sec LN calculated above:

= ( + )( ) x (5.1)

Fc is the coefficient of continuum flow conductance per unit pressure (cm3/atm-sec):

2.49x106 4

= (5.2)

Fm is the coefficient of free molecular flow conductance per unit pressure (cm3/atm-sec):

0.5 3.81x103 3

=

(5.3)

Where:

• D is the leakage hole diameter (cm),

TN-LC-0100 4.6.1-7 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

• a is the leakage hole length (cm),

• µ is the fluid viscosity (cP),

• T is the fluid absolute temperature,

• M is the molecular weight (g/mol),

• Pu is the upstream fluid pressure (atm abs),

• Pd is the downstream fluid pressure (atm abs), and

• Pa is the average stream pressure (atm abs).

The value for LN calculated in Section 4.6.1.3.1.1 is taken equal to the maximum leakage rate LU in equation (5.1). All the parameters are known except the leakage hole diameter D. The diameter is found by iteratively solving the equation (5.1).

For NCT, LU = LN = 3.1 x 10-5 cm3/sec. Using the values given for NCT, the resulting leakage hole diameter D is equal to 5.95 x 10-4 cm.

The Fc and Fm values using this diameter are calculated for helium using equations (5.2) and (5.3) as equal to 1.6 x 10-5 and 9.2 x 10-6 respectively, and LN is verified using the value for D in equation (5.1) as 3.1 x 10-5 cm3/sec.

Using this value of D, the resulting standard air leakage rate is calculated using the same equation (5.1), but with parameters corresponding to air at standard conditions of temperature and pressure. The resulting NCT reference air leakage rate LRN is calculated as:

= 2.4 x 105 ;

= 7.3 x 106 ;

= 1.57 x 105 cm3/sec.

4.6.1.3.2.2 HAC Reference Air Leakage Rate, LRA Similar to what was done in the previous section, the value for LA calculated in Section 4.6.1.3.1.2 is now taken equal to the maximum leakage rate LU in equation (5.1). All the parameters are known except the leakage hole diameter. The diameter is found by iteratively solving equation (5.1).

For HAC, LU = LA = 2.6 x 10-2 cm3/sec.

Using the same iterative calculation method as above for HAC, the resulting leakage hole diameter D is equal to 2.41 x 10-3 cm.

LA is verified using the value for D in equation 5.1 as 2.6 x 10-2 cm3/sec.

Using this value of D, the resulting standard air leakage rate is calculated using the same equation (5.1), but with parameters corresponding to air at standard conditions of temperature and pressure. The resulting HAC reference leakage rate LRA is calculated as:

TN-LC-0100 4.6.1-8 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

= 6.5 x 103 ;

= 4.8 x 104 ;

= 3.48 x 103 cm3/sec.

4.6.1.4 Results The NCT reference air leakage rate LRN is the most restrictive of the values determined for NCT and HAC, therefore, LR = LRN = 1.57 x 10-5 ref cm3/sec.

4.6.1.5 Conclusions The following leak rates are specified for the TN-LC transportation package tests in accordance with ANSI N14.5, Section 7 [3]:

Maintenance and periodic verification tests shall determine that the leak rate for the cask is no greater than LR = 1.57 x 10-5 ref cm3/sec with a test sensitivity better than 7.83 x 10 6 ref cm3/sec.

Pre-shipment verification tests shall demonstrate no detectable leakage when tested to a sensitivity of at least 10-3 ref cm3/sec per Section 7 [3]

4.6.1.6 References

1. 10 CFR 71, Packaging and Transportation of Radioactive Material.

2. Hazelton, R.F., Characteristics of Fuel Crud and its Impact On Storage, Handling and Shipment of Spent Fuel, PNL-6273 Pacific Northwest Laboratories, September 1987.

3. ANSI N14.5-2014,American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, June 2014.

4. Engineering ToolBox - Resources, Tools and Basic Information for Engineering and Design of Technical Applications !, Air - Dynamic and Kinematic Viscosity https://www.engineeringtoolbox.com/air-absolute-kinematic-viscosity-d_601.html?vA=298&units=K#

5. NUREG/CR-6487, Containment Analysis for Type B Packages Used to Transport Various Contents, LLNL, November 1996.

6. Fluid Properties Calculator, http://www.mhtl.uwaterloo.ca/old/onlinetools/airprop/airprop.html.

TN-LC-0100 4.6.1-9 Appendix 4.6.1 is newly added in SAR Revision 9a.

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Chapter 5 Shielding Evaluation NOTE: References in this Chapter are shown as [1], [2], etc. and refer to the reference list in Section 5.5.

This chapter presents the shielding evaluation of the TN-LC transportation package. The dose rates are evaluated per the requirements of 10CFR71.47 and 71.51 for exclusive use transportation in a closed transport vehicle.

The dose rates are evaluated for the four basket types using MCNP5 v1.40 [1]. Information common to all analyses are summarized in the main body of this chapter. The details for each analysis are contained in a separate appendix. The list of appendices is as follows:

TN-LC-MTR Basket: Appendix 5.6.1.

TN-LC-NRUX Basket: Appendix 5.6.2.

TN-LC-TRIGA Basket: Appendix 5.6.3.

TN-LC-1FA Basket: Appendix 5.6.4.

5.1 Description of the Shielding Design 5.1.1 Design Features The TN-LC cask is radially and axially shielded with steel and lead. [

] The neutron shield may be either Resin-F or VYAL-B. [

]

The TN-LC Unit 01 as-fabricated cask body has a reduction in its shielding capability due to localized areas where the radial lead thickness may be as low as 3.10 inches. See Appendices 1.4.1 and 5.6.4 for further details.

The description of the shielding design for the four basket types are contained in the individual appendices for each basket.

5.1.2 Summary Tables of Maximum Radiation Levels Normal conditions of transport (NCT) dose rates are computed for exclusive use transport in a closed transport vehicle. These dose rate limits are as follows:

Surface of the package: 1000 mrem/hr Surface of the transport vehicle: 200 mrem/hr TN-LC-0100 5-1

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 2 m from the surface of the transport vehicle: 10 mrem/hr The transport vehicle is assumed to be 8 ft wide. Because the TN-LC is a long package, the ends of the transport vehicle are conservatively assumed to be at the ends of the impact limiters. The underside (floor) of the vehicle is conservatively assumed to correspond to the radius of the impact limiters. The dose rates on the vehicle roof are not computed, as these dose rates are bounded by the dose rates on the underside of the vehicle.

Dose rates are computed 2 m from the sides and ends of the vehicle. The dose rates in an occupied location are estimated to correspond to 2 m from the ends of the vehicle. As these dose rates exceed the limit of 2 mrem/hr for some baskets, per 10CFR71.47(b)(4), the TN-LC shall be transported by private carrier, and personnel in occupied locations shall wear dosimetry devices.

The NCT dose rates for each basket type are summarized in the following tables:

Table 5-1: Summary of TN-LC-MTR NCT Dose Rates Table 5-2: Summary of TN-LC-NRUX NCT Dose Rates Table 5-3: Summary of TN-LC-TRIGA NCT Dose Rates Table 5-4: Summary of TN-LC-1FA NCT Dose Rates Hypothetical accident condition (HAC) dose rates are computed for each basket. Under HAC, it is conservatively assumed that both the neutron shield and impact limiter wood is lost. Lead slump as a result of an accident is also included. Dose rates are computed 1 m from the surface of the cask body. The HAC dose rate limit is 1000 mrem/hr. The HAC dose rates for each basket type are summarized in the following tables:

Table 5-5: Summary of TN-LC-MTR HAC Dose Rates Table 5-6: Summary of TN-LC-NRUX HAC Dose Rates Table 5-7: Summary of TN-LC-TRIGA HAC Dose Rates Table 5-8: Summary of TN-LC-1FA HAC Dose Rates 5.1.3 Licensing Basis

[

]

TN-LC-0100 5-2

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 5.2 Source Specification The TRITON module of the SCALE6 code package [2] is used to compute the gamma and neutron source terms for MTR, NRU/NRX, TRIGA, BWR and PWR fuel assemblies and rods.

For each fuel type, a bounding source term is developed for NCT analysis. For MTR, NRU/NRX, and TRIGA fuels, the same source term may be used for NCT and HAC analysis This is because the contribution of neutron radiation sources to the total dose rate is either negligible (see, for example, Table 5-1 or Table 5-3), the total dose rate at some or all of the locations of interest are substantially lower than the regulatory limits (see Table 5-2), or both.

For BWR/PWR fuel, the contribution of neutron radiation source to the total dose rate may be large, and separate HAC source terms are developed because the loss of the neutron shield increases the neutron contribution to the dose rate.

A detailed discussion of the source specification for each fuel type is contained in the individual appendices for each basket.

5.2.1 Crud Evaluation for Shielding

[

]

TN-LC-0100 5-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 TN-LC-0100 5-3a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 impact limiter radius (underside of vehicle) and the side of the vehicle, and Locations 1 through 29 are utilized 2 m from the side of the vehicle.

At the ends of the packaging, dose rates are tallied on the impact limiter surfaces and 2 m from the impact limiter surfaces. Eight radial locations are utilized for the surface tallies, as illustrated in Figure 5-6. Location 1 captures any streaming effects from the bottom plug assembly. An off-center tally is used directly over the lid port to capture any streaming effects on the top impact limiter surface. For the dose rates 2 m from the ends, five radial locations are utilized by combining Locations 1,2,3 and 4,5, as shown on Figure 5-6. The effect of end streaming through the lid port and bottom plug assembly is investigated only for the 1FA basket because this basket results in the highest dose rates through the ends of the package. The effects of streaming 2 m from the ends of the transport vehicle are shown to be small.

Because the basket designs are not circumferentially symmetric, the dose rate will vary around the perimeter of the package. This effect is most pronounced close to the package surface, and diminishes with distance. Close to the surface of the package, this variation is approximately 15 percent from the average in most cases. At 2 m from the surface of the vehicle, this variation is typically small (~5 percent). Because the dose rates near the radial surface of the cask are significantly below the dose rate limits, a detailed tally to capture these angular effects is not warranted. Therefore, circumferential average tallies are reported in the radial direction for most dose rate locations, and are supplemented with more detailed mesh tally results only when necessary.

However, because the impact limiter attachments and the shear key penetrate the neutron shield and displace neutron shielding material, there may be neutron streaming at these locations. This effect is captured explicitly using angular mesh tallies. The streaming effect is more pronounced at the location of the shear key because it is at the axial center of the cask. The impact limiter attachments are near the top and bottom of the cask where the neutron source is typically much smaller.

The shear key faces downward and results in a higher than average neutron dose rate on the surface of the package and the underside of the vehicle. For this reason, at these dose rate locations, the circumferential tally results are supplemented by neutron mesh tally results at the shear key. The shear key and impact limiter attachments do not result in gamma streaming because neutron shielding material is replaced with steel, which is a superior gamma shield.

Therefore, these features are conservatively omitted in the gamma models.

Circumferential average dose rates are reported 2 m from the surface of the transport vehicle, which is the limiting location for dose rates. The cask pay load with 1 LWR fuel assembly, referred to as 1 FA throughout this chapter and its appendices, results in 2 meters from side of the cask radial dose rates that are bounding for those due to other payloads of the cask. Because of this, the circumferential averaged tallies are supplemented with mesh tallies at this location in the shielding analysis model of the cask with 1 LWR FA payload. It is demonstrated that at 2 m from the vehicle, the angular fluctuation of the dose rate is small (typically ~5 percent from the average value).

In the HAC models, the neutron shield resin and impact limiter wood is replaced with air, and the tally surfaces are located 1 m from the outer surfaces of the cask. Lead slump is modeled in the cask-body lead at both ends of the cask. Radial lead slump in the lead disks in the lid and TN-LC-0100 5-9

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 bottom ends are modeled in the 1FA models only, although the effect on the dose rates is negligible. The dose rates at the ends of the package are divided into three radial segments, and the dose rates at the side of the package are divided in 25 axial segments of equal width. The tally locations are shown on Figure 5-7.

The detailed results are provided in the individual appendices.

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The TN-LC Unit 01 is the same as the TN-LC with the exception of localized reduced lead thickness on the side of the cask body. The shielding assessment of the TN-LC with a uniformly reduced lead thickness of 3.10 in. is performed in Appendix A.5.4.4.4.5.

TN-LC-0100 5-10

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Appendix 5.6.4 TN-LC-1FA Basket Shielding Evaluation TABLE OF CONTENTS 5.6.4.1 Description of the Shielding Design ...................................................................... 5.6.4-1 5.6.4.1.1 Design Features ............................................................................................. 5.6.4-1 5.6.4.1.2 Summary Tables of Maximum Radiation Levels ............................................ 5.6.4-1 5.6.4.2 Source Specification .............................................................................................. 5.6.4-3 5.6.4.2.1 Fuel Qualification .......................................................................................... 5.6.4-4 5.6.4.2.2 MCNP Response Functions ........................................................................... 5.6.4-5 5.6.4.2.3 Design Basis Gamma and Neutron Source Terms ......................................... 5.6.4-7 5.6.4.2.4 Axial Blankets ................................................................................................ 5.6.4-9 5.6.4.2.5 Validation of the Source Term Methodology ................................................. 5.6.4-9 5.6.4.3 Shielding Model ................................................................................................... 5.6.4-10 5.6.4.3.1 Configuration of Source and Shielding........................................................ 5.6.4-10 5.6.4.3.2 Material properties ...................................................................................... 5.6.4-12 5.6.4.4 Shielding Evaluation ............................................................................................ 5.6.4-13 5.6.4.4.1 Methods ........................................................................................................ 5.6.4-13 5.6.4.4.2 Input and Output Data ................................................................................. 5.6.4-13 5.6.4.4.3 Flux-to-Dose-Rate Conversion .................................................................... 5.6.4-13 5.6.4.4.4 External Radiation Levels ............................................................................ 5.6.4-13 5.6.4.4.5 Reduced Lead Thickness Assessment ......................................................... 5.6.4-18a 5.6.4.5 Appendices ........................................................................................................... 5.6.4-19 5.6.4.5.1 References .................................................................................................... 5.6.4-19 5.6.4.5.2 Sample ORIGEN-ARP Input File................................................................. 5.6.4-20 5.6.4.5.3 Sample MCNP Input File ........................................................................... 5.6.4-21a TN-LC-0100 5.6.4-i

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 LIST OF TABLES Table 5.6.4-1 Summary of TN-LC-1FA NCT Dose Rates .............................................. 5.6.4-34 Table 5.6.4-2 Summary of TN-LC-1FA HAC Dose Rates .............................................. 5.6.4-35 Table 5.6.4-3 PWR Fuel Assembly Data ........................................................................ 5.6.4-36 Table 5.6.4-4 BWR Fuel Assembly Data ........................................................................ 5.6.4-37 Table 5.6.4-5 PWR B&W 15x15 Exposure Zone Materials ........................................... 5.6.4-38 Table 5.6.4-6 BWR GE 7x7 Exposure Zone Materials .................................................. 5.6.4-38 Table 5.6.4-7 Material Element Composition ................................................................ 5.6.4-39 Table 5.6.4-8 Flux Factors............................................................................................. 5.6.4-40 Table 5.6.4-9 Fuel Qualification Table for a PWR Fuel Assembly ............................... 5.6.4-41 Table 5.6.4-10 Fuel Qualification Table for 25 PWR Fuel Rods .................................... 5.6.4-42 Table 5.6.4-11 Fuel Qualification Table for 9 PWR Fuel Rods ...................................... 5.6.4-43 Table 5.6.4-12 Fuel Qualification Table for a BWR Fuel Assembly ............................... 5.6.4-44 Table 5.6.4-13 Fuel Qualification Table for 25 BWR Fuel Rods .................................... 5.6.4-45 Table 5.6.4-14 Fuel Qualification Table for 9 BWR Fuel Rods ...................................... 5.6.4-46 Table 5.6.4-15 Fuel Qualification Table for MOX Fuel Rods ......................................... 5.6.4-47 Table 5.6.4-16 Gamma NCT Response Functions ........................................................... 5.6.4-48 Table 5.6.4-17 Neutron NCT Response Functions........................................................... 5.6.4-48 Table 5.6.4-18 PWR Fuel Assembly Axial Peaking Factors ............................................ 5.6.4-49 Table 5.6.4-19 BWR Fuel Assembly Axial Peaking Factors ............................................ 5.6.4-50 Table 5.6.4-20 PWR Fuel Assembly Design Basis Gamma Source Terms ...................... 5.6.4-51 Table 5.6.4-21 BWR Fuel Assembly Design Basis Gamma Source Terms ...................... 5.6.4-52 Table 5.6.4-22 25 PWR Fuel Rods Design Basis Gamma Source Terms ........................ 5.6.4-53 Table 5.6.4-23 25 BWR Fuel Rods Design Basis Gamma Source Terms ........................ 5.6.4-54 Table 5.6.4-24 25 MOX Fuel Rods Design Basis Gamma Source Terms........................ 5.6.4-55 Table 5.6.4-25 9 PWR Fuel Rods Design Basis Gamma Source Terms .......................... 5.6.4-56 Table 5.6.4-26 9 BWR Fuel Rods Design Basis Gamma Source Terms .......................... 5.6.4-57 Table 5.6.4-27 9 MOX Fuel Rods Design Basis Gamma Source Terms.......................... 5.6.4-58 Table 5.6.4-28 Design Basis Neutron Sources................................................................. 5.6.4-59 Table 5.6.4-29 Fuel Dimensions and Densities for MCNP Models ................................. 5.6.4-60 Table 5.6.4-30 TN-LC-1FA Basket Model Dimensions ................................................... 5.6.4-61 Table 5.6.4-31 Composition of Fuel ................................................................................ 5.6.4-61 Table 5.6.4-32 NCT Dose Rate Summary (mrem/hr) ....................................................... 5.6.4-62 Table 5.6.4-33 PWR Fuel Assembly NCT Side Surface Dose Rates between Impact Limiters (mrem/hr)................................................................................... 5.6.4-63 Table 5.6.4-34 PWR Fuel Assembly NCT Vehicle Underside Dose Rates (mrem/hr) ..... 5.6.4-64 Table 5.6.4-35 PWR Fuel Assembly NCT Vehicle Side Dose Rates (mrem/hr) ............... 5.6.4-65 Table 5.6.4-36 PWR Fuel Assembly NCT 2 m from Vehicle Side Dose Rates (mrem/hr) ................................................................................................. 5.6.4-66 Table 5.6.4-37 PWR Fuel Assembly NCT End Dose Rates (mrem/hr) ............................ 5.6.4-67 Table 5.6.4-38 BWR Fuel Assembly NCT Side Surface Dose Rates between Impact Limiters (mrem/hr)................................................................................... 5.6.4-68 Table 5.6.4-39 BWR Fuel Assembly NCT Vehicle Underside Dose Rates (mrem/hr) ..... 5.6.4-69 Table 5.6.4-40 BWR Fuel Assembly NCT Vehicle Side Dose Rates (mrem/hr) ............... 5.6.4-70 TN-LC-0100 5.6.4-ii

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-41 BWR Fuel Assembly NCT 2 m from Vehicle Side Dose Rates (mrem/hr) ................................................................................................. 5.6.4-71 Table 5.6.4-41a BWR Fuel Assembly NCT End Dose Rates (mrem/hr) ............................ 5.6.4-72 Table 5.6.4-42 25 EPR Rods NCT Side Surface Dose Rates between Impact Limiters (mrem/hr)................................................................................... 5.6.4-73 Table 5.6.4-43 25 EPR Rods NCT Vehicle Underside Dose Rates (mrem/hr) ................ 5.6.4-74 Table 5.6.4-44 25 EPR Rods NCT Vehicle Side Dose Rates (mrem/hr) .......................... 5.6.4-75 Table 5.6.4-45 25 EPR Rods NCT 2 m from Vehicle Side Dose Rates (mrem/hr)........... 5.6.4-76 Table 5.6.4-46 25 EPR Rods NCT End Dose Rates (mrem/hr) ....................................... 5.6.4-77 Table 5.6.4-47 PWR Fuel Assembly HAC Dose Rates, Fuel Shifted Down (mrem/hr) ................................................................................................. 5.6.4-78 Table 5.6.4-48 PWR Fuel Assembly HAC Dose Rates, Fuel Shifted Up (mrem/hr) ........ 5.6.4-79 Table 5.6.4-49 25 MOX Rods HAC Dose Rates (mrem/hr) ............................................. 5.6.4-80 Table 5.6.4-50 PWR Fuel Assembly HAC Dose Rates, Fuel Rubble (mrem/hr) ............. 5.6.4-81 Table 5.6.4-51 BWR Dry keff ............................................................................................ 5.6.4-82 Table 5.6.4-52 PWR Dry keff ............................................................................................ 5.6.4-83 Table 5.6.4-53 C/M Results for Samples from Vandells II ............................................ 5.6.4-84 Table 5.6.4-54 C/M Results for Samples from Takahama and TMI ................................ 5.6.4-85 Table 5.6.4-55 Uncertainties Applied to C/M Data (%) .................................................. 5.6.4-88 Table 5.6.4-56 Isotopic Contribution to Total Dose Rate at Various Cooling Times ..... 5.6.4-88 Table 5.6.4-57 Average and Standard Deviation for C/M Values ................................... 5.6.4-89 Table 5.6.4-58 Dose Rates Computed with C/M Average minus Standard Deviation .... 5.6.4-89 Table 5.6.4-59 NCT Response Functions for 3.10 Lead Thickness - PWR Fuel Assembly ................................................................................................ 5.6.4-89a Table 5.6.4-60 NCT Response Functions for 3.10 Lead Thickness - 25 PWR rods in Pin Can .............................................................................................. 5.6.4-89b Table 5.6.4-61 Fuel Qualification Table for a PWR Fuel Assembly - 3.10 Lead Thickness Cooling Time (years) ............................................................ 5.6.4-89c Table 5.6.4-62 Fuel Qualification Table for 25 PWR Fuel Rods - 3.10 Lead Thickness Cooling Time (years) ............................................................ 5.6.4-89d Table 5.6.4-63 PWR Fuel Assembly Design Basis Gamma Source Terms - 3.10 Lead Thickness....................................................................................... 5.6.4-89e Table 5.6.4-64 25 PWR Rods Design Basis Gamma Source Terms - 3.10 Lead Thickness................................................................................................. 5.6.4-89f Table 5.6.4-65 Design Basis Neutron Sources - 3.10 Lead Thickness ........................ 5.6.4-89g Table 5.6.4-66 NCT Dose Rate Summary (mrem/hr) - 3.10 Lead Thickness .............. 5.6.4-89h LIST OF FIGURES Figure 5.6.4-1 TN-LC-1FA PWR Fuel Assembly MCNP Model, y-z View ..................... 5.6.4-90 Figure 5.6.4-2 TN-LC-1FA BWR Fuel Assembly MCNP Model, y-z View ..................... 5.6.4-91 Figure 5.6.4-3 TN-LC-1FA 25 EPR Rod MCNP Model, y-z View .................................. 5.6.4-92 Figure 5.6.4-4 TN-LC-1FA MCNP Model, x-y View through Shear Key........................ 5.6.4-93 Figure 5.6.4-5 TN-LC-1FA MCNP Model, x-y View through Impact Limiter Attachment Blocks ................................................................................... 5.6.4-94 Figure 5.6.4-6 TN-LC-1FA NCT Radial Surface Tallies ................................................. 5.6.4-95 TN-LC-0100 5.6.4-iii

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Figure 5.6.4-7 TN-LC-1FA NCT End Surface Tallies ..................................................... 5.6.4-96 Figure 5.6.4-8 TN-LC-1FA HAC 1 m Tallies .................................................................. 5.6.4-97 Figure 5.6.4-9 Percent Contribution of La-140 to Total Dose Rates .............................. 5.6.4-98 Figure 5.6.4-10 Total Gamma Dose Rate as Function of Natural Log (Time) ................. 5.6.4-99 Figure 5.6.4-11 Peaking Factors for BWR fuel ............................................................... 5.6.4-100 Figure 5.6.4-12 Peaking Factors for PWR fuel ............................................................... 5.6.4-101 Figure 5.6.4-13 Interpolation for Dry keff ........................................................................ 5.6.4-102 TN-LC-0100 5.6.4-iv

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Package surface: The NCT package side surface dose rates for the PWR fuel assembly are presented in Table 5.6.4-33. Dose rates on the side surfaces of the impact limiters are presented in Table 5.6.4-34 at axial locations 6, 7, 23, and 24. Dose rates on the external flat surfaces of the impact limiters are presented in Table 5.6.4-37. The maximum package surface dose rate occurs on the side of the cask at the shear key, with a dose rate of 508 mrem/hr. This is an NCT dose rate, although it is computed using the HAC source by scaling the NCT dose rate. This dose rate is less than the limit of 1000 mrem/hr.

To estimate the dose rate at the shear key using the HAC source, the neutron scaling factor is the ratio of the HAC to NCT neutron source magnitudes, or 2.256E+09/2.078E +09= 1.1. The gamma scaling factor is estimated by using the PWR fuel assembly response function to compute dose rates for the two in-core gamma sources and taking the ratio, or 2.12/2.58 = 0.82. Using the shear key results for the NCT source from Table 5.6.4-33, the dose rate using the HAC source is estimated as 0.82*23.2 + 1.1*(446+4.79) = 508 mrem/hr.

Vehicle surface: The NCT vehicle underside/impact limiter radius dose rates are presented in Table 5.6.4-34. The maximum dose rate on the vehicle underside occurs at the shear key with a dose rate of 119 mrem/hr. This is an NCT dose rate, although it is computed using the HAC source by scaling the NCT dose rate using the factors shown above. The NCT vehicle side surface dose rates are presented in Table 5.6.4-35 and are bounded by the dose rates on the underside of the vehicle. NCT vehicle end dose rates are presented in Table 5.6.4-37. The maximum vehicle surface dose rate occurs on the impact limiter surface at the bottom center of the package. This dose rate is 92.8 mrem/hr, which bounds the vehicle surface dose rates on the underside, side, and top end. This dose rate is less than the limit of 200 mrem/hr.

2 m from vehicle surface: The NCT dose rates 2 m from the side surface of the vehicle are presented in Table 5.6.4-36. The maximum dose rate of 8.92 mrem/hr is computed using a mesh tally. This dose rate is less than the limit of 10 mrem/hr and bounds the dose rates 2 m from the ends of the vehicle presented in Table 5.6.4-37.

5.6.4.4.4.2 NCT, BWR Fuel Assembly Dose rates are reported for fuel shifted up in all BWR fuel assembly results tables, with the exception of dose rates at the bottom end, which are for fuel shifted down. Radial dose rates are slightly higher when the fuel is shifted up.

Package surface: The NCT package side surface dose rates for the BWR fuel assembly are presented in Table 5.6.4-38. Dose rates on the side surfaces of the impact limiters are presented in Table 5.6.4-39 at axial locations 6, 7, 23, and 24. The maximum package surface dose rate occurs on the side of the cask at the shear key, with a dose rate of 374 mrem/hr. This is an NCT dose rate, although it is computed using the HAC source by scaling the NCT dose rate. This dose rate is less than the limit of 1000 mrem/hr.

To estimate the dose rate at the shear key using the HAC source, the neutron scaling factor is the ratio of the HAC to NCT neutron source magnitudes, or 1.325E+09/1.313E+09 = 1.0. The TN-LC-0100 5.6.4-16

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 HAC dose rate results for 25 BWR MOX rods are provided in Table 5.6.4-49. Because axial shifting of the PWR fuel assembly did not result in peak dose rates near the lead slump regions, the HAC models for 25 MOX rods only have the fuel modeled at the bottom. The maximum radial dose rate using the 25 BWR MOX rods is 353 mrem/hr, which is larger than when the PWR fuel assembly source is utilized. The BWR MOX rods result in slightly larger dose rates at the bottom of the package compared to the PWR source, although the dose rates through the ends are negligible for either scenario.

HAC dose rates are computed with the PWR fuel assembly shifted down when the active fuel region is rubblized during an accident. It is assumed the volume of active fuel region of the assembly reduces by 50% with a corresponding homogenized density increase. The results are provided in Table 5.6.4-50. The maximum radial dose rate is 465 mrem/hr, which is the bounding HAC dose rate for all configurations. Therefore, the HAC dose rates limit of 1000 mrem/hr at 1 meter is met for all TN-LC-1FA sources.

5.6.4.4.5 Reduced Lead Thickness Assessment The gamma shield minimal lead thickness is 3.38 inches; during fabrication, and prior to the installation of the neutron shield, gamma scanning is used to verify the integrity of the poured lead shielding and a minimum thickness is confirmed by comparison of gamma scan results to a calibration block consisting of a known thickness of lead between steel plates of nominal thickness the same as used in the cask fabrication.

This section provides a comprehensive shielding evaluation for a reduced lead thickness, i.e.

lead thickness below the minimal lead thickness of 3.38 inches, for the 1FA basket with 1 PWR fuel assembly and 25 PWR rods in 1 pin can. The assumed lead thickness is 3.10 inches.

The shielding evaluation is consistent with the shielding evaluation performed in Section 5.6.4.4.4 for a minimal lead thickness is of 3.38 inches. The shielding evaluation for the reduced lead thickness consists of updated response functions, FQTs, bounding NCT and HAC sources for the 1 PWR fuel assembly and 25 PWR rods contents and shielding analysis of the 1 PWR fuel assembly and 25 PWR rods contents.

5.6.4.4.5.1 Response Functions The response functions models are identical to those developed in Section 5.5.4.2.2 with the exception of the lead thickness set to 3.10 inches. The updated response functions are developed for the:

- 1 PWR fuel assembly

- 25 PWR rods in 1 pin can Gamma and neutron response functions, at 2 m, for the 3.10 inches lead thickness configuration are provided in Table 5.6.4-59 and Table 5.6.4-60 for respectively 1 PWR fuel assembly and 25 PWR rods in a pin can.

TN-LC-0100 5.6.4-18a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 5.6.4.4.5.2 Fuel Qualification The fuel qualification tables, matrix of burnup, enrichment and cooling time, are generated using the response functions at 2m determined in Section 5.6.4.4.5.1 as the dose rate 2 m from the side of the vehicle is typically the limiting dose rate.

The methodology employed is identical to that described in Section 5.6.4.2.1. The updated FQT are developed for the:

- 1 PWR fuel assembly

- 25 PWR rods in a pin can For PWR fuel assembly, the maximum burnup is 62 GWD/MTU. For fuel rods, the maximum burnup is 90 GWD/MTU. The U-235 enrichment varies between 0.9 and 5.0 wt.%.

FQTs for each of the fuel types are provided in the following tables:

- Table 5.6.4-61, FQT for a single PWR fuel assembly

- Table 5.6.4-62, FQT for 25 PWR fuel rods The cooling times determined in the FQTs ensure the decay heat limits are met and the dose rates 2 m from the side of the vehicle are below 8 mrem/hr for PWR fuel assembly and 7.65 mrem/hr for 25 PWR rods pin can. ORIGEN-ARP models, fuel hardware and irradiation parameters are identical to those described in Section 5.6.4.2.1.

5.6.4.4.5.3 Bounding Gamma and Neutron Source Terms Once FQTs and design basis burnup, enrichment, and cooling time combinations have been established, design basis source terms are developed based on identical ORIGEN-ARP models as described in Section 5.6.4.2.3.

The bounding NCT and HAC gamma source terms for the various fuel types and quantities are summarized in Table 5.6.4-63 and Table 5.6.4-64 as follow:

- Table 5.6.4-63: PWR fuel assembly, NCT assembly average burnup of 61 GWD/MTU, enrichment of 3.2 wt%, and cooling time of 13.3 years; HAC assembly average burnup of 62 GWD/MTU, enrichment of 2.6 percent, and cooling time of 16.1 years

- Table 5.6.4-64: 25 PWR fuel rods, NCT rod peak burnup of 10 GWD/MTU, enrichment of 0.8 wt%, and cooling time of 69.51 days or 0.19 year; HAC rod peak burnup of 90 GWD/MTU enrichment of 3.7 percent, and cooling time of 1.5 years The corresponding neutron sources for the two contents are summarized in Table 5.6.4-65.

TN-LC-0100 5.6.4-18b

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 5.6.4.4.5.4 Shielding Analysis The shielding analysis is performed for the 1 PWR fuel assembly and 25 PWR rods in 1 pin can contents as described in Sections 5.6.4.4.1, 5.6.4.4.2, 5.6.4.4.3 and 5.6.4.4.4. The shielding model is identical to that described in Section 5.6.4.3 with the exception of the lead thickness set to 3.10 inches.

In summary, the following NCT analyses are performed:

- PWR fuel assembly, shifted either up or down. Note that fuel reconfiguration in NCT is not performed as it was shown in Section 5.6.4.4.4.4 that NCT fuel reconfiguration did not result in higher dose rate than NCT dose rate.

- 25 PWR rods axially centered in the long-cavity pin can HAC assumptions are identical to those described in Section 5.6.4.3.1 with the exception of the lead thickness set to 3.10 inches. HAC analysis is performed for the 1 PWR fuel assembly with and without fuel reconfiguration and for the 25 PWR rods pin can.

The summary of the NCT dose rates for the 1 PWR fuel assembly and 25 PWR rods pin can is provided in Table 5.6.4-66.

5.6.4.4.5.4.1 NCT, PWR Fuel Assembly Dose rates are computed for fuel shifted up in all PWR fuel assembly results tables, with the exception of dose rates at the bottom end, which are for fuel shifted down. Radial dose rates are slightly higher when the fuel is shifted up.

Package surface: The maximum package surface dose rate occurs on the side of the cask at the shear key, with a dose rate of 477 mrem/hr. This is an NCT dose rate, although it is computed using the HAC source by scaling the NCT dose rate. This dose rate is less than the limit of 1000 mrem/hr.

Vehicle surface: The maximum dose rate on the vehicle underside occurs at the shear key with a dose rate of 116 mrem/hr. The maximum vehicle surface dose rate occurs on the impact limiter surface at the bottom center of the package. This dose rate is 75.2 mrem/hr, which bounds the vehicle surface dose rates on the underside, side, and top end. This dose rate is less than the limit of 200 mrem/hr.

2 m from vehicle surface: The maximum dose rate of 8.82 mrem/hr is computed using a mesh tally. This dose rate is less than the limit of 10 mrem/hr and bounds the dose rates 2 m from the ends of the vehicle.

TN-LC-0100 5.6.4-18c

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 5.6.4.4.5.4.2 HAC, PWR Fuel Assembly PWR fuel assembly HAC cases are developed for the fuel assembly shifted up or down, as shifting the fuel places the nozzles closer to the lead slump regions that may form under HAC.

The lead slump has little effect on the dose rate because the dose rate is dominated by neutrons and peaks near the axial center. The maximum side dose rate for the PWR fuel assembly is 305 mrem/hr.

HAC reconfiguration is performed with the PWR fuel assembly shifted down when the active fuel region is rubblized during an accident assuming the volume of active fuel region of the assembly reduced by 50% with a corresponding homogenized density increase, see Section 5.6.4.4.4.5.

The maximum radial dose rate is 413 mrem/hr.

5.6.4.4.5.4.3 NCT, 25 PWR rods Pin Can Package surface: The maximum package surface dose rate occurs on the side of the cask at the shear key with a dose rate of 100 mrem/hr. This is an NCT dose rate, although it is computed using the HAC source by scaling the NCT dose rate. This dose rate is less than the limit of 1000 mrem/hr.

Vehicle surface: The maximum dose rate on the vehicle underside occurs at the shear key with a dose rate of 45.6 mrem/hr. This is an NCT dose rate, although it is computed using the HAC source by scaling the NCT dose rate using the factors shown above. The NCT vehicle side surface dose rate is bounded by the dose rate on the underside of the vehicle. NCT vehicle end dose rates are bounded by the dose rates on the underside of the vehicle. The maximum vehicle surface dose rate is less than the limit of 200 mrem/hr.

2 m from vehicle surface: The maximum dose rate of 8.07 mrem/hr is computed using a mesh tally. This dose rate is less than the limit of 10 mrem/hr.

5.6.4.4.5.4.4 HAC, 25 PWR rods Pin Can The maximum HAC dose rate using the 25 PWR rods is 147 mrem/hr.

TN-LC-0100 5.6.4-18d

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-59 NCT Response Functions for 3.10 Lead Thickness - PWR Fuel Assembly Neutron 2nd Gamma Gamma 1G Gamma 2G Gamma 3G Gamma 4G (mrem/h) (mrem/h) (mrem/h) (mrem/h) (mrem/h) (mrem/h) angular 0.25 2.57E-09 2.36E-10 1.06E-14 6.06E-14 4.27E-13 8.14E-13 0.27 2.58E-09 2.35E-10 1.04E-14 6.19E-14 4.42E-13 8.30E-13 0.29 2.60E-09 2.36E-10 1.07E-14 6.10E-14 4.39E-13 8.24E-13 0.31 2.59E-09 2.35E-10 1.05E-14 6.07E-14 4.35E-13 8.18E-13 0.33 2.60E-09 2.36E-10 9.99E-15 5.91E-14 4.25E-13 8.01E-13 0.35 2.59E-09 2.38E-10 1.02E-14 5.90E-14 4.16E-13 7.94E-13 0.37 2.60E-09 2.36E-10 1.02E-14 5.88E-14 4.15E-13 7.87E-13 0.38 2.60E-09 2.36E-10 9.88E-15 5.66E-14 4.10E-13 7.82E-13 0.4 2.58E-09 2.37E-10 1.04E-14 5.81E-14 4.16E-13 7.93E-13 0.42 2.60E-09 2.35E-10 1.03E-14 5.96E-14 4.27E-13 8.02E-13 0.44 2.60E-09 2.32E-10 1.03E-14 6.07E-14 4.28E-13 8.21E-13 0.46 2.62E-09 2.37E-10 1.07E-14 6.14E-14 4.37E-13 8.28E-13 0.48 2.61E-09 2.36E-10 1.05E-14 6.13E-14 4.41E-13 8.35E-13 0.5 2.58E-09 2.34E-10 1.02E-14 6.05E-14 4.30E-13 8.12E-13 0.52 2.58E-09 2.35E-10 1.06E-14 6.15E-14 4.41E-13 8.32E-13 0.54 2.61E-09 2.37E-10 1.06E-14 6.14E-14 4.35E-13 8.24E-13 0.56 2.58E-09 2.36E-10 1.01E-14 5.99E-14 4.28E-13 8.14E-13 0.58 2.59E-09 2.35E-10 1.01E-14 5.94E-14 4.26E-13 8.03E-13 0.6 2.62E-09 2.37E-10 9.97E-15 5.82E-14 4.19E-13 7.93E-13 0.62 2.58E-09 2.36E-10 1.01E-14 5.86E-14 4.17E-13 7.82E-13 0.63 2.60E-09 2.37E-10 9.96E-15 5.78E-14 4.15E-13 7.82E-13 0.65 2.58E-09 2.39E-10 1.01E-14 5.78E-14 4.15E-13 7.89E-13 0.67 2.60E-09 2.36E-10 1.02E-14 5.99E-14 4.25E-13 8.00E-13 0.69 2.58E-09 2.37E-10 1.03E-14 6.10E-14 4.30E-13 8.10E-13 0.71 2.61E-09 2.38E-10 1.08E-14 6.08E-14 4.35E-13 8.29E-13 0.73 2.61E-09 2.38E-10 1.03E-14 6.07E-14 4.41E-13 8.32E-13 0.75 2.58E-09 2.36E-10 1.05E-14 5.97E-14 4.26E-13 8.07E-13 TN-LC-0100 5.6.4-89a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-60 NCT Response Functions for 3.10 Lead Thickness - 25 PWR rods in Pin Can Neutron 2nd Gamma Gamma 1G Gamma 2G Gamma 3G Gamma 4G (mrem/h) (mrem/h) (mrem/h) (mrem/h) (mrem/h) (mrem/h) angular 0.25 2.19E-09 2.47E-10 3.25E-15 2.52E-14 2.32E-13 4.90E-13 0.27 2.22E-09 2.43E-10 3.42E-15 2.55E-14 2.41E-13 5.06E-13 0.29 2.17E-09 2.41E-10 3.33E-15 2.55E-14 2.38E-13 5.07E-13 0.31 2.20E-09 2.43E-10 3.28E-15 2.47E-14 2.31E-13 4.86E-13 0.33 2.18E-09 2.44E-10 3.19E-15 2.34E-14 2.22E-13 4.63E-13 0.35 2.20E-09 2.43E-10 2.93E-15 2.25E-14 2.13E-13 4.47E-13 0.37 2.16E-09 2.45E-10 2.80E-15 2.16E-14 2.06E-13 4.33E-13 0.38 2.20E-09 2.40E-10 2.83E-15 2.21E-14 2.04E-13 4.32E-13 0.4 2.19E-09 2.44E-10 3.00E-15 2.25E-14 2.14E-13 4.49E-13 0.42 2.18E-09 2.41E-10 3.07E-15 2.36E-14 2.21E-13 4.65E-13 0.44 2.16E-09 2.42E-10 3.17E-15 2.49E-14 2.31E-13 4.89E-13 0.46 2.20E-09 2.40E-10 3.35E-15 2.55E-14 2.40E-13 5.04E-13 0.48 2.14E-09 2.46E-10 3.35E-15 2.56E-14 2.39E-13 5.06E-13 0.5 2.17E-09 2.46E-10 3.28E-15 2.49E-14 2.34E-13 4.94E-13 0.52 2.17E-09 2.51E-10 3.32E-15 2.58E-14 2.40E-13 5.06E-13 0.54 2.19E-09 2.43E-10 3.34E-15 2.52E-14 2.40E-13 5.07E-13 0.56 2.15E-09 2.43E-10 3.25E-15 2.47E-14 2.30E-13 4.86E-13 0.58 2.19E-09 2.46E-10 3.10E-15 2.33E-14 2.22E-13 4.68E-13 0.6 2.18E-09 2.46E-10 2.89E-15 2.28E-14 2.13E-13 4.50E-13 0.62 2.17E-09 2.44E-10 2.81E-15 2.18E-14 2.04E-13 4.33E-13 0.63 2.19E-09 2.42E-10 2.79E-15 2.22E-14 2.06E-13 4.32E-13 0.65 2.16E-09 2.46E-10 2.96E-15 2.28E-14 2.14E-13 4.46E-13 0.67 2.16E-09 2.44E-10 3.09E-15 2.33E-14 2.21E-13 4.64E-13 0.69 2.20E-09 2.38E-10 3.27E-15 2.44E-14 2.30E-13 4.89E-13 0.71 2.18E-09 2.45E-10 3.28E-15 2.56E-14 2.39E-13 5.04E-13 0.73 2.20E-09 2.46E-10 3.26E-15 2.57E-14 2.40E-13 5.07E-13 0.75 2.19E-09 2.45E-10 3.37E-15 2.51E-14 2.34E-13 4.88E-13 TN-LC-0100 5.6.4-89b

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-61 Fuel Qualification Table for a PWR Fuel Assembly - 3.10 Lead Thickness Cooling Time (years)

Burn-up, Initial Assembly Averaged 235-U Enrichment, wt.%.

GWd/

MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 3.0 3.0 2.9 2.9 2.8 2.8 2.8 2.8 2.7 2.7 2.7 2.7 2.6 2.6 2.6 2.6 2.6 2.6 2.6 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.4 2.4 2.4 2.4 2.4 20 4.7 4.6 4.5 4.4 4.3 4.2 4.2 4.1 4.0 4.0 3.9 3.9 3.8 3.8 3.8 3.7 3.7 3.6 3.6 3.6 3.6 3.5 3.5 3.5 3.5 3.4 3.4 3.4 3.4 3.4 3.4 3.3 3.3 3.3 3.3 3.3 30 6.7 6.5 6.3 6.2 6.0 5.9 5.7 5.6 5.5 5.3 5.2 5.1 5.1 5.0 4.9 4.8 4.7 4.7 4.6 4.6 4.5 4.5 4.4 4.4 4.3 4.3 4.2 4.2 4.2 4.1 4.1 4.1 4.1 4.0 39 7.1 6.9 6.7 6.6 6.4 6.3 6.2 6.0 5.9 5.8 5.7 5.6 5.5 5.4 5.4 5.3 5.2 5.2 5.1 5.0 5.0 4.9 4.9 4.8 4.8 4.8 4.7 4.7 4.6 4.6 4.6 40 6.4 6.2 6.1 6.0 5.9 5.8 5.7 5.6 5.5 5.4 5.4 5.3 5.2 5.2 5.1 5.1 5.0 5.0 4.9 4.9 4.8 4.8 4.7 4.7 4.7 50 9.6 9.4 9.2 8.9 8.7 8.5 8.3 8.1 7.9 7.8 7.6 7.5 7.3 7.2 7.1 6.9 6.8 6.7 6.6 6.5 6.4 6.3 6.2 6.2 6.1 55 12.0 11.7 11.4 11.1 10.8 10.5 10.3 10.0 9.8 9.6 9.4 9.1 8.9 8.8 8.6 8.4 8.2 8.1 7.9 7.8 7.7 7.5 7.4 7.3 7.2 60 14.8 14.4 14.1 13.7 13.4 13.0 12.7 12.4 12.1 11.8 11.5 11.3 11.0 10.7 10.5 10.3 10.1 9.8 9.6 9.4 9.3 9.1 8.9 8.8 8.6 61 15.4 15.0 14.7 14.3 13.9 13.6 13.3 12.9 12.6 12.3 12.0 11.7 11.5 11.2 10.9 10.7 10.5 10.2 10.0 9.8 9.6 9.4 9.3 9.1 8.9 62 16.1 15.7 15.3 14.9 14.5 14.2 13.8 13.5 13.1 12.8 12.5 12.2 11.9 11.7 11.4 11.1 10.9 10.7 10.4 10.2 10.0 9.8 9.6 9.4 9.3 Enr. wt.% 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 Note: for values not explicitly listed in the table, round burnups up to the first value shown, round enrichments down, and select the cooling time listed. Average assembly burnup listed. Shaded empty area of the table indicates fuel not analyzed for loading TN-LC-0100 5.6.4-89c

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-62 Fuel Qualification Table for 25 PWR Fuel Rods - 3.10 Lead Thickness Cooling Time (years)

Burn-up, Initial Assembly Averaged 235-U Enrichment, wt.%.

GWd/

MTU 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 10 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 0.19 20 0.22 0.22 0.22 0.22 0.22 0.22 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 0.21 30 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 0.23 39 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 40 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 0.24 45 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.26 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 0.25 50 0.31 0.30 0.30 0.30 0.30 0.30 0.30 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.29 0.28 0.28 0.28 0.28 0.28 0.28 0.28 0.28 0.28 0.28 55 0.37 0.37 0.37 0.36 0.36 0.35 0.35 0.35 0.35 0.34 0.34 0.34 0.34 0.33 0.33 0.33 0.33 0.33 0.33 0.32 0.32 0.32 0.32 0.32 0.32 60 0.50 0.49 0.48 0.47 0.47 0.46 0.45 0.45 0.44 0.44 0.43 0.43 0.42 0.42 0.41 0.41 0.41 0.40 0.40 0.40 0.39 0.39 0.39 0.38 0.38 61 0.53 0.52 0.51 0.50 0.50 0.49 0.48 0.47 0.47 0.46 0.46 0.45 0.44 0.44 0.44 0.43 0.43 0.42 0.42 0.42 0.41 0.41 0.41 0.40 0.40 62 0.56 0.55 0.54 0.53 0.53 0.52 0.51 0.50 0.49 0.49 0.48 0.48 0.47 0.46 0.46 0.45 0.45 0.44 0.44 0.44 0.43 0.43 0.43 0.42 0.42 65 0.56 0.55 0.55 0.54 0.53 0.53 0.52 0.51 0.51 0.50 0.50 0.49 0.49 0.49 70 0.71 0.70 0.69 0.68 0.67 0.67 0.66 0.65 0.65 0.64 0.63 0.62 0.62 0.61 75 0.87 0.86 0.85 0.84 0.83 0.82 0.80 0.79 0.79 0.78 0.77 0.76 0.75 0.75 80 1.04 1.03 1.01 1.00 0.99 0.97 0.96 0.95 0.94 0.93 0.92 0.91 0.90 0.89 85 1.24 1.22 1.20 1.18 1.16 1.15 1.13 1.12 1.10 1.09 1.08 1.06 1.05 1.04 90 1.47 1.44 1.41 1.39 1.37 1.34 1.32 1.30 1.29 1.27 1.25 1.23 1.22 1.20 Enr. wt.% 0.7 0.8 0.9 1.2 1.5 2.0 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 4.0 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5.0 Note: for values not explicitly listed in the table, round burnups up to the first value shown, round enrichments down, and select the cooling time listed. Average assembly burnup listed. Shaded empty area of the table indicates fuel not analyzed for loading TN-LC-0100 5.6.4-89d

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-63 PWR Fuel Assembly Design Basis Gamma Source Terms - 3.10 Lead Thickness Bottom Nozzle In-core Plenum Total Emax (MeV) Top Nozzle (/s)

(/s) (/s) (/s) (/s)

NCT 0.05 1.28E+11 1.11E+15 2.58E+11 7.52E+10 1.11E+15 0.10 2.08E+10 2.98E+14 4.57E+10 1.44E+10 2.98E+14 0.20 7.17E+09 2.16E+14 1.24E+10 3.47E+09 2.16E+14 0.30 4.03E+08 6.33E+13 6.53E+08 1.73E+08 6.33E+13 0.40 8.34E+08 4.00E+13 1.05E+09 2.26E+08 4.00E+13 0.60 1.15E+10 1.17E+14 7.51E+09 1.60E+07 1.17E+14 0.80 7.76E+09 2.21E+15 1.29E+10 1.51E+09 2.21E+15 1.00 2.17E+09 7.02E+13 9.33E+09 1.74E+09 7.02E+13 1.33 6.03E+12 1.35E+14 1.33E+13 4.19E+12 1.59E+14 1.66 1.70E+12 2.89E+13 3.75E+12 1.18E+12 3.56E+13 2.00 1.19E+02 1.10E+11 7.73E+01 1.76E-02 1.10E+11 2.50 4.07E+07 1.17E+10 8.98E+07 2.83E+07 1.18E+10 3.00 3.48E+04 1.01E+09 7.67E+04 2.42E+04 1.01E+09 4.00 1.79E-05 1.74E+08 9.18E-05 1.48E-05 1.74E+08 5.00 3.07E-33 4.07E+07 1.53E-33 0.00E+00 4.07E+07 6.50 8.84E-34 1.64E+07 4.42E-34 0.00E+00 1.64E+07 8.00 1.12E-34 3.21E+06 5.62E-35 0.00E+00 3.21E+06 10.00 1.50E-35 6.81E+05 7.50E-36 0.00E+00 6.81E+05 Total 7.91E+12 4.29E+15 1.74E+13 5.46E+12 4.32E+15 HAC 0.05 8.79E+10 9.98E+14 1.86E+11 5.46E+10 9.98E+14 0.10 1.49E+10 2.66E+14 3.28E+10 1.03E+10 2.66E+14 0.20 4.65E+09 1.88E+14 8.61E+09 2.49E+09 1.88E+14 0.30 2.55E+08 5.52E+13 4.48E+08 1.25E+08 5.52E+13 0.40 4.84E+08 3.50E+13 6.80E+08 1.63E+08 3.50E+13 0.60 5.65E+09 6.38E+13 3.70E+09 1.20E+07 6.38E+13 0.80 4.73E+09 2.05E+15 1.10E+10 1.52E+09 2.05E+15 1.00 1.94E+09 4.27E+13 9.20E+09 1.61E+09 4.27E+13 1.33 4.33E+12 9.20E+13 9.53E+12 3.00E+12 1.09E+14 1.66 1.22E+12 1.77E+13 2.69E+12 8.48E+11 2.24E+13 2.00 1.13E+02 9.14E+10 7.35E+01 1.62E-02 9.14E+10 2.50 2.92E+07 5.85E+09 6.44E+07 2.03E+07 5.96E+09 3.00 2.50E+04 5.44E+08 5.50E+04 1.73E+04 5.44E+08 4.00 1.67E-05 1.52E+08 8.52E-05 1.37E-05 1.52E+08 5.00 3.50E-33 4.85E+07 1.75E-33 0.00E+00 4.85E+07 6.50 1.01E-33 1.95E+07 5.04E-34 0.00E+00 1.95E+07 8.00 1.28E-34 3.82E+06 6.41E-35 0.00E+00 3.82E+06 10.00 1.71E-35 8.10E+05 8.55E-36 0.00E+00 8.10E+05 Total 5.67E+12 3.81E+15 1.25E+13 3.92E+12 3.83E+15 TN-LC-0100 5.6.4-89e

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-64 25 PWR Rods Design Basis Gamma Source Terms - 3.10 Lead Thickness Bottom Nozzle In-core Plenum Total Emax (MeV) Top Nozzle (/s)

(/s) (/s) (/s) (/s)

NCT 0.05 1.28E+11 2.30E+12 1.65E+16 1.95E+12 1.65E+16 0.10 2.08E+10 2.04E+11 4.68E+15 2.38E+11 4.68E+15 0.20 7.17E+09 3.01E+11 7.30E+15 2.14E+11 7.30E+15 0.30 4.03E+08 5.16E+10 1.16E+15 4.44E+10 1.16E+15 0.40 8.34E+08 2.77E+12 1.07E+15 1.65E+12 1.07E+15 0.60 1.15E+10 9.37E+11 1.13E+16 1.81E+12 1.13E+16 0.80 7.76E+09 2.62E+13 3.88E+16 1.78E+13 3.88E+16 1.00 2.17E+09 4.98E+12 7.13E+14 5.51E+12 7.27E+14 1.33 6.03E+12 1.31E+13 4.38E+14 2.69E+13 4.87E+14 1.66 1.70E+12 3.41E+12 9.42E+14 7.55E+12 9.55E+14 2.00 1.19E+02 1.26E+10 2.76E+13 2.68E+10 2.77E+13 2.50 4.07E+07 1.40E+08 9.87E+13 2.19E+08 9.87E+13 3.00 3.48E+04 1.16E+05 2.77E+13 1.84E+05 2.77E+13 4.00 1.79E-05 1.70E-06 3.24E+11 8.70E-06 3.24E+11 5.00 3.07E-33 1.06E-42 7.91E+05 5.28E-43 7.91E+05 6.50 8.84E-34 3.04E-43 3.17E+05 1.53E-43 3.17E+05 8.00 1.12E-34 3.92E-44 6.21E+04 1.96E-44 6.21E+04 10.00 1.50E-35 5.61E-45 1.32E+04 2.80E-45 1.32E+04 Total 5.42E+13 8.30E+16 6.38E+13 1.69E+13 8.31E+16 HAC 0.05 2.45E+12 1.72E+16 2.64E+12 4.60E+11 1.72E+16 0.10 1.38E+11 5.63E+15 2.92E+11 9.07E+10 5.63E+15 0.20 9.58E+10 5.33E+15 1.11E+11 2.22E+10 5.33E+15 0.30 7.93E+09 1.45E+15 7.73E+09 1.16E+09 1.45E+15 0.40 5.66E+10 1.13E+15 4.01E+10 1.48E+09 1.13E+15 0.60 3.42E+11 1.10E+16 2.44E+11 9.51E+09 1.10E+16 0.80 6.16E+11 1.46E+16 4.12E+11 2.25E+09 1.46E+16 1.00 3.96E+12 3.93E+15 7.67E+11 2.26E+12 3.93E+15 1.33 3.78E+13 1.24E+15 8.33E+13 2.63E+13 1.39E+15 1.66 1.07E+13 5.00E+14 2.35E+13 7.42E+12 5.42E+14 2.00 2.78E+08 3.89E+13 5.07E+08 1.45E+08 3.89E+13 2.50 2.64E+08 9.97E+13 5.68E+08 1.78E+08 9.97E+13 3.00 2.25E+05 2.89E+12 4.85E+05 1.52E+05 2.89E+12 4.00 4.99E-05 2.66E+11 2.55E-04 4.11E-05 2.66E+11 5.00 1.57E-31 2.17E+08 7.85E-32 0.00E+00 2.17E+08 6.50 4.52E-32 8.73E+07 2.26E-32 0.00E+00 8.73E+07 8.00 5.75E-33 1.71E+07 2.88E-33 0.00E+00 1.71E+07 10.00 7.67E-34 3.63E+06 3.84E-34 0.00E+00 3.63E+06 Total 5.62E+13 6.22E+16 1.11E+14 3.65E+13 6.24E+16 TN-LC-0100 5.6.4-89f

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-65 Design Basis Neutron Sources - 3.10 Lead Thickness Neutron Source for Source Total Neutron Fuel and Analysis Type Average Assembly Multiplication Source (n/s)

Burnup (n/s) Factor PWR Fuel Assembly NCT 1.18E+09 1.152/(1-0.1843) 1.67E+09 PWR Fuel Assembly HAC 1.41E+09 1.152/(1-0.1625) 1.94E+09 Neutron Source for Source Total Neutron Fuel and Analysis Type Peak Rod Burnup Multiplication Source (n/s)

(n/s) Factor PWR 25 Rod NCT 2.37E+07 1.000/(1-0.1350) 2.73E+07 PWR 25 Rod HAC 7.01E+09 1.000/(1-0.1350) 8.10E+09 TN-LC-0100 5.6.4-89g

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 5.6.4-66 NCT Dose Rate Summary (mrem/hr) - 3.10 Lead Thickness 2m Vehicle Vehicle 2 m from 2m from Package Vehicle Vehicle from Fuel Type Bottom Top Vehicle Vehicle Surface Underside Side Vehicle End End Bottom End Top End Side PWR Fuel 477 116 35.4 75.2 49.4 8.82 4.02 3.04 Assembly 25 PWR 100 45.6 25.0 Note 1 Note 1 8.07 Note 1 Note 1 rods Limit 1000 200 10 (1) Bounded by the maximum EPR rod dose rate at this location, see Table 5.6.4-32 TN-LC-0100 5.6.4-89h

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 6.10.4.2 Fissile Material Contents The fissile materials are a single PWR or BWR fuel assembly. Additionally, PWR, BWR, EPR and MOX fuel pins are allowed in the 25 pin can.

The PWR fuel assemblies and their parameters are provided in Table 6.10.4-2. The KENO model fuel assemblies are constructed using these parameters. Note that WE 16x16 fuel class is not specifically analyzed as this fuel class is similar to WE 17x17 fuel class. WE 16x16 fuel class is a 235 fuel rods design (16x16 - 21 guide/instrument tubes) with fuel characteristics (pellet OD, clad thickness and clad OD) similar to those of WE 17x17 LOPAR. WE 17x17 fuel class is expected to bound WE 16x16 fuel class.

Similarly, the BWR fuel assembly parameters are provided in Table 6.10.4-3 and Table 6.10.4-

30. As stated, no credit is taken for burn up of fuel in the calculations. A maximum enrichment of 5.0 wt. percent U-235 is used for all fuel assemblies listed in Table 6.10.4-2, Table 6.10.4-3, and Table 6.10.4-30, with the following exception. For the CE 15x15 class assemblies, the maximum enrichment is 3.7 wt. percent U-235.

Each fuel assembly listed for the PWR assemblies is modeled using nominal dimensions within the cask to obtain a limiting assembly with the highest ks for subsequent analyses. For BWR fuel, the most reactive fuel assembly for each lattice group is obtained. In addition, since the BWR LaCrosse fuel assemblies have a much smaller active fuel length than the other 10x10 assemblies, both are evaluated individually. The two LaCrosse fuel assemblies are Allis Chalmers and Exxon/ANF.

The ABB fuel assemblies evaluated are provided in Table 6.10.4-30. The table shows three array types. However, the SVEA 96Opt fuel has two fuel pellet and fuel clad options; so using the maximum pitch of 0.502", two cases are evaluated individually: the fuel pellet OD of 0.346" and fuel clad OD of 0.406" and fuel pellet OD of 0.323" and fuel clad OD of 0.379". The fuel assemblies are modeled with and without fuel channels. The thickness of the fuel channel is set at 0.025", 0.08", and 0.12".

In order to qualify individual fuel rods for transport, the fuel rods from the most reactive PWR and BWR assembly calculations are inserted in the fuel rod tubes located in the 25 pin can. The MOX and EPR fuel rods are modeled according to the parameters provided in Table 6.10.4-4 and Table 6.10.4-5, respectively. Additionally, a generic UO2 fuel model is considered in the 25 pin can with parameters shown in Table 6.10.4-5. The plutonium isotopic vector provides a bounding ks and the analysis is performed with three different plutonium concentrations: 6.0, 8.0 and 10.2 wt. percent plutonium.

TN-LC-0100 6.10.4-4

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 subcritical and below the USL with the PRA configuration as shown in Figure 6.10.4-13. The number of PRAs required is 8, each at a minimum diameter of 0.88 cm. The maximum allowable U-235 enrichment is 5.00 weight percent. All rotationally symmetric configurations of the absorber rods are also acceptable.

WE 17x17 Class Assemblies:

The most reactive WE 17x17 assembly evaluated is the WE 17x17 OFA fuel assembly, as shown in Table 6.10.4-8. These class of assemblies will remain subcritical and below the USL with the PRA configuration as shown in Figure 6.10.4-15. The number of PRAs required is 8, each at a minimum diameter of 0.88 cm. The maximum allowable U-235 enrichment is 5.00 weight percent. All rotationally symmetric configurations of the absorber rods are also acceptable.

Results for WE 17x17 class assembly are applicable to WE 16x16 class assembly.

BW 15x15 Class Assemblies:

The most reactive BW 15x15 assembly is the BW 15x15, Mark B11 fuel assembly as shown in Table 6.10.4-8. The number of PRAs required is 8, each at a minimum diameter of 0.88 cm. The maximum allowable U-235 enrichment is 5.00 weight percent. The configuration of PRA location is as shown in Figure 6.10.4-15. All rotationally symmetric configurations of the absorber rods are also acceptable.

CE 14x14 Class Assemblies:

The most reactive CE 14x14 assembly is the Framatome CE 14x14 fuel assembly as shown in Table 6.10.4-8. The number of PRAs required is 5, each at a minimum diameter of 1.02 cm. This translates to a linear density of 0.618 g/cm. The maximum allowable U-235 enrichment is 5.00 weight percent. The configuration of PRA location is as shown in Figure 6.10.4-12.

CE 15x15 Class Assemblies:

For CE 15x15 Class assemblies that have just one location for PRA insertion, the maximum enrichment is reduced to 3.70 weight percent U-235. The analysis is performed with a PRA diameter of 0.76 cm or linear density of 0.343 g/cm B4C per PRA.

TN-LC-0100 6.10.4-12a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 6.10.4-2 PWR Fuel Assembly Parameters (Part 3 of 3)

Parameter WE 14x14 WE 16x16 WE 17x17 CE 16x16 STD, STD, STD, OFA 16ACE7 OFA KOFA, 17ACE7 HIPER17 KSFA, GUARDIAN KOFA KOFA V5H, RFA WEC/ WEC/ ABB-CE/

Manufacturer WEC KNF WEC WEC/KNF KNF KNF KNF KNF KNF Array 14x14 16x16 17x17 16x16 Fuel Assembly 159.76 159.97 159.76 159.97 159.97 178.25 178.25 Length(8) (in.)

Number of fuel rods per 179 235 264 236 assembly Pitch(9) (in.) 0.556 0.485 0.496 0.506 Fuel Pellet Outer Ø(9) 0.3659 0.3444 0.3225 0.3088 0.3225 0.3250 0.3225 (in.)

Clad thickness(9) 0.0243 0.0243 0.0225 0.0225 0.0250 0.0225 (in.)

Outer Cladding Ø(9) 0.422 0.400 0.374 0.360 0.374 0.374 0.382 0.374 (in.)

Max Fuel Assembly 520 589 554.7 667.5 673.1 673.8 651.92 639.39 642.5 Weight (kg)

Max U weight (kg) per 357.4 407.9 376.8 461.5 461.3 429.03 430.78 430.8 Assembly (8) Fuel assembly length of RFA (159.975) is the longest for all the fuels (9) As-fabricated nominal value TN-LC-0100 6.10.4-26

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 6.10.4-26 Summary of PRA Requirements Under all Conditions of Transport for PWR Fuel Assembly Classes Minimum Max U-235 Number of Diameter of B4C Content Enrichment Assembly Class PRAs PRAs (cm) (g/cm) (wt %)

WE 17x17 8 0.88 0.613 5.00 WE 16x16 8 0.88 0.613 5.00 CE 16x16 5 1.02 0.824 5.00 BW 15x15 8 0.88 0.613 5.00 CE 15x15 1 0.76 0.475 3.70 WE 15x15 8 0.88 0.613 5.00 CE 14x14 5 1.02 0.824 5.00 WE 14x14 8 0.88 0.613 5.00 BW 17x17 8 0.76 0.475 5.00 TN-LC-0100 6.10.4-44

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Chapter 7 Package Operations TABLE OF CONTENTS 7.1 TN-LC Package Loading .................................................................................................. 7-1 7.1.1 TN-LC Cask Preparation for Loading ......................................................................... 7-1 7.1.2 TN-LC Cask Wet Loading ............................................................................................ 7-3 7.1.3 TN-LC Cask Dry Loading ............................................................................................ 7-5 7.1.4 TN-LC Cask Preparation for Transport ...................................................................... 7-6 7.2 TN-LC Package Unloading .............................................................................................. 7-7 7.2.1 Receipt of Loaded TN-LC Package from Carrier ........................................................ 7-7 7.2.2 Removal of Contents from TN-LC Cask ...................................................................... 7-8 7.3 Preparation of Empty Package for Transport ................................................................ 7-10 7.4 Other Operations ............................................................................................................ 7-11 7.4.1 Assembly Verification Leakage Testing of the Containment Boundary .................... 7-11 7.5 References ....................................................................................................................... 7-14 7.6 Glossary .......................................................................................................................... 7-15 7.7 Appendices ...................................................................................................................... 7-16 7.7.1 TN-LC-NRUX Basket Wet and Dry Loading and Unloading .................................... 7-16 7.7.2 TN-LC-MTR Basket Wet and Dry Loading and Unloading ...................................... 7-16 7.7.3 TN-LC-TRIGA Basket Wet and Dry Loading and Unloading ................................... 7-16 7.7.4 TN-LC-1FA Wet and Dry Loading and Unloading ................................................... 7-16 LIST OF TABLES Table 7-1 Applicable Fuel Specification for Various Fuel Types......................................... 7-17 Table 7-2 Appendices Containing Loading Procedures for Various TN-LC Baskets .................................................................................................................. 7-17 Table 7-3 Appendices Containing Unloading Procedures for Various TN-LC Baskets .................................................................................................................. 7-17 LIST OF FIGURES Figure 7-1 TN-LC Packaging Torquing Patterns .................................................................. 7-18 Figure 7-2 Assembly Verification Leakage Test ..................................................................... 7-19 TN-LC-0100 7-i

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

3. Remove the tamper-indicating seals.

4. Remove the impact limiters from the cask.

5. Prior to removing the lid, sample the cask cavity atmosphere. If removing the lid at this stage, inspect the lid seals and sealing surfaces and verify that the O-ring seals have been replaced within the last 12 months.

6. Remove the skid tie-down assembly.

7. Take contamination smears on the outside surfaces of the cask. If necessary, decontaminate the cask.

8. The lid, bottom plug and all drain/vent/test ports incorporate O-ring seals. O-ring seals may be reused. Prior to installation, the seals and sealing surfaces shall be inspected.

Verify that the seals have been replaced within the last 12 months.

9. Remove the trunnion and pocket trunnion plugs.

10. Install the two lifting trunnions in place of the front trunnions plugs. Install the trunnion bolts and torque them to the torque specified on drawing 65200-71-01, Appendix 1.4.1, following the torquing sequence shown in Figure 7-1.

11. For the specific payload to be transported as part of the TN-LC package, verify that the basket type (TN-LC-NRUX, TN-LC-MTR, TN-LC-TRIGA, or TN-LC-1FA) and spacers, if required, are appropriate for the fuel to be transported.

12. The candidate intact fuel assemblies/elements or fuel rods to be transported in a specific basket must be evaluated to verify that they meet the fuel qualification requirements of the applicable fuel specification as listed in Table 7-1.

13. Prior to being placed in service, the cask is to be cleaned or decontaminated, as necessary.

14. Remove the bottom plug assembly, inspect the seals and sealing surfaces, verify that the O-ring seals have been replaced within the last 12 months, lubricate and reinstall the bottom plug assembly, torquing the bolts to the torque specified on drawing 65200-71-01, Appendix 1.4.1.

15. Remove the two test ports, the drain port and the vent port, inspect the seals and sealing surfaces, verify that the O-ring seals have been replaced within the last 12 months, reinstall each port (hand tight). The vent port on the lid may be left partially threaded to facilitate draining operations in step 14. The ports covers may be reinstalled over the two test ports at this time.

16. Engage the cask trunnions with the cask lifting yoke.

17. Rotate the cask to a vertical orientation, lift the cask, and place the cask in the designated preparation area.

TN-LC-0100 7-2

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 NOTE: Alternatively, the cask may be lifted in a horizontal orientation and placed on an onsite transfer trailer or upending frame; or the cask/transportation skid may be lifted together and placed in the appropriate location.

18. Install the shear key plug assembly and the pocket trunnion plugs.

19. If the cask lid has not already been removed, remove the bolts from the lid and lift the lid from the cask. Inspect the seals and sealing surfaces and verify that the O-ring seals have been replaced within the last 12 months,

20. Depending on the basket design being loaded, verify that a cask spacer of appropriate height is placed at the bottom of the cask cavity and/or bolted to the underside of the lid.

21. Insert the basket appropriate for the fuel to be transported into the cask, as listed in Table 7-2.

Notes:

• Install a bottom spacer in the basket if required by Chapter 1, Appendix 1.4.1 basket drawings.

• If loading a BWR fuel assembly in a TN-LC-1FA basket, place a BWR sleeve with a BWR hold down ring in the basket as shown in drawing 65200-71-96.

• If loading fuel rods in a TN-LC-1FA basket, place a pin can inside the BWR sleeve with a hold down ring in the basket as shown in drawing 65200-71-102.

7.1.2 TN-LC Cask Wet Loading NOTE: The wet loading procedure described in this section is applicable only when using the TN-LC cask for loading fuel from a fuel pool into any one of the baskets listed in Chapter 1, Table 1-2.

Site-specific conditions or procedures may require the use of different equipment and ordering of steps than those described below to accomplish the same objectives or to meet acceptance criteria which ensure the integrity of the package.

CAUTION: Radioactive particulate matter may float to the surface of the water during underwater loading of the cask. Precautionary measures (filters, etc.) should be considered to minimize the radiation dose to personnel during these operations.

1. Fill the cask cavity with fuel pool water.

2. Lift the cask and position it over the cask loading area of the fuel pool.

3. Lower the cask into the fuel pool.

4. Place the cask in the fuel pool cask loading area.

5. Disengage the lifting yoke from the cask trunnions and move the yoke clear of the cask.

TN-LC-0100 7-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

6. The operations for loading fuel into a specific basket type are described in detail in Appendices 7.7.1 through 7.7.4 as listed in Table 7-2.

7. Lower the lid into place

8. Visually verify that the lid is properly seated in the cask.

9. Raise the cask to the pool surface using the cask trunnions and the lifting yoke.

10. Verify that the lid is properly seated on the cask. If not, lower the cask and reposition the lid. Repeat the above steps as necessary.

11. Continue to raise the cask from the pool until the top region of the cask is accessible.

12. Perform a radiological survey of the cask as it is raised out of the pool.

13. Install at least two lid bolts hand tight.

14. When the cask drain plug is accessible, open the drain plug and drain the cavity until no appreciable water is noted. Optionally, the cavity may be drained after securing the cask in the site work area. For loading a PWR/BWR fuel assembly, consistent with NUREG-1536 [6] guidance, helium at 1-3 psig is used to backfill the cask cavity with an inert gas (helium) as water is being removed from the cask cavity.

15. Move the cask to the site designated preparation area and secure the cask, as required.

The cask is now ready to be prepared for downending as described in Section 7.1.2.1 below.

7.1.2.1 Preparing the TN-LC Cask for Downending

1. Torque the drain port plug to the torque specified on drawing 65200-71-01, Appendix 1.4.1. The drain port plug cover may be installed at this time.

2. Not used.

3. Install the cask lid remaining bolts. Follow the torquing sequence shown in Figure 7-1, torque the lid bolts to the torque specified on drawing 65200-71-01, Appendix 1.4.1.

4. Not used.

CAUTION: During vacuum pump system operation, personnel should be in the area of loading operations, or in nearby low dose areas, in order to take proper action in the event of a malfunction.

5. Using the vacuum pump system, evacuate the cask cavity to 3 torr or less to ensure that the cask cavity has no free standing water and is essentially dry.

6. Backfill with helium to 2.5 +/- 1.0 psig.

7. Perform the assembly verification leakage test following the procedure given in Section 7.4.1.

TN-LC-0100 7-4

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.1.2.2 TN-LC Cask Downending NOTE: Alternate procedures may be developed for plants with unique requirements.

1. Remove the shear key plug assembly and the pocket trunnion plugs from the cask.

2. Lift the cask over the transportation skid.

3. Lower and rotate the cask from vertical to horizontal and secure it to the skid.

4. Prepare the cask for transportation in accordance with the procedure described in Section 7.1.4.

7.1.3 TN-LC Cask Dry Loading NOTE: The dry loading procedure described in this section is applicable only when using the TN-LC cask for loading fuel from a hot cell into any one of the baskets listed in Chapter 1, Table 1-2.

The procedure for loading the cask from a hot cell is highly dependent on the design of the facility. The procedure described below is intended to show the type of operations that will be performed and is not intended to be limiting. Site-specific conditions or procedures may require the use of different equipment and ordering of steps than those described below to accomplish the same objectives or to meet acceptance criteria which must be met to ensure the integrity of the package.

1. Place the cask in the location of the hot cell designated as the cask loading area or mate the cask opening with the hot cell portal. Note that this may require downending the cask into a horizontal orientation.

2. If the cask lid has not already been removed, remove the bolts from the lid and, using appropriate slings and/or the cask yoke with appropriate slings, lift the lid from the cask.

Inspect the seals and sealing surfaces and verify that the O-ring seals have been replaced within the last 12 months,

3. Disengage the lifting yoke or other lifting device from the trunnions and move the yoke clear of the cask.

4. The operations for loading fuel into a specific basket type are described in detail in Appendices 7.7.1 through 7.7.4 as listed in Table 7-2.

5. Install the lid on the cask. Verify that the lid is properly seated in the cask. If not, reposition the lid.

6. Install and torque the lid bolts to the torque specified on drawing 65200-71-01, Appendix 1.4.1, following the torquing sequence shown in Figure 7-1.

7. Remove the loaded cask from the hot cell or disengage it from the hot cell portal. Perform a radiological survey of the cask as it is removed.

TN-LC-0100 7-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

8. Move the cask to the site designated preparation area and secure the cask, as required.

The cask is now ready to be back-filled with helium and down-ended as described starting from step 6 in section 7.1.2.1 above.

7.1.4 TN-LC Cask Preparation for Transport Once the TN-LC cask has been loaded using either the wet loading procedure described in Section 7.1.2 or the dry loading procedure described in Section 7.1.3 above, the following tasks are performed to prepare the cask for transportation. The cask is assumed to be seated horizontally on the transportation skid. Alternate procedures may be developed for plants with unique requirements.

1. Verify that the cask surface removable contamination levels meet the requirements of 49CFR173.443 [2] and 10CFR71.87 [3].

2. Verify that the assembly verification leakage testing specified in Section 7.4.1 has been performed.

7.1.4.1 Placing the TN-LC Cask onto the Conveyance The procedure for placement of the cask on the conveyance is given in this section. If the cask is on the transportation skid, but the skid is not on the conveyance, rig the cask/skid, lift and place them onto the conveyance.

1. Install the transportation skid tie-down straps.

2. Install the pocket trunnion plugs.

3. Remove the two trunnions, and install the trunnion plugs.

4. Install the impact limiters on the cask, torqueing the attachment bolts to the torque specified on drawing 65200-71-01, Appendix 1.4.1.

5. Remove the impact limiter hoist rings and replace them with hex bolts.

6. Install the tamper-indicating seals.

7. Perform a final radiation survey to ensure the cask radiation levels do not exceed 49CFR173.441 [2] and 10CFR71.47 [3] requirements.

8. Place the TN-LC Package in the shipping container.

9. Verify that the temperature on all accessible surfaces is < 185º F.

10. Verify that placards, labels, markings and seals are in place and correct.

11. Close the transport container.

12. Prepare the final shipping documentation and release the TN-LC Package for shipment.

TN-LC-0100 7-6

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.2 TN-LC Package Unloading Unloading the TN-LC Package after transport involves removing the cask from the conveyance and removing the fuel from the cask. The cask is designed to allow the fuel to be unloaded from the cask into a hot cell and provisions exist to allow wet unloading in a fuel pool. The necessary procedures for these tasks are essentially the reverse of those described in Section 7.1.

7.2.1 Receipt of Loaded TN-LC Package from Carrier Procedures for receiving the TN-LC Package after shipment are described in this section.

Procedures for preparing an empty package are provided in Section 7.1.1.

1. Upon arrival of the loaded TN-LC Package at the receiving site, perform receipt inspection. Inspect for damage, verify tamper-indicating seal is intact and perform radiation survey.

2. Open the transport container, and remove the TN-LC package.

3. Remove the tamper-indicating seals.

4. Remove the hex bolts from the impact limiters and replace them with the impact limiter hoist rings provided.

5. Remove the impact limiters from the cask.

6. Remove the transportation skid tie-down straps.

7. Prior to removing the lid, sample the cask cavity atmosphere CAUTION: Check for the presence of damaged or oxidized fuel. This can be achieved by first obtaining gas sample at the vent or siphon port fittings, which permits an evaluation of the atmosphere within the cask cavity before removal of the cask lid.

If the cask has retained the helium atmosphere and no airborne radioactive particulates have been detected, then operations may proceed normally with fuel removal. However, if air or airborne radioactive particulates are present within the cask cavity, then appropriate filters should be used to preclude the uncontrolled release of any potential airborne radioactive particulate from the cask. This will protect both personnel and the operations area from potential contamination. For the accident case, personnel protection in the form of respirators or supplied air should be considered in accordance with licensees Radiation Protection Program.

8. Take contamination smears on the outside surfaces of the cask. If necessary, decontaminate the cask.

9. Remove the pocket trunnion plugs and trunnion plugs.

10. Install the two trunnions in place of the trunnion plugs, torquing the trunnion bolts to the torque specified on drawing 65200-71-01, Appendix 1.4.1, in the sequence shown in Figure 7-1.

TN-LC-0100 7-7

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.2.2 Removal of Contents from TN-LC Cask 7.2.2.1 Unloading the TN-LC Cask in a Fuel Pool The procedure for unloading the cask in a fuel pool is summarized in this section. Site-specific conditions and requirements may require the use of different equipment and ordering of steps than those descried below to accomplish the same objectives or to meet acceptance criteria to ensure the integrity of the package.

CAUTION: Radioactive particulate matter may float to the surface of the water during underwater unloading of the cask. Precautionary measures (filters, etc.) should be considered to minimize the radiation dose to personnel during these operations.

1. Verify that the TN-LC cask receipt process in Section 7.2.1 has been completed.

2. Using the cask port tool, install a pressure gauge, isolation valve and vent line to the site radwaste system on the vent port. Open the cask cavity vent to the site radwaste system and sample the cask cavity atmosphere. Flush the cask cavity if necessary.

3. Remove the drain port plug and install an appropriate fitting in the drain port.

Alternatively, a cask port tool may be used to perform flooding and draining activities.

4. Install a pressure gauge, isolation valve, check valve, and a supply of clean water to the drain port/fitting.

5. Slowly feed water to enter the cask cavity.

6. Maintain the pressure in the cask cavity below 20 psig.

7. When the cask is filled with water, remove the vent and supply lines.

8. Loosen the lid bolts, leaving the threads engaged. Reverse the torquing sequence shown in Figure 7-1.

9. Slowly lower the cask into the pool until the lid is just above the surface.

10. Remove the lid bolts and lower the cask to its unloading position in the pool.

11. Detach the yoke from the trunnions and lift the lid from the cask.

12. Follow the fuel-specific unloading procedure as listed in Table 7-3 and described in Appendix 7.7.1 through 7.7.4.

13. Following removal of the fuel assemblies or fuel rods and lift the cask from the spent fuel pool.

14. Open the drain and drain the pool water from the cavity. Continue draining the cavity until no appreciable water is noted. Optionally, the cavity may be drained after securing the cask body in the site work area.

TN-LC-0100 7-8

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.2.2.2 Unloading the TN-LC Cask to a Hot Cell The procedure for unloading the cask to a hot cell is highly dependent on the design of the dry cell. The procedure described below is intended to show the type of operations that will be performed and is not intended to be limiting. Site-specific conditions and requirements may require the use of different equipment and ordering of steps than those descried below to accomplish the same objectives or acceptance criteria which must be met to ensure the integrity of the package.

NOTE: See Section 7.2.2.3 for dry unloading of a pin can.

1. Verify that the TN-LC cask receipt process in Section 7.2.1 has been completed.

2. Using the cask port tool, install a pressure gauge, isolation valve and vent line to the site radwaste system on the vent port. Vent the cask cavity to the site radwaste system and sample the cask cavity atmosphere. Flush the cavity gases if necessary.

3. Place the cask on a designated transfer cart if required.

4. Move the cask to the hot cell cask unloading area or mate the cask opening with the hot cell portal.

5. Remove the lid from the cask. Reverse the torquing sequence shown in Figure 7-1.

6. Follow the specific unloading procedure as listed in Table 7-3 and described in Appendices 7.7.1 through 7.7.4.

7. Retrieve the cask from the hot cell loading area and decontaminate the cask if necessary.

8. Move the cask to the site-designated preparation area and secure the cask, as required.

7.2.2.3 Horizontal Unloading of a Pin Can from the TN-LC Cask This procedure is for handling a TN-LC cask with a pin can in a 1FA basket at a facility with a hot cell. The procedure described below is intended to show the type of operations that will be performed and is not intended to be limiting. Site-specific conditions and requirements may require the use of different equipment and ordering of steps than those descried below to accomplish the same objectives or acceptance criteria which must be met to ensure the integrity of the package.

1. Verify that the TN-LC cask receipt process in Section 7.2.1 has been completed.

2. Using the cask port tool, install a pressure gauge, isolation valve and vent line to the site radwaste system on the vent port. Vent the cask cavity to the site radwaste system and sample the cask cavity atmosphere. Flush the cavity gases if necessary.

3. Lift the cask and downend it to a horizontal position on an unloading cradle.

4. Move the cask to the hot cell and mate the cask to the hot cell.

TN-LC-0100 7-9

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.4 Other Operations 7.4.1 Assembly Verification Leakage Testing of the Containment Boundary The procedure for leakage testing of the cask containment boundary prior to shipment is given in this section. Assembly verification leakage testing shall conform to the requirements of ANSI N14.5 [1]. A flow chart of the assembly verification leakage testing is provided in Figure 7-2.

The order in which the leakage test of the various seals are performed may vary. If more than one leakage detector is available, then more than one seal may be tested at a time. Personnel performing the leakage testing shall be specifically trained in leakage testing in accordance with SNT-TC-1A [4].

The acceptance criterion for pre-shipment leakage rate testing shall be either (a) a leakage rate of not more than the reference air leakage rate, or (b) no detected leakage when tested to a sensitivity of at least 10-3 ref-cm3/s.

The following steps present one method of performing the pre-shipment verification leakage testing. Alternate methods and order of testing are acceptable as long as the above criteria is satisfied for the TN-LC containment boundary seals.

Vent Port Plug Seal Leakage Test

1. Remove the vent port plug cover if previously installed. Install the cask port tool in the vent port.

2. Open the vent port plug.

3. Attach a suitable vacuum pump to the cask port tool.

4. Reduce the cask cavity pressure to below 1.0 psia.

5. Fill the cask cavity with helium to atmospheric pressure.

6. Close the vent port plug, torquing it to the torque specified on drawing 65200-71-01, Appendix 1.4.1.

7. Remove the helium-saturated cask port tool and install a clean (helium free) cask port tool.

8. Connect a leak detector to the cask port tool.

9. Evacuate the vent port until the vacuum is sufficient to operate the leakage detection equipment.

10. Perform the pre-shipment leak test in accordance with Section 8.2.2. If either O-ring was replaced, the maintenance leak test in Section 8.2.2 shall be performed.

NOTE: Upon removing the vent port plug and seal, it will be necessary to reduce the cask cavity pressure below 1.0 psia and refill with helium through the vent port.

11. Remove the leakage detection equipment.

TN-LC-0100 7-11

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

12. Remove the cask port tool and replace the vent port plug cover.

Lid O-ring Leakage Test

13. Remove the lid test port plug cover if previously installed.

14. Install the cask port tool in the lid test port.

15. Open the lid test port plug.

16. Connect the vacuum pump to the cask port tool.

17. Connect the leak detector to the cask port tool.

18. Evacuate the lid test port until the vacuum is sufficient to operate the leakage detection equipment per the manufacturer's recommendations.

19. Perform the pre-shipment leak test in accordance with Section 8.2.2. If either O-ring was replaced, the maintenance leak test in Section 8.2.2 shall be performed.

NOTE: Upon removing and reinstalling the cask lid, it will be necessary to reduce the cask cavity pressure below 1.0 psia and refill with helium through the vent port. The vent port assembly verification leakage test must also be redone as described above.

20. Remove the leak detection equipment.

21. Close lid test port plug and tighten it to the torque specified on drawing 65200-71-01, Appendix 1.4.1. Remove the cask port tool from the lid test port and replace the lid test port plug cover.

Drain Port Plug Seal Leakage Test

22. Remove the cask drain port plug cover if previously installed.

23. Verify that the cask drain port is closed and torqued to the torque specified on drawing 65200-71-01, Appendix 1.4.1.

24. Install the cask port tool in the cask drain port.

25. Connect the vacuum pump to the cask port tool.

26. Connect the leak detector to the cask port tool.

27. Evacuate the drain port until the vacuum is sufficient to operate the leakage detection equipment.

28. Perform the pre-shipment leak test in accordance with Section 8.2.2. If either O-ring was replaced, the maintenance leak test in Section 8.2.2 shall be performed.

TN-LC-0100 7-12

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 NOTE: Upon removing the drain port plug and seal, it will be necessary to reduce the cask cavity pressure below 1.0 psia and refill with helium through the vent port. The vent port assembly verification test must also be redone as described above.

29. Remove the leak detection equipment.

30. Remove the cask port tool from the cask drain port and replace the drain port plug cover.

Bottom Plug O-ring Leakage Test

31. Remove the bottom test port plug cover if previously installed.

32. Install the cask port tool in the bottom test port.

33. Open the bottom test port plug.

34. Connect the vacuum pump to the cask port tool.

35. Connect the leak detector to the cask port tool.

36. Evacuate the bottom test port until the vacuum is sufficient to operate the leakage detection equipment per the manufacturer's recommendations.

37. Perform the pre-shipment leak test in accordance with Section 8.2.2. If either O-ring was replaced, the maintenance leak test in Section 8.2.2 shall be performed.

NOTE: Upon removing and reinstalling the bottom plug, it will be necessary to reduce the cask cavity pressure below 1.0 psia and refill with helium through the vent port. The vent port assembly verification leakage test must also be redone as described above.

38. Remove the leak detection equipment.

39. Close bottom test port plug and tighten it to the torque specified on drawing 65200-71-01, Appendix 1.4.1. Remove the cask port tool from the bottom test port and replace the bottom test port plug cover.

This concludes the assembly verification leakage test procedure.

TN-LC-0100 7-13

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.5 References

1. ANSI N14.5-2014, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, American National Standards Institute, Inc., New York, 2014.

2. Title 49, Code of Federal Regulations, Part 173 (49 CFR 73), Shippers - General Requirements for Shipments and Packaging.

3. Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material.

4. SNT-TC-1A, American Society for Nondestructive Testing, Personnel Qualification and Certification in Nondestructive Testing.

5. Not used.

6. USNRC, NUREG-1536, Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility, Final Report, Revision 1.

TN-LC-0100 7-14

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.6 Glossary The terms used in the above procedures are defined below.

cask lifting yoke: Passive lifting yoke used for vertical lifts and upending of the cask.

conveyance: Any suitable conveyance such as a railcar, heavy haul trailer, barge, ship, etc.

ram: rod with threaded end used to insert/withdraw pin can to/from hot cell.

TN-LC-0100 7-15

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Table 7-1 Applicable Fuel Specification for Various Fuel Types Applicable Fuel Specification from Basket Design Chapter 1 TN-LC-NRUX Table 1.4.2-1 and 1.4.2-2 TN-LC-MTR Table 1.4.3-1 thru Table 1.4.3-3 TN-LC-TRIGA Table 1.4.4.1 thru 1.4.4-5 TN-LC-1FA Table 1.4.5-1 thru 1.4.5-14 Table 7-2 Appendices Containing Loading Procedures for Various TN-LC Baskets Bottom Basket Type Subbasket Type Appendix Spacer Required?

7.7.1, Sections Yes 7.7.1.1-2 TN-LC-NRUX 7.7.1, Sections Yes 7.7.1.1-2 7.7.2, Sections TN-LC-MTR Yes 7.7.2.1-2 7.7.3, Sections TN-LC-TRIGA Yes 7.7.3.1-2 1-PWR 7.7.4, Sections Yes 7.7.4.1-2 1-BWR 7.7.4, Sections TN-LC-1FA Yes 7.7.4.1-2 Pin Can 7.7.4, Sections No 7.7.4.1-2 Table 7-3 Appendices Containing Unloading Procedures for Various TN-LC Baskets Basket Type Subbasket Type Appendix 7.7.1, Sections 7.7.1.3-4 TN-LC-NRUX 7.7.1, Sections 7.7.1.3-4 TN-LC-MTR 7.7.2, Sections 7.7.2.3-4 TN-LC-TRIGA 7.7.3, Sections 7.7.3.3-4 1-PWR 7.7.4, Sections 7.7.4.3-4 TN-LC-1FA 1-BWR 7.7.4, Sections 7.7.4.3-4 pin can 7.7.4, Sections 7.7.4.3-4 TN-LC-0100 7-17

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 20 1 11 9 8 17 15 6 3 14 13 4 5 16 18 7 10 12 2 19 LID 1 6 8 4 3 7 5 2 TRUNNION AND BOTTOM PLUG ASSEMBLY Figure 7-1 TN-LC Packaging Torquing Patterns TN-LC-0100 7-18

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Figure 7-2 Assembly Verification Leakage Test TN-LC-0100 7-19

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Appendix 7.7.4 TN-LC-1FA Basket Wet and Dry Loading and Unloading NOTE: References in this chapter are shown as [1], [2], etc., and refer to the reference list in Section 7.5. A glossary of terms used in this chapter is provided in Chapter 7, Section 7.6.

Site-specific conditions and requirements may require the use of different equipment and ordering of steps than those described below to accomplish the same objectives or to meet acceptance criteria to ensure the integrity of the package.

7.7.4.1 TN-LC-1FA Basket Wet Loading The starting condition for the following steps assumes completion of the preparation steps in Section 7.1.1 and steps 1-14 of Section 7.1.2 of Chapter 7.

The wet loading procedure described in this section is applicable only when using the TN-LC cask for loading LWR spent fuel assemblies (SFA) or fuel rods from a fuel pool into the TN-LC cask with TN-LC-1FA basket, which is submerged in a fuel pool.

A TN-LC cask with a TN-LC-1FA basket may be configured in one of three configurations:

A TN-LC-1FA basket for transporting a PWR SFA, A BWR sleeve and hold-down ring placed inside the TN-LC-1FA basket when transporting a BWR SFA, or A TN-LC-1FA pin can placed inside the BWR sleeve when transporting individual LWR fuel rods.

Spacers may be required for shorter SFAs/rods. The TN-LC-1FA pin can may be loaded prior to placement in the cask or loaded while in the cask.

1. Verify that the TN-LC-1FA basket is configured appropriately for one of the three payloads as listed above.

2. The potential for fuel misloading is essentially eliminated through the implementation of procedural and administrative controls. The controls instituted to ensure that acceptable SFAs or rods are placed into the TN-LC cask consist of the following:

A package loading plan is developed to verify that candidate SFAs or rods meet the fuel qualification requirements of the applicable sections as listed in step 12 of Section 7.1.1.

The loading plan, including the number of locations of PRAs if required, is independently verified and approved before fuel load.

A fuel movement schedule is then written, verified, and approved based upon the loading plan. All fuel movements from any rack location are performed under strict compliance with the fuel movement schedule.

TN-LC-0100 7.7.4-1

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20

3. Prior to loading SFAs or fuel rods into a TN-LC-1FA basket, the identity of the spent fuel is to be verified by two individuals using a video camera or other means. Read and record the identification number from the SFA/fuel rods, if applicable, and check this identification number against the site loading plan which indicates which SFAs/fuel rods are acceptable for transport.

4. Position the SFA or fuel rod for insertion into the TN-LC-1FA basket compartment:

If loading an SFA, Load the PWR or BWR SFA as applicable into the TN-LC-1FA basket.

Record the identity of the SFA.

If loading fuel rods, Load the fuel rods to be transported into the TN-LC-1FA pin can.

After the TN-LC-1FA pin can has been loaded, check and record the identity of the fuel rods.

Place a fuel can spacer as required.

Install the pin can lid.

Following completion of above listed steps, the TN-LC cask is ready for draining as described in step 15, Section 7.1.2 of Chapter 7.

TN-LC-0100 7.7.4-2

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.7.4.2 TN-LC-1FA Basket Dry Loading The starting condition for the TN-LC-1FA basket assumes completion of the preparation steps in Section 7.1 and steps 1-4 of Chapter 7, Section 7.1.3.

A TN-LC cask with a TN-LC-1FA basket may be configured in one of three configurations:

A TN-LC-1FA basket for transporting a PWR SFA, A BWR sleeve and hold-down ring placed inside the TN-LC-1FA basket when transporting a BWR SFA, or A TN-LC-1FA pin can placed inside the BWR sleeve when transporting individual LWR fuel rods.

Dry loading of a PWR/BWR fuel assembly must be conducted in a helium environment. Spacers may be required for shorter SFAs/rods. The TN-LC-1FA pin can may be loaded prior to placement in the cask or loaded while in the cask. This procedure assumes the TN-LC-1FA pin can is loaded with fuel rods prior to loading the can into the TN-LC cask.

1. Verify that the TN-LC-1FA basket is configured appropriately for one of the three payloads as listed above.

2. The potential for fuel misloading is essentially eliminated through the implementation of procedural and administrative controls. The controls instituted to ensure that acceptable SFAs or rods are placed into the TN-LC cask consist of the following:

A package loading plan is developed to verify that candidate SFAs or rods meet the fuel qualification requirements of the applicable sections as listed in step 12 of Section 7.1.1.

The loading plan is independently verified and approved before fuel load.

A fuel movement schedule is then written, verified, and approved based upon the loading plan. All fuel movements from any rack location are performed under strict compliance with the fuel movement schedule.

3. Prior to loading SFA or fuel rods into a TN-LC-1FA basket, the identity of the spent fuel is to be verified by two individuals using a video camera or other means. Read and record the identification number from the SFA/fuel rods, if applicable, and check this identification number against the site loading plan which indicates which SFAs/fuel rods are acceptable for transport.

4. Position the SFA or TN-LC-1FA pin can for insertion into the TN-LC-1FA basket compartment:

If loading an SFA, TN-LC-0100 7.7.4-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Load the PWR or BWR SFA as applicable into the TN-LC-1FA basket.

Record the identity of the SFA.

If loading a TN-LC-1FA pin can, Load the fuel rods to be transported into the TN-LC-1FA pin can.

After the TN-LC-1FA pin can has been fully loaded, check and record the identity of the fuel rods.

Place a fuel can spacer as required.

Install the pin can lid.

Position the TN-LC-1FA pin can for insertion into the TN-LC cask. Note that this step is done with the TN-LC cask oriented horizontally.

The TN-LC-1FA pin can is inserted into the cask cavity using site transfer equipment.

Assemble the loaded TN-LC cask as required prior to upending.

Upend the TN-LC cask.

Following completion of the above listed steps, the TN-LC-1FA basket dry loading is continued in step 5, Section 7.1.3 of Chapter 7.

TN-LC-0100 7.7.4-4

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.7.4.3 TN-LC-1FA Basket Wet Unloading The starting condition for the following steps assumes completion of the unloading preparation steps in Section 7.2.1 and steps 1-12 of Section 7.2.2.1 of Chapter 7.

The TN-LC-1FA contents may be unloaded directly from the cask or the TN-LC-1FA basket may be removed from the cask, staged, and unloaded away from the cask. The sequence below assumes that the contents are unloaded directly from the cask.

1. If unloading BWR or PWR SFAs:

If necessary, remove fuel spacer.

Remove TN-LC-1FA BWR hold-down ring if unloading a BWR SFA.

Unload the PWR or BWR SFA as applicable using the appropriate grapple.

Place the SFA in the appropriate location in the pool.

Replace the BWR hold-down ring, if necessary.

2. If unloading fuel rods:

Remove TN-LC-1FA BWR hold-down ring if necessary to gain access to the TN-LC-1FA pin can lid.

Remove the TN-LC-1FA pin can lid.

Unload the fuel rods from the TN-LC-1FA pin can using the appropriate handling tool.

After all fuel rods have been removed, replace the TN-LC-1FA pin can lid.

Replace the TN-LC-1FA BWR hold-ring if necessary.

Following completion of the above steps, the TN-LC-1FA basket wet unloading is continued in step 13, Section 7.2.2.1 of Chapter 7.

TN-LC-0100 7.7.4-5

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 7.7.4.4 TN-LC-1FA Basket Dry Unloading The starting condition for the following steps assumes completion of the unloading preparation steps in Section 7.2.1 and steps 1-6 of Section 7.2.2.2 or Section 7.2.2.3 of Chapter 7 for SFA or fuel rod unloading, respectively.

1. If unloading BWR or PWR SFAs:

If necessary, remove fuel spacer.

Remove the TN-LC-1FA BWR hold-down ring if unloading a BWR SFA.

Unload the PWR or BWR SFA as applicable using the appropriate grapple.

Place the SFA in the appropriate location in the hot cell.

Replace the BWR hold-down ring, if necessary.

Following completion of the above listed steps, the TN-LC-1FA dry unloading is continued in step 7, Section 7.2.2.2 of Chapter 7.

2. If unloading fuel rods from the TN-LC-1FA pin can:

Note that the TN-LC-1FA pin is unloaded while the TN-LC cask is horizontal.

Install temporary shielding around the bottom plug and remove the bottom plug.

Attach the transfer system ram to the base of the TN-LC-1FA pin can and push the TN-LC-1FA pin can out of the TN-LC cask.

Stage the TN-LC-1FA pin can as required for removal of the fuel rods.

Retract the TN-LC-1FA pin can back into the TN-LC cask. Alternately, the ram may be disconnected, and the TN-LC-1FA pin can may be left in the hot cell work area.

Remove the transfer system ram as necessary to allow reattachment of the bottom plug.

Reattach the bottom plug.

Following completion of the above listed steps, the TN-LC-1FA dry unloading is continued in step 7 Section 7.2.2.3 of Chapter 7.

TN-LC-0100 7.7.4-6

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Alternatives to the Code are described in Chapter 2, Section 2.1.4 and Appendix 2.13.13.

The TN-LC impact limiter welds are designed, fabricated, and inspected in accordance with the AWS Structural Welding Code - Stainless Steel [17] or the ASME Code,Section III Subsection NF [1], and Section IX [5].

TN-LC-0100 8-1a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 8.1.3 Structural and Pressure Tests 8.1.3.1 Load Tests One set of trunnions is provided for the TN-LC transport package lifting. The trunnions have a single shoulder (single failure proof). The trunnions are fabricated and tested in accordance with ANSI N14.6 [3]. A load test of 3.0 times the design lift load (for single failure proof trunnions) is applied to the trunnions for a period of ten minutes to ensure that the trunnions can perform satisfactorily.

A force equal to 1.5 times the impact limiter weight will be applied to the hoist rings of each impact limiter for a period of ten minutes. At the conclusion of the test, the impact limiter hoist rings will be examined visually for defects and permanent deformation.

8.1.3.2 Pressure Tests A pressure test is performed on the TN-LC cask at a pressure between 45.0 and 50.0 psig. This is well above 1.5 times the maximum normal operating pressure of 16.9 psig (Chapter 3, Table 3-8). The test pressure is held for a minimum of ten minutes. The test is performed in accordance with ASME B&PV Code,Section III, Subsection NB, Paragraph NB-6200 or NB-6300. All visible joints and surfaces are examined visually for possible leakage after application of the pressure.

In addition, a bubble leakage test is performed on the neutron shield enclosure. The purpose of this test is to identify any potential leakage paths in the enclosure welds.

8.1.4 Containment Boundary Leakage Tests 8.1.4.1 TN-LC Cask Leakage Tests Leakage tests are performed on the TN-LC cask containment boundary prior to first use, typically at the fabricators facility. The fabrication verification leakage test can be separated into the following five tests: 1) cask leakage integrity, 2) vent port plug seal integrity, 3) drain port plug seal integrity, 4) lid seal integrity, and 5) bottom plug seal integrity. These tests are usually performed using the helium mass spectrometer method. Alternative methods are acceptable provided that the required sensitivity is achieved. The leakage test is performed in accordance with ANSI N14.5 [4]. The personnel performing the leakage test are qualified in accordance with SNT-TC-1A [2].

8.1.4.1.1 Cask Leakage Integrity Test Prior to lead pour and final machining of the inner shell, the containment boundary, including containment boundary base metal and welds, will be leakage tested in accordance with the requirements of ANSI N14.5 using temporary closures and seals, as necessary, for the bottom plug and lid. As the inner shell will not be accessible for leakage testing after lead is poured, leakage testing will be performed during the fabrication process as permitted by ANSI N14.5 Table 1. As one means of performing a portion of this test, the interior of the cask cavity may be flooded with a helium atmosphere while a vacuum is drawn on the lead cavity to TN-LC-0100 8-2

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 determine the leakage rate. If a leak is discovered, the source will be determined and repaired, and the shell will be retested to ensure that the measured leakage rate is less than 1 x 10-7 ref cm3/s.

Similarly, the lid forging will not be accessible for leakage testing after lead is poured into the forging or after a machined lead piece is installed and captured with the gamma shielding cap.

The lid forging will be leak tested using temporary closures and seals. If leakage is found, the leak will be repaired and the forging retested, as described above for the inner shell, prior to lid gamma shielding installation.

The leakage tests will be performed in conjunction with the non-destructive examination of the inner shell welds in accordance with ASME B&PVC Code,Section III, Subsection NB. Liquid penetrant examination of all final machined weld surfaces of the inner shell will be performed per the Code.

8.1.4.1.2 Fabrication Verification Leakage Tests The fabrication verification leakage tests include the following:

• Cask vent port plug seal integrity

• Cask drain port plug seal integrity

• Cask lid seal integrity

• Bottom Plug seal integrity The tests will be performed as described in Chapter 7, Section 7.4.1, in accordance with ANSI N14.5. The acceptance criterion requires each component to be individually leak tight, that is, the leakage rate must be less than 1 x 10-7 ref cm3/s.

8.1.5 TN-LC Cask Component and Material Tests 8.1.5.1 Valves, Rupture Discs, and Fluid Transport Devices There are no valves, rupture discs, or couplings in the containment of the TN-LC packaging.

8.1.5.2 Gaskets The lid and all the other containment penetrations are sealed using O-ring seals. Leakage testing of the seals is described in Section 8.1.4.1.

8.1.5.3 Impact Limiter Leakage Test Prior to initial use, after all the wood blocks have been installed and the seal welds have been completed, the following test will be performed on the impact limiter to verify that the impact limiter wood is completely enclosed, thereby preventing any moisture exchange with the ambient environment.

TN-LC-0100 8-3

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Each impact limiter container will be pressurized to a pressure between 2.0 and 3.0 psig. All the weld seams and penetrations will be tested for leakage using a soap bubble test. If bubbles are detected, the weld will be repaired and the test re-performed.

8.1.5.4 Functional Tests The following functional tests will be performed prior to the first use of the TN-LC package.

Generally these tests will be performed at the fabrication facility.

(a) Installation and removal of the lid, bottom plug, vent and drain port plugs, and other fittings will be observed. Each component will be checked for difficulties in installation and removal. After removal, each component will be visually examined for damage. Any defects will be corrected prior to the acceptance of the cask.

(b) Each TN-LC-1FA basket as well as each TN-LC-MTR, TN-LC-TRIGA and TN-LC-NRUX fuel assembly/element compartment will be checked by gauge to demonstrate that the fuel assemblies or elements, as applicable, will fit in the basket.

8.1.6 Shielding Tests Chapter 5 presents the analyses performed to ensure that the TN-LC package shielding design is adequate.

8.1.6.1 Gamma Shield Test The TN-LC cask poured lead gamma shielding shall be inspected via gamma scanning at the intersections of a grid no larger than 6 x 6 inches on the outside of the shell prior to installation of the neutron shield. Lead gamma shielding that is packed into place shall be inspected for uncontrolled voids. This shall be done either by gamma scanning or by dimensional inspection and weighing to determine the average density of each part.

The acceptance criterion for the gamma scan is based on dose rate measurements of a test block constructed to replicate the layers of stainless steel, lead, and stainless steel in the TN-LC cask.

The thickness of each stainless steel layer in the test block shall be no less than the minimum specified thickness of the corresponding cask shell, and the thickness of the lead layer in the test block shall be no less than the minimum thickness of lead specified for the cask. The dose rate measured using the test block shall be the maximum acceptable reading for the inspected cask.

The source/detector distance used in the cask inspection shall be the same as that used in establishing the maximum dose rate limit. Inspection results which exceed this limit will be evaluated to ensure that the regulatory dose rate limits will not be exceeded.

TN-LC-0100 8-4

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 (b) Change of the billet production process, e.g., from vacuum hot pressing to cold isostatic pressing followed by vacuum sintering, (c) Change in the nominal matrix alloy, (d) Changes in mechanical processing that could result in reduced density of the final product, e.g., for PM or thermal spray MMCs that were qualified with extruded material, a change to direct rolling from the billet, (e) For MMCs using a magnesium-alloyed aluminum matrix, changes in the billet formation process that could increase the likelihood of magnesium reaction with the boron carbide, such as an increase in the maximum temperature or time at maximum temperature, (f) Changes in powder blending or melt stirring processes that could result in less uniform distribution of boron carbide, e.g., change in duration of powder blending, and (g) For MMCs with an integral aluminum cladding, a change greater than 25% in the ratio of the nominal aluminum cladding thickness (sum of two sides of cladding) and the nominal matrix thickness could result in changes in the mechanical properties of the final product.

In no case shall process changes be accepted if they result in a product outside the limits established in Sections 8.1.7.8.1 and 8.1.7.8.4.

8.1.8 Thermal Tests The thermal evaluation of the TN-LC package described in Chapter 3 is performed using very conservative and bounding assumptions. Gaps between the components are modeled in the thermal analysis to account for possible gaps expected during fabrication. Gaps are assumed to be present during NCT and HAC post-fire cases when calculating heat flow out of the cask, and gaps are assumed closed when calculating heat flow into the cask (i.e., during the HAC fire).

The calculated cladding temperatures are much lower than the cladding temperature limit, assuring a large margin of safety.

For these reasons, thermal acceptance testing is not required for the TN-LC package 8.1.9 Impact Limiter Wood Test Mechanical properties of the energy absorbing wood used in the impact analysis are listed in the following table:

Specified Properties of Impact Limiter Wood Wood Type Density Moisture Content Crush Stress Parallel to Grain Minimum (lb/ft3) (%) (psi) Lockup Strain (in/in)

Balsa 7-12 6-12 1498-1900 0.8 Redwood 18.7-27.5 6-12 5000-7000 0.6 TN-LC-0100 8-13

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 8.2 Maintenance Program 8.2.1 Structural and Pressure Tests Within 14 months prior to any lift of a TN-LC package, the trunnions shall be subject to either of the following:

• A test load equal to 300% of the maximum service load per ANSI N14.6 [3], paragraph 7.3.1(a) for single failure proof trunnions.

• Dimensional testing, visual inspection and nondestructive examination of accessible critical areas of the trunnions including the bearing surfaces in accordance with Paragraph 6.3.1(b) of ANSI N14.6 [3].

8.2.2 Leakage Tests The following containment boundary components shall be subject to periodic maintenance, and preshipment leakage testing in accordance with ANSI N14.5 [4]:

• Lid and seals

• Bottom Plug and seals

• Vent Port Plug Seal

• Drain Port Plug Seal Leakage Tests for NRUX, TRIGA and MTR Shipments Typical Method (ANSI N14.5 Test Frequency Acceptance Criteria TABLE A.1 [4])

(He)

Within 12 months prior to Each component individually Periodic A.5.3 shipment 1x10-7 ref cm3/s A.5.4 A.5.1 Before each shipment, after No detected leakage, sensitivity A.5.2 Pre-shipment the contents are loaded and of 10-3 ref cm3/s or better, unless A.5.8 the package is closed seal is replaced.

A.5.9 After maintenance, repair, (He) or replacement of Each component individually Maintenance A.5.3 containment components, 1x10-7 ref cm3/s A.5.4 including inner seals TN-LC-0100 8-14

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 Leakage Tests for 1FA Shipments Typical Method (ANSI N14.5 Test Frequency Acceptance Criteria TABLE A.1 [4])

A.5.1 Each component individually Within 12 months prior to A.5.2 Periodic 1.57x10-5 ref cm3/s with a test shipment A.5.8 sensitivity of 7.83x10-6 ref cm3/s A.5.9 A.5.1 Before each shipment, after No detected leakage, sensitivity A.5.2 Pre-shipment the contents are loaded of 10-3 ref cm3/s or better, unless A.5.8 and the package is closed seal is replaced.

A.5.9 After maintenance, repair, A.5.1 Each component individually or replacement of A.5.2 Maintenance 1.57x10-5 ref cm3/s with a test containment components, A.5.8 sensitivity of 7.83x10-6 ref cm3/s including inner seals A.5.9 No leakage tests are required prior to shipment of an empty TN-LC packaging.

8.2.3 Component and Material Tests The TN-LC cask shall be inspected in accordance with the requirements of 10 CFR Part 71.87, Routine determinations, part (b) prior to each shipment. Any defects or signs of degradation discovered by these inspections for any component (including accessible welds and fasteners) or feature would be repaired and brought into compliance with the licensing drawings prior to shipment of the loaded package.

TN-LC-0100 8-14a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 8.2.3.1 Fasteners All threaded fasteners and port plugs shall be inspected at the time of use for deformed or stripped threads. Damaged parts shall be evaluated for continued use and replaced as required.

At a minimum, the TN-LC lid and bottom plug bolts shall be replaced at least every 75 shipments (round trip) to ensure adequate fatigue strength is maintained.

8.2.3.2 Impact Limiters A visual examination of the impact limiters before each shipment will be performed to ensure that the impact limiters have not degraded. If there is no evidence of weld cracking or other damage which could result in water in-leakage, the wood performance is assured. If there is evidence of weld cracking or any other damage that could result in water in-leakage, the entire block of wood in the affected damaged segment(s) will be replaced, and the entire impact limiter will be leakage tested using the same acceptance criteria as the original leak tests. Impact limiters will be leakage tested once every five years using the same acceptance criteria as the original leak tests to ensure that water has not entered the impact limiters. If the leakage test indicates that the impact limiters have a leak, a humidity test will be performed to verify that there is no free water in the impact limiters.

8.2.3.3 Valves, Rupture Discs, and Gaskets on Containment Vessel If the bottom plug or the lid is removed, the seals shall be inspected prior to transport of a loaded TN-LC package. The seals will be leakage tested after retorquing the bolts in accordance with Chapter 7, Section 7.4.

O-ring seals may be reused for transport of an empty TN-LC packaging. O-ring seals shall be inspected prior to each shipment of a loaded TN-LC package and replaced at least every twelve months.

There are no valves, rupture discs, or couplings on the containment of the TN-LC packaging.

8.2.3.4 Shielding There are no periodic tests or inspections required for the TN-LC gamma or neutron shielding.

As described in Chapter 7, radiation surveys will be performed on the package exterior to ensure that the limits specified in 10 CFR 71.47 are met prior to each shipment.

The material composition of the VYAL-B or Resin-F neutron shielding resin employed in the shielding calculations is based on minimum guaranteed values that are determined by extensive tests under various (including extreme) environmental conditions. These tests indicate that the neutron shielding resin does not degrade under normal conditions and is durable over extended periods of time. The shielding calculations employed are based on conservative models and design basis source terms and demonstrate that the dose rate criteria are satisfied with sufficient margin. The comparisons of calculated and measured dose rate have indicated that the calculated dose rates are highly conservative. The 10 CFR Part 71 dose rate compliance TN-LC-0100 8-15

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 measurements serve to indicate the shielding effectiveness of the package. Therefore, periodic tests for the neutron shielding resin are not necessary.

TN-LC-0100 8-15a

TN-LC Transportation Package Safety Analysis Report Revision 9a, 04/20 8.3 References

1. ASME Boiler and Pressure Vessel Code,Section III and Appendices, 2004 Edition including 2006 addenda.

2. SNT-TC-1A, American Society for Nondestructive Testing, Personnel Qualification and Certification in Nondestructive Testing.

3. ANSI N14.6-1993, American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials.

4. ANSI N14.5-2014, American National Standard for Leakage Tests on Packages for Shipment of Radioactive Materials.

5. ASME Boiler and Pressure Vessel Code,Section IX, 2004 Edition including 2006 addenda.

6. Aluminum Standards and Data, 2003, The Aluminum Association.

7. Natrella, Experimental Statistics, Dover, 2005.

8. ASTM E1225, Thermal Conductivity of Solids by Means of the Guarded-Comparative-Longitudinal Heat Flow Technique.

9. ASTM E1461, Thermal Diffusivity of Solids by the Flash Method.

10. Sung, C., Microstructural Observation of Thermally Aged and Irradiated Aluminum/Boron Carbide (B4C) Metal Matrix Composite by Transmission and Scanning Electron Microscope, 1998.

11. Boralyn testing submitted to the NRC under docket 71-1027, 1998.

12. ASTM B557, Standard Test Methods of Tension Testing Wrought and Cast Aluminum and Magnesium-Alloy Products.

13. ASTM E290, Standard Methods for Bend Testing of Materials for Ductility.

14. ASTM E94, Recommended Practice for Radiographic Testing.

15. ASTM E142, Controlling Quality of Radiographic Testing.

16. ASTM E545, Standard Method for Determining Image Quality in Thermal Neutron Radiographic Testing.

17. AWS D1.6/D1.6M, Structural Welding Code - Stainless Steel.

TN-LC-0100 8-17

Enclosure 4 to E-56457 AFFIDAVIT PU RS UA NT TO 10 CF R 2.390 TN Americas LLC )

St ate of Maryland ) ss.

County of Howard )

I, Prakash Narayanan, "epo and say that I am Chief Techni cal Officer of TN Americas LL C, du ly au th or ize d to execute th is af fid a 'it, and have reviewed or caus ed to ha ve reviewed the info11nation which is identified as proprietary and r fe ri need in the paragraph immediat ely below. I am submitting this affidavit in confor111ance with the provisions of 0 CFR 2.390 of the Commission's regulations for withholding this in for ,nation.

Th e info1mation for which proprietary treatment is sought is co below: ntained in Enclosure 2 and is lis ted

• Portions of certain chapters and appendices of the Safety Analysis Re port (SAR) for Ce rti fic ate of Compliance No. 9358 TN-LC, Revision 9a, Do cket 71-9358 (Proprietary an d SUNSI Version)

Th is document has been appropriately designated as proprietar y.

I have personal knowledge of the criteria and procedures utilized by

• TN Americas LLC in designating

• info1111ation as a trade secret, privileged, or as confidential comm ercial or financial info1mation..

Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Co m m iss io n's reg11lations, the following is furnished for consideration by the Commission in de ter111ining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

I) The info1mation sought to be withheld from public disclo sure involves certain design details associated wi th the SAR analyses, calculations, and SAR drawings for the TN-LC System, which are owned and ha ve been he ld in confidence by TN Americas LL C.

2) The information is of a type customarily held in confidence by TN Americas LLC an d not customarily disclosed to the public. TN Americas LLC has a rational basis for dete11r1ining the types of info1mation customarily held in confidence by it.

3) Public disclosure of the infonnation is likely to ca11se substan tial harm to the competitive position of TN Americas LLC because the information consists of descriptions of the design and analysis of a radioactive material transportation system, the application of which provide a competitive economic advantage. Th e availability of such info1n1ation to competitor s would enable th em to modify their product to better compete with TN Americas LLC, take marketing or other actions to improve their product's position or impair the position of TN America LLC' s prod uct, and avoid developing similar data and analyses in support of their processes, methods or appa ratus.

Further the deponent sayeth not.

Prakash Narayanan Chief Technical Officer, TN Americas LLC e this 22nd day of April, 2020.

RO, OAJONES NOTARY PUBLIC MONTGOMERY COUNTY MARYLA D COMMISSION EXPIRES OCT. 16, 2023 Page 1 of 1