ML22060A044

Response to Request for Additional Information for Application for Revision 6 of Certificate of Compliance No. 9358 for the Model No. TN-LC

ML22060A044
Person / Time
Site:	07109358
Issue date:	02/28/2022
From:	Shaw D Orano TN Americas, TN Americas LLC
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
Shared Package
ML22060A043	List: ML22060A044
References
E-60264
Download: ML22060A044 (79)

Text

February 28, 2022 E-60264 Orano TN 7160 Riverwood Drive U. S. Nuclear Regulatory Commission Suite 200 Columbia, MD 21046 Attn: Document Control Desk USA One White Flint North Tel: 410-910-6900 11555 Rockville Pike Fax: 434-260-8480 Rockville, MD 20852

Subject:

Response to Request for Additional Information for Application for Revision 6 of Certificate of Compliance No. 9358 for the Model No.

TN -LC, Docket No. 71-9358

References:

(1) TN Letter E -59073 Application for Revision 6 of Certificate of Compliance No. 9358 for the Model No. TN -LC, Docket No.

71-9358, dated August 9, 2021.

(2) Letter to Don Shaw (TN) from Pierre Saverot (NRC), Request for Additional Information for the Review of the Model No. TN-LC Package, dated December 21, 2021, Docket Number 71-9358, Enterprise Project Identifier (EPID) No. L-2021-LLA -0148 (3) NRC Certificate of Compliance for the Model No. TN -LC, USA/9358/B(U)F-96, Rev ision 5 In accordance with 10 CFR 71.38, TN Americas LLC (TN) made a submission of an application to revise Certificate of Compliance (CoC) No. 9358 for the TN -LC packaging [1]. The NRC requested additional information (RAI) needed to continue the review of the application [2].

Enclosure 1 provides the responses to the RAI. In addition to this enclosure providing the response to the RAI, details for a change to the material specification for the containment boundary O-rings that is not related to the RAI is included as Enclosure 2.

Preliminary changed SAR pages are provided as Revision 10B in Enclosure 3. A consolidated SAR Revision 10 will be submitted upon completion of the NRC review.

The changed pages are indicated by Revision 10, 07 /21 in the header of the page.

Each changed page includes a revision bar adjacent to the changed content and the changes made relating to Revision 10B are gray shaded to distinguish them from the Revision 10A changes to the SAR. A public version of the Revision 10B SAR changed pages with proprietary information redacted is provided for public availability as Enclosure 4.

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

E-60264 Document Control Desk Page 2 of 2

The NRC Electronic Information Exchange (EIE) system is used for submission of this application.

A set of computer calculation files is included as Enclosure 6. Enclosure 5 provides a listing of these computer files. Because the Enclosure 6 computer calculation files exceed the size limit allowed by the NRC EIE application process, they are provided separately on one computer disk.

Proposed changes to the NRC Certificate of Compliance [3] are annotated and provided as.

Certain portions of this submittal include proprietary information. In accordance with 10 CFR 2.390, TN Americas is providing an affidavit (Enclosure 8 ) requesting that this proprietary information be withheld from public disclosure.

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 Licensing Manager TN Americas LLC

Enclosures:

1. Responses to RAIs

2. Summary of Changes to TN-LC SAR Revision 10B Not Related to RAIs

3. TN-LC Transportation Package Safety Analysis Report Revision 10B Changed Pages (Proprietary Version)

4. TN-LC Transportation Package Safety Analysis Report Revision 10B Changed Pages (Public Version)

5. Listing of Computer Files Contained in Enclosure 6

6. Computer Files Associated with UFSAR Revision 10B (Proprietary)

7. Proposed Certificate of Compliance No. 9358, Revision 6 Markup

8. 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 Kamran Travassoli, Project Manager, TN Americas LLC

RAIs and Responses Enclosure 1 to E-60264

Chapter 2 - Structural Evaluation:

RAI 2-1 :

Evaluate applicable Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) scenarios for the Fuel Assembly Can (FAC).

In Section 1.4.5.2 of the application, the applicant discussed the TN -LC -1FA Basket contents which includes the new damaged fuel assembly can (FAC) intended for use with damaged fuel assemblies. The applicant only addressed thermal expansion of the FAC in this amendment request, particularly in Section 2.13.10.2 of the application. The staff could not find any additional evaluations for NCT or HAC for the FAC or a justification why it remains bounded by other analyses already performed.

It is not clear to the staff how the FAC will behave under NCT and HAC scenarios, mainly drop events. In particular, the applicant needs to clarify if impact loads have any effects (i.e.,

deformations) on the FAC which could then have potential effects on the fuel geometry and therefore on the criticality evaluations. The applicant needs to evaluate how the fuel assemblies, contained within the FAC, will continue to remain subcritical under the applicable NCT and HAC scenarios.

This information is required for the staff to determine compliance with 10 CFR 71.55(e), 71.71 and 71.73.

Response to RAI 2-1 :

The fuel assembly can (FAC) is described in SAR Section 1.4.5.2.1 and Drawing 65200-71-91 of the TN-LC SAR.

Per SAR Section 1.4.5.2.1, the thickness of the FAC top lid (Item 5 of Drawing 65200-71-91) is 1.25 inches, and the thickness of the bottom cover ( Item 2 of Drawing 65200-71-91) i s 1.25 inches (including the thickness of the bottom spacer bars ), both of which are greater than the axial gap of 1.00 inch between the basket length of 181.50 inches and cask cavity length of 182.50 inches. Therefore, both top lid and bottom cover are always confined within the fuel compartment, and cannot move outside the fuel compartment on either end, regardless of the basket axial location within the cask cavity during NCT or HAC.

The top lid and bottom cover are design features of the packaging that ensure that any fuel material is always confined within the fuel compartment square boundary, and fuel material cannot be released outside the fuel compartment square boundary into the space between the 1FA basket and TN-LC inner cavity. Figure 2-1 -1 shows the top of the cask, where the fuel compartment square boundary is shown in green, the basket is shown in grey, and the FAC top lid is shown in blue. No fuel material can be released into the volumes marked in red because of the presence of the FAC top cover plate. A similar configuration with a gap between the basket and inner cavity could exist at the bottom of the cask if the basket were to slide to contact the TN-LC lid.

Page 1 of 7 RAIs and Responses Enclosure 1 to E-60264

Figure RAI 2-1 -1 Fuel Compartment Square Boundary (not to scale)

The function of the top lid and bottom cover would be accomplished even in the event of a failure of the welds that connect the top lid (Item 5 of Drawing 65200-71-91) to the top lid centering liner (Item 4 of Drawing 65200-71-91), or of the weld that connects the bottom cover (Item 2 of Drawing 65200-71-91) to the liner (Item 1 of Drawing 65200-71 -91), or of the liner itself. The thicknesses of the top lid and bottom cover alone without the liner are sufficient to achieve the intended function of the FAC. The liner with bottom cover is an operational feature to allow for fuel handling with the FAC attached to the fuel assembly. The top lid and bottom cover perform the safety function of confining fuel material within the 1FA basket fuel compartment, and the liner is not required for the contents to remain subcritical.

The liner is not important in keeping the top lid and bottom cover in place to achieve the intended function of the FAC during transportation. For side drops, both the top lid and bottom cover are self-supported and self-loaded. Also, both plates are not slender; therefore, there is no risk of buckling. For end drop on the lid, the top cover plate is loaded with the fuel assembly through bearing on the TN-LC lid. There is no deformation of the top lid that would allow debris to escape from the fuel compartment. For bottom end drop, the bottom cover of the FAC is loaded by the fuel assembly, which may lead to some deformation of the plate because the spacer bars provide a point of contact for bending. During an end drop or side drop the debris is restricted from escaping the fuel compartment. This ensures debris will not accumulate in the gap between the basket and TN -LC ends. Therefore, the top lid and bottom cover perform the intended safety function described above during both NCT and HAC impacts.

Page 2 of 7 RAIs and Responses Enclosure 1 to E-60264

The criticality evaluation for the 1FA basket with damaged PWR fuel assembly is provided as Appendix 6.10.5 in the SAR. The FAC structure is not required to demonstrate that the pressurized water reactor (PWR ) contents remain subcritical during routine, normal, or accident transport conditions. The determination of the most reactive damaged fuel configuration from a cask-drop accident is performed in SAR Section 6.10.5.4.2.3. It assumes fresh water intrusion in conjunction with extended damaged fuels condition and fuel reconfiguration including single-ended, double-ended rod shear, missing rods and pitch expansion. The bent or bowed fuel rods scenario is bounded by the pitch expansion configuration, which assumes that the fuel rods remain within the lattice but not in its nominal fuel rod pitch. It is possible that the fuel rods may be crushed inward or bow outward to a certain degree. Rearrangement of the PWR fuel assembly has been evaluated by varying the fuel rod pitch from a minimum pitch that is limited by the clad outer diameter, and a maximum pitch that is limited by the 1FA basket fuel compartment inner width (i.e., the FAC liner is not modeled). Further cladding failure is considered assuming fully de-cladded fuel rods.

The presence of the FAC does not change the assumptions for fuel rearrangement that are used in the criticality evaluation. Any deformation of the FAC liner does not change the assumptions for fuel rod pitch, and the presence of the top lid and bottom cover confine the fuel assembly rods and any fuel debris from a damaged fuel assembly to the geometry of 1FA basket fuel compartment.

Drawing No. 65200-71-91 has been revised on sheet 1 to show construction details for Item 6, to show Items 1, 4, and 6 as not-important to safety, to allow Item 7 to be optional under certain conditions, and to make an editorial change. Sheet 2 of the drawing has been revised to show details between Items 1 and 6.In addition, Appendix 1.4.1 has been updated to reflect the drawing revision change to Revision 10.

Impact:

SAR Drawing No. 65200-71-91 and Appendix 1.4.1 have been revised as described in the response.

Page 3 of 7 RAIs and Responses Enclosure 1 to E-60264

Chapter 6 - Criticality Evaluation:

RAI 6-1 :

Revise the application to address the potential for fuel assembly misload considering the burnup, enrichment, and cooling time of the total population of discharged PWR fuel assemblies for the burnup credit analysis of the TN -LC package.

Section 6.10.5.8 of the SAR discusses the analy sis to determine misload loading curves for the package which are shown in Tables 6.10.5-28 and 6.10.5-29 for the WE 14x14 and WE 17x17 fuel assembly classes, respectively.

However, the applicant does not discuss how the misload loading curves relate to the probability of having a misload, by comparing the loading curves to the characteristics of the discharged fuel population. NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, recommends that a severely underburned assembly for misload analysis should be chosen such that the misloaded assemblys reactivity bounds 95 percent of the discharged PWR fuel population considered unacceptable for loading in the transportation package with 95-percent confidence. Demonstration that the misload loading curves bound the discharged fuel population is typically accomplished by comparing that curve with the characteristics of discharged PWR fuel assemblies allowed for loading in the package. The NRC finds the results of the most recent U.S. Energy Information Administrations Nuclear Fuel Data Survey (RW -859) or later similar fuel data sources (i.e., GC -859), acceptable to estimate the discharged fuel population characteristics.

This information is required for the s taff to ensure that the Model No. TN -LC package will meet the criticality safety requirements of 10 CFR 71.55 when loaded with the contents described in the application.

Response to RAI 6-1 :

SAR Section 6.10.5.8 has been updated to add a discussion and new figures of misload loading curves vs. discharged spent fuel inventory (Figure 6.10.5-25 and Figure 6.10.5-26). Department of Energy (DOE) compiled fuel data is employed for this purpose; the data consists of existing and projected spent nuclear fuel assemblies that are potentially available for long-term storage at a repository. The data encompasses more than one hundred thousand entries of burnup and enrichment combinations from RW-859.

The study shows the misload loading curves bound 99%+ of the discharged population, and is shown in new Table 6.10.5-30. Note that the set of misload curves in Table 6.10.5-28 and Table 6.10.5-29 has been updated using ORIGEN-APR library with control rods insertion since the set of base loading curves with ORIGEN-APR library with control rods insertion is more penalizing (slightly lower enrichments) compared to that obtained with ORIGEN -APR library with BPR (see the Response to RAI 6-4 for further details ).

Impact:

SAR Section 6.10.5.8 and Tables 6.10.5-28 and 6.10.5-29 have been revised as described in the response.

SAR Table 6.10.5-30 and Figures 6.10.5-25 and 6.10.5-26 have been added as described in the response.

Page 4 of 7 RAIs and Responses Enclosure 1 to E-60264

RAI 6-2 :

Revise the application to include procedural steps in the package operating procedures to compare the irradiation parameters of the fuel assembly being loaded to those used in the burnup credit isotopic depletion analysis.

Table 6.10.5-27 of the SAR provides a summary of the fuel irradiation parameter s employed in the isotopic depletion analysis for fuel contents analyzed using burnup credit.

However, there is no discussion in the SAR regarding if these conditions are bounding of all PWR fuel. NUREG-2216 recommends that contents specifications tied to the actual reactor operating conditions may be needed unless the operating condition values used in the evaluation can be justified as those that produce the maximum k eff values for the proposed contents. Staff notes that while the irradiation parameter v alues included in Table 6.10.5-27 are reasonably bounding for most PWR fuel assemblies, there may be assemblies in the discharged fuel population that exceed one or more of these parameter values, and which may produce a higher keff than considered in the applicants analysis.

This information is required for the staff to ensure that the Model No. TN -LC package will meet the criticality safety requirements of 10 CFR 71.55 when loaded with the contents described in the application.

Response to RAI 6 -2 :

SAR Table 6.10.5-27 shows the depletion parameters employed in the pressurized water reactor (PWR) burnup credit approach for the criticality analysis performed in Appendix 6.10.5.

The depletion parameters shown in Table 6.10.5 -27 ensure the overall conservatism of the depletion model and are therefore the bounding set of isotopic concentrations employed in the criticality analysis. Section 6.10.5.6 of Appendix 6.10.5 has been updated to provide additional discussions related to the depletion parameters. A note has been added to Table 6.10.5-27 for clarification regarding the upper-end values shown for fuel temperature, moderator temperature, soluble boron concentration, and specific power. A requirement to compare the reactor operating parameters to values shown in Table 6.10.5-27 is added in Appendix 7.7.4.

Additional administrative controls have been added to SAR Section 7.7.4.1 TN -LC -1FA Basket Wet Loading. These administrative controls include a requirement to compare the reactor operating parameters for the irradiation period of the fuel assembly against those shown in Table 6.10.5-27 to ensure compliance with the isotopic depletion analysis.

Impact:

SAR Section 6.10.5.6 and Table 6.10.5-27 and Section 7.7.4.1 have been revised as described in the response.

Page 5 of 7 RAIs and Responses Enclosure 1 to E-60264

RAI 6-3 :

Revise the application to include additional administrative procedures in the package operating procedures to ensure that the TN -LC package will be loaded with fuel that is within the specifications of the approved contents.

The operating procedures for the TN -LC package do not include any procedures specific to loading fuel assemblies under the burnup credit requirements of Tabl e 1.4.5-4a of the SAR.

NUREG-2216 recommends procedures the applicant may consider in order to protect against misloads in transportation packages that rely on burnup credit for criticality safety, including the following:

• verification of the location of high-reactivity fuel (i.e., fresh or severely underburned fuel) in the SNF pool, both before and after loading,

• qualitative verification that the assembly to be loaded is burned (visual or gross measurement),

• quantitative measurement of any fuel assembli es without visible identification numbers,

• independent, third-party verification of the loading process, including the fuel selection process and generation of the fuel move instructions

This information is required for the staff to ensure that the Model No. TN-LC package will meet the criticality safety requirements of 10 CFR 71.55 when loaded with the contents described in the application.

Response to RAI 6 -3 :

SAR Section 7.1.1 of Chapter 7 and Section 7.7.4.1 of Appendix 7.7.4 have been updated for the package operations when burnup credit is employed for criticality safety.

Impact:

SAR Sections 7.1.1 and 7.7.4. 1 have been revised as described in the response.

Page 6 of 7 RAIs and Responses Enclosure 1 to E-60264

RAI 6-4 :

Revise the application to clarify the irradiation conditions used in generating OR IGEN-ARP reactor libraries for the burnup credit analysis with respect to control rod or burnable poison rod assembly (BPRA) exposure.

Section 6.10.5.6 of the SAR discusses the generation of ORIGEN -ARP reactor libraries for use with the STARBUCS code sequence. This section states that two sets of libraries for each fuel assembly class were developed - one with BPRAs inserted into the fuel assembly guide tubes for the full irradiation period, and another with control rods inserted into the fuel assembly gui de tubes for the first 15 GWd/MTU of irradiation.

However, the SAR does not include any discussion of which of these reactor libraries results in a higher keff, or which is used to generate the final burnup credit loading curves for each assembly class. The SAR should be revised to include results of reactivity comparisons using each reactor library, and demonstration that the reactor library that results in the highest keff is the one used to determine the loading curve for each fuel assembly class.

This information is required for the staff to ensure that the Model No. TN -LC package will meet the criticality safety requirements of 10 CFR 71.55 when loaded with the contents described in the application.

Response to RAI 6 -4 :

The loading curves shown in SAR Table 6.10.5-8 and Table 6.10.5-9 were developed, respectively, for WE 14x14 and WE 17x17 fuel classes using the ORIGEN -ARP library assuming burnable poison rods (BPR) inserted in fuel assembly during the entire irradiation history. A clarification has been added to SAR Section 6.10.5.4.3 for this purpose.

Two additional loading curves for WE 14x14 and WE 17x17 fuel classes using the ORIGEN-ARP library assuming control rod ( CR) insertion as described in Section 6.10.5.6 are generated and shown in new SAR Tables 6.10.5-8A and 6.10.5-9A. Note that the maximum initial enrichments obtained with ORIGEN-ARP library assuming CR insertion are reduced compared to those with ORIGEN-ARP library assuming BPR. The maximum reduction in initial enrichment is less than 0.2 wt% U-235.

New SAR Table 1.4.5-4b has been added to Appendix 1.4.5 for the maximum planar average initial enrichment and minimum burnup combination (1 FA PWR) with control rod insertion.SAR Section 7.1 and Table 7-1, and Section 6.10.5.1.2 have been updated regarding new Table 1.4.5-4b, accordingly.

Impact:

SAR Appendix 1.4.5, Section 6.10.5.4.3, Section 6.10.5.1.2, Section 7.1, and Table 7 -1 have been revised as described in the response.

SAR Tables 1.4.5-4b, 6.10.5-8A, and 6.10.5-9A have been added as described in the response.

Page 7 of 7 Enclosure 2 to E-60264

Summary of Changes to TN-LC SAR Revision 10B Not Related to RAIs

Change 1:

Changed SAR (Chapter, Section, Appendix, Table, Drawing):

SAR Appendix 1.4.1

Description of Change:

Editorial correction to drawing revision numbers.

Justification:

Drawing revision numbers for drawings that were previously approved were corrected in Appendix 1.4.1 to reflect the latest drawings revisions. The previously approved CoC references the correct drawing revisions. There was no tracking used for this and other editorial changes.

Change 2:

Changed SAR (Chapter, Section, Appendix, Table, Drawing):

Drawing 65200-71-01

Description of Change:

Revise Drawing 65200-71-01 to update N ote 17 to allow the following additional seal materials in addition to the current one: VM835-75, VM125-75 or VX065-75.

Justification:

V1289 compound for the seals is discontinued, and because there is no equivalence allow ance, three compounds that exhibit similar temperature, elongation, and compression properties have been added as alternate seal materials.

Relevant properties for the proposed replacement seal specifications are compiled in the table below.

All three proposed alternate materials are acceptable for use at temperatures down to -40 °C.

All three alternate compounds are also rated up to 400 °F maximum steady state, and have been tested for the same short-term maximum temperature (482 °F) as the V1289 compound.

All three exhibit similar properties changes after 70h at that temperature (note: for the TN -LC the temperatures only exceed 400 °F for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />), except the elongation change for VX065 and VM125, but the drop in elongation is not deemed relevant because both materials start out with a better elongation than V1289 to begin with.

Finally, all three alternate compounds exhibit either better or similar compression set than the V1289 compound.

Based on this, these three alternate compounds are acceptable for use for the TN-LC seals.

Page 1 of 2 Enclosure 2 to E-60264

Summary of Changes to TN-LC SAR Revision 10B Not Related to RAIs

Temp. range Heat aging - 70hr Compression

(°F) Elongation @250°C/482°F Set, % of Material Type A (ASTM D573) Original Hardness (% min) Hardness Elongation Deflection Min. Max. Change Change (Max.)

(pts) ( %)

V1289 75 -50 400 151 -1 7 11 (discontinued)

VX065 75 -65 400 174 0 -13 5 VM125 75 -40 400 280 -2 -6 10 VM835 75 -40 400 215 3 8 13

See the following five pages for related information from Parker Hannifin Corporation.

Page 2 of 2 Parker Hannifin Corporation 11/03/21 O-Ring & Engineered Seals Division 2360 Palumbo Drive Lexington, KY 40509

Office 859 269 2351

Reference:

V1289-75 Obsolete

Dear Valued Customer,

We regret to share that the polymer used to make V1289-75 has been discontinued by our supplier.

V1289 was used across several industries and also carries the pedigree of QPL (Qualified Producers List) for AMS7379.

V1289 is a 75 durometer fluorocarbon, ASTM D1418 type 3 classification for low temperature flexibility. At present there is not a straightforward replacement available but is in development and will be finalized in the coming months.

Until a replacement is commercialized, alternate low temperature fluorocarbon options are listed below V1289 in the table:

Material Type A Temperature range Tg hardness V1289 75 -50 to 400°F -40°F (-40°C)

VX065 75 -65 to 400°F -49°F (-45°C)

VG286 80 -45 to 400°F -31°F (-35°C)

VG109 90 -50 to 400°F -33°F (-36°C)

VM125 75 -40 to 400°F -22°F (-30°C)

VM835 75 -40 to 400°F -24°F (-31°C)

If you have any questions regarding the performance of V1289 compared to the other material options, please contact an Applications Engineer by calling 859-335-5101, or emailing OESmailbox@parker.com.

Sincerely,

Dorothy Kern Applications Engineering Manager O-Ring & Engineered Seals Division dkern@Parker.com Compound Data Sheet O-Ring Division United States

LABORATORY TEST REPORT*

• Data compiled from previous reports

TITLE: Evaluation of Parker Compounds V1289-75, VX065-75, and VM125-75

Parker O-Ring Division 2360 Palumbo Drive Lexington, Kentucky 40509 (859) 269-2351

REPORT DATA

BASIC PROPERTIES V1289-75 VX065-75 VM125-75 Hardness, Shore A, pts. 76 75 73 Tensile Strength, psi. 1766 1658 2301 Elongation, % min. 151 174 280

HEAT AGING:

70 HRS. @ 250 C, ASTM D573 Hardness chg. pts. -1 0 -2 Tensile Strength chg, % -9 -12 -13 Elongation chg, % max. +7 -13 -6

COMPRESSION SET (Plied):

22 HRS. @ 200 C, ASTM D395 Percent of Original Deflect, Max. +11 +5 +10

FLUID RESISTANCE (Service Fluid 101):

70 HRS. @ 200 C, ASTM D471 Hardness chg. pts. -6 0 -5 Tensile Strength chg, % -15 -14 -14 Elongation chg, % max. -8 -13 +10 Volume chg, % +9 +6 +10

Low Temperature TR-10. °C, ASTM D1329 Report -38 -45 -29

COMPOUND DATA SHEET Parker O-Ring Division, North America

MATERIAL REPORT

Report Number: 92880 Date: 3/13/2013

Title:

Evaluation of Parker Compound VM835-75

Elastomer Type: Fluorocarbon (FKM)

Purpose:

To obtain typical test data.

Specification: ASTM D2000 M2HK710 A1-10 B38 E078 Z1 (Shore A Hardness 75 +/-5), Z2 Elongation 125% min, Z3 (Specific Gravity), Z4 (TR-10)

Color: Black

Recommended Temperature Range: -40°F to 400°F

Recommended For: Mineral oil and grease, ASTM No. 1 oil, IRM 902 oil, IRM 903 oil, non-flammable hydraulic fluids, silicone oils and greases, aliphatic hydrocarbons (propane, butane, natural gas), aromatic hydrocarbons (benzene, toluene), chlorinated hydrocarbons (trichloroethylene and carbon tetrachloride), gasoline, high vacuum, ozone, weather, and aging resistance.

Not Recommended For: Glycol based brake fluids, ammonia gas, amines, alkalis, superheated steam, and low molecular weight organic acids (formic and acetic acids).

Additional Approvals: N/A

Parker O-Ring Division 2360 Palumbo Drive Lexington, Ky 40509 1 of 2 (859) 269-2351 REPORT DATA

Test Spec Test Original Physical Properties Method Limits Results (Z1) Hardness, Shore A, pts. ASTM D2240 75 +/-5 78 Tensile Strength, PSI (Mpa) ASTM D412 1450 (10) 3059 (Z2) Ultimate Elongation, % ASTM D412 125 215 (Z3) Specific Gravity ASTM D297 as received 1.84

Fluid Resistance (Basic Requirement)

IRM 903, 70 hrs @ 302°F Volume Change, % ASTM D471 +10 +2

(A1-10) Heat Age 70 hrs. @ 482°F Hardness Change, pts. ASTM D573 +10 +3 Tensile Strength Change, % -25 -22 Ultimate Elongation Change, % -25 +8

(B38) Compression Set (Plied) 22 hrs. @ 392°F Percent of Original Deflection, Max ASTM D395 Method B 50 13

(E078) Fluid Resistance Service Fluid 101, 70 hrs @ 392°F Hardness Change, pts. ASTM D471 -15 to +5 -8 Tensile Strength Change, % -40 -6 Ultimate Elongation Change, % -20 -1 Volume Change, % 0 to +15 +11

(Z4) Low Temperature Resistance TR-10, temperature °F, C ASTM D1329 report -22 (-30)

"Purchaser use only. Reproduce only in full. Data pertains to items referenced only."

"The recording of false, fictitious, or fruaudulent statements or entries in this report may be punishable as a felony under federal law." Parker O-Ring Division

2360 Palumbo Drive Lexington, Ky 40509 2 of 2 (859) 269-2351 Enclosure 3 to E-60264

TN-LC Transportation Package Safety Analysis Report Revision 10B Changed Pages (Proprietary Version)

Withheld Pursuant to 10 CFR 2.390

Enclosure 4 to E-60264

TN-LC Transportation Package Safety Analysis Report Revision 10B Changed Pages (Public )

TN-LC Transportation Package Safety Analysis Report Revision 10, 07/21

Appendix 1.4.1 TN-LC Transport Package Drawings

Drawing Number Title 65200-71-01 Revision 10 TN-LC Cask Assembly (11 sheets) 65200-71-20 Revision 5 TN-LC Impact Limiter Assembly (2 sheets) 65200-71-21 Revision 2 TN-LC Transport Packaging Transport Configuration (1 sheet)

Drawing Number Title 65200-71-40 Revision 4 TN-LC-NRUX Basket Basket Assembly (5 sheets) 65200-71-50 Revision 4 TN-LC-NRUX Basket Basket Tube Assembly (5 sheets)

Drawing Number Title 65200-71-60 Revision 4 TN-LC-MTR Basket General Assembly (4 sheets) 65200-71-70 Revision 4 TN-LC-MTR Basket 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 7 TN-LC-1FA Basket (5 sheets)

TN-LC Transportation Cask 65200-71-91 Revision 0 TN-LC-1FA PWR Fuel Basket Damaged Fuel Assembly Can (FAC) (3 sheets) 65200-71-96 Revision 5 TN-LC-1FA BWR Sleeve and Hold-Down Ring (2 sheets) 65200-71-102 Revision 7 TN-LC-1FA 21 Pin Can Basket (4 sheets)

TN-LC -0100 1.4.1-1

TN-LC Transportation Package Safety Analysis Repo rt Revision 10, 07/21

Alternatively, in the absence of PRAs, burnup credit restrictions, as shown in Table 1.4.5-4a or Table 1.4.5-4b are required for 1FA PWR fuel transportation. The burnup/enrichment/cooling times are determined using PWR burnup credit approach and are required while transporting intact or damaged PWR fuel assembly in order to ensure that the maximum reactivity is subcritical and below the upper subcritical limit (USL). Note that burnup credit is not applicable to BW 15x15 fuel class.

The damaged fuel assembly can (FAC) is a stainless steel, square-section structure made of a sheet metal liner, a bottom closure welded to this liner, and a top closure lid, which simply sits on top of the can. There is a square structure welded to the bottom of the top lid, which slides inside the damaged fuel can liner. Because of this, the top closure lid can freely slide along the axis of the cask (within the cask cavity), but is captured by the body of the can in any direction perpendicular to the axis of the cask cavity (see Figure 1.4.5 -7), and the top lid cannot come off of the damaged fuel can once the cask is cl osed because the axial gap between the damaged fuel can and the cask cavity is smaller than the length of the square structure welded under the lid.

The damaged fuel can top closure lid, liner, and bottom closure all have multiple drain holes to allow for efficient draining of the water from the can during operations.

Finally, the top and bottom closures are sufficiently thick (in relation to the available axial gaps within the cask cavity) to ensure that they are always constrained by the basket in any direction perpendicular to the cask axis.

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 or damaged 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. (4,578.1 mm) or 169.55 in. (4,306.6 mm). 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.

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Table 1.4.5-1 PWR Fuel Specification for the Fuel to be Transported in the TN-LC-1FA Basket

PHYSICAL PARAMETERS:

Fuel Class (1)(2) Intact or damaged unconsolidated B&W 17x17, WE 17x17, 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 + damaged FAC (as 1850 lbs (839 kg) applicable) Weight Assemblies are PWR assemblies containing fuel rods with known or suspected cladding defects greater that hairline cracks or pinhole leaks. The extent of damage in the fuel assembly, including non-cladding Damaged Fuel damage, is to be limited so that a fuel assembly maintains its configuration for normal conditions.

Damaged fuel assemblies shall also contain top and bottom end fittings. Damaged fuel assemblies may also contain missing or partial fuel rods.

Fissile Material UO2 Maximum Initial Uranium Content(4) 490 kg/assembly Maximum Unirradiated Assembly Length 178.3 inches (4,528.8 mm)

THERMAL/RADIOLOGICAL PARAMETERS:

Fuel Assembly Average Burnup, Enrichment and Per Table 1.4.5-8 and Table 1.4.5-8a (applicable to Minimum Cooling Time TN-LC Unit 01)

Maximum Planar Average Initial Enrichment 5.0(3) wt.% U-235 Maximum Decay Heat(5) 3.0 kW per Assembly

• 16.7 mg/cm2 (Natural or Enriched Boron Aluminum Alloy / Metal Matrix Composite Minimum B-10 content in poison plates loading (MMC))

• 20.0 mg/cm2 (Boral)

Minimum number of absorber rods per PRA as a Per Table 1.4.5-4 (Note that the use of PRAs is function of assembly class optional)

Burnup Credit Restriction in the absence of PRAs Per Table 1.4.5-4a or Table 1.4.5-4b

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Table 1.4.5-4b Maximum Planar Average Initial Enrichment/Minimum Burnup Combination - 1FA PWR - With Control Rod Insertion (1)

WE 17x17, WE 16x16, WE 15x15, CE 14x14, CE 15x15 and CE 16x16 Fuel Assembly Classes Fresh Fuel 2.90 wt. % U-235 Cooling Time 5 Years 10 Years 15 Years 20 Years Burnup Fuel Initial Enrichment (wt. % U-235)

(GWd/MTU) 5 2.97 2.99 3.00 3.01 10 3.29 3.31 3.34 3.36 15 3.54 3.60 3.64 3.66 20 4.21 4.38 4.45 4.53 25 4.75 4.91 4.98 5.00 30 5.00 5.00 5.00

WE 14x14 Fuel Assembly Class Fresh Fuel 2.95 wt. % U-235 Cooling Time 5 Years 10 Years 15 Years 20 Years Burnup Fuel Initial Enrichment (wt. % U -235)

(GWd/MTU) 5 3.20 3.20 3.21 3.23 10 3.57 3.57 3.59 3.59 15 3.81 3.86 3.90 3.90 20 4.48 4.62 4.71 4.78 25 5.00 5.00 5.00 5.00 (1) Fuel assemblies with accumulated control rod insertion through the first 15 GW d/MTU. Fuel assemblies with accumulated control rod insertion greater than the first 15 GWd/MTU are not authorized.

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The criticality analysis is performed using the bounding Westinghouse (WE) 17x17 and WE 14x14 fuel classes identified in Table 6.10.4-2. The results of the WE 17x17 fuel class bound those of the WE 16x16, WE 15x15, Combustion Engineering (CE) 14x14, CE 16x16 and CE 15x15 fuel classes. The Babcock &

Wilcox (BW) 15x15 fuel class is not included in the PWR burnup credit criticality analysis.

Intact or damaged fuel assembly is authorized for transportation in the 1FA PWR basket. Criticality calculations are performed to determine the minimum assembly average burnup as a function of initial enrichment and cooling time for the two bounding fuel assembly ( FA) classes, which are listed in Table 1.4.5-4a or Table 1.4.5-4b. The calculations determine keff with the CSAS5 control module of SCALE 6.1.3 [3] for each assembly class and initial enrichment, including all uncertainties to assure criticality safety under all credible conditions.

The results of the evaluation demonstrate that the maximum keff, including statistical uncertainty, and appropriate biases associated with the burnup credit methodology is less than the USL determined from a statistical analysis of benchmark criticality experiments. The statistical analysis procedure includes a confidence band with an administrative safety margin of 0.05.

6.10.5.1.3 Criticality Safety Index

For the PWR fuel assembly payload, no HAC array models are developed (2N = 1). Therefore, per 10 CFR 71.59, N=0.5, and the CSI is 50/N = 100 for this payload. In the NCT array cases for the PWR fuel assembly payload, 5N=2.5 and three packages are modeled.

6.10.5.2 Package Fuel Loading

The 1FA PWR basket is capable of transporting one PWR fuel assembly. A detailed listing of the contents of the 1FA PWR basket is provided in Table 6.10.4-2.

For all the FA classes, control components ( CCs) are also included as authorized contents. The only change to the package fuel loading to evaluate the addition of these C Cs would be to replace the water in the guide tubes with 11B4C. Since these CCs displace moderator in the assembly guide and or instrument tubes, an evaluation is not needed to determine the potential impact of storage of CCs that extend into the active fuel region on the system reactivity. The presence of these CCs such as control rod assemblies (CRAs), control element assemblies (CEAs) and burnable poison rod assemblies ( BPRAs) will result in a reduction in the reactivity of the FAs. CCs that do not exten d into the active fuel region of the assembly do not have any effect on the reactivity of the system as evaluated. Therefore, CCs are not included in any of the criticality models.

The criticality analysis is performed using two fuel assembly types, WE 14x14 STD, and WE 17x17 Robust Fuel Assembly (RFA)/LOPAR representing WE 14x14 and WE 17x17 fuel classes, respectively.

6.10.5.3 Model Specification

The evaluations are performed using SCALE 6.1.3 [ 3] and ENDF/B-VII nuclear data. The SCALE 6.1.3 capabilities used include automated sequences to produce problem-dependent multi -group cross -section data and analysis sequences for Monte Carlo neutron transport (CSAS5) and burnup-credit criticality safety (STARBUCS). The 238-group cross -section library based on the ENDF/B -VII nuclear data and the resonance cross-section methodology employing CENTRM are used.

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6.10.5.4.3 Criticality Results

The study to determine the most reactive configuration considered four di fferent scenarios: double shear, single shear, missing rods and pitch expansion. The results of the study are presented in Table 6.10.5-6 and Table 6.10.5-7 for WE 14x14and WE 17x17 fuel assemblies, respectively. The results show that the configuration with fully expanded pitch is the most reactive configuration in both cases. Therefore, the fully expanded pitch is used for developing the loading curves. (It should be noted that the k eff values in these cases are higher. These models are used only for comparative study and hence the value of keff need not be compared to any USL values).

Loading curves are generated for the WE 14x14 and WE 17x17 fuel assembly classes loaded in TN -LC cask. The BECT combinations are generated such that the keff is below the kBUC values. This includes maximum allowable enrichment values corresponding to burnup values ranging from 0 to 30 GWd/MTU, cooling times 5, 10, 15 and 20 years.

The BECT for WE 14x14 fuel class are determined for single/array of package under HAC and are tabulated in Table 6.10.5-8 and Table 6.10.5-8A using respectively ORIGEN -ARP cross-section libraries assuming BPR and CR insertions, as described in Section 6.10.5.6. The BECT for WE 17x17 fuel class are determined for single/array of package under HAC and are tabulated in Table 6.10.5-9 and Table 6.10.5 -9A using ORIGEN -ARP cross-section libraries assuming BPR and CR insertions, respectively, as described in Section 6.10.5.6. The results of the WE 17x17 fuel class bound those of the WE 15x15, CE 14x14, CE 16x16 and CE 15x15 fuel classes as WE 17x17 fuel class bounds those fuel classes per Table 6.10.4-8. The BW 15x15 fuel class is not included in the PWR burnup credit criticality analysis.

The criterion for sub-criticality is that:

keff + (i + ki) + kx < USL

where USL is the upper subcriticality limit established by an analysis of benchmark criticality experiments,

kx is the code bias due to minor actinides and fission products, 0.1 x 0.015 = 0.0015,

And ( i + ki) is the burnup -dependent bias and bias uncertainty for isotopics validation shown in Table 6-3 of [ 2].

kKENO + 2 KENO + (i + ki) + 0.0015 < 0.9424

6.10.5.5 Critical Benchmark Experiments and applicable biases

This section summarizes evaluations performed for benchmarking the various computer codes utilized in the criticality analysis. A description of the benchmarking analyses performed in support of the criticality analyses for the TN-LC transport cask where f resh fuel is considered is provided in Section 6.10.5.5.1.

The description of the benchmarking analyses performed in support of burnup credit is provided in Section 6.10.5.5.2. For both fresh fuel and burnup credit evaluations, the USL is also determined.

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Additionally, a sensitivity and uncertainty (S/U) tool is used to generate a parameter that quantifies the similarity of the two systems. From [3]: The technique compares the detailed sensitivity data for the two systems, giving greater weight to comparisons of sensitivities for nuclides and reactions with the highest nuclear data uncertainties. The correlation coeffici ent, ck obtained from this process indicates the degree of similarity by the following standards: c k greater than or equal to 0.9 indicates similar systems, ck between 0.8 and 0.9 indicates marginally similar and less than 0.8 is not recommended for use. T he system application is the TN-LC loaded with 1FA PWR basket, while the experiment is the GBC -32.

The TSUNAMI-3D module of SCALE 6.1.3, [ 3], has been used for performing the similarity analysis.

TSUNAMI-3D calculations are performed for the application and experiment models. Isotopic number densities obtained from the STARBUCS outputs are used for the TSUNAMI-3D models as well. Direct perturbation calculations are used to confirm the adequacy of the sensitivity data files. Direct perturbation calculations involve varying the composition information around the nominal value and using the resulting keff value variations to calculate the total sensitivity. The direct perturbation results are compared with the TSUNAMI sensitivity results to confirm the adequacy of the sensitivity data. Finally, TSUNAMI-3D generates sensitivity data files (.sdf), which contains the energy -dependent sensitivity coefficients for each value of burnup.

The.sdf files generated by TSUNAMI-3D for the two systems in the previous step are used by TSUNAMI-IP to determine a c k value at each burnup. The TSUNAMI-IP module of SCALE 6.1.3, [ 3],

was used to calculate detailed keff uncertainty information for the application model. The correlation factor, Ck quantifies correlations in uncertainties by propagating the tabulated cross -sect ion-uncertainty information to the calculated keff value of a given system via the energy-dependent sensitivity coefficients.

This evaluation demonstrates similarity by comparing the global parameters, as well as by determining the sensitivity and uncertainty. The c k parameters generated from the sensitivity and uncertainty calculation, which indicate high degree of similarity, are provided in Table 6.10.5-15.

The results satisfy the requirements of Reference [2 ] thereby allowing the user to adopt the results from Table 3 of [2], in preparing system-specific loading requirements with burnup credit.

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6.10.5.9 Evaluations under NCT and HAC

This section describes the evaluations under HAC and NCT performed for the TN-LC transport package with the 1FA PWR basket.

6.10.5.9.1 Package Arrays under HAC s

As the CSI is 100, the criticality analysis performed to demonstr ate the compliance with the sub-criticality requirements of 10 CFR 71.55 (e), Section 6.10.5.9.3, also meets the sub-criticality requirements of 10 CFR 71.59 (a) (2).

6.10.5.9.2 Package Arrays under Normal Conditions of Transport

As the CSI is 100, the criticality analysis performed for an array of three undamaged packages described in Section 6.10.5.4.2.4 demonstrates the compliance with the sub-criticality requirements of 10 CFR 71.59 (a) (1). The analysis is conducted for WE 14x14 and WE 17x7 fuel classes. Note that BW 15x15 fuel class is added the analysis for conservatism. The results of the analysis are shown in Table 6.10.5-10.

The highest keff is 0.2303.

6.10.5.9.3 Single Package Evaluation

The criticality analysis performed for a single package damaged PWR fuel under HAC in a fully flooded 1FA PWR basket is employed to demonstrate compliance with the sub -criticality requirements of 10 CFR 71.55 (e). The damaged PWR fuel under HAC model simulates a fuel reconfiguration scenario where fuel cladding, guide and instrument tubes, end fittings, and spacer grids are removed with the fuel rods in the most reactive configuration. The loading curves generated in Section 6.10.5.4.2.4 for the WE 14x14 and WE 17x17 fuel classes determined the maximum allowable enrichment values associated with burnup values ranging from 0 to 30 GWD/MTU, cooling times 5, 10, 15 and 20 years allowed for transportation.

The analysis demonstrates the compliance with the sub-criticality requirements of 10 CFR 71.55 (e), 10 CFR 71.55 (d) and 10 CFR 71.55 (b) for the 1FA PWR basket in the TN -LC transport cask. The highest keff plus biases is 0.9418.

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6.10.5.10 References

1. 10 CFR 71, Packaging and Transportation of Radioactive Materials.

2. U.S. NRC, NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, August 2020.

3. Oak Ridge National Laboratory, RSIC Computer Code Collection, SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, Version 6.1, June 2011.

4. International Handbook of Evaluated Criticality Safety Benchm ark Experiments (IHECSBE), NEA-1486/15, NEA Nuclear Science Committee, September 2016.

5. U.S. NRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, January 2001.

6. U.S. NRC, NUREG/CR-7203, A quantitative Impact Assessment of Hypothetical Spent Fuel Reconfiguration in Spent Fuel Storage Casks and Transportation Packages, September 2015.

7. U.S. NRC, NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis, March 2003.

8. U.S. NRC, NUREG/CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analysis - Criticality (keff) Predictions, J. M.

Scaglione, D.E. Mueller, J.C. Wagner, and W.J. Marshall, April 2012.

9. U.S. NRC, NUREG/CR-6665, Review and P rioritization of Technical Issues Related to Burnup Credit for LWR Fuel, February 2000.

10. YAEC 1937, Axial Burnup Profile Database for Pressurized Water Reactors, R.J.

Cacciaputi, Van S. Volkinburg, May 1997.

11. Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors, ANSI/ANS-8.1 -1998; R2007, an American National Standard, published by the American Nuclear Society, LaGrange Park, IL, 1998.

12. U.S. NRC, Interim Staff Guidance-10, Revision 0, Justification for Minimum Margin of Subcriticality for Safety, ISG-10, Revision 0, Division of Fuel Cycle Safety and Safeguards.

13. U.S. NRC, NUREG/CR-6979, Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data, September 2008.

14. US NRC, NUREG/CR-7108, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions, April 2012.

15. US NRC, NUREG/CR-6951, Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit, January 2008.

16. DOE Purchase Order DE-AF28-04RW12278, Pool Inventories in Compact Disk Attachment, May 21, 2004.

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17. U.S. NRC, Interim Staff Guidance-8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

18. U.S. NRC, NUREG/CR-6800, Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs, March 2003.

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Table 6.10.5-8A Loading Curve for Damaged WE14 Fuel - Single/Array Package HAC - With Control Rod Insertion (1)

Damage Fuel under HAC

Cooling Time Burnup Enrichment kkeno keff keff+kx+ki (GWd/MTU) (wt.% U-235)

0 years (Fresh Fuel) 0 2.95 0.9354 0.0008 0.9370 0.9370 5 3.20 0.9241 0.0005 0.9251 0.9416 10 3.57 0.9246 0.0005 0.9256 0.9419 5 years 15 3.81 0.9233 0.0005 0.9243 0.9415 20 4.48 0.9238 0.0004 0.9246 0.9415 25 5.00 0.9205 0.0005 0.9215 0.9384 5 3.20 0.9234 0.0005 0.9244 0.9409 10 3.57 0.9243 0.0005 0.9253 0.9416 10 years 15 3.86 0.9234 0.0005 0.9244 0.9416 20 4.62 0.9236 0.0005 0.9246 0.9415 25 5.00 0.9147 0.0005 0.9157 0.9326 5 3.21 0.9235 0.0005 0.9245 0.9410 10 3.59 0.9245 0.0005 0.9255 0.9418 15 years 15 3.90 0.9237 0.0005 0.9247 0.9419 20 4.71 0.9235 0.0005 0.9245 0.9414 25 5.00 0.9121 0.0005 0.9131 0.9300 5 3.23 0.9237 0.0005 0.9247 0.9412 10 3.59 0.9242 0.0005 0.9252 0.9415 20 years 15 3.90 0.9227 0.0005 0.9237 0.9409 20 4.78 0.9233 0.0005 0.9243 0.9412 25 5.00 0.9090 0.0005 0.9100 0.9269

(1) Fuel assemblies with accumulated control rod insertion through the first 15 GWd/MTU. Fuel assemblies with accumulated control rod insertion greater than the first 15 GWd/MTU are not authorized.

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Table 6.10.5-9A Loading Curve for Damaged WE17 Fuel - Single/Array Package HAC - With Control Rod Insertion (1)

Damage Fuel under HAC

Cooling Time Burnup Enrichment kkeno keff keff+kx+ki (GWd/MTU) (wt.% U-235)

0 years (Fresh Fuel) 0 2.90 0.9380 0.0008 0.9396 5 2.97 0.9240 0.0005 0.9250 0.9415 10 3.29 0.9248 0.0005 0.9258 0.9421 5 years 15 3.54 0.9229 0.0005 0.9239 0.9411 20 4.21 0.9233 0.0005 0.9243 0.9412 25 4.75 0.9237 0.0005 0.9247 0.9416 30 5.00 0.9077 0.0005 0.9087 0.9263 5 2.99 0.9234 0.0005 0.9244 0.9409 10 3.31 0.9242 0.0005 0.9252 0.9415 10 years 15 3.60 0.9235 0.0005 0.9245 0.9417 20 4.38 0.9234 0.0005 0.9244 0.9413 25 4.91 0.9237 0.0005 0.9247 0.9416 30 5.00 0.8987 0.0005 0.8997 0.9173 5 3.00 0.9241 0.0005 0.9251 0.9416 10 3.34 0.9246 0.0005 0.9256 0.9419 15 years 15 3.64 0.9230 0.0005 0.9240 0.9412 20 4.45 0.9233 0.0005 0.9243 0.9412 25 4.98 0.9234 0.0005 0.9244 0.9413 30 5.00 0.8928 0.0005 0.8938 0.9114 5 3.01 0.9244 0.0005 0.9254 0.9419 10 3.36 0.9248 0.0005 0.9258 0.9421 20 years 15 3.66 0.9233 0.0005 0.9243 0.9415 20 4.53 0.9242 0.0005 0.9252 0.9421 25 5.00 0.9213 0.0005 0.9223 0.9392

(1) Fuel assemblies with accumulated control rod insertion through the first 15 GWd/MTU. Fuel assemblies with accumulated control rod insertion greater than the first 15 GWd/MTU are not authorized.

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Table 6.10.5-27 Operational Conditions Employed in T-DEPL Calculation (1)

Parameters Value Average Fuel Temperature (K) 1100 Average Moderator Density (g/cc) 0.63 Average Moderator Temperature (K) 610 Average Clad Temperature (K) 640 Soluble Boron Concentration (ppm) 1000 Specific Power (MW/MTU) 40 Down Time (Days) 0

(1) Bounding upper values for fuel temperature, moderator temperature, soluble boron concentration and specific power for depletion parameters

Table 6.10.5-28 Misload Curve Developed for 5 Years Cooling Time - WE 14x14 Fuel Class Cooling Time Burnup Enrichment kkeno keff (GWd/MTU) (wt.% U-235) 0 years (Fresh Fuel) 0 3.35 0.9668 0.0008 0.9685 5 3.67 0.9539 0.0005 0.9549 5 Years 10 4.12 0.9536 0.0005 0.9546 15 4.40 0.9531 0.0005 0.9541 20 5.00 0.9448 0.0005 0.9458

Table 6.10.5-29 Misload Curve Developed for 5 Years Cooling Time - WE 17x17 Fuel Class Cooling Time Burnup Enrichment kkeno keff (GWd/MTU) (wt.% U-235) 0 years (Fresh Fuel) 0 3.25 0.9676 0.0008 0.9691 5 3.42 0.9543 0.0005 0.9553 10 3.80 0.9544 0.0005 0.9554 5 Years 15 4.11 0.9529 0.0005 0.9539 20 4.93 0.9539 0.0005 0.9549 25 5.00 0.9342 0.0005 0.9352

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Table 6.10.5-30 Misload Curve Vs DOE Inventory DOE Inventory = 103475 PWR Fuel Assemblies Case Qualified # Assemblies Percentage Qualified WE 14x14 103417 99.96 WE 17x17 103395 99.92

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Chapter 7 Package Operations

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 Section 7.6.

This Chapter contains TN-LC cask loading and unloading procedures that are intended to show the general approach to cask operational activities. The procedures in this chapter are intended to show the types of operations that will be performed and are not intended to be limiting. Site-specific conditions and requirements may require the use of different equipment and ordering of steps to accomplish the same objectives or to meet acceptance criteria to ensure the integrity of the package.

A separate operations manual (OM) will be prepared for the TN -LC cask to describe the operational steps in greater detail. The OM, along with the information in this chapter, will be used to prepare the site-specific procedures that will address the particular operational considerations related to the cask.

7.1 TN-LC Package Loading The use of the TN-LC cask to transport fuel offsite involves (1) preparation of the empty cask for use; (2) verification that the fuel assemblies or fuel rods to be loaded in the TN-LC cask with the appropriate fuel-specific basket meet the criteria set forth in this document; (3) installation of a basket into the cask; and (4) loading fuel or placing loaded fuel buckets or pin cans in an empty TN-LC cask with the appropriate fuel -specific basket.

Offsite transport involves (1) preparation of the loaded cask fo r transport; (2) assembly verification leakage-rate testing of the package containment boundary; (3) placement of the cask onto a transportation vehicle; (4) installation of the impact limiters and (5) closure of the transportation container.

During shipment, the package contains any one of the TN-LC basket designs with its authorized contents as described in Chapter 1, Appendices 1.4.2 through 1.4.5. TN -LC Unit 01 shall only be loaded with the TN-LC-1FA basket with one PWR fuel assembly (Table 1.4.5-8a or Table 1.4.5-8b) (including a damaged fuel assembly can if transporting a damaged PWR fuel assembly) or one pin can with up to 21 PWR fuel rods (Table 1.4.5-10a). Procedures are provided in this section for (1) transport of the cask directly from a site s pent fuel pool and (2) transport of the cask directly from a site hot cell. Appendix 7.7 contains a sub-appendix for each basket design detailing its loading procedures.

7.1.1 TN-LC Cask Preparation for Loading Procedures for preparing the cask for use after receipt at the loading site are provided in this section and are applicable for shipment of casks loaded with any one of the basket designs and its respective approved contents.

1. Upon arrival of the empty 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 empty TN-LC package.

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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 to rque 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.

NOTE: TN -LC Unit 01 shall only be loaded with TN-LC-1FA basket.

12. The candidate 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. For the transportation of fuel within the TN-LC-1FA where burnup credit is employed for criticality safety, additional administrative controls to prevent misloading are also outlined in Appendix 7.7.4.

NOTE: TN -LC Unit 01 shall only be loaded with TN-LC-1FA basket (equipped with the damaged fuel assembly can if loading a damaged PWR fuel assembly) with one PWR fuel assembly or one fuel rod pin can with up to 21 PWR or BWR intact or damaged fuel rods.

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 of Section 7.1.2. 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 10, 07/21

Table 7-1 Applicable Fuel Specification for Various Fuel Types

Basket Design Applicable Fuel Specification from 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 TN-LC - 1FA in Unit 01 Table 1.4.5 - 1, 1.4.5 - 2, 1.4.5 - 4, or 1.4.5-4a or 1.4.5-4b and either 1.4.5-8a (PWR fuel assembly) or 1.4.5-10a (up to 21 rods in pin can)

Table 7-2 Appendices Containing Loading Procedures for Various TN-LC Baskets

Bottom Basket Type Subb asket Type Appendix Spacer Required?

7ctions Yes TN-LC -NRUX 7 -

7ctions Yes 7 -2

- -M 7ctions Yes 7 -2

- -TRI 7ctions Yes 7 -2 1-PWR 7ctions Yes 7 -2

- -1 1-R 7ctions Yes 7 -2 Pinan 7, Section 7 -2

Tl3 Appendices Containg UreduresariousN-LCasks

Basket Type Subbasket Type Appendix TN-LC -NRUX 7ction s 3 -

7ction s 3 -4

- -M 7ction s 3 -4

- -TRI 7ction s 3 -4 1-PWR 7ction s 3 -4

- -1 1-R 7ction s 3 -4 can 7ction s 3 -4

TN-LC -0100 7-18 TN-LC Transportation Package Safety Analysis Report Revision 10, 07/21

• Additional Administrative Controls for Burnup Credit When burnup credit is employed for demonstration of criticality safety, additional administrative controls are required for verification of fuel assembly burnup and to prevent misloading. Fuel loading plans developed above shall also include these additional requirements:

A requirement to compare the reactor operating parameters for the irradiation period of the fuel assembly against those shown in Table 6.10.5-27 to ensure compliance with the isotopic depletion analysis A requirement for no fresh fuel in pool at time of loading, or verification that fuel being loaded into the cask is not fresh, either visually or by qualitative measurement A pool audit prior to loading, including visual verification of assembly identification numbers Identification of highly underburned and high reactivity (Table 6.10.5-28 and Table 6.10.5-29) fuel assemblies in the pool both prior to and after loading. Alternatively, the licensee can perform a misload evaluation using the methodology and criteria described in Section 6.10.5.8 to identify these highly underburned and high reactivity fuel assemblies. This evaluation will be subject to NRC review and approval A requirement that assemblies without visible identification number must have a quantitative confirmatory measurement prior to loading.

As described by the procedures above, multiple barriers are included to preclude misloading events.

• 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.

• If loading damaged fuel assemblies, verify that the fuel assembly can (FAC) is installed in the TN-LC-1FA Basket.

TN-LC -0100 7.7.4-1a Enclosure 5 to E-60264

Listing of Computer Files Contained in Enclosure 6

Disk ID Discipline System/Component File Series Number No. (size) (topics) of Files

Folder: \\TRITON-CR-WE14 Folder for TRITON inputs for WE14x14 fuel class with CR insertion during depletion (input files) - See description in Section 6.10.5.6 - 17 Note that these TRITON inputs with CR insertion were not provided in the initial submission.

TN-LC_TRITON_CR (TRITON)

Folder: \\TRITON-CR-WE17 Folder for TRITON inputs for WE17x17 fuel class with CR insertion during depletion (input Enclosure files) - See description in Section 6.10.5.6 - 17 6 Note that these TRITON inputs with CR insertion were not provided in the initial One Criticality submission.

Computer Hard Drive Total Folder: \\HAC-STARBUCS-CR-WE14 (94.5 MB) Samples of SCALE6.1 STARBUCS input/output files for HAC single package 12 using ARP library with CR insertion (input and output files) - New Table 6.10.5-8A

TN-LC_STARBUCS_CR

Folder: \\HAC-STARBUCS-CR-WE17 Samples of SCALE6.1 STARBUCS input/output files for HAC single package 12 using ARP library with CR insertion (input and output files) - New Table 6.10.5-9A

1 of 1 Enclosure 7 to E-60264

Proposed Certificate of Compliance No. 9358, Revision 6 Markup

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Table 11a Maximum Planar Average Initial Enrichment/Minimum Burnup Combination - PWR Fuel Assembly Classes

WE 17x17, WE 16x16, WE 15x15, CE 14x14, CE 15x15 and CE 16x16 Fuel Assembly Classes Fresh Fuel 2.90 wt. % U-235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U -235)

(GWd/MTU) 5 3.04 3.05 3.06 3.08 10 3.37 3.40 3.42 3.44 15 3.66 3.70 3.74 3.76 20 4.43 4.53 4.61 4.65 25 4.87 5.00 5.00 5.00

WE 14x14 Fuel Assembly Class Fresh Fuel 2.95 wt. % U -235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U-235)

(GWd/MTU) 5 3.26 3.26 3.27 3.28 10 3.65 3.65 3.66 3.68 15 3.92 3.96 4.00 4.03 20 4.67 4.80 4.86 4.93 25 5.00 5.00 5.00 5.00 Table 11b Maximum Planar Average Initial Enrichment/Minimum Burnup Combination - PWR Fuel Assembly Classes - With Control Rod Insertion (1)

WE 17x17, WE 16x16, WE 15x15, CE 14x14, CE 15x15 and CE 16x16 Fuel Assembly Classes Fresh Fuel 2.90 wt. % U-235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U -235)

(GWd/MTU) 5 2.97 2.99 3.00 3.01 10 3.29 3.31 3.34 3.36 15 3.54 3.60 3.64 3.66 20 4.21 4.38 4.45 4.53 25 4.75 4.91 4.98 5.00 30 5.00 5.00 5.00

WE 14x14 Fuel Assembly Class Fresh Fuel 2.95 wt. % U -235 Cooling 5 Years 10 Years 15 Years 20 Years Time Burnup Fuel Initial Enrichment (wt. % U-235)

(GWd/MTU) 5 3.20 3.20 3.21 3.23 10 3.57 3.57 3.59 3.59 15 3.81 3.86 3.90 3.90 20 4.48 4.62 4.71 4.78 25 5.00 5.00 5.00 5.00 (1) Fuel assemblies with accumulated control rod insertion through the first 15 GWd/MTU. Fuel assemblies with accumulated control rod insertion greater than the first 15 GWd/MTU are not authorized.

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Enclosure 8 to E-60264 AFFIDAVIT PURSUANT TO 10 CFR 2.390

TN Americas LLC )

State of Maryland SS.

County of 1-!o~ard )

I, Prakash Narayanan, depose and say that I am Chief Technical Officer of TN Americas LLC, duly authorized to execute this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below. I am submitting this affidavit in confonnance with the provisions of 10 CFR 2.390 of the Commission ' s regulations for withholding this information.

The information for which proprietary treatment is sought is listed below:

• Enclosure 3 - Portions of certain chapters and appendices of the Safety Analysis Report (SAR) for Certificate of Compliance No. 9358 TN-LC, Revision 10B, Docket 71-9358 (Proprietary Version)

• Enclosure 6 - Certain computer files associated with CoC 9358 (Proprietary)

This document has been appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by TN Americas LLC in designating information as a trade secret, privileged, or as confidential commercial or financial infonnation.

Pursuant to the provisions of paragraph (b) ( 4) of Section 2.390 of the Commission ' s regulations, the following is furnished for consideration by the Commission in detennining whether the infonnation sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

1) The infonnation sought to be withheld from public disclosure involves certain design details associated with the SAR analyses and SAR drawings for the TN-LC System, which are owned and have been held in confidence by TN Americas LLC.

2) The information is of a type customarily held in confidence by TN Americas LLC and not customarily disclosed to the public. TN Americas LLC has a rational basis for determining the types of information customarily held in confidence by it.

3) Public disclosure of the infonnation is likely to cause substantial 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. The availability of such infonnation to competitors would enable them 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 product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.

Further the deponent sayeth not.

Prakash Narayanan Chief Technical Officer, TN Americas LLC Subscribed and sworn before me this ~,J day of February, 2022.

( HVNESYA TAYLOR i G~:-= Notary Public Howard County Notary Public Maryland My Commission Expires _fu__; _2_/ -2.S_,~ -,,,.,mission Expires Oct. 5, 2025

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