ML21214A054

Rev 21 a to Safety Analysis Report (Non-Proprietary Version)

ML21214A054
Person / Time
Site:	07109356
Issue date:	07/31/2021
From:	NAC International
To:	Office of Nuclear Material Safety and Safeguards
Shared Package
ML21214A057	List: ML21214A054 ML21214A055
References
ED20210107
Download: ML21214A054 (76)

Text

July 2021 Revision 21 A MAGNATRA (Modular Advanced generation N

Nuclear All-purpose TRANsport)

SAFETY ANALYSIS REPORT NON-PROPRIETARY VERSION A NAC ii INTERNATIONAL Atlanta Corporate Headquarters: 3930 East Jones Bridge Road, Norcross , Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com

Enclosure 1 to ED20210107 Page 1 of 1 Enclosure 1 RAI Responses

• No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion RAI Response Submittal MAGNATRAN SAR, Revision 21A

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000 Enclosure 1 NACINTERNATION AL RESPONSE TO THE UNITED STATES NUCLEAR REGULATORY COMMISSION REQUEST FOR ADDITIONAL INFORMATION

• JULY2021 FOR REVIEW OF THE CERTIFICATE OF COMPLIANCE NO. 9356, REVISION NO. 3 (EPID No. L-2021-LLA-0000, DOCKET NO. 71-9356)

JULY2021

• Page 1 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• TABLE OF CONTENTS STRUCTURAL EVALUATION ........................................................................................................................... 3 THERMAL EVALUATION ................................................................................................................................ 5 SHIELDING EVALUATION ............................................... :............................................................................... 8 PACKAGE OPERATIONS ............................................................................................................................... 10 THERMAL EVALUATION .............................................................................................................................. 12 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ................................................................................. 13

• Page 2 of 13

Enclosure 1 to ED202 l O107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000 NAC INTERNATIONAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION STRUCTURAL EVALUATION 2-1 Provide responses to the following questions related to Section 2.11 .4, "Side Drop Evaluation,"

of the SAR, Revision 20C:

(a) Describe how the maximum stresses were calculated in the table shown on page 2.11.4-2 of the SAR, Revision 20C, and (b) Justify the use of a dynamic load factor (DLF) of 1.75.

In Section 2.11 .4, "Side Drop Evaluation," of the SAR, Revision 0 (Reference 2.1 ), the applicant provided the table for the maximum stress in the fuel rods based on the design basis acceleration of 60g for cask side drop, as shown below:

Calculated Maximum Stress in the Fuel Rods (SAR, Revision 0) ---I Fuel Rod Maximum Stress (ksi) Factor of Safety CE14x14 37.1 1.88  ;

i WE15x15 48.1 1.45 I

WE17x17 46.3 1.50 I 1 Note: Allowable Stress= Yield Strength= 69.6 ksi at 752°F I In the SAR, Revision 20C, the applicant recalculated and reduced the maximum stress using an acceleration of 45.5g for cask side drop and a factor of 1.25, which is based on a guidance provided in Section 2.3.3 ofNUREG-2224 (Reference 2.2), as shown below:

Calculated Maximum Stress in the Fuel Rods Table in SAR, Revision 20C)

Fuel Assembly Maximum Stress Maximum Stress (ksi)

Factor of Safety T e ksi with DLF=1.75 CE14x14 29.2 51.1 1.36 WE15x15 35.5 62.1 1.12 WE17X17 34.5 60.4 1.15 Note: Allowable Stress = Yield Stren th = 69.6 ksi 752°F The applicant stated that the maximum stresses were recalculated by applying a maximum acceleration of 45 .5 g for cask side drop and a factor of 1.25 to the rod moment of inertia based on a suggestion in Section 2.3.3 ofNUREG-2224. However, in Section 2.11.1, "PWR [pressurized-water reactor] Fuel Rod Evaluation," it states that there were two finite element (FE) models (ANSYS and LS-DYNA) for the PWR fuel rod evaluations, where the ANSYS FE model used elastic material properties and the LS-DYNA model used bilinear material properties. The staff is not clear how the acceleration of 45 .5 g and the factor of 1.25 were applied to the calculations for rigidity (E and I) and maximum stresses. The SAR, Revision 20C does not provide detailed information that explains the reduction in maximum stress values from Revision 0. Describe how

• the maximum stresses were calculated to arrive at the values in Revision 20C .

Page 3 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• In addition, the applicant used a DLF of 1. 75 to calculate the factor of safety (FS). However, Section 2.3.5.2 ofNUREG-22 24 (Reference 2.2) provides guidance to use a DLF of 2.0 to account for uncertainties involved in natural frequency, load duration, and load time history shape, which depend on the physical characteristics of the fuel assembly, the rod, and the cask.

The staff recalculated a FS using the DLF of 2.0 and found that the FS for the fuel rod assembly (WE 15xl5) is less than 1.0, indicating that there would be a concern with the safety of the fuel rods.

Provide the technical justification for using a DLF of 1.75 and not using the DLF of2.0 recommended in NUREG-2224.

This information is needed to determine compliance with 10 CFR 71.73(c)(l).

References:

2.1 NAC International, MAGNA TRAN Safety Analysis Report, Revision 0, USNRC Docket Number 71-9356, April 2019.

2.2 NUREG-2224, Dry Storage and Transportation of High Bumup Spent Nuclear Fuel (Final Report), Office of Nuclear Material Safety and Safeguards, November 2020.

NAC International Response to Thermal Evaluation RAI 2-1:

The model description in SAR Section 2.11.1 was for the end drop condition. The

• ANSYS model for the side drop is described in SAR Section 2.11.4. The 45.5g is applied in the same manner as the 60g lateral loading as described on SAR Page 2.11.4-1. As described on SAR Page 2.11.4-2, the 1.25 is applied to the moment of inertia, which would have been used to factor the moment of inertia property of the beam element forming the finite element model referenced on SAR Page 2.11.4-1.

The stress reported at the top of SAR Page 2.11.4-2 would have been reduced to the value shown in the lower table on SAR Page 11.4.4-2, since the accelerations were reduced from 60g (used in the table at the top of the page) to 45.5g, and the moment of inertia would have been increased by 1.25. The stresses reported in the table are a direct output from the ANSYS static solution.

The DLF of2.0 reported in NUREG-2224 corresponds to a suddenly applied load. In the side drop, the peak acceleration of 45.5g cannot physically be developed in a suddenly applied manner. The acceleration data in SAR Figure 2.6.7-13 shows that the acceleration increases from Oto its peak value over a finite amount of time, not instantly. The use of the DLF associated with sine pulse is more appropriate and the maximum DLF value that can be achieved with a sine pulse is 1.75.

Page 4 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• THERMAL EVALUATION NAC INTERNATIONAL RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION 4-1 Clarify the penultimate sentence in the last paragraph of Section 4, "Containment," on page 4-1 of the application to describe that leak testing the entire containment boundary, rather than the containment boundary seals, specifically to the American National Standards Institute (ANSI)

N14.5-1997 1 leaktight criterion assures that the containment does not leak.

Section 4. "Containment," of the application describes that the leakage testing of the containment boundary seals assures that the containment doesn't leak. That statement is inconsistent with the statement in Section 6.1.1, "Design Features," of the application that states, "Based on a no credible leakage TSC [transportable storage canister] boundary and a leaktight transport cask boundary, moderator is not present in the TSC while it is being transported.," which necessitates a leaktight transportation package boundary to provide reasonable assurance that moderator is excluded in the TSC. That statement in Section 4 of the application on leakage testing the containment boundary seals is also inconstant with the another statement that is in Section 4, "Containment," of the safety analysis report (SAR) that states, "The containment boundary is tested to ANSI Nl 4.5-1997 ... ," (e.g. bottom inner forging, inner shell, top forging, cask lid, lid metal inner O-ring, cover plate, and cover plate metal inner O-ring). Also, ANSI N14.5-1997 describes that the entire containment boundary should be leak tested during a fabrication leakage

• rate test, and the containment boundary seals should also be tested during the periodic, and pre-shipment leakage rate tests. In addition, Section 6.4.1, "Configuration/Discussion," of the application describes that the MAGNA TRAN system is designed with two independent boundaries, one of which is the entire transport cask, not only the containment boundary seals.

The penultimate sentence in the last paragraph of Section 4 of the application also appears starting on page 1.3-8 of the application, and as the first sentence of the second paragraph in Section 5(a)(2) of the Certificate of Compliance (CoC), and therefore should also be clarified in each of the locations.

This information is necessary to determine compliance with 10 CFR 71.51(a)(l) and (2), and 71.55(c).

Reference:

1. American National Standards Institute ANSI N14.5, American National Standard for Radioactive Materials -Leakage Tests on Packages for Shipment, New York, NY, 1997.

NAC International Response to Thermal Evaluation RAI 4-1:

As indicated by the reviewer and stated in SAR Section 4.1, as shown in Figure 4.1-1, the containment boundary extends beyond boundary seals.

"The MAGNATRAN transport cask containment boundary is defined by the following components: (1) bottom inner forging; (2) inner shell; (3) top forging; (4) cask lid

• and lid inner O-ring; and (5) lid port coverp/ate and lid port coverplate inner 0-ring."

Page 5 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• Also, within the weld subsection of SAR Section 4.1.2 states the following:

"Upon completion of containment vessel fabrication, the cask containment boundary is hydrostatically tested in accordance with ASME Code requirements to ensure the integrity of the welds and containment components as described in Section 8.1.3.

During and following fabrication, the containment boundary of each cask is leakage tested in accordance with Section 8.1.4.

7 The post-fabrication leakage rate test is to leaktight criteria of 1 x 10- ref cm3!sec, per ANSI NU.5-1997. Test equipment and methods are selected to ensure a minimum test sensitivity of one-half the reference leak rate, or 5 x 1 o-s ref cm3!sec. The 7

equivalent allowable helium leak rate at reference conditions is 2 x 10- cm3!sec 7

(helium), with a minimum helium leak test sensitivity of 1 x 10- cm3!sec (helium)."

This existing SAR text specifies that the entire, as defined containment boundary, is tested to the required leakage standards. There does not appear to be any inconsistency present in the SAR that would indicate that the entire cask will not meet the leakage criteria. In addition, SAR Section 8 also states that all aspects of the containment boundary are tested. Nevertheless, the following text is added to the end of SAR Section 4, Page 4-1, as part of responding to this RAI.

"The entire transport cask containment boundary is tested to the American National Standards Institute (ANSI) N14. 5-1997 leaktight criteria during post-fabrication testing as described in Section 8.1.4. Periodic, maintenance, and pre-shipment testing to the American National Standards Institute (ANSI) NU.5-1997 leaktight criteria are performed on the containment boundary closures per Section 8.2.2."

In addition, text on SAR Page 1.3-8 is revised to also invoke the entire containment boundary. Revised text states:

"Leakage testing of the cask containment seal, in conjunction with the post-fabrication leakage testing of the entire containment boundary, assures that the containment does not leak. "

• Page 6 of 13

Enclosure 1 to ED202 l O107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• THERMAL EVALUATION NAC INTERNATIONAL RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION 4-2 Clarify Section 6.1.1, "Design Features," of the application to also address the leakage tests described in Section 8.1.4 of the application, in addition to Section 8.2.2 of the application that is already described in the application.

Section 6.1.1, "Design Features," of the application states that, "Containment boundary integrity is checked via leakage tests described in Section 8.2.2." However, the fabrication leakage rate test is described in Section 8 .1.4 of the SAR, in addition to the maintenance, periodic, and pre-shipment leakage rate tests are described in Section 8.2.2 of the application.

This information is necessary to determine compliance with 10 CFR 71.5l(a)(l) and (2), and 71.55(c).

NAC International Response to Thermal Evaluation RAl 4-2:

SAR Section 6.1.1 is revised to refer to both the acceptance & maintenance and pre-shipment testing, as described in SAR Sections 8.1.4 and 8.2.2.

The revised SAR text reads as follows:

• "Containment boundary integrity is checked via fabrication acceptance tests described in Section 8.1.4 and periodic, maintenance, and pre-shipment leakage tests described in Section 8.2.2."

Page 7 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPIDNo. L-2021-LLA-0000

• SIDELDING EVALUATION NAC INTERNATIONAL RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION 5-1 It is not clear that removing restrictions on the placement of high bumup fuel within the TSC in the proposed amendment does not impact the ability of the MAGNATRAN package to meet the regulatory requirements of 10 CFR 71.4 7.

The contents include high bumup spent fuel (maximum assembly average bumup exceeding 45 GWd/MTU) for both PWR and boiling-water reactor (BWR) up to a maximum assembly average bumup of 60 GWd/MTU. In revisions 0, 1 and 2, of the CoC, the PWR fuel contents contained a condition that "All fuel with bumup >45,000 MWd/MTU is treated as damaged fuel and is placed into damaged fuel cans" and BWR fuel was limited to a maximum bumup of 45,000 MWd/MTU. Therefore, any limitations for damaged fuel (i.e., additional cool time) and for the remaining package contents when damaged fuel is present also apply to high bumup fuel, regardless of the actual condition of the high bumup fuel. It is not clear from the information submitted in the consolidated application and subsequent supplements 1 for Revision 1 of the application, that fuel with a bumup greater than 45,000 MWd/MTU was considered for placement in all fuel assembly locations and not just in the damaged fuel locations. The evaluation to support elimination of this condition (i.e., stating that previous revisions included high bumup fuel in all fuel locations) was not given in the Revision 20C of the application. It was not clear to NRC staff whether previous analyses bound the new fuel loading of placing high bumup fuels in any location of TSC in order to meet the regulatory requirements for dose rates in 10 CFR 71.47.

This information is needed to determine compliance with 10 CFR 71.47.

NAC International Response to Thermal Evaluation RA1 5-1:

In the case of the PWR fuel, the previously NRC reviewed SAR Chapter 5 was consistently based on a maximum 60,000 MWd/MTU shielding evaluations for all fuel assemblies, whether damaged or undamaged. The evaluations documented were performed and maximum dose rates were reported for cases up to 60,000 MWd/MTU where the bumup range was searched for state points that produced maximum dose rates.

NAC calculation 71160-5508, in which this calculation was performed, was provided to the NRC in 2016 as part of the RAl-3 response package. That an undamaged fuel system (i.e., a full cask load of high bumup fuel was addressed) can most easily be seen in the results shown in undamaged fuel bounding source summary Table 5.1-9 of the SAR, which reports the bounding source for various conditions. Note, the 60,000 MWd/MTU PWR source was also documented in SAR Section 5 .2. Maximum bumup is, and was in previous condition, 60,000 MWd/MTU for the majority of state points. While high bumup fuel was not licensed in the undamaged fuel configuration, 1 See consolidated application dated July 1, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19186A385), as supplemented on July 15, 2019 (ADAMS Accession No. ML19203A252), and August 7, 2019 (ADAMS Accession No. ML19221B591)

Page 8 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• the results shown bound those of the 45,000 MWd/MTU system and there was no need to generate low burnup specific limit tables. As noted, the high burnup fuel was permitted in damaged fuel system. For the damaged fuel system, all assemblies, including the undamaged assemblies in the damaged fuel basket, were evaluated up to the 60 GWd/MTU limit. .

BWR evaluations were revised for this amendment request to increased burnup. It's important to recognize that SAR Rev. 0, Table 5.1-9 already contains high burnup results for the BWR system. This is because it used data generated prior to the NRC limiting the BWR system to 45 GWd/MTU during the initial approval of MAGNA TRAN. In other words, as a result of obtaining the original NRC approval of MAGNATRAN, SAR Section 5.8.4, which contains the detailed BWR shielding results, was re-written to limit the burnup to 45 GWd/MTU. As part of this amendment request, SAR pages were revised to account for the higher bumup. In particular, SAR Section 5.8.4.3 was revised for higher burnup limiting dose rates and bounding sources.

In summary, PWR results were already generated for undamaged fuel up to 60 GWd/MTU, as indicated by the maximum dose/source tables presented in the SAR and the BWR results were updated to reflect the higher allowed burnups. This in conjunction with the source description already present in the SAR as part of the initial NRC approval, extending to 60 GWd/MTU, using existing analysis methods and models in the SAR, resulted in minimum page changes within Chapter 5 .

• Page 9 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• PACKAGE OPERATIONS NAC INTERNATIONAL RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION 7-1 Provide details for the methodology and acceptance criteria for the condition assessment of MAGNATRAN TSCs that demonstrate that TSCs that may have experienced aging-related degradation during prior storage are capable of excluding water during all conditions of transportation.

Section 7 .1.2 of the application (Loading of Contents) described that TS Cs containing spent nuclear fuel that are to be retrieved from storage for off-site transport in the MAGNATRAN transport cask will be evaluated to ensure that the specific TSC stored in the storage overpack, which may have been subject to 10 CFR 72 normal, off-normal, accident and natural phenomena events, retains its ability to satisfy functional and performance requirements of the MAGNATRAN packaging certified content conditions.

The application also states that dry storage systems that have been maintained in an aging management program (AMP) will include a system specific review and an assessment of this AMP information record will be conducted as part of the off-site transport evaluation to ensure that the MAGNA TRAN packaging certified content conditions are validated. The application states that the TSCs containing spent nuclear fuel and experiencing only normal or off-normal events during storage will be evaluated for potential corrosion and cracking at the welds and any damage caused by removal of the TSC from the storage overpack.

However, the application does not include any specific details on how to determine whether a TSC that has been in storage and subject to aging mechanisms, is capable of excluding water during a transportation accident. It is not clear to the staff what parameters will be evaluated, the methods used to characterize those parameters, or the acceptance criteria that will be used to make this determination.

The staff also notes that an evaluation of AMP records can provide valuable information to inform the assessment; however, such a program is still undefined. As a result, the reliance on a future, undefined, AMP for storage provides only a limited basis to demonstrate that TSCs are free of aging-related degradation effects (e.g., corrosion and cracking) such that TSCs can maintain moderator exclusion.

This information is needed to determine compliance with 10 CFR 71.SS(c), 10 CFR 71.85(a) and 10 CFR 71.87(b).

NAC International Response to Thermal Evaluation RAI 7-1:

The SAR text noted by the reviewer is text that was part of the original NRC approval of MAGNATRAN. However, at that time the TSC was not credited with excluding water during a transportation accident. In response to this RAI, a new condition is added to the proposed CoC ( see proposed Condition 6(c)) limiting the use of canisters

• under moderator exclusion to be only those canisters who are within the initial term of MAGNASTOR (i.e., 20 years) or are brand new TSCs that have not been used in Page 10 of 13

Enclosure 1 to ED20210107 , MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• storage.

"(c) For TSCs to be shipped under the moderator exclusion option of this certificate, only TSCs that are within their initial term for storage or are new and haven 't been loaded and placed into storage are authorized for use under moderator exclusion."

Page 11 of 13

Enclosure 1 to ED20210107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• THERMAL EVALUATION NAC INTERNATIONAL RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION 7-2 Revise the operating procedures to provide instructions for opening the package when it is transported using moderator exclusion to ensure that the package is not placed in a pool.

NAC proposed revisions to the contents for shipments containing moderator exclusion that would increase the maximum enrichment to 5 weight percent uranium-235, obviate the use ofbumup credit, and other changes to the contents, which, consistent with moderator exclusion, have not been evaluated for an optimally moderated package. However, it is not clear from the package operations chapter, how a licensee receiving the package would know whether the contents were packaged using moderator exclusion or not. One cannot rely on the fact that any shipment of spent fuel from a storage general licensee would not be transported to a licensee that has a pool for opening the package, in the event that the spent fuel inside the canister would need to be repackaged for future storage or disposal.

This information is needed to determine compliance with 10 CFR 71.89.

NAC International Response to Thermal Evaluation RAI 7-2:

• The receipt facility is expected to receive a fuel inventory description for each TSC .

Comparing the received assembly enrichments and PWR assembly bumups to those shown in SAR Chapter 6 for moderator evaluations will allow a receipt facility to determine if the package was shipped using moderator exclusion. In response to this RAI, a caution note is added to Step 21 of SAR Section 7.2.2 that describes the process.

"Caution: Section 6.1.2 of the SAR contains, or contains references to Chapter 6 sections, that contain enrichment limits and/or bumup credit or other requirement assuring criticality control of the system. Options are providedfor transport assuming moderator (water) intrusion into the TSC cavity or for applying moderator exclusion. Enrichment limits that apply moderator within the TSC will bound a TSC unloading scenario into an unborated spent fuel pool. Increased enrichment limits exist for transport applying moderator exclusion. Receipt of a TSC containing.fuel in excess of the moderated evaluation threshold indicates use of moderator exclusion for transport. For fuel requiring moderator exclusion unloading of the TSC from the transport cask into a spent fuel pool environment is not permitted unless unloading facility specific enrichment limits have been licensed and are met by the payload [e.g.,

soluble boron of sufficient level to prevent criticality is available in the spent fuel pool water at the unloading facility]. "

Page 12 of 13

Enclosure 1 to ED202 l O107 MAGNATRAN Docket No.: 71-9356 EPID No. L-2021-LLA-0000

• NAC INTERNATIONAL RESPONSE TO REQUEST FOR SUPPLEMENTAL INFORMATION ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 8-1 Either revise Chapter 8 to state that all TSCs containing spent fuel (whether loaded from storage, or loaded with spent fuel on-site prior to transport) are leak tested in accordance with Section 10.1.3 of the MAGNAS TOR SAR, Revision 9 (ADAMS Accession No. ML17293A085) or revise the MAGNATRAN SAR to incorporate TSC leak testing requirements.

MAGNA TRAN packages that are transported using moderator exclusion rely on the TSC entire confinement boundary as a special design feature in addition to the MAGNATRAN containment boundary that both together prevent a single packing error from permitting water in-leakage into the TSC and allowing contact with the fissile material, as required by 10 CFR 71.SS(c).

However, it is not clear in the MAGNA TRAN SAR that TSCs, which are loaded and not placed into storage (load and go scenario), have the same leak test requirements at fabrication as TSCs that are loaded and placed into storage under the CoC No. 1031 for the MAGNASTOR storage system. Section 8.1.4 of the MAGNATRAN SAR only includes fabrication leak test requirements for MAGNATRAN containment boundary and does not include leak testing of the TSC shell weldment after completion of the TSC shell seam and shell to bottom plate weld, the TSC composite closure lid, and the TSC vent and drain port inner port covers and welds, which are included in the MAGNASTOR leak testing requirements .

This information is needed to determine compliance with IO CFR 71.SS(c).

NAC International Response to Thermal Evaluation RAl 8-1:

MAGNATRAN SAR Chapter 8 is specific to the transportation cask and not the TSC.

While the CoC does invoke SAR Chapter 8, the requested TSC leak testing requirements is placed in the proposed CoC changes (see proposed CoC Condition 6(d)) in order to make it clear to the shipper that the TSC must be leak tested in accordance with MAGNAS TOR FSAR, Section 10.1.3 requirements.

"(d) For TSCs to be shipped under the moderator exclusion option of this certificate, the TSC confinement boundary shall have been leak tested in accordance with MAGNASTOR FSAR, Section 10.1.3 leakage test requirements. "

• Page 13 of 13

Enclosure 2 to ED20210107 Page 1 of 1 Enclosure 2 Proposed CoC Changes

• No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion RAI Response Submittal MAGNATRAN SAR, Revision 21A

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1.

2.

a. CERTIFICATE NUMBER PREAMBLE 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 1 OF PAGES 42

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

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

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

a. ISSUED TO (Name and Address) b. TITLE AND IDENTIFICATION OF REPORT OR APPLICATION NAG-International NAC International, Inc., application dated 3930 East Jones Bridge Road July 1, 2019, as supplemented.

Norcross, GA 30092

4. CONDITIONS _*~ g:, ~" .

This certificate is conditional upon fulfilling the requir&{ffis,i~o cli'fy~,p~AtiJ.,raP,plic.able, and the conditions specified below.

tr' ~-\;p.

• """"' ~'.:11 °3/4~q'~

,)

5. y ~

(a) Packaging (1) Model No.:

(2)

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

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

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

• 1. a. CERTIFICATE NUMBER 5.(a)(2) 9356 Description ( continued)

FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 2 OF PAGES 42 The package lid is a 7.75-inch-thick stainless steel disk used to close the package. The lid is attached to the top forging by forty-eight, 2-8 UN-2A socket head cap screws. The socket head cap screws screw into the tapped holes in the upper forging. The package lid is sealed by two concentric O-rings, as is the coverplate for the lid port, using inner metallic and outer ethylene propylene diene monomer (EPDM) O-rings. The MAGNATRAN package contains a lid port that is closed by a bolted Type 304/304L stainless steel coverplate with dual O-rings.

There are four stainless steel coverplate bolts. The lid port provides access to the port opening and the quick-disconnect fitting for backfilling and sampling the cavity gas during loading and unloading.

The neutron shield is comprised of NS-4-FR encased in stainless steel enclosures. The neutron shield material and its enclosure have two thicknesses, 5.8 inches and 6.4 inches, and is attached onto the outer shell aloR_~ 1~e l~l1Qtl'1 ~fi.ltl),e active fuel region around the circumference of the package cavity. t, i:J""~,li \\ u u ~ (l'?:;i Ui tr' "'., .,"" '-l:,.,_,,;;f~

Two diametrically opJi>~iAf lifting trunnions are bolted td tb~putside of the top forging to lift the transport package.,C:ffi_~ior to transport, the lifting trunnions a'-'\emoved and replaced with trunnion plugs. 1wo1\@J~tion trunnions are located on the0jufe]d$shell near the bottom of the package to permif.r~tatfo~:B" een the horizontal_,,. * *~icaf,p,o_sitions and to provide longitudinal tiedMvn restral h~ aft~ection/ ,~ation trur:mions are located approximately 51/2!nches off s~ cen~~iToe to,.. e that thJf;,~sk rotates in the proper direction. <ft: I

• Q r=> a,m A cavity spacer ,is usecl::,forf &;,_ , rt ;r;s~,s.t I ~

ari,cfys.ttpporftn'e canister and to minimize

. Iong1*t-t'u'd*

excessive u 1na' I ;;::\1~""*"

• • * ,_ enwn, &,t/.rnt,.. *H!* ,rn. .P s.~Jf§fty, wjwh is sized to accommodate the long TSC. ()_ . ij 1( /._:f '-..:

~1"

~~9  ;~{

The MAGNATRAN,sp'ackage has~cu~,; * /~\\.1'*°~ ,"p~imf§ae_t limiters~qonsisting of a combination of

"> ,:) . *f/J/Jil '**-l-'i' ~ =

redwood and balsa \fi{~P,9 encased imat~ajmless-sfeel sh~lf*~he impact limiters are bolted over each end of the packar~;t~ limit the g-loads acting on,;!l.tJ,p'ackage during a package drop event. The impact limiters ar~.,-~ttac_hed to the lid}c;l!.Jd o>ottom plate via 16 tapped holes for retaining rods and nuts. ,,~- is:--1~ ...}'.,.- <-i]S.

. .t-*~4 y'-1. ,

The TSC is constructed of a stainless steel cylindrical shell, bottom-end plate, closure lid, closure ring, and redundant port covers. The TSC confines the fuel basket structure and the spent fuel or the Greater-Than-Class C (GTCC) waste basket liner and GTCC waste. The TSC cylindrical shell is dual certified 304/304L stainless steel with a 72-inch diameter and is 1/2 inch thick and either 191.8 or 184.8 inches long, depending on the contents. The bottom end plate is welded onto the lower end of the TSC shell and is 2.75 inches thick. The closure lid is 9 inches thick and is either a solid stainless steel closure lid or stainless steel/carbon steel closure lid.

The closure lid is welded onto the upper end of the TSC shell. The dual port covers provide a dual-welded closure system for the vent and drain ports. The GTCC TSC is similar in design and construction to the TSC's for spent fuel, but instead of a basket, it contains a GTCC waste liner.

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1.

5.(a)(2)

a. CERTIFICATE NUMBER 9356 Description (continued)

FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U) F-96 PAGE 3 OF PAGES 42 The PWR fuel basket design is an arrangement of 21 square, stainless steel fuel tubes held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The 21 tubes develop 37 positions within the basket for the PWR spent fuel.

Each PWR basket fuel tube has a nominal 8.86-inch square opening. Each developed cell fuel position has a nominal 8.76-inch square opening. The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube. Each neutron absorber sheet is covered by a thin stainless steel sheet to protect the neutron absorber during fuel loading and to keep it in position. The neutron absorber and stainless steel cover are secured to the fuel tube using weld posts distributed across the width and along the length of the fuel tube.

The PWR damaged fuel basket is designed to store up to four damaged fuel cans in the damaged fuel basket assembly in the short TSC. The damaged fuel basket assembly has a capacity of up to 37 undamaged PW~ ftteJ ass~tnblies, which includes the four damaged fuel can l?cations. A damaged./uf):itn~rfiayJb'e~pl~_qe~JlJ?*ea_ch of the four d_amaged fuel can basket locations. The arrange11;1ent,0ftubes and fuel pos1t10Q§.$1s the same as m the standard fuel basket, but the desig~,-<:-~f"Efach of the four corner supp~rf >>'"eldments is modified with additional structural support tq::pr6vide an enlarged position for a damaged

<~,,~.,_,, ~... '\

fuel can at the outermost corners of the fuel f>'asket. Each damaged fuel can locatioJJ bas a nominal 9.80-inch square opening. A damif'ged

/j',,.

ffi~G or an undamaged fu

~

<Efmbfy:roay be loaded in a damaged

~~

fuel can location>?t/

i~~ ' ~{* .

Similar to the R~R basket,. ./ R b~~k~ cd~ f 45 stain~s steel fuel tubes that develop 87 ba~Mt locafi~ristd'.)t;y~~BWRls~e e: acj;l BW~~;::isket fuel tube has a nominal 5.86-!nch squ9r~-~opel1!,l_~~~ Ea~Jt~he)~P1r1?rH .~iii~1n ha~~1_nominal 5.77-in~h squa~e opening. The B~R ba;~~1\}W,~1-tQ~~~:~~ j~r,1~ 1 n9,~;~fcular~&ylmder configuration by side and corner supROitt welciments that.::are) olte:d~to the outer fuel fabes. The fuel tubes support an 11 11 enclosed neutrot~J;>sort5~~]i~~ ' '8tlf8J~19p~ides oJje fuel tube for criticality control.

Each neutro~ abs~Jb~r s~e'et 1s\§ov /NJ/J . .0-~?'l-~~\pf stainli~ steel to protect the n~utron absorber during fuel,;19:ijl~mg and to i<(lE;1~},!tlll:ipos1t10n. T~eiAeutron absorber and stainless steel cover are secured to 'tf.\~f~el tube using weld posts di§!@Wted across the width and along the length of the fuel tube. ,;, \~-ti. ..)1_ ~

Th e d amage d f ue I can con f .mes r*.('thy\_f-.

• e

• ue'1*"7

• ma-t~~

erra I yth*

w1 1n th e can t o mm1m1ze. . . th e po ten t*1a Ifor dispersal of the fuel material into the TSC cavity. The side plates that form the upper end of the damaged fuel can are 0.15-in thick and the tube body walls are 0.048-in thick (18-gage sheet).

The damaged fuel can lid plate and bottom thicknesses total 11/16 inches and the lid overall height is 2.32 inches. The damaged fuel can bottom plate thickness is 5/8 (0.625) inch. The damaged fuel can is designed in two lengths: an overall length of 166.9 inches with a nominal cavity length of 164.0 inches; or an overall length of 171.8 inches with a nominal cavity length of 169.0 inches (shorter fuel assemblies may be accommodated with a fuel assembly spacer to limit axial movement). For the shorter damaged fuel can, a spacer is used in the damaged fuel basket assembly or alternatively fixed to the damaged fuel can bottom plate to provide an overall height of 171.5 inches. The damaged fuel can (DFC) lid and bottom include screened drain holes .

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

• 1.

5.(a)(2)

a. CERTIFICATE NUMBER 9356 Description (Continued)

FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 4 OF PAGES 42 The stainless steel GTCC waste basket liner is designed to hold GTCC waste and dimensionally fit in a TSC. The GTCC waste basket liner is 173 inches long with a 1-inch-thick bottom plate welded onto it. The GTCC liner stainless steel shell is 2 inches thick for structural and gamma shield functions, and has lifting lugs welded on the inside diameter of the shell. The liner design also includes an outer ring and a middle support under the bottom plate and drain holes in the bottom plate to facilitate free flow drainage from the liner. The GTCC TSC includes a sump location in the bottom plate and the closure lid includes a drain tube assembly to enable draining and drying of the loaded TSC.

The package has approximate dimensions and weight as follows:

• 5.(a)(3) 71160-512, Rev. 1 71160-530, Rev. 1 71160-531, Rev. 2P 71160-551, Rev. 10P 71160-559, Rev. 0 71160-571, Rev. 10P 71160-572, Rev. 9P 71160-574, Rev. 6 71160-575, Rev. 11 P 71160-581, Rev. 5 71160-584, Rev. 8 71160-585, Rev. 13 71160-591, Rev. SP

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

• 1.

5.(a)(3)

a. CERTIFICATE NUMBER 9356 Drawings (Continued)

FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 5 OF PAGES 42 71160-598, Rev. 7P Basket Support Weldments, MAGNASTOR - 87 BWR 71160-599, Rev. 8P Basket Assembly, MAGNASTOR - 87 BWR 71160-600, Rev. 5P Basket Assembly, MAGNASTOR - 82 BWR 71160-601, Rev. O Damaged Fuel Can (DFC), Assembly, MAGNASTOR 71160-602, Rev. 1 Damaged Fuel Can (DFC), Details, MAGNASTOR 71160-620, Rev. 1P Top Fuel Spacer, MAGNASTOR 71160-671, Rev. 2P Details, Neutron Absorber, Retainer, For OF [Damaged Fuel]

Corner Weldment, MAGNASTOR - 37 PWR 71160-673, Rev. 1 Damaged Fuel Can (DFC), Spacer, MAGNASTOR 71160-674, Rev. 4P DF Corner Weldment, MAGNASTOR 71160-675, Rev. 3P OF Basket Assembly, 37 Assembly PWR, MAGNASTOR 71160-681, Rev. 1 DF, Shell Weldment, TSC, MAGNASTOR 71160-684, Rev. 2 Detail,_S;,. DFrcClps[Jr:,e Lid, MAGNASTOR

--~ ,,_,, *c:,I I""' Ii "' R 71160-685, Rev. 8 ,~~Ji:f.$1& A!ssembJ:yi<<.~~GNASTOR 71160-711, Rev. 1 "1,<:ncc Waste Basket lsit:]~_9J MAGNASTOR 71160-781, Rev.1 Shell Weldment, GTCC TSG~MAGNASTOR 71160-785, Rev. GTCC TSC, Assembly, MAG@STOR

}~~

t:-"' f\

.5.(b) Contents .:!k (1)

~

(i) Undamage§!J?W

. ft, ,D ~-*-~_~A C .* rr**--*-,*,**~***-**---r***~--* .....,

• 'Jli~P.?l<et as~mbly /Jli!h.Q_l!l.fl29fil1.l!J1

...__:.,-. ,;.,#::"------:::w.:~,::r..~o~,:. *r--r,1 *,-rrr damag~ fuel;Jtspent _nuclear fuel that does not have any v1s112.I~, defo~~g~,~~0tme~~~ I~ wing t~!,occurs m the reactor, asse~blies_that dqtp?,~ hav~ mi~,1,in~~~~F~rE ~~t~pblies jttt;J missing rods that are replaced by solid stainless sfo~;'.f?f zirconium fiVf&{jf~~s that dIspla~e~fwolume equal to or greater than the original rods and asie~blies that do not contain stf.~~tural defects that adversely affect radiological and/or criticality s~Jety and/or result ir;tynsupported fuel rod lengths in excess of 60 inches and that can be hariqlediJJ¥- 00r:r;r1{1l,.p,e~r.is. Undamaged PWR fuel is loaded into the 1 1 short TSC, except for Combustion Engifi6erilig (CE) 16x16 fuel assemblies, which may be loaded into either length TSC.

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

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 6 OF PAGES 42 The burn up credit loading curve in Table 3, must be used for the 37 assembly loading profile.

WE15x15 fuel may use the burn up credit loading curve in Table 4, with the 33, 35 or 36 assembly loading scheme provided the required cell locations for that profile shown in Figure 1 are left empty, at a minimum. Fuel assembly burn up, minimum initial average enrichment 1 , and cool time requirements are provided in Table 9 and Table 11, for PWR baskets with Type 2 neutron absorbers and Table 12 and Table 14 for baskets with Type 1 neutron absorbers.

t?-~.([1U~7g"e~/t~vr;ye,~.we or,!y apP,ltcapJe,to ~yste.i;;ns ~dtcre~i!ing.;;i;npcterc1t9(~xc_l,usiqn:,.Jn.!Efil

~drichrnemfup tO*~ ~ ..% 235 U, with no burnup"reguirement>iS permitted wlien crediting mmctergtor ~~clus1on-!

Unirradiated fuel and unenriched fuel are not authorized for loading, except that unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than 6 inches. An unenriched rod max be71use_~ ~~ i~,,Seplacement rod to return a fuel assembly to an undamaged condition. 1-_;, 11)\ ~ o N t~ 'iJ) iJ fl 0

~ _\.._~U ~ !J fF ~i~~sf' ~ {~-;;/:(Ii *&

Undamaged PWR fueP=ats*emblies may contain nonfuel*hcfraware. Fuel assemblies with an instrument tube tie 4:t>~"(epair shall be loaded with fuel ins~~jand/or top spacers to ensure proper spacing and *suRROrt of the fuel assembly. Fuel ins.ertsJ~nd/or top spacers are not required when uf~ th-~~l~,~~ed fuel tube basket ~~e t~§.(,t,op nozzle is ad~qu_a~ely supported. T~~l~~nfuel ha!l~f"8, may be load1'/4~Jeomplet r~.~sembly or as 1nd1v1dual 7 components, mchv1dual nonf1/4(;!J~J;c;icls m ~ f4!1-l.~J1~~ rods or 11>~_rt1al-length rods/rodlets.

Partial-length rio]ls/rodlets a1feit§'r&iitted 1in ~uide , .. ,.~ providedfguide tube plug devises are insta_lled. No~tld§I hard~arffsJ'> \

• t4'~t'#e:ei~:.::~2"'%*,i~~d1c,,ool tiiffei or cobalt-60 activity requirements 1trr1ables0.6~8. i"l_ ernblle t '~.ti 9 oonfuel!hardware must meet the 1/eitro~~.0sorbers),

additional cool~tirfle rei((~* 1/r ll /' n I' H Typ~* and Table 15 (for Type 1 neutron ~~orbe,t~'  ;:;,1

"' "-"1'

\!I~'ll ............

(/if} . ..'.1'1:5' 4'?_,;:;,,

. !;;?¥ Hafnium absorber.assemblies (ljlF q,' 1,. all.ovved for \@es"tinghouse (WE) assemblies and may have a maxi~'Dr.i],:exposure of 41..'if./1':,,,\,,-. -fl:Jahd mu_ s'f:_ *b~ve a minimum cool time of

'tli'fJ,, "l/\\,,w- ft';3/4 16 years. Fuel assemblies {/ ,;

may contain any number of.:_1:1.nirradiated ~->')

nonfuel solid filler fuel replacement rods. Activated ~1¥Iinless steel fuel r~glacement rods are limited to 5 steel rods per assembly, 1 assembly per ba~k,et;,-,-;a'nd,.a\mc:))(ip,rfga{steel rod exposure of 32.5 GWd/MTU. Fuel assemblies with activated stainless**ste~l*~o&frnust be cooled for either a minimum of 21 years or the loading table minimum cool time (as adjusted for additional cool times for nonfuel hardware, as applicable) plus 1 year, whichever is greater.

fuel a?.~e[llql}es lo.ad_edwith in-core iQstru111~ntthJmbl"~J mll,~t r:i~et th~, 9ddition_al coo} t~rrui requirements in Table 5 or Table 15, as appropriate, for BPRAs dr GTRDs, whichever is.

bqµriding, for Westingt:Jouse and:,B&Wfuel

(,:i,\,*,\* ,*** * , , , ,, .,,;* ',,*,_/' , ,. .,--t , *,*\" types.1:ind

'< .* .,*,*/

for *\Reactor,conti:ol compooents (RCGs),

• ,.*\,,;:., 1**'~,... ,}\, * ,< ' ' ' ,*, * , ' ,.*~

for CE fuel types:\The additjonal'cool time requirements for assembli_es witn nohfuel h.ardware

~te a_d9ed totAl)'(fidditipnal ~ool_time r~quir~men,t? dl,I~ to dar:iaged fue),al5:p being loaded_ ir(_.

th~ same TSC._.Reador' control _components (RCCs) are restricted to fuel storag_e Jocc3tionsJ 1_,,

1 Assembly average fuel enrichment is the enrichment value determined by averaging the entire fuel region (U02) of an individual assembly, including axial blankets, if present.

235 U wt% enrichment over the

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1)(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER Contents - Type and Form of Material (continued) 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 7 OF PAGES 42 iti.13,*.18,

}~,:>_,. ** ':,\.*:Jf 19.,,20,-25,26

• .+[>'.* *'"\ ,,, "'

al"}Q27 s'/ ** ,

in Figure

• 0 ,,,,:_*\:,

1. OrJly one;neutrqn source

~*** * ** .;},'(**, ":*\[/~

assembly

• ,>_*.o,,}-',,'" :/I<*'\,~.

(NSA)ii

,,',,. "'}J!~/. **,.=*~,.,..-/_-:"'\

• p*errl)itted to bet1*oadedJn_ a IS~infuel storaQe .locations 11)12, 1;}~ ;t 8,_ 1.9,> 20,:,?5; 26jjr 27, as shown .on *Figure_1 J NSAs may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rodlets and/or thimble plug rodlets. NSAs, guide tube thimble plug devices (GTPDs), and burnable poison rod assemblies (BPRAs) are not authorized for CE fuel assemblies. In addition, the following un-irradiated nonfuel hardware may be loaded with the fuel assemblies:

stainless steel rods inserted to displace guide tube "dashpot" water, instrument tube tie components, and guide tube anchors or similar devices. Axial power shaping rods are not allowed contents .

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 9356

b. REVISION NUMBER FOR RADIOACTIVE MATERIAL PACKAGES 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 8 OF PAGES 42 Table 1 - PWR Hybrid Fuel Assembly Characteristics No. of Min Min Max Max No. of Guide Max Max Hybrid Hybrid Clad Clad Pellet Active Vendor Array Fuel Tubes Pitch Load Assembly Group OD Thick. OD Length Rods (See (in.) (MTU)

Note 1 (in.) (in.) (in.) (in.)

BW BW15H1 H1 15x15 208 17 0.5680 0.4300 0.0265 0.3686 144.0 0.4807 BW BW15H2 H2 15x15 208 17 0.5680 0.4300 0.0250 0.3735 144.0 0.4807 BW BW15H3 H3 15x15 208 17 0.5680 0.4280 0.0230 0.3742 144.0 0.4807 BW BW15H4 H4 15x15 208 17 0.5680 0.4140 0.0220 0.3622 144.0 0.4690 BW BW17H1 H1 17x17 264 25 0.5020 0.3770 0.0220 0.3252 144.0 0.4681 CE CE14H1 H1 14x14 176 5 0.5800 0.4400 0.0260 0.3805 137.0 0.4115 16x16 ,~* 11! 11r0.~ro63 0.3820 0.0250 0.3250 150.0 0.4463 CE CE16H1 H1 2~§. R; ' \'i"\\5 L_ t1 ...... ,~ 1 WE WE14H1 H1 14x14 - ~17§:,.Ki' ~-17 0.556©1 ,#0.4000 0.0162 0.3674 145.2 0.4144

,:;;',.;:;:/t .,,_,,Jl ,.,\,.. ..¢".A~

"'.r-~H l">

WE WE15H1 H1 15xjs{: ':e::!J204 21 0.5630 0.4:2i0. 0.0242 0.3669 144.0 0.4671 WE WE15H2 H2 15xsl jy r.:~~~., 204 21 0.5630 0.417{[.' ? 0.0265 0.3570 144.0 0.4469

• ~

WE WE17H1 H1 ,;:;1.,Zx1 (~~ :---.,_264 25 OJ64J;) j0.p205 0.3232 144.0 0.4671 WE WE17H2 H2  ;;71,;7x17 * \i .

b, .......... -f"':

oo <Of©225 0.3088 144.0 0.4327 Notes: ~

0

1. Combined number of guidt-2fund instru 0 I"- .

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER Contents - Type and Form of Material (continued) 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 9 OF PAGES 42 Table 2 - PWR Fuel Class Assembly Characteristics Fuel Class Characteristic 14x14 14x14 15x15 15x15 16x16 17x17 2 BW, SPC, Base Fuel Type CE, SPC WE, SPC WE, SPC BW, FCF CE WE, FCF Max Initial Enrichment (wt. % 235 U) 3 5.0 5.0 5.0 5.0 5.0 5.0 Min Initial Enrichment (wt. % 235 U) 3 1.3 1.3 1.3 1.3 1.3 1.3 Number of Fuel Rods 4 176 179 204 208 236 264 Max Assembly Average Burnup 60,000 60,000 60,000 60,000 60,000 60,000 (MWd/MTU) 5 Min Cool Time (years) 4 4 4 4 4 4 Max Weight per Storage Location -~1;See\N0te

• r:,) !Qt TT:::.

J'f: li'~~Nt See Note{\ r*-~""' 9,,e 1 See Note 1 See Note 1 See Note 1 (lbs.)

Max Decay Heat per Fuel Location

" 'If~

~,,1;;,-.0'>>"

Seq-;Note 2 See Note 2

,~_1 ef 1/

~I"\

,~,

A!i See Not~-* ~pee Note 2 See Note 2 See Note 2 (Watts) 6 ~-"-. ,-,,~-4 2

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

3 All reported enrichment values are nominal preirradiation fabrication values.

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

5 Assembly average burnup is the burn up value determined by averaging the burnup over the entire fuel region (U02) of

. , ar:i_ in 9. iY.ig_~ c:31_~~-~~9:-~!Y-2.. L~s!!;l,s!Lr:i~ _,a.,xJ.~I -~'cl~ kets' if present. ;l\ll__f(,J gJ,:tt]!.bJ?)i rn_ldl?.?.15 ,QQQ}!9V1{d1-MJll J.!!>JFQ§tQ_d ~:;i-~ d~magg~

fuel_and_1s_pJaced into d9maged fuel*cans:

6 Maximum uniform heat load per storage location.

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

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 10 OF PAGES 42 Table 3 - Maximum Initial Enrichment Assembly Undamaged Fuel 15 Year Minimum Cool Time Zero (0) Max Initial Enrichment (wt% 235 U) 109 Burnup = C4 x Burnup (GWd/MTU) + Cs Assembly Absorber Maximum Burnup 18 :S Burnup Burnup ID (g/cm 2 ) Enrichment (GWd/MTU) < 18 (GWd/MTU) :S 30 (GWd/MTU) > 30 (wt%)

C4 Cs C4 Cs C4 Cs BW15 1.9 0.0501 1.69 0.0693 1.65 0.0748 1.60 BW17 1.9 0.0502 1.72 0.0687 1.70 0.0742 1.66 CE14 2.1 0.0473 2.04 0.0675 2.03 0.0759 1.93 CE16 0.036 2.1 0.0464 2.03ft,.,,,_, 0.0657 2.06 0.0733 1.99

• -*  ;;'-1.....

..l' 1'-"' it~{~~

WE14 2.2 0. Q-1~~.~--\~ 2~©8:.. * : !JrJtq,672 2.21 0.0725 2.29 1ft ,,~,,.<:,;'" ~y I.'

WE15 1.9 P

,. <I 0;:(:)494 1.74 0'.-0.§.§_?, ~ 1.72 0.0742 1.67 WE17 1.9 ~,"'.::: *-0.0494 1.71 0.06$5./~ 1.68 0.0749 1.61 BW15 1 ~-- 0.0507 1.61 0.0687 i .:-:;¥.A 1.59 0.0745 1.48

• BW17 1~;,\\ ~ ~-Q0503 1.66 0.0§,8_3q,:' ?'""*\63 0.0733 1.59 CE14 2'ii&

"11 It 1.95 .,0r:*-5.,.,.f ?1':97 0.0738 1.90 f'

I' V

CE16 0.030 11i:J ,'.'."71;;9.,._5, ,; 9 rzf:~9 0.0727 1.90 WE14 :4,',1 .J )2~0~' I:\ -_:So ~?1,0 0.0728 2.19 WE15 ..~f~9 ), . ..lc:i16§]j._  ; QJ,75 f:[6 0.0747 1.54 WE17 BW15 r-***

_l,9

\g1~8 ~'"'I *

-<<l I

1(P.., .. WT!

o!M~ .

i 3d !l'lll

• J.64 lf;_j

,l

~ d68,q~

0.Q¢,_$J{w I

,*-:r:--

J~5B

t

52 0.0737 0.0754 1.53 1.41 BW17 llf~l ~ ,-~- ~. O.~Qi{31"  :'1'::59 0.0748 1.47 CE14 2:f\ l~:G1Qef66 e~~92 0.0729 1.87 CE16 0.027 2.1:/ I0.04q_~ ../1 -~~  :<:10'[0657 f',;.t'1.92 0.0747 1.75 WE14 2.1*4;,~~) 0.0499 t1. .~~ ~

$'" **,llf\0'.0667 A: '::r,"" 2.10 0.0743 2.07 V {/I, 1.63 WE15 1.9 ,,,,,.~0.0503 V 0.09-z~sl::c,J 1.60 0.0749 1.46 WE17 1.9 . V0.0497 . ~,' 1.60 11s.9,068tf' 1.54 0.0749 1.41

~

}..::t>-

\/ 1? <<[;i: V"*

-c*!

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1)(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 11 OF PAGES 42 Table 4 - Maximum Initial Enrichment - Undamaged Fuel Configuration WE15x15 0 tional Confi urations - 20 Year Minimum Cool Time Zero (0) Max Initial Enrichment (wt% 23 su)

Number of 10s Burnup = C4 x Burnup (GWd/MTU) + Cs Assemblies Absorber Maximum Burnup 18 :5 Burnup Burnup Loaded (g/cm 2 ) Enrichment GWd/MTU < 18 GWd/MTU :5 30 GWd/MTU > 30 235 (wt.% U) C4 Cs C4 Cs C4 Cs 36 2.0 0.0497 1.93 0.0681 1.99 0.0747 2.00 35 0.036 2.1 0.0507 1.97 0.0673 2.08 0.0730 2.12 33 2.2 0.0504 2.12 0.0664 2.29 0.0745 2.32 36 2.0 0.0494 1.87 0.0687 1.90 0.0737 1.93 35 0.030 2.0 o.04,~9ra; ~1-,@2.~ 0.0688 1.97 0.0740 1.99 u 2.'"661~,r t,.. (f04~i7*U 1

33 2.1 \i'.'/.1,ll@"l<

(/Q'.0686 1'-,

2.15 0.0724 2.29

--~_,.,..,,,,f 36 2.0 P"1i.0501 1.83 0.0)77:~ ,, .

1.87 0.0741 1.84 35 0.027 0.0494 1.89 0.0675 0.0735 1.96 33 2.03 0.0730 2.21 Core Assembl NSA CE14 WE14 1.1 WE15 1.3 BW15 0.2 CE16 WE17 1.4 BW17 "'0.3 0.2

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356 Contents - Type and Form of Material (continued)

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 12 OF PAGES 42 Table 6 - Nonfuel Hardware Max Exposure and Required Cool Times Years)

Maximum Minimum Cool Time (Years}

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

BPRA 70 8.0 8.0 8.0 8.0 8.0 GTPD 180 8.0 8.0 8.0 8.0 8.0 Note: 1. Specified minimum cool times for BPRAs are independent of the required minimum cool times for the fuel assembly containing the BPRA

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

3. The maximum exposure and minimum cooling time limits for NSAs without absorber rods are the same a~ those for GTPDs ':'.:'h~ thfeJ11pi_,Q2um exposure and minimum cooling time limits for NSAs with absorber r~d~:,a~~Jg*~ sarme~i::,f~-fje17for BPRAs. .

4. Only GTPDs that do n9t. 1q9Jude absorber, or po1so41~/:9ds or water displacement rods are allowed contents. ,,_ <1;..., ~ "'iJ

~

-~

• Hardware BPRA GTPD B&W 17x17 27.0 107.8 Note:

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER Contents - Type and Form of Material (continued) 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U) F-96 PAGE 13 OF PAGES 42 Figure 1 - Undamaged Fuel Basket Loading Profile8 (2)

(5) (6) (7)

• (26) (27) it TSC~ALIG NM ENT MAAK

\f

.,0 t:i!J/

3?7"assemb

• ~ u 36 asQmb~¥r_(oad_i.ng: _remo_v__ e l?r ~~--

.t:::r *,;*./\ .,..} Jin- ~~-<..

35 assembl/ load!"ngJir:~m¢v,e 1-tr 18 33 assembly loading: remove 19, 18, 20, 12 Note: The 33, 35 and 36-Assembly patterns also apply to the damaged fuel basket.

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

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000)

CERTIFICATE OF COMPLIANCE 10 CFR 71 FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U) F-96 14 OF 42 5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 9-Loading Table for PWR Fuel - 23 kW/Package 1 Minimum Initial Assembly Average Burnup ~ 30 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt% 23su E 14x14 14x14 15x15 15x15 16x16 17x17 17x17 2.1 ~ E < 2.3 5.7 5.8 6.7 6.9 6.3 6.8 6.8 2.3 ~ E < 2.5 5.7 5.8 6.6 6.9 6.2 6.7 6.7 2.5 ~ E < 2.7 5.6 5.7 6.6 6.8 6.1 6.6 6.6 2.7 ~ E < 2.9 5.5 5.6 6.5 6.7 6.0 6.6 6.6 2.9~E<3.1 5.6 5.6 6.4 6.7 6.0 6.5 6.5 3.1 ~E<3.3 5.4 5.6 6.4 6.6 6.0 6.5 6.5 3.3 ~ E < 3.5 5.4 5.5 tr l~i:~

,,,,, ~'.\\ l{I

[fa3= '°')j ~-6 5.9 6.4 6.4 3.5 ~ E < 3.7 5.3 y-5k5~ .o/J (615

,ti rJ 5.9 6.4 6.4 3.7 ~ E < 3.9 5.3r ,;:;>~'5.4 6.2 6.5>._,;? f ~5.9 6.3 6.3 3.9 ~ E < 4.1 ,_5~3\'i!iJ 5.4 6.2 6.5 , c1,5 8 6.3 6.3 efr ~

4.1 ~ E < 4.3 ~~ 5.4 ,5.8 6.3

~- 6.1 6.4

-;ac\

6.3 4.3 ~ E < 4.5 5-~ 5.3 6.1 6.4 ':;11 *(r 6.2 6.2 4.5~E<4.7

~-........

G~,ti 5.2>t,' 6.1 5.8~{_ 6.2 6.2 4.7~ E < 4.9 ~ ' 5.1 6.0 5.7 6.1 6.1 E;;::4.9 ~ r.;flfe)

-,,, -* ' 6.1 6.1 i seiflbly~ve)'ia'g*<- d/MTU Minimum lniti~

Assembly Avg1~

Enrichment (J~ WE B&W wt% 235 U E 17x17 17x17 2.1 ~E<2.3 2.3 ~ E < 2.5 9.1 9.1 2.5 ~ E < 2.7 9.0 9.0 2.7 ~ E < 2.9 -~;9

.... *'fl 7.1 8.9 8.8

~'~

2.9 ~ E < 3.1 3.1 ~ E < 3.3 e.a? - ~

7.0 8.8 8.6 8.7 8.6 6.8 'JJ:~O A 8.6 , ji-9::0 J\..,.

- s,;;' 8:<5-<:/i,,;' . ""\g~o A

3.3 ~ E < 3.5 6.7 6 . 9"'*

)~;i". 7.7 8.6 8.6

[/'"i 6.9 ~

,~.

i.

3.5 ~ E < 3.7 6.7 8.4 8.9 7.6 8.5 8.5 3.7~E<3.9 6.6 6.8 8.3 8.9 7.5 8.4 8.4 3.9 ~ E < 4.1 6.5 6.7 8.2 8.8 7.5 8.4 8.4 4.1 ~E<4.3 6.5 6.7 8.2 8.7 7.4 8.3 8.3 4.3 ~ E < 4.5 6.4 6.6 8.1 8.7 7.4 8.2 8.2 4.5 ~ E < 4.7 6.4 6.6 8.1 8.6 7.3 8.2 8.2 4.7 ~ E < 4.9 6.4 6.6 8.0 8.6 7.3 8.1 8.1 E;;::4.9 6.3 6.5 8.0 8.5 7.2 8.1 8.1 1 . '-' means not allowed

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c: DOCKET NUMBER Contents - Type and Form of Material (continued) 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U) F-96 PAGE 15 OF PAGES 42 Table 9-Loading Table for PWR Fuel - 23 kW/Package 1 (continued)

Minimum Initial 35 < Assembly Average Burnup 5; 40 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt% 23su E 14x14 14x14 15x15 15x15 16x16 17x17 17x17 2.1 5; E < 2.3 2.3 5; E < 2.5 2.5 5; E < 2.7 9.7 11.9 13.5 14.7 11.6 13.7 13.7 2.7 5; E < 2.9 9.5 10.1 13.3 14.4 11.5 13.4 13.4 2.95;E<3.1 9.3 9.8 13.1 14.1 11.3 13.2 13.2 3.1 5; E < 3.3 9.1 9.7 12.8 14.0 11.1 13.0 13.0 3.3 5; E < 3.5 9.0 9.5 ~' 1,2,6c" 13.8 10.9 12.8 12.8

e'9 'A)'- .* i If?' I(

,

3.5 5; E < 3.7 3_7 5; E < 3.9 8.9 ~-- *u"'"'-"~

8.8 ' ~\J.3 1'2*~5"" I 12.3 i~J~, 10.8 12.7 12.5 12.6 12.5 13.§,,.*<>>- 10.7 3.95;E<4.1 8~7,'v 9.1 12.1 13.3 "' 7 ;1:0.5 12.3 12.3

<--~ii 4.1 5; E < 4.3 t~a-5 9.0 12.0 13.2 ¥t~\\.-,.:

12.2 12.2

~-5 4.3 5; E < 4.5 11.9 13.1 JOi3~ 12.1 12.1

,h § 4.5 5; E < 4.7 11.8 10.2/4, 12.0 12.0 4.7 5; E < 4.9 11.7 10.2 () 12.0 11.9 E~4.9 ~7) 10.1 "' 11.9 11.9 Minimum Initial~ errf!SI Av ' ~~-- d/MTU J,. I *,

Assembly Avg/= n Enrichment ffi WE B&W wt% 23su E flf:l'.' 17x17 17x17 2.1 5; E < 2.3 ""Z)

~ ....

2.3 5; E < 2.5 2.5 5; E < 2.7

-0*

-~

~

2.7 5; E < 2.9 ~i~5,th/,, f~~ 20.0 20.0 2.9 5; E < 3.1 3.15;E<3.3 3.3 5; E < 3.5 3.5 5; E < 3.7 1~,6:,)

13.!

13.1 12.9 111-.6 Zi~1,,.A

'l. ~

1'9.3 19.0 1 1 8.7

'<:;: ?""**

13.8 \~ t,'f8.6 jr ' 20.2

. )\. -~

20.7

,.""-~OA

~*6.8 16.5 16.3 16.0 19.6 19.4 19.1 18.8 19.6 19.3 19.1 18.8 3.7 5; E < 3.9 12.7 13.7 18.3 19.9 15.8 18.7 18.6 3.9 5; E < 4.1 12.5 13.5 18.1 19.7 15.6 18.4 18.4 4.1 5; E < 4.3 12.3 13.3 17.9 19.6 15.4 18.3 18.3 4.3 5; E < 4.5 12.1 13.1 17.7 19.4 15.3 18.1 18.0 4.5 5; E < 4.7 12.0 13.0 17.6 19.2 15.2 18.0 17.9 4.7 5; E < 4.9 11.9 12.8 17.4 19.0 15.0 17.7 17.8 E;::: 4.9 11.8 12.7 17.3 19.0 14.9 17.6 17.6

1. '-' means not allowed

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000)

CERTIFICATE OF COMPLIANCE 10 CFR 71 FOR RADIOACTIVE MATERIAL PACKAGES

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U) F-96 16 OF 42 5.(b)(1 )(i) Contents - Type and Form of Material (continued)

Table 10-Loading Table for High Burnup PWR Fuel- 21.85 kW/Package 1 Minimum Initial 45 < Assembly Average Burnup :5; 50 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W WE B&W wt% 23su E 14x14 14x14 15x15 15x15 17x17 17x17 2.1 :5; E < 2.3 2.3 s E < 2.5 2.5 :5; E < 2.7 2.7 s E < 2.9 29.5 2.9 :5; E < 3.1 21.7 25.2 28.4 30.3 28.9 28.8 3.1 :5; E < 3.3 21.3 22.9 28.2 30.1 28.7 28.6 3.3 s E < 3.5 21.1 r~2\12i.) ,8.0 29.8 28.4 28.4

"""f,:::."i, 3.5 s E < 3.7 20.7 I ~ li2t.,9'i 2,7117.' Ff

- ** t.,J, iJ ,,,29.7 28.1 28.1 3.7 :5; E < 3.9 20f§\,,. ~~ 21.7 27.5 . l{29~4 28.0 28.0

~\ (} ., ....,,,,,~

3.9 s E < 4.1 20~'2" 21.4 27.3 29.2,. 27.8 27.7

~)" fl' 4.1 :5; E < 4.3 ~0.0 21.2 27.1 29.0 27.6 27.5 4.3 s E < 4.5 ":t.9.7 21.0 26.9 28.8 I y

a

./P, 21.4 27.4 4.5 s E < 4.7 26.7 -c.4'f7.3 27.2 4.7 :5; E < 4.9 5 26.5 **;,.

27.1 27.0 E ;;::4_9 4C ~ F'26.;t (l;,i,7_0 26.9 Minimum lniti eifoix ~ve'r 55 G.Wd/MTU Assembly Av ear~;,i Enrichment 1*

.wE B&W wt% 23su E )

<f51 .;t'Zx17 17x17 2.1

5; E < 2. /Yci?'j?

,,./_, ~- r:,;;;;;_, _

_\~y

~

2.5:5;E<2.7 ;JJl' - ~"lt).

'L&~i:;,,-

,..f. l 2.7 :5; E < 2.9 2.9 :5; E < 3.1 t'~

-c---**-~

,._ >,J'

3/4\

<c,,, ' ~-. ~t,....\"\

3.1 :5; E < 3.3 26<<8' 31.7 3'5:8 34.9 34.9 3.3 :5; E < 3.5 26.4 rf~2-91~'? ~ *

,, -t.' *,{~ '*

35.5 34.7 34.6 3.5 :5; E < 3.7 26.2 213:'4 } 35.3 34.5 34.4 3.7:5;E<3.9 25.9 27.9 35.1 34.4 34.2 3.9 s E < 4.1 25.7 27.6 32.9 34.9 34.1 34.1 4.1 :5; E < 4.3 25.4 27.4 32.8 34.8 34.0 33.9 4.3 :5; E < 4.5 25.1 27.2 32.5 34.6 33.9 33.8 4.5 :5; E < 4.7 25.0 26.9 32.4 34.5 33.7 33.7 4.7 :5; E < 4.9 24.7 26.7 32.3 34.3 33.5 33.4 E;;::4.9 24.5 26.6 32.0 34.2 33.4 33.3 1 . '-' means not allowed

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356 Contents - Type and Form of Material (continued)

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U) F-96 PAGE 17 OF PAGES 42 Table 10-Loading Table for High Burnup PWR Fuel - 21.85 kW/Package 1 (continued)

Minimum Initial 55 < Assembly Average Burnup::; 60 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W WE B&W wt% 23su E 14x14 14x14 15x15 15x15 17x17 17x17 2.1::; E < 2.3 2.3::; E < 2.5 2.5::; E < 2.7 2.7::; E < 2.9 2.9::; E < 3.1 3.1::; E < 3.3 39.8 39.7 39.6 39.5 39.3 39.1 39.0 38.9 38.7 Max.

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

[MWd/MTU] [wt% 23sU] [Years]

10,000 1.3 4.0 15,000 1.5 4.0 20,000 1.7 4.4 25,000 1.9 5.5 30,000 2.1 6.9

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 18 OF PAGES 42 Table 12-Loading Table for PWR Fuel - 22 kW/Package 1 Minimum Initial Assembly Average Burn up::: 30 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt o/o 235LJ E 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1:::; E < 2.3 6.0 6.1 7.1 7.4 6.6 7.2 7.2 2.3:::; E < 2.5 5.9 6.0 7.0 7.3 6.6 7.0 7.1 2.5:::; E < 2.7 5.9 6.0 7.0 7.2 6.5 7.0 7.0 2.7:::; E < 2.9 5.8 5.9 6.9 7.2 6.4 6.9 6.9 2.9:::; E < 3.1 5.8 5.9 6.8 7.1 6.4 6.9 6.9 3.1:::; E < 3.3 5.7 5.8 ..**:, 7.0 6.3 6.9 6.9 3.3:::; E < 3.5 5.7 \'e'.9-,~c..,."' ~i 9~!

6\8l {E L.

1"""'.

l'J7.01,7. 6.3 6.8 6.8 3.5:::; E < 3.7 5.6~' *~"~5'7 . 6.7

.., ' *? 1 1.0 ,.,, ~6l 6.8 6.8

,.\y V.61:;2**

3.7:::; E < 3.9 ,5~6 5.7 6.7 6.9 ,,, ;;P"--~

6.7 6.7 t;;.. i;y 3.9:5E<4.1

~-55 5.7 6.6 6.9 6.1U J 6.7 6.7

• ~>>

4.1 :5E<4.3 ~5~ 6.6 -6?,!l*. . r.~,;; ~6.7 6.7 4.3:::; E < 4.51tt)" 5.5 6.6 .0 "°6.6 6.6 4.5:::; E < 4J"::::;;;;;. .0 ,---

t(g.6 6.6 4.7:::; E < 4l!3~ ~6 6.6 E 2': 4.9 lt~,-. ,.0  :::6::6 6.6 Minimum lnilif} ~ 35 GWd/MTU Assembly Avg. ears~

,; } t] j I '

"'--=i.WE EnrichmentQ wt% 235 U E ,"'-"': i 14Xl . . :¥:i .,_._

Yf ~ \ .

lll**~15 6,1 ~f7X17 B&W 17X17 2.1:::; E < 2.3 \l'1 ~ -

Y

{l 1~~'~1t{1

-g  !:>

2.3:::; E < 2.5 t,;f:z,,,1 (?tf,,§>

7.9 10.vu 818"-~/ 10.0 10.0

{~<)

2.5:::; E < 2.7 7J:5f') 7.8 9.8 10.6 ,,c-.~8:i

t.\;'-:.>

9.9 9.9 2.7:::; E < 2.9 7_4# "tll7 01, 9.7 10,4r "'8.6 9.7 9.7

_t.,..' ('J"' A 2.9:::; E < 3.1 7.3 7.6 ),..., l" -s9~5--<: :r1*012' 8.5 9.6 9.6 J,**v, .i 3.1:::; E < 3.3 7.2 7.5 9.4 10.1 8.4 9.5 9.5 3.3:::; E < 3.5 7.2 7.4 9.3 10.0 8.3 9.4 9.4 3.5:::; E < 3.7 7.1 7.4 9.2 9.9 8.2 9.3 9.3 3.7:::; E < 3.9 7.0 7.3 9.1 9.8 8.1 9.3 9.2 3.9:5E<4.1 7.0 7.2 9.1 9.7 8.1 9.1 9.2 4.1 :::; E < 4.3 6.9 7.2 9.0 9.6 8.0 9.1 9.1 4.3:::; E < 4.5 6.9 7.1 9.0 9.6 8.0 9.0 9.0 4.5:::; E < 4.7 6.9 7.0 8.9 9.5 7.9 9.0 9.0 4.7:::; E < 4.9 6.8 7.0 8.8 9.5 7.9 9.0 9.0

• 1.

E 2':4.9

'-' means not allowed 6.8 7.0 8.8 9.4 7.9 8.9 8.9

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U) F-96 PAGE 19 OF PAGES 42 Table 12-Loading Table for PWR Fuel - 22 kW/Package 1 (continued}

Minimum Initial 35 < Assembly Average Burnup S 40 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W CE WE B&W wt% 235LJ E 14X14 14X14 15x15 15x15 16X16 17X17 17X17 2.1 :;; E < 2.3 2.3:;; E < 2.5 2.5:;; E < 2.7 10.7 11.9 15.2 16.6 13.1 15.4 15.4 2.7:5E<2.9 10.5 11.2 14.9 16.2 12.9 15.2 15.1 2.9 :5 E < 3.1 10.3 11.0 14.7 16.0 12.6 14.8 14.8 3.1 :5 E < 3.3 10.1 10.8,~- 15.8 12.4 14.7 14.7 3.3:;; E < 3.5 9.9 1~tr'

,_, l,w 1!42;,,. G,1$ ~, 6 .., 12.2 14.4 14.5 f101~\?1 3.5 :5 E < 3.7 9-8~ ,,, 1-,:J'IJ 14.1

-* //

15.4~ i12.0 14.3 14.2

• t~~\~ " 10.4 ' ~

3.7:5E<3.9 ,J~7 10.3 13.9 15.3  ;!;-1',9 14.2 14.1 3.9:;; E < 4.1 <~*9.6 10.1 13.7 15.1 1lsti., 14.0 14.0 4.1 :;; E < 4.3 (

~

13.6 -1,ci

--:;::::-1~11. 7 . 13.9 4.3 :5 E < 4.5!li 1 9.4 13.5 ,-* 11.6 ,:: 13.8

- """~fr 4.5 :;; E < 4-f:::-.::.

4.7 :5 E < 4r.9;f' H

-~

E ~ 4.9 it.........

~;:3~

~-

• 13.3 ' 1"9'3

~ -6 3.7

-:,:::,1:6.5 13.6 13.6 13.5 Minimum Initial 1t:;45 ...,__........ d/MTU

,'!) (J

J Assembly Avg..._

(j,-\\ '

Enrichment~/.\~ . WE B&W wt3/4 235 U E \}) ""14X1'. 7X17 17X17 2.1 :5 E < 2.3 2.3 :5 E < 2.5 2.5 :5 E < 2.7 ,~ - '\ -

1. '~{

2.7 :5 E < 2.9 2.9:5E<3.1 15.7 15.3

~~ H"-

' *,,-"'(

16.7 1i]t:~~- 1'

\.

23~5 23.2 19.2 18.8 22.1 21.8 22.1 21.8 3.1 :;; E < 3.3 15.0 16.2 21.1 22.9 18.6 21.5 21.5 3.3 :5 E < 3.5 14.8 15.9 20.9 22.6 18.3 21.3 21.3 3.5:;; E < 3.7 14.5 15.7 20.7 22.4 18.0 21.1 21.0 3.7 :5 E < 3.9 14.2 15.5 20.4 22.2 17.8 20.8 20.8 3.9 :5 E < 4.1 14.0 15.3 20.2 22.0 17.6 20.6 20.6 4.1 :5 E < 4.3 13.9 15.0 20.0 21.8 17.5 20.5 20.4 4.3 :5 E < 4.5 13.7 14.8 19.8 21.6 17.3 20.3 20.3 4.5 :5 E < 4.7 13.6 14.7 19.7 21.5 17.1 20.1 20.1 4.7 :5 E < 4.9 13.5 14.5 19.6 21.3 17.0 20.0 19.9 E ~ 4.9 13.4 14.4 19.5 21.2 16.9 19.8 19.9

1. '-' means not allowed

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 20 OF PAGES 42 Table 13-Loading Table for High Burnup PWR Fuel - 20.9 kW/Package1 Minimum Initial 45 < Assembly Average Burnup :5 50 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment CE WE WE B&W WE B&W wt % 23su (E) 14X14 14X14 15x15 15x15 17X17 17X17 2.1 ::;; E < 2.3 2.3::;; E < 2.5 2.5::;; E < 2.7 2.7::;; E < 2.9 31.0 2.9::;; E < 3.1 23.8 25.2 30.7 32.7 31.3 31.2 3.1 ::;;E<3.3 23.5 lt24t'Z* 3'(:)'.~~ ~ 32.5 31.0 31.0 I" tri ~-~J: I fJ.

3.3::;; E < 3.5 23t2 ~ :,1rt4.4 c:11 ,32.2 30.8 30.8

'4' "

3.5::;; E < 3.7 .?"'> """'""'

~2~ff 24.1 30.0 "32.~' f/;

• 7 ~

30.6 30.5 3.7::;; E < 3.9 *022.6 23.9 29.8 31.9 e ,:l"""~3o 4 30.3

~3*0:2 3.9::;; E < 4.1 22.4 29.6 31.7 . V1--.,:,.,

30.1

~ , 1 30?b 4.1::;; E < 4';> -~2 29.4 ~

29.9 4.3$ E<Mst 3.2 , - 29.3 2~Q>> 29.8 4.5::;; Et7 29~"% 29.6 4.7$E~9 29~6J 29.5

                                                                                                                                      ""')

E;:: 4f.g""-" I , 29(,f; 29.3 Minimum 1~i\ial v~ G)'l{,d/MTU Assembly ~v,g.

                                    .       \(.,...-)

rs~ Enrichment,~ \_},,; B&W wt% 235 U E..,._...,,;.\) 17X17 2.1::;; E < 2.3 '\,1 2.3::;; E < 2.5 2.5::;; E < 2.7 2.7::;; E < 2.9 2.9$E<3.1 3.1 $E<3.3 28.9 31.7 36.1 38.1 37.3 37.2 3.3::;; E < 3.5 28.7 30.7 35.8 38.0 37.1 37.0 3.5::;; E < 3.7 28.3 30.4 35.7 37.8 36.9 36.8 3.7::;; E < 3.9 28.1 30.2 35.4 37.6 36.8 36.6 3.9::;; E < 4.1 27.9 29.9 35.2 37.4 36.6 36.5 4.1::;; E < 4.3 27.6 29.7 35.1 37.3 36.4 36.3 4.3::;; E < 4.5 27.4 29.5 34.8 37.1 36.3 36.2 4.5$E<4.7 27.2 29.3 34.7 37.0 36.2 36.1

• 4.7::;; E < 4.9 E ;::4_9

1. '-' means not allowed 27.1 26.9 29.1 28.9 34.6 34.4 36.8 36.7 36.0 35.8 35.9 35.7

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 21 OF PAGES 42 Table 13-Loading Table for High Burnup PWR Fuel - 20.9 kW/Package 1 (continued)

Minimum Initial 55 < Assembly Average Burnup :5 60 GWd/MTU Assembly Avg. Minimum Cooling Time (years) Enrichment CE WE WE B&W WE B&W wt % 23su (E) 14X14 14X14 15x15 15x15 17X17 17X17 2.1 s E < 2.3 2.3 s E < 2.5 2.5 s E < 2.7 2.7 s E < 2.9 2.9 s E < 3.1 3 1<E<33 ~ -~ f'f ct~

                                * -         *             -, ~:, b~tt                tt;I-~:tJ /J , , -

3.3 s E < 3.5 ;ttf1~ ,;-fl 37.6 41.4 * ~3,:~ 42.2 42.1 3.5 s E < 3.7 -=-~:33~8 36.1 41.3 4t'5' , 42.0 41.9 3.7 s E < 3.9~;'.;33.6 35.9 41.1 43.4 41.8 41.7 3.9SE<4J,. ~-t4 35.7 41.0 43.3~ "j'.4l8 41.6 4.1 s E <  ;/'9 J4!1 35:* .5 40. ,11 All-.6 41.5 4.3SE~5 ---.. 41.3 4.5 s E ~4,7 41.2

                                       ~

4.7SE~9 41.1 E ~ 4.,~IP'll 41.0

1. '-' means'hot al

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(i) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356 Contents - Type and Form of Material (continued)

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 22 OF PAGES 42 Table 14-Low Burnup PWR Fuel Loading Table - 22 kW/Package Max.

Min. Assembly Assembly Minimum Avg. Initial Avg. Cool Time Enrichment Burnup [Years] [wt3/4 235 U] [MWd/MTU] 10,000 1.3 4.0 15,000 1.5 4.0 20,000 1.7 4.5 25,000 1.9 5.7 30,000 2.1 7.4

                                                         " ~JR HU:

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1.

5.(b)(1)

a. CERTIFICATE NUMBER 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 23 OF PAGES 42 (ii) Undamaged and damaged PWR assemblies A combination of damaged and undamaged PWR fuel assemblies in the 37 PWR damaged fuel basket, shown in Figure 2, in a short TSC. Undamaged, low burnup fuel assemblies must meet the description for PWR fuel in 5.(b)(1)(i). _Up to four damaged fuel assemblies, ,hig!Lburnl.Jpfe,eI g~se,mqlieS:!{mgxJ1i1J.1m,asscinb)Y, ;;iver;m:ie .t:J!-!fnllp ::.,*4 5,])0.0.M'Nd/MJU),,: or fuel material that is less than, or equivalent to, one undamaged PWR fuel assembly must be placed in a damaged fuel can and must be placed in locations 4, 8, 30 and 34 in the PWR damaged fuel basket.

Undamaged, low burnup fuel may also be placed in the 4 damaged fuel locations, without the use of a damaged fuel can. Prior to irradiation, the damaged and undamaged fuel assemblies must be within the dimensions and specifications of the hybrid assemblies listed in Table 1 and meet the fuel class assembly specifications listed in Table 2. For the 33 non-damaged fqel{~af1c(\ti~kf:~ 1'.fi~tap;1aged fuel basket, the fuel must meet the class enrichment, post-ifa1~iatibn cooling time, burntfi;t~tedit loading curves, and the TSC neutron absorber she~ 0B density in Table 16. For thetp~jng profiles up to the 33, 35 and 36 assembly loading Ral.t~rn, the PWR fuel must meet the burif{_yp loading curves in Table 17. A short-loaded 33, ~?§r~~::-@__ssembly loading profile mayytilJ us~the burnup credit curve in Table 17 provided the J~fuireai~h!ocations for that profil~~n iK,J!igure 1 are left empty, at a minimun:i. _For c!""'TI§Q_~it~~,l' ~~ed fu~as~et'j~~bly ~h~f'~oes ~ot ~ontain any dama_ged fuel :9LbJg!:!J2J!F~Q..ft!fil, th rfi* c~ss~Jenm~~,~~l?£§"!-1rrad1at1~~cooling time, burn up credit loading cur:-res?~}he_J~j _ ,~.~-J.!3igr~~[.:~Dg;t~~-~. ~,$,~~!l~@?Tables_3 ~~,dnt,m~x~~t:-~sed for ~H _!ocat1onj"".O?urnQQ.creQ1 curyes*ar~tor}!

• a
• llcable tq_§Y~i"D§~not_ g~_gJ.!!1:!.g modernt0ri

                        ~~~                                                                                                              ~

Fuel assembly b("y)iup, \

                                                               ~

1

                                                                ~Zm,,_,-

1 h la ~r~ e,_enric

                                                                                                                                       ~
                                                                                                                           ' ancl'lool time requirements are provided in Tabl~§__,~})-1 f'r[.

~~hM1 ,. 11 ~ltl1~1S~P,"t~*

                                                                 ...,,Ii,;.~                      ef1'1;;l:..,,,,,,,

neutro&absorbers

                                                                                                                           ,;;..,;:~ llf;,#       r-\

and Tables. 12-14 for 71 baskets with Type~1:1neutron absorb r~s. ~w,gontainin"'@;-clamaged

                                                                                                .                                                 Qf..:J:l!gh ournug fuel, all fuel assemblies in thtf_pC must m~ \li litdditioh'al cool;tidl'erequirements in Table 18 for the assembly type that is"'l~d.,ed in the damaged fuel can.~J}~fw1o types of fuel assemblies are loaded in different damag:d fy1~I cans in a single i(~C, 'the longest additional fuel cooling time applies to all fuel _assemblie~*in{th~;}s~2 Tbi~{ad9itional cool time requirements in Table 18 7

apply to assemblies loaded m TSC oasl<ets with Type 1 or Type 2 neutron absorbers. Damaged 'and ;high burriug CE 16x 16 fuel assemblies are not authorized for shipment. The fuel assemblies consist of uranium dioxide pellets with zirconium alloy-clad fuel rods and zirconium alloy instrument and guide tubes. Empty fuel rod positions for undamaged fuel assemblies are to be filled with a solid filler rod or a solid neutron absorber rod that displaces a volume equal to or greater than the original rod. PWR fuel assemblies containing nonfuel hardware may be loaded in the TSC . 9 235 U

• Assembly average fuel enrichment is the enrichment value determined by averaging the wt% enrichment over the entire fuel region (U02) of an individual assembly, including axial blankets, if present.

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(ii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER Contents - Type and Form of Material (continued) 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 24 OF PAGES 42 Unirradiated fuel and unenriched fuel are not authorized for loading, except that unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than 6 inches. An unenriched rod may be used as a replacement rod to return a fuel assembly to an undamaged condition. Damaged '.or, high.burnug fuel located in a damaged fuel can location in the damaged fuel basket must have a minimum burnup of 5 GWd/MTU, a maximum enrichment of 4.05 wt.% 235 U, and a minimum cool time of 15 years. PWR fuel assemblies loaded in a damaged fuel can must not contain nonfuel hardware with the exception of instrument tube tie l~~m~o_n~_nt~,pui~~ Jy,be}n0?!,~ o~~i~!)a~3.~~ic~~*- and steel i_ns~rts; l'\ppli~,~ti?P ?t-m<ro~rafori

                     ~~clus1on" allows *increasing* the':max1111um Irnt1al ennchrri_ehf:!o 5 .Wt.':(cc:.:.~O.~!:tt!:.I!Q.,Qldffi.!dJ:!
                     'rELQUirement:l Undamaged PWR fuel assemblies may contain nonfuel hardware, while damaged PWR fuel assemblies shall not, with the exce~rtipn @f \peJ9llowing unirradiated nonfuel hardware:

1 9r instrument tube tie compo~e~J~;fgyLai~ ttJ15et'a~~i~~ 1 similar d~vi~e_s, and steel inserts. The nonfuel hardware may qfi}>l0,.9aea as a complete assemJ>[Y or as 1nd1v1dual components, individual nonfuel roqs~):f1'ay be full-length rods or partialilemgth rods/rodlets. Partial-length rods/rodlets are per_~)t~d in guide tubes provided guide fu©plug devises are installed. Fuel assemblies with an~s.trument tube tie rod repair shall be load@d with fuel inserts and/or top

• spacers to ensur~pro~~ cing and support of th,~,,-tu~i.s(nJgly. Fuel inserts and/or top spacers are notl11e.q" a ~

uired ~ ad~q_uately ~up13~rted: No act1v1ty requirements listed

                                       ~..
                                      ~=

HFRAs are only.i~llowe, , , fil\r:.11 i:Jsing

                                                                  '\

the extendec!Ji:ra1;fcibe E>les 6.t.:8. __ 1 '11

                                                                     ,. houser~

I~

                                                                      ;,~. i"l lfli q ,, M p     "iifi?,,.. """
                                                               \h8fdw~us(m 1 e expos&:re and cool time or cobalt-60
                                                                                                 ,j'<":
                                                                                                            ,F Hl i"/"/,

1 basket IP~,, because the top nozzle is es ar<d may~ve a maximum exposure of 4.0 GWd/MTU\:i~d mu *  :, 1nimum,*, 1 co .

• 0 1:fyQt6~years. ;l§uel assemblies loaded with

                                                                   ~ '"'ij 11 *1

• ii 1 H 1 r-'"'Y ,, * **-*~

nonfuel hardwar;\.--.::, e"rJ1us . 1 e_ ;;.,J c!,Glti ition_, al;

                                                                        ..\ 1I 'I II /j
                                                                               )1 time('re~t'.iirements I    r.            ~-~

of Table 5 (for Type 2 neutron absorbe~~)\an V1rF ~;~,t~f©r..1 ** , --:-:~5absorb@~- Fuel assemblies may contain any number of uni~d}atea_ n. ~elt2,,;;,1,1, i 1 1\"\~UJ;I. ,;placem~.Fods. Activated stainless steel fuel replacement ro<!~~re limited to !Df~tei~oas ~rlasseJJ:!J:Hy, 1 assembly per basket, and a maximum steel rod efp<1$y,re of 32.5 G'Nd/MTU. Fue~se¥'mblies with activated stainless steel rods must be cooled foleither ~- minimum of 21 Yf~rs or'the loading table minimum cool time (as adjusted for additional codl~tim~sJ9_pq.oQ_fu,~1~iiifrdware and the presence of damaged fuel in the TSC, as applicable) plus 1 year:'1hiehever* is greater. Fuel assemblies loaded with in-core instrument thimbles must meet the additional cool time requirements in Table 5 or Table 15, as appropriate, for BPRAs or GTPDs, whichever is bounding, for Westinghouse and B&W fuel types and for RCCs for CE fuel types. The additional cool time requirements for assemblies with nonfuel hardware are added to any additional cool time requirements due to damaged fuel also being loaded in the same TSC. RCCs are restricted to fuel storage locations 11, 12, 13, 18, 19, 20, 25, 26 and 27 in Figure 1.

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER One NSA is permitted to be loaded in a TSC in fuel storage locations 11, 12, 13, 18, 19, 20, 25, USA/9356/B(U)F-96 PAGE 25 OF PAGES 42 26 or 27 in Figure1. NSAs may contain source rods attached to hardware similar in configuration to guide tube plug devices (thimble plugs) and burnable absorbers, in addition to containing burnable poison rod lets and/or thimble plug rodlets. NSAs, GTPDs, and BPRAs for CE fuel types are not allowed contents. In addition, the following unirradiated, nonfuel hardware may be loaded with the fuel assemblies: stainless steel rods inserted to displace guide tube "dashpot" water, instrument tube tie components, and guide tube anchors or similar devices.

Axial power shaping rods are not allowed contents. Under-burned Westinghouse 15x15 assemblies (assemblies with a maximum enrichment greater than that dictated by the burnup credit loading curve) may be loaded provided that an RCCA is inserted in the assembly, the enrichment is equal to or less than 4.05 wt. % 235 U, and the assembly burnup is greater than or equal to 12,000 MWd/MTU. When loading under-burned fuel, the RCCAs must be full length Ag-In-Cd RCCAs comprised of stainless steel clad rods constructed with 80% Ag, 1~%t1n a11'!,d 1~/~d~ absorber pellets and having an exposure equal to or less than 200,000~~,d/rvFfU~liif:ie,,!tratk~t must include absorber sheets with an effective 10 B areal densiw:,~~@?d36 g/cm 2

• Any as\~fubljes loaded without an RCCA inserted mus~_l1\E:Jet .m~0~-~~~ug~c~cfo !D.~d)Q9J;U~~,f~~'.1~e app~d!~l~c1~em_bly,.fo~_~i!WP!.9ftLe.J3-~~i pred1t cyrves, c1nd1the;c:nJ1cc1htytreqwrement '.fqr,.R c;;gA;ms~rt!<;m ,; c1re' on)Y ciPR]Jcaple Jo',sy.~!fil!:1§

                    'dofcfeciiting 2moclera'ttir"exclu'sfo'h;::rriitiaN*Hirit~me'nt*lfp fc)".5'y.Jt.'o/o 'i35 lJ't*wfth'./io:i1tMiuQfo'rl

• ~;z .\-'/'~ >"";:-::}

is R@QAreg*'Uirement;' n~rmitted'When*cte'aiting~moderatdr*excf(JsiQh j

                                -*-1.ro*~-t:.....~
                                                   ~
                                                                                                            -~
                                                        ,;:.::::;;>~**~~t*,., . ,>*~-\.'~--,}-(-- v;:;.~'***3{~'

n**--------

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(ii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 26 OF PAGES 42 Table 16-Maximum Initial Enrichment - 37 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt% 23 su) 109 Zero (0) = C4 x Burnup (GWd/MTU) + Cs Assembly Burnup Burnup Absorber 18 :5 Burnup 30 < Burnup 50 < Burnup ID Max. Enr. (GWd/MTU) <

(g/cm~ (GWd/MTU) :5 30 (GWd/MTU) :5 50 (GWd/MTU) (wt%) 18 C4 Cs C4 Cs C4 Cs C4 Cs BW15 1.6 0.0453 1.42 0.0681 1.29 0.0750 1.03 0.0750 0.736 BW17 1.6 0.0476 1.45 0.0668 1.37 0.0712 1.17 0.0712 0.891 CE14 1.9 0.0504 1.79 0.0696 1.75 0.0751 1.60 0.0751 1.60 CE16 0.036 1.9 0.0484 ~~ 1rl~ ~Pf.Q,6J~.i 1.74 0.0758 1.52 0.0758 1.52 WE14 1.9 0.05.¥2* I '1'.M

                                                           , = ,~~(::;l>
                                                                                      . 01f72'9 < .:1'.85J{           ...--"\.

0.0794 1.75 0.0794 1.75

                                                                                                          ......,.-t:*Xtf WE15                            1.6        ,. {l l;\0l)482          1.43          0.0692        1.2~

• 0.0738 1.08 0.0738 0.767

                                                 ~  ,, 11                                                                      .*   ?

WE17 1.6,,;,~- 0.0439 1.45 0.0657 1.35 l.,,,;P-9732 1.00 0.0732 0.700

                                                  -~

BW15 1,{~11/\) <:::~$.0487

                                                         ' .............      1.31          0.0660        1..;.2~ ~Qf0J40                                      0.896 0.0740      0.614
                                                                                                                     .,g<
                                        ,, - q BW17                           il.f§l                                               0.06                                       O~OT45  ,.,~,

0.937 0.0745 0.655 CE14 ,~'l:;8 .7~ ~ 7

                                                                                      <t,,_   ,

0.07.&1 1.37 0.0781 1.37

                                                                                                                                                  !',"' .,,"9 CE16               0.030                                                                                                       0.0%-24                  1.52 0.0724       1.52 1

WE14 0.0?2~ 1.50 0.0821 1.50 WE15 O.QJ~-6 0.859 0.0746 0.575

                                                                                                                                             ~\'\-

WE17 0.0-l,10 0.968 0.0710 0.691 4--,.....,.. BW15 0{0725

                                                                                                                                     ,,;._                     0.857 0.0725      0.581 BW17 CE14 1.5 1.8                   !)9486          1.

27 6'1:~., l

                                                                                                               *\.,,;,-..._\,
                                                                                                               .64-
                                                                                                                              -   ' ~c/0724
                                                                                                                                    ~~.

0.0778 0.918 1.32 0.0724 0.0778 0.639 1.32 CE16 0.027 1.8 ~-049Jt. 0.0761 1.33 0.0761 1.33 WE14 1.8 0.05351 y .75 0.0805 1.52 0.0805 1.52 WE15 1.5 0.0465 1.33 0.0664 1.24 0.0710 0.968 0.0710 0.685 WE17 1.5 0.0447 1.31 0.0647 1.25 0.0714 0.846 0.0714 0.564

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(ii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356 Contents - Type and Form of Material (continued)

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 27 OF PAGES 42 Table 17-Maximum Initial Enrichment-WE 15x15 Assembly Damaged Fuel Configuration 20 Year Minimum Cool Time Max Initial Enrichment (wt% 23 su) 109 Zero (0) = C4 x Burnup (GWd/MTU) + Cs Number of Burnup Burnup 18 S Burnup 30 < Burnup 50 <

Absorber Assemblies Max. Enr. (GWd/MTU) (GWd/MTU) (GWd/MTU) Burnup (g/cm 2 ) (wt%) < 18 S 30 S50 GWd/MTU Cs Cs C4 Cs 36 1.6 0.0483 1.53 0.0721 1.35 0.0750 1.17 0.0750 0.851 35 0.036 1.7 0.0532 1.51 0.0722 1.45 0.0778 1.14 0.0778 1.14 33 1.7 0.0524 1.60 0.0734 1.52 0.0791 1.22 0.0791 1.22 36 1.6 o.0483 <<,,,1.4;~i> t ,,,o,,,Q.J07 1.32 0.0739 1.15 0.0739 0.811 35 0.030 1.6 1.20 0.0733 0.847 33 1.7 ,. . 1.19 0.0780 1.19 36 1.02 0.0731 0.693 35 0.027 1.13 0.0738 0.775

• 33 1.09 0.0784 R Fuel Contents 15x15 b.

I Time WE 17x17 b. Cool Time 1.09 ears ears 2.5 N/A 0.8 0.3 3.3 2.8 40 1.2 0.8 0.0 0.0 2.7 ~ # 4.5 4.2 2.9 2.7 2.2 45 3.1 0.7 0.1 3.3 0.0 0.0 0.0 2.7 N/A N/A 4.8 N/A 2.9 3.6 2.8 3.5 2.8 50 3.1 1.7 2.8 1.2 0.5 3.3 0.0 1.2 0.0 0.0 3.1 4.2 2.9 4.0 3.6 55 3.3 2.2 3.0 1.9 1.5 3.5 0.2 2.0 0.0 0.0 3.1 N/A N/A 5.0 N/A 3.3 4.6 3.0 4.9 4.1 60 3.5 3.1 3.1 2.9 2.1 3.7 1.3 2.8 0.8 0.0 3.9 0.0 0.9 0.0 0.0

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(ii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356 Contents - Type and Form of Material (continued)

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 28 OF PAGES 42 Figure 2-Damaged Fuel Basket Loading Profile (2)

(5) (6) DFC designated locations may contain a loaded DFC or an undamaged PWR fuel assembly .

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (B-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1.

5.(b)(1)

a. CERTIFICATE NUMBER 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 29 OF PAGES 42 (iii) Undamaged BWR assemblies Undamaged BWR fuel assemblies within the 87 BWR basket assembly shown in Figure 3.

Undamaged fuel is spent nuclear fuel that does not have any visible deformation other than uniform bowing that occurs in the reactor, assemblies that do not have missing rods, and assemblies with missing rods that are replaced by solid stainless steel or zirconium filler rods that displace a volume equal to or greater than the original rods and assemblies that do not contain structural defects that adversely affect radiological and/or criticality safety and/or result in unsupported fuel rod lengths in excess of 60 inches and that can be handled by normal means. BWR/2-3 assemblies are to be loaded into short TSCs, and BWR/4-6 assemblies are to be loaded into long TSCs . BWR fuel assemblies may be unchanneled, or channeled with zirconium-based alloy channels. BWR fuel assemblies with stainless steel channels are not authorized. [fJ:ie SJ-Assembly configyration* is ttfe result of criticality .coristr:aints on maximum 0enr,jchrn~n!]

                       \fVhen cr~diting nioderat~r exclusior1thi5. c~'pfiguration is not requif~Q a5.JyJJ:gc:1gc:i9iJfrnZ2 Assembly) is*permitt~d_ ii ari: inJtial enrichment\ip fo 5-wf "o/o_ 2~ 5 UJ 10                                                                                                    235

• Assembly average fuel enrichment is the enrichment value determined by averaging the U wt% enrichment over the entire fuel region (U02) of an individual assembly, including axial blankets, if present.

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(iii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER Contents - Type and Form of Material (continued) 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 30 OF PAGES 42 Table 19-BWR H brid Fuel Assembl Characteristics Geometry 2 *3 Number Min Min Max Max Number Max Max Assembly of Partial Clad Clad Pellet Active of Fuel Pitch Loading Type Length OD Thick. OD Length Rods (inch) (MTU)

Rods 1 (inch) (inch) (inch) (inch) B7_48A 48 N/A 0.7380 0.5700 0.03600 0.4900 144.0 0.1981 B7_49A 49 N/A 0.7380 0.5630 0.03200 0.4880 146.0 0.2034 B7_49B 49 N/A 0.7380 0.5630 0.03200 0.4910 150.0 0.2037 B8_59A 59 N/A 0.6400 0.4930 0.03400 0.4160 150.0 0.1828 B8_60A 60 N/A 0.6417 0.4840 0.03150 0.4110 150.0 0.1815 B8_60B 60 NIA 0.6400 0.*:HP9 0.03000 0.4140 150.0 0.1841 B8_61B 61 N/A 0.6400 ~ fr\\0.~!t3IF<<~ ,~-93000 0.4140 150.0 0.1872 B8_62A 62 B8_63A 63 N/A -1! "\~.,_@,6420 0.4840 0.0272~ 0.4195 150.0 0.1985 B8_64A 64 N/A ,~ I 0.6420 0.4840 0.02725v r;i 0.4195 150.0 0.1996 B8_64B4 64 N/A'~~ _ 0.6090 0.4576 0.02900 ~:::,~0,,3913 150.0 0.1755 Notes:

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(iii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 31 OF PAGES 42 Table 20-BWR Fuel Class Assembly Characteristics Fuel Class Characteristic 7x7 8x8 9x9 10x10 Base Fuel Type 11 SPC,GE SPC, GE SPC, GE SPC, GE, ABB Max Initial Enrichment (wt% 235 U) 4.5 4.5 4.5 4.5 59 72 9113 60 7413 48 61 9213 Number of Fuel Rods 76 9613. 14 49 62 79 63 10014 6412 80 Max Assembly Average Burnup (MWd/MTU) ,60,QPQ B6,r6oo

                                                                                                                                   §_O,OQQ         ~6,,QJ)Q Min Cool Time (years)                                    <1'    6~ 'f?{4 ~ lt (( ~HI).              4                           4               4 Min Average Enrichment (wt% 235 U) 15          h.:.,'\,,,.~

1

                                                                    ;flt) 1.3        ,_ \\&' { 1.3 A,,,-T1l 1.3             1.3 Max Weight (lb) per Storage Locatio.n:\"\"~""'                     See Note 1            See-N6!;~                        See Note 1     See Note 1 Max Decay Heat (Watts) per Fuelz,;01,ation                              253                      253             (J           253             253 11 Indicates assembly vendor/type referenced for fuel input data. Fuel acceptability for loading is not restricted to the indicated vendor/type provided that the fuel assembly meets the limits listed in Table 6.2.1-1. Table 6.2.1-2 contains vendor information by fuel rod array. Abbreviations are as follows: General Electric/Global Nuclear Fuels (GE),

Exxon/Advanced Nuclear Fuels/Siemens Power Corporation (SPC). 12 May be composed of four subchannel clusters . 13 Assemblies may contain partial-length fuel rods.

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

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (B-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

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b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 32 OF PAGES 42 Table 21-Undamaged BWR Fuel Assembly Loading Criteria Enrichment Limits for Fuel With Axial Blankets Max. Initial Enrichment16 (wt% 235 U)

Fuel Absorber17 0.027 10 B /cm 2 Absorber17 0.0225 10 B /cm 2 Absorber7 0.02 10 B /cm 2 Type 87 -Assy 82-Assy 87 -Assy 82-Assy 87 -Assy 82-Assy Basket Basket Basket Basket Basket Basket B7 48A 4.0% 4.5% 3.7% 4.5% 3.6% 4.4% B7 49A 3.8% 4.5% 3.6% 4.4% 3.5% 4.3% B7 49B 3.8% 4.5% 3.6% 4.4% 3.5% 4.2% B8 59A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3% B8 60A 3.8% 4.5% fr',9. ?PLo) Si':"' , ""- 4.4% 3.5% 4.2% B8 60B B8 61B 3.8% 3.8% 4.5%,

                                                  ~('o/;~-
                                                                   ~~- :G:\.l;D' g U1'6 q:Si
                                                                       ,!        3.61/o 3.6%
                                                                                                        £4.3%
                                                                                                       '4:;._,¢

i1c.~_?/o,,,

3.5% 3.5% 4.2% 4.2%

                                                                                                               .,. r)f' "O B8 62A               3.8%                "'"'
                                              ~,~U.5%                            3.6%                     4.3°,1; .0-                  3.5%              4.1%
                                             ~~

B8 63A 3.8% " , ~.5% 3.6% 3.4% 4.2% B8 64A 3.8% \'-'.:-"';; 4.5._o 3.5% 4.2% B8 64B 3.6% 3.3% 4.0% B9 72A 3.8% 3.4% 4.1% B9 74A 3.7% 3.4% 4.0% B9 76A 3.5% 3.3% 3.9% B9 79A 3.7% 3.3% 4.0% B9 80A 3.8% 3.5% 4.2% B10 91A 3.7% 3.5% 4.1% B10 92A 3.8% 3.5% 4.1% B10 96A 3.7% 3.4% 4.0% 4.0%

    ~----,- *?--...,..----*;-;--~-c----,-*           ..        ... 1~..,,.4,--;}l_J;_ ~~-                     ..              ..      - *------. -*-

Note_:.Wl)en c~~d1t1pg n;ioct~rator;e~c;lufJon,.the.;nax1mu1JtalJowes;I 1rnt1al..ejrnchment rs  !:i wt%_f~ 5 U}orcal1 b9 sket/absorber combina.!J..ons. 16 17 Maximum planar average. Borated aluminum neutron absorber sheet effective areal 10 B density.

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• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(iii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER Contents - Type and Form of Material (continued) 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U) F-96 PAGE 33 OF PAGES 42 Table 22-Undamaged BWR Fuel Assembly Loading Criteria Enrichment Limits for Fuel Without Axial Blankets Max. Initial Enrichment18 (wt% 235 U)

Fuel Absorber 19 0.027 10 8 /cm 2 Absorber19 0.0225 10 8 /cm 2 Absorber 19 0.02 10 8 /cm 2 Type 87-Assy 82-Assy 87 -Assy 82-Assy 87 -Assy 82-Assy Basket Basket Basket Basket Basket Basket 87 48A 3.9% 4.5% 3.7% 4.5% 3.6% 4.3% 87 49A 3.7% 4.5% 3.6% 4.3% 3.4% 4.1% 87 498 3.7% 4.5% 3.6% 4.3% 3.5% 4.2% 88 59A 3.8% 4.5% 3.7% 4.4% 3.5% 4.3% 88 60A 3.7% 3.5% 4.1% 88 608 3.7% 3.4% 4.1% 88 618 3.7% 3.5% 4.1% 88 62A 3.6% 3.4% 4.1% 88 63A 3.7% ~~.4% 3.5% 3.4% 4.1% 88 64A 3.7% 3.4% 4.1% 88 648 3.6% 3.3% 4.0% 89 72A 3.7% 3.4% 4.1% 89 74A 3.6% 3.3% 4.0% 89 76A 3.5% 3.2% 3.8% 89 79A 3.5% 3.2% 3.9% 89 BOA 3.7% 3.5% 4.1% 810 91A 3.7% 4.1% 810 92A 3.7% 4.1%

        .____ -.,-.- ---~----- _____r~~lLA_;_.,.~,~:!l1t                                       ---,---.r----.....---    _--,-,,~--,,....~-~-..

Note: __ \/Vhe.n;ye9!t.in8~mod~rat6_r,. ~~f lusion ,. tt;i;e; ma~i.m.u,m_ aljowed initialenrichrnent is_ 5 wt%.~,35 U for all basket/absorber cohibiriatiohs. 18 19 Maximum planar average. Borated aluminum neutron absorber sheet effective areal 10 8 density.

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1)(iii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356 Contents - Type and Form of Material (continued)

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 34 OF PAGES 42 Figure 3-Undamaged Fuel Basket 87 Assembly Loading Profile TSC ALIGNMENT MARK X = Designated NonFuel Location

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

• 1. a. CERTIFICATE NUMBER 5.(b)(1)(iii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356 Contents - Type and Form of Material (continued)

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 35 OF PAGES 42 Figure 4-Undamaged Fuel Basket 82 Assembly Loading Profile

                                       / / (1)                  (2) (3)     ~
                                /

(4) (5) (6) (7) (8) (9) (10) X l (11) (12) (13) (14) (15) (16) (17) (18) (19) (20: (21) :22) (23) (24) (25) (2s: (27)

• (28) (29: (30) (31) X (33) X (35) ()6) (3?: (38)

(39: (40) ( 41) (42) : 43) X (45) (46) (47: (48-) '49' (5D) (5l) (52) :s3) X (55) X (57) (58) (59: (60)

                                                                                                    '-    J (61) (62: (63) ~64) (65) (66) (67) (68~ (69)
                           \X         (70) (7*1) (72) (73) (74) (75) (76) (77)

(78: (79) ~80) (81) (82) (83) (84:

                                                                                         !/
                                '                  ~ (85) (86) (87)

TSC ALIGNMENT MARK X = Designated Nonfuel Location

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

• 1. a. CERTIFICATE NUMBER 5.(b)( 1)(iii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 36 OF PAGES 42 Table 23-Loading Table for BWR Fuel - 22kW/Package 1 Minimum Initial Assembly Average Burnup~30 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 23su E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 ~ E < 2.3 6.5 12.3 5.8 13.7 5.3 13.0 13.5 2.3 s E < 2.5 6.3 11.6 5.7 13.0 5.2 12.3 12.8 2.5 s E < 2.7 6.3 11.0 5.7 12.3 5.1 11.7 12.2 2.7 s E < 2.9 6.2 10.3 5.6 11.8 5.1 11.1 11.6 2.9 s E < 3.1 6.1 9.8 5.6 11.2 5.0 10.5 11.1 3.1 s E < 3.3 6.0 9.3 5.5 ,,,,, 10.7 5.0 10.0 10.6 3.3 s E < 3.5 6.0 8.8 ~ ~ \i5)5 1

                                                                              ~     Prl:Gro~                              4.9                  9.6    10.0 3.5 s E < 3.7       6.0
                                                          'I\

48*.4¢,," 5.4 -~9~c I{. fJJ '\,.

                                                                                                             ~

4.9 9.1 9.6

                                                    ~-"ir'o v*7.7                          .

3.7 s E < 3.9 5.9 54 9.4 ii _,..,r-'4.9 8.8 9.2

                                                ~

3.9 s E < 4.1 5.9 ~ 5.3 "~8 8.4 8.9

                                               ~

4.1 s E < 4.3 5.9~

                                                  *                                                                          .:-;t)            8.0     8.5
                                                    ~?:~                                                         ~4.§~

4.3 s E < 4.5 5.w f 4.8 7.7 8.2 4.5 s E < 4.7 5.18:,,,_, 7.5 7.9 e~ 4.7 s E < 4.9 7.2 7.6 E:?:4.9 6.9 7.4 Minimum Initial /MTU Assembly Avg. rs~ Enrichment 2::3°.ll" BWR/4-6 BWR/4-6

                                                                                                             ;:;                  ~

wt% 23su E 9x£!~ 9x9 10x10 2.1 s E < 2.3 <?:!;/' 2.3 s E < 2.5 8.9 t~Jt cate 4),

                                                                                                                         ))

15.0 15.5 2.5 s E < 2.7 8.8 14.8 14.1 14.6

                                                                                                                 ~~)
                                                                                                                      """6 .5 2.7  s E < 2.9      8.6                                       /.3              14~0;             ~         6.4                13.4    13.9 j,. ;-:;:,':,_
                                                                       ~ ')lj-r 2/:-7 2.9  s E < 3.1      8.5                                   """'; . t;,,,, )itBf4                            6.3                12.7    13.2 3.1  s E < 3.3      8.4                11.4                     7.2            12.8                        6.3                12.1    12.6 3.3 s E < 3.5       8.3                10.8                     7.1            12.2                        6.2                11.5    12.0 3.5 s E < 3.7       8.2                10.3                     7.0            11.7                        6.1                11.0    11.5 3.7sE<3.9           8.1                9.8                      6.9            11.2                       6.0                 10.6    11.0 3.9 ~ E < 4.1       8.0                9.3                      6.9            10.8                        6.0                10.1    10.6 4.1 s E < 4.3       8.0                8.9                      6.9            10.4                       6.0                  9.7    10.1 4.3 s E < 4.5       8.0                8.7                      6.8            10.0                       6.0                  9.3     9.8 4.5 s E < 4.7       7.9                8.6                      6.8             9.6                        5.9                 8.9     9.4 4.7 s E < 4.9       7.8                8.6                      6.7             9.3                        5.9                 8.6     9.1

• 1.

E:?: 4.9 7.8

                  '-' means not allowed 8.6                      6.7             9.0                        5.9                 8.3     8.8

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 5.(b)(1 )(iii) 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

Contents - Type and Form of Material (continued)

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U)F-96 PAGE 37 OF PAGES 42 Table 23-Loading Table for BWR Fuel - 22kW/Package 1 {continued}

Minimum Initial 35 < Assembly Average Burnup::;; 40 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 23su E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1::;; E < 2.3 2.3::;; E < 2.5 2.5::;; E < 2.7 14.6 16.9 12.2 18.0 10.0 17.3 17.7 2.7 s E < 2.9 13.3 15.8 10.7 17.0 8.7 16.3 16.7 2.9 s E < 3.1 13.1 14.9 10.5 16.0 8.5 15.4 15.8 3.1 s E < 3.3 12.9 14.1 J.~.3!Fi .""' 15.2 8.4 14.6 15.0 3.3 s E < 3.5 12.6 13.9 ~~io.1

                                                                 \\     ".

l : (G4f~,,

                                                                                                \\: (

8.3 13.8 14.3 3.5 s E < 3.7 12.5 (11~.'v,,-,,,

                                                     "'I; .* .,.Ji 10.0         13.8 4          tjJ    8.2            13.2       13.6 3.7sE<3.9                                                                                    'j, ~)a.a 12.4          "~)13.6                      9.9         13.3               q' h-'\,,

12.6 13.0 3.9 s E < 4.1 12.2 "t,.'r/13.5 9.8 12.7 12.1 12.5 3/4'.81~

                                                                                                                    ,:5!:

4.1 s E < 4.3 12.0it}J 9.7 12.3 7.9 A 12.0 11.9 4.3 s E < 4.5 11.S.t, 7.9

                                                                                                                         &        12.0       11.5 4.5sE<4.7             11$9.::.                                                                      7.8            11.9       11.2 4.7sE<4.9             1M3"'                                                                         7.8            11.8       11.2
                                        *-:.1:;

E ::C:4.9 H.8, 7.7 ~1 11.8 11.1 Minimum Initial 5 GW'tf/MTU Assembly Avg. ars:'~ Enrichment R/2'";:3';} BWR/4-6 BWR/4-6

                                                                                                                         ~

wt% 23su E it*~~ 9x9'... 9x9 10x10 2.1 s E < 2.3 2.3 s E < 2.5 iJ

                                                                                              ~;f41;},
                                                                                               - .;J 2.5 s E < 2.7 2.7sE<2.9             22.3                                ,    _,119~        22~9-                                 22.7       21.5 2.9 s E < 3.1         19.7                                  iv.~;?. {:{261"                        14.8            20.0       19.4 3.1 s E < 3.3         18.9                  20.5                15.4         19.1                  12.3            18.8       18.2 3.3 s E < 3.5         18.7                  20.2                15.2         18.8                  11.9            18.6       17.4 3.5 s E < 3.7         18.5                  20.0                15.0         18.7                  11.7            18.3       17.2 3.7::;; E < 3.9       18.2                  19.9                14.7         18.5                  11.5            18.0       17.1 3.9sE<4.1             18.1                  19.6                14.6         18.2                  11.4            17.9       16.9 4.1 s E < 4.3         17.8                  19.5                14.3         18.1                  11.3            17.7       16.7 4.3 s E < 4.5         17.8                  19.4                14.3         18.0                  11.2            17.7       16.5 4.5 s E < 4.7         17.6                  19.2                14.1         17.8                  11.1            17.5       16.5 4.7sE<4.9             17.4                  19.0                14.0         17.8                  11.0            17.4       16.3

• E ::c:4.9

1. '-' means not allowed 17.3 18.9 13.8 17.7 10.9 17.3 16.2

NRG FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE

• 1. a. CERTIFICATE NUMBER 9356 FOR RADIOACTIVE MATERIAL PACKAGES

b. REVISION NUMBER 3

c. DOCKET NUMBER 71-9356

d. PACKAGE IDENTIFICATION NUMBER USA/9356/B(U) F-96 PAGE 38 OF PAGES 42 Minimum Initial 45 < Assembly Average Burnup ~ 50 GWd/MTU Assembly Avg. Minimum Coolin~ Time (vears)

Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235 U(El 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 s E < 2.3 - - - - - - - 2.3 s E < 2.5 - - - - - - - 2.5 s E < 2.7 - - - - - - - 2.7 s E < 2.9 - - - - - - - 2.9sE<3.1 29.6 31.5 27.3 30.2 24.9 30.2 29.0 3.1 s E < 3.3 27.8 29.6 24.7 27.9 22.2 27.6 26.3 3.3 s E < 3.5 27.6 29.3 23.6 27.7 19.6 27.4 26.1 3.5 s E < 3.7 27.4 29.0 23.2 27.4 19.0 27.1 25.9 3.7 s E < 3.9 27.2 28.9 23.0 27.3 18.7 26.9 25.6 3.9sE<4.1 26.9 28.6 22.8 27.0 18.5 26.7 25.5 4.1 s E < 4.3 26.8 28.6 22.6 27.0 18.4 26.5 25.2 4.3 s E < 4.5 26.6 28.3 22.3 26.8 18.2 26.5 25.1 4.5 s E < 4.7 26.4 28.1 22.3 26.6 17.9 26.3 25.0 4.7 s E < 4.9 26.2 28.0 22.1 26.4 17.9 26.1 24.8 E 24.9 26.0 27.8 22.0 26.4 17.8 25.9 24.7 Minimum Initial 50 < Assembly Average Burnup ~ 55 GWd/MTU Assembly Avg. Minimum Cooling Time (years) Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt % z3su (E) 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 s E < 2.3 - - - - - - - 2.3 s E < 2.5 - - - - - - - 2.5 s E < 2.7 - - - - - - - 2.7 s E < 2.9 - - - - - - - 2.9sE<3.1 - - - - - - - 3.1 sE<3.3 36.4 38.4 34.1 37.2 31.8 37.2 35.9 3.3 s E < 3.5 34.0 35.8 31.7 34.6 29.2 34.6 33.4 3.5 s E < 3.7 33.3 35.0 29.1 33.4 26.6 33.1 31.8 3.7 s E < 3.9 33.1 34.8 28.8 33.3 24.3 32.8 31.4 3.9sE<4.1 32.9 34.6 28.6 33.1 24.0 32.7 31.4 4.1 s E < 4.3 32.7 34.5 28.5 32.9 23.9 32.5 31.1 4.3 s E < 4.5 32.5 34.3 28.2 32.7 23.6 32.4 30.9 4.5 s E < 4.7 32.5 34.2 28.0 32.6 23.5 32.2 30.8 4.7sE<4.9 32.3 34.0 27.8 32.4 23.3 32.0 30.6

• 1.

E 24.9 32.1

                    '-' means not allowed 33.9          27.6         32.4          23.2             31.8     30.5

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

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 39 OF 42 if able 2~Ioading Table for BWR Fuel..:. 20.9kW/Package (continued)

Minimum Initial 55 < Assembly Average Burnup ~ 60 GWd/MTU Assembly Avg. Minimum Coolin Time ears Enrichment BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/2-3 BWR/4-6 BWR/4-6 wt% 235 U E 7x7 7x7 8x8 8x8 9x9 9x9 10x10 2.1 s E < 2.3 2.3 s E < 2.5 2.5 s E < 2.7 2.7 s E < 2.9 2.9 s E < 3.1 3.1 s E < 3.3 3.3 s E < 3.5 42.3 44.6 40.0 43.4 38.2 43.5 42.3 3.5 s E < 3.7 39.8 42.3 37.7 41.0 35.7 41.1 39.8 3.7sE<3.9 38.3 40.0 35.3 38.7 33.3 38.7 37.5 3.9 s E < 4.1 38.1 40.2 33.7 38.3 30.9 38.0 36.5 4.1 s E < 4.3 37.8 40.0 33.6 38.2 29.0 37.8 36.4

• 1.

4.3 s E < 4.5 4.5 s E < 4.7 4.7 s E < 4.9 E~4.9 37.8 37.7 37.6 37.4

                    '-' means not ~lLgwed 39.8 39.6 39.5 39.4 33.4 33.2 33.1 32.9 38.2 38.0 37.9 37.8 28.9 28.7 28.4 28.4 37.8 37.7 37.5 37.4 36.4 36.2 36.0 35.8 0            *-.-,.*                                             ,;
                             *             ~                 11    A                        .        -~             _f~

iTable 25-t~w Burn up BWRij - :2-~

                           '~ '-*-*--***      * ~ .
                                                  .i:J
                                                                     **~      '(jJ/fI ' . adin§S:iTable
                                                                                       *-i
                                                                                           ~  ,i,,;_ti/fl         ~
                                                                                                                .A, W/Package Max.

Assembly Min. Assembly Minimum Avg. Avg. Initial Cool Burnup Enrichment Time [MWd/MTU] [wt% 235 U] [Years] 10,000 1.3 6.3 15,000 1.5 8.6 20,000 1.7 10.3 25,000 1.9 11.9 30,000 2.1 13.7

NRC FORM 618 U.S. NUCLEAR REGULATORY COMMISSION (8-2000) 10 CFR 71 CERTIFICATE OF COMPLIANCE iii-----------..---F_O_R_R_A_D_I_O_A,_C_T_IV_E_M_A_T_E_R_IA__,LP"'"P_A_C_K_A_G_E_S_ _ _ _ __,__ _ _ _ _~ 11

• 1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U) F-96 40 OF 42 5.(b)(1 )(iii) Contents - Type and Form of Material (continued)

(iv) Greater Than Class C Waste GTCC waste consisting of solid, irradiated, and contaminated hardware, provided the quantity of fissile material does not exceed a Type A quantity and does not exceed the mass limits of 10 CFR 71.15, within a GTCC waste basket liner transported in a GTCC TSC with a welded closure lid. The specific Curie content source of the GTCC waste shall be limited to a maximum specific activity of 2.7 Ci 6 °Co/lb averaged over the GTCC waste, with a maximum localized peak specific activity of 16.1 Ci 6 °Co/lb and a total 6 °Co activity of 85,760 Ci at transport. The maximum allowed weight of this waste is 55,000 lbs. 5.(b)(2) Maximum quantity of material per package (i) (ii) (iii) (iv) 5.(c) Criticality Safety Index Undamaged PWR and BWR Fuel 0.00 Damaged PWR Fuel 100.00

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

(a) The package must be prepared for shipment and operated in accordance with the Operating Procedures in Chapter 7 of the application, as supplemented. (b) Each packaging must be acceptance tested and maintained in accordance with the Acceptance Tests and Maintenance Program in Chapter 8 of the application, as supplemented, except that the minimum component thicknesses for the mockup in Section 8.1.6.1 and the minimum shielding effectiveness configuration for calculating the dose rates used as acceptance criteria

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

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U) F-96 41 OF 42 for the tests in Sections 8.1.6.3 and 8.2.3 are defined by the component dimensions and tolerances in the drawings listed in Condition 5.(a)(3).

(c) For TSCs to be shipped under the moderator exclusion option of this certificate, only TSCs that are within their initial term for storage or are new and haven't been loaded and placed into storage are authorized for use under moderator exclusion. (d) For TSCs to be shipped under the moderator exclusion option of this certificate, the TSC confinement boundary shall have been leak tested in accordance with MAGNASTOR FSAR, Section 10.1.3 leakage test requirements.

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

8. 9. 10. 11. 12.

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

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

1. a. CERTIFICATE NUMBER b. REVISION NUMBER c. DOCKET NUMBER d. PACKAGE IDENTIFICATION NUMBER PAGE PAGES 9356 3 71-9356 USA/9356/B(U)F-96 42 OF 42

14. Expiration date: April 30, 2024 REFERENCES NAC International, Inc., Application dated Qece&mber..31,:20291 FOR THE U.S. NUCLEAR REGULATORY COMMISSION

• Date: _ _ _ _ _ __

Enclosure 3 to ED20210107 Page I of2 Enclosure 3 List of SAR Changes

• No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion RAI Response Submittal MAGNATRAN SAR, Revision 21A

Enclosure 3 to ED20210107 Page 2 of2

• List of Changes for the MAGNATRAN SAR, Revision 21A

          .Chapter/Page/

Figure/fable Description of Change Note: The List of Effective Pages and the Chapter Table of Contents, List of Figures and List of Tables have been revised accordingly to reflect the list of changes detailed below, if needed. Chanter 1 Pages 1.3-8 thru 1.3-9 Added text to the paragraph at the bottom of page 1.3-8 thru the top of page 1.3-9 where indicated. Pages 1.3-10 thru 1.3-11 Text flow changes. Chanter2 No Changes Chanter 3 No Changes Chanter 4 Page 4-1 Added paragraph to the end of Section 4 where indicated. Chanter5

• No Changes Chanter6 Page 6.1.1-1 Added text in the middle of the fourth paragraph of Section 6.1.1 where indicated.

Page 6.1.1-2 Text flow changes. Chanter7 Page 7.2-5 Added "Caution" text following Item 21 in Section 7.2-2 where indicated. Page 7.2-6 Text flow changes. Chanter8 No Changes

Enclosure 4 to ED20210107 Page 1 of 1 Enclosure 4

• LOEP and SAR Page Changes No. 71-9356 for the MAGNATRAN Cask Moderator Exclusion RAI Response Submittal MAGNATRAN SAR, Revision 21A

MAGNATRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21 A

• Chapter 1 Page Page List of Effective Pages 1-i thru I-ii ................................. Revision 0 I-iii ........................................ Revision 20C Page 2.6.2-1 thru 2.6.2-10 .................. Revision 0 Page 2.6.3-1 ........................................ Revision 0 Page 2.6.4-1 thru 2.6.4-2 .................... Revision 0 Page 1-1 .............................................. Revision 0 Page 2.6.5-1 thru 2.6.5-5 .................... Revision 0 Page 1.1-1 thru 1.1-8 ........................... Revision 0 Page 2.6.6-1 ........................................ Revision 0 Page 1.2-1 thru 1.2-5 ........................... Revision 0 Page 2.6. 7-1 ........................................ Revision 0 Page 1.3-1 thru 1.3-7 ........................... Revision 0 Page 2.6.7.1-1 thru 2.6.7.1-9 .............. Revision 0 Page 1.3-8 thru 1.3-11.. .................. Revision 2 IA Page 2.6.7.2-1 thru 2.6.7.2-5 :............. Revision 0 Page 1.3-12 .................................... Revision 20C Page 2.6. 7.3-1 thru 2.6. 7 .3-9 .............. Revision 0 Page 1.3-13 thru 1.3-20 ....................... Revision 0 Page 2.6.7.4-1 ..................................... Revision 0 Page 1.3-21 thru 1.3-32 .................. Revision 20C Page 2.6.7.5-1 thru 2.6.7.5-34 ............ Revision 0 Page 1.3-33 thru 1.3-37 ....................... Revision 0 Page 2.6.7.6-1 thru 2.6.7.6-5 .............. Revision 0 Page 1.3-3 8 .................................... Revision 20C Page 2.6.7.7-1 thru 2.6.7.7-19 ............ Revision 0 Page 1.3-39 thru 1.3-43 ....................... Revision 0 Page 2.6.8-1 ........................................ Revision 0 Page 1.3-44 .................................... Revision 20C Page 2.6.9-1 ........................................ Revision 0 Page 1.3-45 thru 1.3-46 ....................... Revision 0 Page 2.6.10-1 ...................................... Revision 0 Page 1.3-47 .................................... Revision 20C Page 2.6.11-1 thru 2.6.11-5 ................ Revision 0 Page 1.3-48 ......................................... Revision 0 Page 2.6.12-1 thru 2.6.12-2 ................ Revision 0 Page 1.3-49 thru 1.3-50 .................. Revision 20C Page 2.6.12.1-1 ................................... Revision 0 Page 1.3-51 ......................................... Revision 0 Page 2.6.12.2-1 thru 2.6.12.2-7 .......... Revision 0 Page 1.4-1 thru 1.4-4 ........................... Revision 0 Page 2.6.12-3-1 thru 2.6.12-3-4 .......... Revision 0 Page 2.6.12.4-1 thru 2.6.12.4-12 ........ Revision 0 36 drawings (see Section 1.4.3) Page 2.6.12.5-1 thru 2.6.12.5-6 .......... Revision 0 Page 2.6.12.6-1 thru 2.6.12.6-4 .......... Revision 0 Chapter 2 Page 2.6.12-7-1 thru 2.6.12-7-3 .......... Revision 0 Page 2-i ............................................... Revision 0 Page 2.6.12.8-1 thru 2.6.12.8-10 ........ Revision 0 Page 2-ii ......................................... Revision 20C Page 2.6.12.9-1 thru 2.6.12.9-6 .......... Revision 0 Page 2-iii thru 2-xxiii .......................... Revision 0 Page 2.6.12.10-1 ................................. Revision 0 Page 2-xxiv .................................... Revision 20C Page 2.6.12.11-1 ................................. Revision 0 Page 2-1 .............................................. Revision 0 Page 2.6.12.12-1 thru Page 2.1-1 ........................................... Revision 0 2.6.12.12-3 ............................. Revision 0 Page 2.1.1-1 thru 2.1.1-6 ..................... Revision 0 Page 2.6.12.13-1 thru Page 2.1.2-1 thru 2.1.2-11 ................... Revision 0 2.6.12.13-3 ............................. Revision 0 Page 2.1.3-1 thru 2.1.3-2 ..................... Revision 0 Page 2.6.12-14-1 thru Page 2.1.4-1 thru 2.1.4-5 ..................... Revision 0 2.6.12.14-2 ............................. Revision 0 Page 2.2-1 ........................................... Revision 0 Page 2.6.13-1 thru 2.6.13-2 ................ Revision 0 Page 2.2.1-1 thru 2.2.1-17 ................... Revision 0 Page 2.6.13.1-1 thru 2.6.13.1-2 .......... Revision 0 Page 2.2.2-1 thru 2.2.2-9 ..................... Revision 0 Page 2.6.13.2-1 thru 2.6.13.2-13 ........ Revision 0 Page 2.2.3-1 ........................................ Revision 0 Page 2.6.13.3-1 thru 2.6.13.3-3 .......... Revision O Page 2.3-1 ........................................... Revision 0 Page 2.6.13.4-1 thru 2.6.13.4-14 ........ Revision 0 Page 2.3 .1-1 ........................................ Revision 0 Page 2.6.13.5-1 thru 2.6.13.5-2 .......... Revision 0 Page 2.3 .2-1 ........................................ Revision 0 Page 2.6.13.6-1 thru 2.6.13.6-5 .......... Revision 0 Page 2.4-1 thru 2.4-2 ........................... Revision 0 Page 2.6.13.7-1 thru 2.6.13.7-4 .......... Revision 0 Page 2.5-1 ........................................... Revision 0 Page 2.6.14-1 ...................................... Revision 0 Page 2.5.1-1 thru 2.5.1-12 ................... Revision 0 Page 2.6.14.1-1 thru 2.6.14.1-8 .......... Revision 0 Page 2.5.2-1 thru 2.5.2-18 ................... Revision 0 Page 2.6.14.2-1 thru 2.6.14.2-2 .......... Revision 0 Page 2.6-1 ........................................... Revision 0 Page 2.6.14.3-1 thru 2.6.14.3-14 ........ Revision 0 Page 2.6.1-1 thru 2.6.1-15 ................... Revision 0 Page 2.6.14.4-1 thru 2.6.14.4-3 .......... Revision 0 Page 1 of 4

MAGNA TRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21A List of Effective Pages (cont'd) Page 2.6.15-1 thru 2.6.15-2 ................. Revision 0 Page 2.7.12.2-1 thru 2.7.12.2-6 .......... Revision 0 Page 2.6.15.1-1 thru 2.6.15.1-2 ........... Revision 0 Page 2.7.12.3-1 thru 2.7.12.3-4 .......... Revision 0 Page 2.6.15.2-1 thru 2.6.15.2-13 ......... Revision 0 Page 2.7.12.4-1 thru 2.7.12.4-6 .......... Revision 0 Page 2.6.15.3-1 thru 2.6.15.3-2 ........... Revision 0 Page 2.7.12.5-1 ................................... Revision 0 Page 2.6.15.4-1 thru 2.6.15.4-15 ......... Revision 0 Page 2.7.12.6-1 thru 2.7.12.6-2 .......... Revision 0 Page 2.6.15.5-1 ................................... Revision 0 Page 2.7.13-1 ...................................... Revision 0 Page 2.6.15.6-1 thru 2.6.15.6-5 ........... Revision 0 Page 2.7.13.1-1 thru 2.7.13.1-7 .......... Revision 0 Page 2.6.15.7-1 thru 2.6.15.7-3 ........... Revision 0 Page 2.7.13.2-1 thru 2.7.13.2-21 ........ Revision 0 Page 2.6.16-1 ...................................... Revision 0 Page 2.7.14-1 thru 2.7.14-8 ................ Revision 0 Page 2.6.16.1-1 ................................... Revision 0 Page 2. 7 .15-1 ...................................... Revision 0 Page 2.6.16.2-1 thru 2.6.16.2-6 ........... Revision 0 Page 2.7.16-1 thru 2.7.16-4 ................ Revision 0 Page 2.6.16.3-1 thru 2.6.16.3-3 ........... Revision 0 Page 2.8-1 ........................................... Revision 0 Page 2.6.16.4-1 thru 2.6.16.4-7 ........... Revision 0 Page 2.9-1 ........................................... Revision 0 Page 2.6.16.5-1 thru 2.6.16.5-4 ........... Revision 0 Page 2.10-1 ......................................... Revision 0 Page 2.6.16.6-1 thru 2.6.16.6-3 ........... Revision 0 Page 2.11-1 .................................... Revision 20C Page 2.6.16.7-1 thru 2.6.16.7-4 ........... Revision 0 Page 2.11.1-1 ................................. Revision 20C Page 2.6.16.8-1 thru 2.6.16.8-6 ........... Revision 0 Page 2.11.1-2 thru 2.11.1-12 .............. Revision 0 Page 2.6.16.9-1 thru 2.6.16.9-6 ........... Revision 0 Page 2.11.2-1 thru 2.11.2-4 ................ Revision 0 Page 2.6.16.10-1 ................................. Revision 0 Page 2.11.3-1 thru 2.11.3-6 ................ Revision 0 Page 2.6.16.11-1 ................................. Revision 0 Page 2.11.4-1 ...................................... Revision 0 Page 2.6.16.12-1 ................................. Revision 0 Page 2.11.4-2 ................................. Revision 20C Page 2.6.16.13-1 thru 2.6.16.13-3 ....... Revision 0 Page 2.11.4-3 ...................................... Revision 0 Page 2.6.17-1 ...................................... Revision 0 Page 2.11.5-1 ...................................... Revision 0 Page 2.6.17.1-1 .................................. Revision 0 Page 2.11.6-1 thru 2.11.6-2 ........... Revision 20C Page 2.6.17.2-1 thru 2.6.17.2-4 ........... Revision 0 Page 2.12-1 ......................................... Revision 0 Page 2.6.17.3-1 thru 2.6.17.3-7 ........... Revision 0 Page 2.12.1-1 thru 2.12.1-4 ................ Revision 0 Page 2.7-1 ........................................... Revision 0 Page 2.12.1-5 ................................. Revision 20C Page 2.7.1-1 thru 2.7.1-3 ..................... Revision 0 Page 2.12.2-6 thru 2.12.2-82 .............. Revision 0 Page 2.7.1.1-1 thru 2.7.1.1-7 ............... Revision 0 Page 2.7.1.2-1 thru 2.7.1.2-4 ............... Revision 0 Chapter 3 Page 2.7.1.3-1 thru 2.7.1.3-7 ............... Revision 0 Page 3-i thru 3-iii ................................ Revision 0 Page 2. 7 .1.4-1 ..................................... Revision 0 Page 3-1 .............................................. Revision 0 Page 2.7.1.5-1 thru 2.7.1.5-3 ............... Revision 0 Page 3.1-1 thru 3.1-3 .......................... Revision 0 Page 2.7.1.6-1 ..................................... Revision 0 Page 3.2-1 thru 3.2-8 .......................... Revision 0 Page 2.7.1.7-1 thru 2.7.1.7-6 ............... Revision 0 Page 3.3-1 thru 3.3-2 .......................... Revision 0 Page 2.7.1.8-1 thru 2.7.1.8-6 ............... Revision 0 Page 3.4-1 thru 3.4-35 ........................ Revision 0 Page 2.7.2-1 ........................................ Revision 0 Page 3 .5-1 thru 3 .5-17 ........................ Revision 0 Page 2.7.3-1 thru 2.7.3-11 ................... Revision 0 Page 3.6-1 thru 3.6-2 .......................... Revision 0 Page 2.7.4-1 thru 2.7.4-6 ..................... Revision 0 Page 2.7.5-1 ........................................ Revision 0 Chapter 4 Page 2.7.6-1 ........................................ Revision 0 Page 4-i thru 4-ii ................................. Revision 0 Page 2.7.7-1 thru 2.7.7-3 ..................... Revision 0 Page 4-1 ......................................... Revision 21A Page 2.7.8-1 thru 2.7.8-32 ................... Revision 0 Page 4.1-1 thru 4.1-6 .......................... Revision 0 Page 2.7.9-1 thru 2.7.9-15 ................... Revision 0 Page 4.2-1 thru 4.2-2 .......................... Revision 0 Page 2.7.10-1 thru 2.7.10-12 ............... Revision 0 Page 4.3-1 thru 4.3-2 .......................... Revision 0 Page 2.7.11-1 thru 2.7.11-14 ............... Revision 0 Page 4.4-1 thru 4.4-3 .......................... Revision 0 Page 2.7.12-1 thru 2.7.12-2 ................. Revision 0 Page 4.5-1 thru 4.5-19 ........................ Revision 0 Page 2.7.12.1-1 thru 2.7.12.1-2 ........... Revision 0 Page 2 of 4

MAGNATRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21 A

• Chapter 5 List of Effective Pages (cont'd)

Page 5-i thru 5-viii .............................. Revision 0 Page 5-ix ........................................ Revision 20C Chapter 6 Page 6-i .......................................... Revision 20C Page 6-ii thru 6-iii.. ............................. Revision 0 Page 5-x .............................................. Revision 0 Page 6-iv thru 6-v .......................... Revision 20C Page 5-1 ......................................... Revision 20C Page 6-vii thru 6-viii ........................... Revision 0 Page 5-2 thru 5-3 ................................. Revision 0 Page 6-ix ........................................ Revision 20C Page 5.1-1 thru 5.1-5 ........................... Revision 0 Page 6-1 .............................................. Revision 0 Page 5.1-6 ...................................... Revision 20C Page 6.1.1-1 thru 6.1.1-2 ............... Revision 21A Page 5.1-7 thru 5.1-9 ........................... Revision 0 Page 6.1.1-3 thru 6.1.1-5 .................... Revision 0 Page 5 .1-10 .................................... Revision 20C Page 6.1.2-1 thru 6.1.2-6 ............... Revision 20C Page 5.1-11 thru 5.1-12 ....................... Revision 0 Page 6.1.3-1 ........................................ Revision 0 Page 5.2-1 thru 5.2-13 ......................... Revision 0 Page 6.2.1-1 thru 6.2.1-5 .................... Revision 0 Page 5.3-1 thru 5.3-3 ........................... Revision 0 Page 6.2.2-1 ........................................ Revision 0 Page 5.3-4 ...................................... Revision 20C Page 6.2.3-1 ........................................ Revision 0 Page 5.3-5 thru 5.3-6 ........................... Revision 0 Page 6.3.1-1 thru 6.3.1-7 .................... Revision 0 Page 5.4-1 thru 5.4-5 ........................... Revision 0 Page 6.3.2-1 thru 6.3.2-3 .................... Revision 0 Page 5.5-1 thru 5.5-14 ......................... Revision 0 Page 6.3 .3-1 ........................................ Revision 0 Page 5.6-1 thru 5.6-25 ......................... Revision 0 Page 6.3 .4-1 ........................................ Revision 0 Page 5.6-26 .................................... Revision 20C Page 6.4-1 ........................................... Revision 0 Page 5.6-27 ......................................... Revision 0 Page 6.4.1-1 thru 6.4.1-4 ............... Revision 20C Page 5.7-1 thru 5.7-3 ........................... Revision 0 Page 6.4.2-1 thru 6.4.2-2 .................... Revision 0 Page 5.8-1 ........................................... Revision 0 Page 6.5.1-1 ........................................ Revision 0 Page 5.8.1-1 thru 5.8.1-4 ..................... Revision 0 Page 6.5.2-1 thru 6.5.2-2 .................... Revision 0 Page 5.8.2-1 thru 5.8.2-2 ..................... Revision 0 Page 6.6.1-1 ........................................ Revision 0 Page 5 .8.2-3 ................................... Revision 20C Page 6.6.2-1 thru 6.6.2-2 .................... Revision 0 Page 5.8.2-4 thru 5.8.2-11 ................... Revision 0 Page 6.6.3-1 thru 6.6.3-4 .................... Revision 0 Page 5.8.3-1 thru 5.8.3-25 ................... Revision 0 Page 6.7-1 ........................................... Revision 0 Page 5.8.4-1 ........................................ Revision 0 Page 6.8.1-1 thru 6.8.1-70 .................. Revision 0

 ' Page 5.8.4-2 thru 5.8.4-3 ................ Revision 20C                  Page 6.8.2-1 thru 6.8.2-3 .................... Revision 0 Page 5.8.4-4 thru 5.8.4-9 ..................... Revision 0               Page 6.9- 1 thru 6.9-3 .......................... Revision 0 Page 5.8.4-10 ................................. Revision 20C             Page 6.10.1-1 thru 6.10.1-112 ............ Revision 0 Page 5.8.4-11 thru 5.8.4-14 ................. Revision 0                 Page 6. 10.2-1 thru 6. 10.2-18 .............. Revision 0 Page 5. 8.4-15 ................................. Revision 20C            Page 6.10.3-1 thru 6.10.3-86 .............. Revision 0 Page 5.8.4-16 ...................................... Revision 0          Page 6.10.4-1 thru 6.10.4-3 ........... Revision 20C Page 5.8.4-17 ................................. Revision 20C Page 5.8.4-18 thru 5.8.4-19 ................. Revision 0                 Chapter 7 Page 5.8.4-20 thru 5.8.4-25 ............ Revision 20C                    Page 7-i ............................................... Revision 0 Page 5.8.5-1 thru 5.8.5-11 ................... Revision 0                Page 7-1 thru 7-2 ................................ Revision 0 Page 5.8.6-1 thru 5.8.6-7 ..................... Revision 0               Page 7.1-1 thru 7.1-5 .......................... Revision 0 Page 5.8.7-1 thru 5.8.7-96 ................... Revision 0                Page 7 .1-6 thru 7. l-16 ................... Revision 20C Page 5.8.8-1 ........................................ Revision 0         Page 7 .2-1 thru 7 .2-4 .......................... Revision 0 Page 5.8.9-1 thru 5.8.9-5 ..................... Revision 0               Page 7.2-5 thru 7.2-6 ..................... Revision 21A Page 5.8.10-1 thru 5.8.10-75 ............... Revision 0                  Page 7.3-1 ........................................... Revision 0 Page 5.8.11-1 thru 5.8.11-18 ............... Revision 0                  Page 7.4-1 ........................................... Revision 0 Page 5.8.12-1 thru 5.8.12-20 ............... Revision 0 Page 5.8.13-1 thru 5.8.13-5 ................. Revision 0 Page 5.8.14-1 thru 5.8.14-5 ................. Revision 0 Page 3 of 4

MAGNA TRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21 A Chapter 8 List of Effective Pages Page 8-i thru 8-ii ................................. Revision 0 Page 8-1 .............................................. Revision 0 Page 8.1-1 thru 8.1-29 ......................... Revision 0 Page 8.2-1 thru 8.2-8 ........................... Revision 0 Page 8.3-1 thru 8.3-2 ........................... Revision 0 Page 4 of 4

                       "NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR                                                           April 2019 Docket No. 71-9356                                                                     Revision 0 Port and Coverplate The MAGNA TRAN has a lid port that is .closed by a bolted Type 304/304L stainless steel coverplate with dual O-rings. The four coverplate bolts are SA-193, Grade B6, Type 410 stainless steel, socket head cap screws. The bolts are countersunk flush with the top of the coverplate. The basic configuration of the lid port and coverplate includes a 5.32-inch-diameter opening to recess the coverplate and for access to the port opening and the quick-disconnect installed there. Two concentric O-rings are located on the bottom face of the coverplate, an inner metal O-ring and an outer EPDM O-ring. The inner O-ring provides the containment boundary seal for the lid port. The outer O-ring and a test port located between the two O-rings provide the means to leak test the containment boundary seal. After the leak test is completed, the seal test port is closed by a threaded plug fitted with a metal boss seal.

Lifting Trunnions and Rotation Trunnions The two lifting trunnions on the MAGNATRAN are Type 17-4 PH stainless steel, which are bolted into recesses in the top forging at diametrically opposite locations around the cask circumference. Each lifting trunnion is bolted to the top forging by nine SB63 7, GR N07718, nickel alloy bolts. The basic diameter of the lifting trunnions is 6.6 inches and the load-bearing width is 3.75 inches. A retainer, or flange, on the outer end of each lifting trunnion acts as a safety stop to ensure that proper engagement with the lift yoke is maintained. The MAGNATRAN lifting trunnions are designed and load tested in accordance with the requirements of ANSI N14.6 and 10 CFR 71.45(a). Two rotation trunnions, located 17.65 inches above the bottom of the cask and circumferentially in line with the two lifting trunnions, are offset approximately 5 inches from the cask centerline to ensure that the cask rotates in the proper direction. The rotation trunnions also serve as the cask tiedown restraint in the aft longitudinal direction. Each rotation trunnion support is Type XM-19 stainless steel housing a 17-4 PH pin and is welded to the outer shell and bottom outer forging. The neutron shield assemblies are shaped to accommodate the location and operation of the rotation trunnions. Transport Impact Limiters The MAGNATRAN transport cask is equipped with removable impact limiters that are bolted over each end of the cask to ensure that the design impact loads for the cask are not exceeded for any of the normal conditions of transport and hypothetical accident conditions defined in 10 CFR 71

• NAC International 1.3-7

                     "NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR                                                             July 2021 Docket No. 71-9356                                                                    Revision 21 A The lower impact limiter is bolted to the cask bottom plate by 16 equally spaced retaining rods and nuts. The upper impact limiter is similarly bolted to the cask lid.

Transportable Storage Canisters The transportable storage canister (TSC) and the closure lid are dual-certified Type 304/304L stainless steel. The canister holds the fuel basket or GTCC waste basket liner assembly and contains the contents. A schematic of a typical TSC with a fuel basket is shown in Figure 1.3-1. There are two different length TSCs (short and long) to accommodate the various PWR and BWR fuel assembly lengths, damaged fuel (only short), and GTCC waste (only short). The TSC body (shell and bottom) and the closure lid provide confinement, shielding and lifting capability for the TSC. The loaded TSCs include a solid stainless steel closure lid or stainless steel/carbon steel closure lid assembly with a closure ring and dual port covers to provide a dual-welded closure system. The closure lid is positioned inside the TSC on the lifting lugs above the basket assembly following fuel loading, or on the top of the GTCC waste basket liner following GTCC waste loading. After the closure lid is placed on the TSC, the TSC is moved to a workstation and the closure lid is welded to the TSC. The vent and drain ports are penetrations through the lid, which provide access for auxiliary systems to drain, dry and helium backfill the TSC. Following completion of backfilling, the dual port covers are installed and welded in each port. Removable lifting fixtures installed in the closure lid are used to lift and lower the loaded TSC. The design characteristics of the TS Cs are summarized in Table 1.3-2. The fuel TSC is designed, fabricated, tested and inspected to the requirements of the ASME Boiler and Pressure Vessel Code (ASME Code), Section III, Division 1, Subsection NB, to the extent practical, except as noted in the Alternatives to the ASME Code provided in Table 2.1.4-1. The GTCC waste TSCs are fabricated using ASTM materials and are fabricated in accordance with ASME Code, Section III, Division 1, Subsection NF. Criticality evaluations were performed for conditions both crediting and not crediting the TSC sealed boundary for moderator exclusion from the fissile material region. In the context of moderator exclusion, the TSC is credited with serving the 10 CFR 71.55(c) function of being a special design feature that prevents a single packing error from permitting leakage into the fissile material region. Leakage testing of the cask containment seal, in conjunction with the post-NAC International 1.3-8

                      "NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR                                                             July 2021 Docket No. 71-9356                                                                   Revision 21 A fabrication leakage testing of the entire containment boundary, assures that the containment does not leak. Regardless of credit applied to the TSC confinement boundary to prevent water in-leakage, the containment function is retained by the transport cask body.

Fuel Baskets Each TSC containing spent fuel includes a PWR, PWR-DF or BWR fuel basket that positions and supports the contents (fuel). Consistent with the TSC design, there are two different length fuel baskets (the two lengths are the same for both the PWR and the BWR fuel baskets). As described in the following sections, the design of the basket is similar for the PWR and BWR configurations. The fuel basket for each fuel type is designed, fabricated and inspected to the requirements of the ASME Code, Section III, Division 1, Subsection NG, to the extent practical, except as noted in the Alternatives to the ASME Code provided in Table 2.1.4-1. The structural components of the PWR, PWR-DF and BWR baskets are fabricated from ASME SA537, Class 1, carbon steel. To minimize corrosion and preclude significant generation of combustible gases during fuel loading, the assembled basket is coated with electroless nickel plating using an immersion process. Following plating of the structural components, the neutron absorber panels and the stainless steel retainers are installed on the basket structure as shown on

• the License Drawings. The principal dimensions and materials of fabrication of the fuel basket and PWR damaged fuel cans are provided in Table 1.3-3 and Table 1.3-4, respectively.

The fuel basket designs minimize horizontal surfaces that could entrain water. Open paths for water flow to the drain tube and sump in the bottom of the TSC are provided. The fuel baskets are supported from the TSC bottom plate by 3-in high standoffs at the corner of the fuel tubes enabling the TSC to fill and drain evenly. Fuel spacers may be used in the TSCs to reduce axial gaps for the spent fuel assemblies, non-fuel PWR Fuel Baskets The PWR fuel basket design is an arrangement of square fuel tubes held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube. The neutron absorber sheets, in conjunction with minimum TSC cavity water boron levels, provide criticality control in the basket. Each neutron absorber sheet is covered by a thin stainless steel sheet to protect the neutron absorber during fuel loading and to keep it in position. The neutron absorber and stainless steel cover are secured to the fuel tube using weld NAC International 1.3-9

                     "NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR                                                             July 2021 Docket No. 71-9356                                                                   Revision 21 A posts distributed across the width and along the length of the fuel tube. The neutron absorber sheets may be replaced by commercial aluminum sheets on the two outside surfaces of the eight outermost fuel tubes of the PWR fuel basket (refer to sheet 3 of Drawing 71160-575). The design parameters for the two lengths of PWR fuel baskets are provided in Table 1.3-3.

Each PWR fuel basket has a capacity ofup to 37 fuel assemblies in an aligned configuration. Square tubes are assembled in an array where the tubes function as independent fuel positions and as sidewalls for the adjacent fuel positions in what is called a developed cell array. Consequently, the 37 fuel positions are developed using only 21 tubes. The array is surrounded by side and corner weldments that serve both as sidewalls for some perimeter fuel positions and as the structural load path to the TSC shell. Each PWR basket fuel tube has a nominal 8.86-inch square opening. Each developed cell fuel position has a nominal 8.76-inch square opening. The system is also designed to store up to four damaged fuel cans (DFCs) in the DF Basket Assembly in the short TSC. The DF Basket Assembly has a capacity ofup to 37 undamaged PWR fuel assemblies, including four DFC locations. DFCs may be placed in up to four of the DFC locations. The arrangement of tubes and fuel positions is the same as in the standard fuel basket, but the design of each of the four corner support weldments is modified with additional structural support to provide an enlarged position for a damaged fuel can at the outermost corners of the fuel basket. Each DFC location has a nominal 9.80-in square opening. A DFC or an undamaged fuel assembly may be loaded in a DFC location. BWR Fuel Basket The BWR fuel basket design is an arrangement of square fuel tubes held in a right-circular cylinder configuration by side and corner support weldments that are bolted to the outer fuel tubes. The fuel tubes support an enclosed neutron absorber sheet on up to four interior sides of the fuel tube for criticality control in the basket during fuel loading/unloading. Each neutron absorber sheet is covered by a sheet of stainless steel to protect the neutron absorber during fuel loading and to keep it in position. The neutron absorber and stainless steel cover are secured to the fuel tube using weld posts distributed across the width and along the length of the fuel tube. The neutron absorber sheets may be replaced by commercial aluminum sheets on the three outer surfaces of the outermost fuel tubes of the BWR fuel basket (refer to sheet 3, Drawing 71160-599). Each BWR fuel basket has a capacity of 87 fuel assemblies in an aligned configuration. Square tubes are assembled in an array where the tubes function as independent fuel positions and as sidewalls for the adjacent fuel positions in what is called a developed cell array. Consequently, the 87 fuel positions are developed using only 45 tubes. The array is surrounded by weldments NAC International 1.3-10

                      "NAC PROPRIETARY INFORMATION REMOVED" MAGNATRAN Transport Cask SAR                                                            July 2021 Docket No. 71-9356                                                                 Revision 21 A that serve both as sidewalls for some perimeter fuel positions and as the structural load path to the TSC shell wall. Each BWR basket fuel tube has a nominal 5.86-in square opening. Each developed cell fuel position has a nominal 5.77-in square opening.

GTCC Waste Basket Liner An ASTM A240, Type 304, stainless steel GTCC waste basket liner is designed to hold GTCC waste and dimensionally fit in a TSC. The waste basket liner design includes a shell for structural and gamma shield functions, a welded bottom plate, and lifting lugs welded on the inside diameter of the shell so that the liner may be loaded with GTCC waste prior to being inserted into a TSC (Table 1.3-5). The liner design also includes an outer ring and a middle support under the bottom plate and drain holes in the bottom plate to facilitate free flow drainage from the liner. The GTCC TSC includes a sump location in the bottom plate and the closure lid includes a drain tube assembly to enable draining and drying of the loaded TSC. The GTCC waste basket liner and TSC are designed, fabricated and inspected in accordance with ASME Code, Section III, Division 1, Subsection NF. The lifting features of the GTCC components are designed for noncritical lifting in accordance with NUREG-0612 and ANSI N14.6, with safety factors of 3 on material yield strength and 5 on material ultimate strength applied. Damaged Fuel Can The MAGNASTOR Damaged Fuel Can (DFC), shown in Figure 1.3-3, is provided to accommodate damaged PWR fuel assemblies. The DFC may also contain PWR fuel assemblies in an undamaged condition or fuel debris equivalent to, or less than, one PWR fuel assembly. The primary function of the DFC is to confine the fuel material within the can to minimize the potential for dispersal of the fuel material into the TSC cavity. In normal operation, the DFC is in a vertical orientation. The DFC is fabricated from Type 304 stainless steel and has an 8.7-in square inside dimension (see Figure 1.3-3). The DFC is designed in two lengths: an overall length of 166.9 inches with a nominal cavity length of 164.0 inches; or an overall length of 171.8 inches with a nominal cavity length of 169.0 inches (shorter fuel assemblies may be accommodated with a fuel assembly spacer to limit axial movement). For the shorter DFC, a DFC spacer is used in the DF basket assembly or alternatively fixed to the DFC bottom plate to provide an overall height ofDFC and spacer of 171.5 inches. The side plates that form the upper end of the DFC are 0.15-in thick and the tube body walls are 0.048-in thick (18-gage sheet). The DFC lid plate and bottom NAC International 1.3-11

MAGNATRAN Transport Cask SAR December 2020 Docket No. 71-9356 Revision 20C thicknesses total 11/16 (0.688) inch and the lid overall height is 2.32 inches. The DFC bottom plate thickness is 5/8 (0.625) inch. The DFC lid and bottom include screened drain holes. The DFC is designed, fabricated, tested and inspected to the requirements of the ASME Code, Section III, Division 1, Subsection NG, to the extent practical, except as noted in the Alternatives to the ASME Code provided in Table 2.1.4-1. Cask Cavity Spacer An ASME SA240, Type 304, stainless steel cask cavity spacer is used in the upper end of the MAGNA TRAN cavity to limit the axial movement of the short TSCs. The spacer consists of six concentric rings welded to a flat plate. The depth of the rings, i.e., the length of the spacer, is approximately seven inches, which represents the difference in length between the short and long TS Cs. The spacer is bolted through the flat plate to the underside of the cask lid. Containment System MAGNATRAN provides a containment system to retain the radioactive material and gas contents during transport operations. The cask design pressure is 120 psig. The MAGNATRAN containment system components include the bottom inner forging, inner shell, top forging, cask lid and lid bolts, metal inner O-ring (lid), coverplate and bolts, and metal inner O-ring (coverplate ). The cask lid is sealed by two concentric O-rings, as is the coverplate for the lid port. In both cases, the metal inner O-ring is the containment boundary and the outer EPDM O-ring forms the annulus to facilitate leakage testing of the containment seal following installation of the lid and coverplate. A test port is provided in each annulus to enable the performance of the leakage tests. After the leakage tests are completed, each test port is closed by a stainless steel plug and boss seal. A sketch of the containment boundary is shown in Figure 1.3-2. All of the containment boundary components are defined on the License Drawings in Section 1.4-3. NAC International 1.3-12

MAGNATRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21 A 4 CONTAINMENT The MAGNATRAN transport cask containment boundary is designed and analyzed to ensure the containment of the cask contents in accordance with 10 CFR 71 (71.43 and 71.51). The containment boundary is tested to ANSI N14.5-1997 leaktight criteria and is designed, fabricated and inspected in accordance with ASME Code, Section III, Subsection NB, with the exception of code stamping. The cask is designed to facilitate leakage testing of the containment boundary penetrations (i.e., lid and lid port cover) prior to transport to confirm the containment boundary. The transportable storage canister (TSC), while not a component of the cask containment system, is evaluated for maximum pressure under normal and a HAC conditions. The sealed TSC represents the expected transport configuration, and bounds any other potential TSC condition as the sealed TSC contains a high pressure helium backfill that can, and is, considered to be released into the cask cavity. No credit is taken for the TSC as a pressure boundary under any transport condition. Criticality evaluation were performed for conditions both crediting and not crediting the TSC sealed boundary for moderator exclusion from the fissile material region. In the context of moderator exclusion, the TSC is credited with serving the 10 CFR 71.55(c) function of being a special design feature that prevents a single packing error from permitting leakage into the fissile material region. Leakage testing of the cask containment seals assures that the containment does not leak. Regardless of credit applied to the TSC confinement boundary to prevent water in-leakage, the containment function is retained by the transport cask body. The entire transport cask containment boundary is tested to the American National Standards Institute (ANSI) N14.5-1997 leaktight criteria during post-fabrication testing as described in Section 8.1.4. Periodic, maintenance, and pre-shipment testing to the American National Standards Institute (ANSI) N14.5-1997 leaktight criteria are performed on the containment boundary closures per Section 8.2.2."

• NAC International 4-1

MAGNATRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21 A

• 6.1 6.1.1 Description of Criticality Design Design Features MAGNA TRAN consists of a TSC (Transportable Storage Canister) and a transport cask overpack. The system is designed to safely transport up to 37 PWR fuel assemblies, up to 87 BWR fuel assemblies, or GTCC materials. The system is also designed to transport up to four damaged fuel cans (DFCs) in the DF Basket Assembly. The DF Basket Assembly has a capacity of up to 37 undamaged PWR fuel assemblies, including four DFC locations. DFCs may be placed in up to four of the DFC locations. Each DFC may contain an undamaged PWR fuel assembly, a damaged PWR fuel assembly, or PWR fuel debris equivalent to one PWR fuel assembly. Undamaged PWR fuel assemblies may be placed directly in the DFC locations of a DF Basket Assembly.

The TSC is comprised of a stainless steel canister and a fuel basket. Both the PWR and BWR system include two TSC lengths to transport fuel assemblies. Spacers inside the TSC may be employed to facilitate loading or unloading operations. Spacer use is not required by the criticality analysis presented in this chapter. Axial movement evaluated does not credit the presence of spacers. Should spacers be used they stay in the TSC during transportation. Fuel is

• loaded into the TSC contained within a transfer cask under water in the spent fuel pool. Once loaded with fuel, the TSC is drained, dried, backfilled with helium and welded closed. The welded TSC boundary is designed to withstand all normal conditions and hypothetical accident events and to retain a no credible leakage boundary. A single transport cask accommodates all of the PWR and BWR TSCs. An axial spacer is used inside the transport cask cavity for the short TSCs.

Based on a no credible leakage TSC boundary and a leaktight transport cask boundary, moderator is not present in the TSC while it is being transported. The structural evaluations of the MAGNA TRAN cask demonstrate that the cask containment is maintained during all conditions of transport. Containment boundary integrity is checked via fabrication acceptance tests described in Section 8 .1.4 and periodic, maintenance, and pre-shipment leakage tests described in Section 8.2.2. With no credible leakage into the fissile material region, the under moderated system will remain significantly subcritical during all normal and hypothetical accident conditions. Evaluations in this chapter address allowable payload limitations when invoking moderator exclusion and for configurations when the cask containment an TSC interiors are flooded in the criticality evaluations. Discussion in Sections 6.2 to 6.6 are relevant to the flooded TSC configuration. Moderator exclusion evaluations are included in Section 6.10. Maximum

• NAC International 6.1.1-1

MAGNATRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21 A reported reactivity for moderator exclusion is simply a function of maximum fissile material mass and initial enrichment. System criticality control is achieved through a combination of neutron absorber sheets on the interior faces of the fuel tubes and in the case of moderator intrusion for the PWR system the use of actinide-and fission product burnup credit. Individual fuel assemblies are supported in place by the fuel tubes, by developed cells formed by the fuel tubes, or by a combination of fuel tubes and side or corner weldments. The baseline neutron absorber modeled is a borated aluminum sheet with effective JOB loadings of 0.036 g/cm 2 and 0.027 g/cm 2 for the PWR and BWR system, respectively. The system is also evaluated for effective JOB loadings of 0.030 and 0.027 g/cm 2 for PWR baskets and 0.0225 and 0.020 g/cm 2 for BWR baskets. The minimum as-manufactured loading of the neutron absorber sheets depends on the effectiveness of the absorber and the minimum effective absorber areal density. Effectiveness of the absorber is influenced by the uniformity and quantity of the JOB nuclide within the absorber base material. Depending on the absorber type, a 75% or 90% effectiveness is credited. Any material meeting the JOB areal density and physical dimension requirements will produce similar reactivity results. See Table 6.1.1-1 for effective versus "credit" adjusted absorber areal densities. A combination of stainless steel cover sheets and weld posts holds the neutron absorber sheets in

• place. The PWR basket design includes 21 fuel tubes forming 37 fuel-assembly-sized openings, while the BWR basket contains 45 fuel tubes forming 89 fuel-assembly-sized openings. The PWR damaged fuel basket design includes 17 fuel tubes and four corner weldments forming 37 openings. A sketch of a cross-section of the damaged fuel basket is shown in Figure 6.1.1-2.

The combination of 45 BWR fuel tubes with four corner and four side weldments forms 89 fuel-assembly-sized openings; however, two openings are below the vent and drain ports and are not loaded. For simplicity and cask symmetry, all 89 slots are modeled as filled with fuel. An optional "82-assembly" configuration of the BWR basket is evaluated, where five center openings in an "X" pattern are left unoccupied (the basket model fills the openings below the port cover and, therefore, contains 84 assemblies). See Figure 6.1.1-1 for the loadable basket locations in the 82-assembly basket configuration. NAC International 6.1.1-2

MAGNATRAN Transport Cask SAR July 2021 7 I Docket No. 71-9356 Revision 21 A

11. Position the transfer adapter on the top of the transport cask and connect the shield door ancillary hydraulic actuation system.

12. Install the TSC lifting system swivel hoist rings and lifting slings (or other appropriate TSC lifting system meeting the facility's heavy load program) to the threaded holes in the TSC closure lid and torque to the value specified in Table 7 .1-1.

13. Using the MTC lift yoke, lift the empty MTC and place it on the transfer adapter on top of the transport cask. Ensure that the connector assemblies are in the engaged position.

Remove the door stops.

14. Install a stabilization system for the MTC, if required by the facility heavy load handling or seismic analysis programs.

15. Disengage the MTC lift yoke and remove it from the area.

16. Open the MTC shield doors using the ancillary hydraulic actuation system.

17. Connect the handling crane to the TSC lifting sling set(s) or the site-specific approved lifting system meeting the facility's heavy load program. Verify that the MTC retaining device is in the engaged position and lift the loaded TSC from the transport cask cavity into the MTC.

18. Using the ancillary hydraulic actuation system, close the MTC shield doors and set the TSC down on the doors. Install the shield door stops .

19. Disengage the TSC lifting sling set(s) from the cask handling crane or disengage the site-specific approved lifting system meeting the facility's heavy load program.

20. Using the MTC lift yoke, engage the lift yoke to the MTC lifting trunnions. Remove the stabilization system from the MTC (if used).

21. Lift the MTC containing the loaded TSC and move it to the designated location for further processing, on-site storage or unloading.

Caution: Section 6.1.2 of the SAR contains, or contains references to Chapter 6 sections, that contain enrichment limits and/or bumup credit or other requirement assuring criticality control of the system. Options are provided for transport assuming moderator (water) intrusion into the TSC cavity or for applying moderator exclusion. Enrichment limits that apply moderator within the TSC will bound a TSC unloading scenario into an unborated spent fuel pool. Increased enrichment limits exist for transport applying moderator exclusion. Receipt of a TSC containing fuel in excess of the moderated evaluation threshold indicates use of moderator exclusion for transport. For fuel requiring moderator exclusion unloading of the TSC from the transport cask into a spent fuel pool environment is not permitted unless unloading facility specific enrichment limits have been licensed and are met by the payload [e.g., soluble boron of sufficient level to prevent criticality is available in the spent fuel pool water at the unloading facility] .

• NAC International 7.2-5

MAGNATRAN Transport Cask SAR July 2021 Docket No. 71-9356 Revision 21 A

22. Attach lifting slings to the transfer adapter, lift the transfer adapter from the top of the transport cask and move the adapter to a designated storage area.

23. Connect the handling slings and remove the transfer shield ring from the transfer cask lid recess. Store the transfer shield ring in a designated storage area.

24. Take removable contamination smears of the cask inner shell to verify that the package will meet the definition for an empty package in accordance with Department of Transportation (DOT) regulations per 49 CFR 173.428(d).

25. Install the two transport cask lid alignment pins in their designated hole locations (#s 14 and 36).

26. Using the cask lid lifting slings (or equivalent site-specific approved lid lifting system) lift and install the transport cask lid. Remove the lid lifting components.

Note: It is not necessary to replace the metallic seals on the cask lid and lid port coverplate for an empty shipment. The metallic seals will be replaced prior to the next loaded transport. Note: Depending on the next planned shipment contents, the cask cavity spacer may be left installed on the lid or removed and shipped separately, as appropriate.

27. Remove the cask lid alignment pins.

28. Inspect the 48 lid bolts for damage and replace, as required, with approved spares. Lubricate the bolt with nuclear-grade Never-Seeze, or equivalent, and install the lid bolts to hand tight.

In a minimum of four passes of increasing torque, torque the 48 lid bolts to the final value specified in Table 7.1-1 for an empty cask system transport following the torquing sequence pattern marked on the lid.

29. Install the lid port coverplate and bolts on the cask lid, and torque the bolts to the value specified in Table 7 .1-1.

30. Remove scaffolding or work platforms from areas around the top of the cask.

NAC International 7.2-6

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