ML20071H192

Enclosure 2: Safety Evaluation Report for Certificate of Compliance No. 9355, Revision No. 3

ML20071H192
Person / Time
Site:	07109355
Issue date:	03/12/2020
From:	John Mckirgan Storage and Transportation Licensing Branch
To:	Al-Daouk A US Dept of Energy (DOE)
Garcia-Santos N
Shared Package
ML20071H189	List: ML20071H190 ML20071H192 ML20071H515
References
EPID L-2018-LLA-0314
Download: ML20071H192 (51)

Text

SAFETY EVALUATION REPORT Docket No. 71-9355 Model No. 435-B Certificate of Compliance No. 9355 Revision No. 3 Enclosure 2

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TABLE OF CONTENTS

SUMMARY

................................................................................................................................... 1 1.0 GENERAL INFORMATION .............................................................................................. 1 1.1 Packaging .............................................................................................................. 2 1.1 Package ................................................................................................................ 2 1.1.1 Lodgments for Disposal Canisters and IBL 437 Shielded Devices............ 2 1.2 Drawings ............................................................................................................... 3 1.3 Contents ................................................................................................................ 4 1.3.1 Disposal Canisters (Heavy, Medium, and Light)........................................ 4 1.3.2 IBL 437 Large Shielded Device: Type 1 and Type 2 ................................. 6 1.3.3 Hopewell Shielded Devices (including Transport Shield) .......................... 7 1.4 Evaluation Findings ............................................................................................... 8 2.0 STRUCTURAL EVALUATION ......................................................................................... 8 2.1 Description of Structural Design ............................................................................ 8 2.2 Structural Evaluation ............................................................................................. 9 2.2.1 Hopewell Shielded Devices ....................................................................... 9 2.2.2 Disposal Canister..................................................................................... 10 2.2.3 Disposal Canister Lodgment and IBL 437 Lodgment .............................. 10 2.2.4 Fabrication and Materials Requirements ................................................. 11 2.2.5 Applicable Codes and Standards ............................................................ 11 2.2.6 Weld Design and Specification ................................................................ 12 2.3 Evaluation Findings ............................................................................................. 12 3.0 THERMAL EVALUATION .............................................................................................. 12 3.1 Description of Thermal Design ............................................................................ 13 3.2 General Considerations for Thermal Evaluations ................................................ 13 3.3 Thermal Evaluation under Normal Conditions of Transport ................................ 14 3.3.1 Heat and Cold .......................................................................................... 14 3.3.2 Differential Thermal Expansion................................................................ 14 3.4 Thermal Evaluation under Hypothetical Accident Conditions .............................. 15 3.4.1 Initial Conditions and Fire Test Conditions .............................................. 15 3.4.2 Maximum Temperatures .......................................................................... 15 3.4.3 Differential Thermal Expansion................................................................ 17 3.5 Hopewell Device Payload .................................................................................... 17

3.6 Evaluation Findings ............................................................................................. 17 4.0 CONTAINMENT EVALUATION ..................................................................................... 18 4.1 Description of the Containment System .............................................................. 18 4.2 Containment under Normal Conditions of Transport (Type B Packages) ........... 18 4.3 Containment under Hypothetical Accident Conditions (Type B Packages) ......... 18 4.4 Leakage Rate Tests for Type B Packages .......................................................... 18 4.5 Evaluation Findings ............................................................................................. 19 5.0 SHIELDING EVALUATION ............................................................................................ 19 5.1 Description of Shielding Design .......................................................................... 20 5.1.1 Design Features ...................................................................................... 20 5.1.2 Summary Table of Maximum Radiation Levels ....................................... 22 5.2 Radiation Source Specifications .......................................................................... 23 5.2.1 GC-3000 and GC40, Gamma Source...................................................... 23 5.2.2 IBL 437, Gamma Source ......................................................................... 24 5.2.3 Disposal Canister, Gamma Source ......................................................... 25 5.2.4 Neutron Source........................................................................................ 26 5.3 Shielding Model ................................................................................................... 27 5.3.1 Configuration of Shielding and Source .................................................... 27 5.4 Shielding Evaluation ............................................................................................ 32 5.4.1 Methods ................................................................................................... 32 5.4.2 Flux-to-Dose Rate Conversion ................................................................ 35 5.4.3 External Radiation Levels ........................................................................ 35 5.5 Evaluation Findings ............................................................................................. 36 6.0 CRITICALITY .................................................................................................................. 37 7.0 PACKAGE OPERATIONS ............................................................................................. 37 7.1 Package Loading ................................................................................................. 37 7.2 Loading of Contents ............................................................................................ 38 7.2.1 Disposal Canisters ................................................................................... 38 7.2.2 IBL 437 Shielded Device ......................................................................... 38 7.2.3 Hopewell Shielded Devices ..................................................................... 38 7.3 Preparation for Transport .................................................................................... 38 7.3.1 Preparation for Transport - Inner Container............................................. 39 7.4 Package Unloading ............................................................................................. 39 7.5 Preparation of Empty Package for Transport ...................................................... 39 7.6 Evaluation Findings ............................................................................................. 39 ii

8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW ........................... 39 8.1 Material Properties .............................................................................................. 39 8.1.1 Mechanical/Thermal Properties ............................................................... 39 8.1.2 Fracture Resistance................................................................................. 40 8.1.3 Corrosion and Chemical Reactions ......................................................... 40 8.1.4 Protective Coatings.................................................................................. 41 8.1.5 Radiation Effects...................................................................................... 41 8.2 Leakage Rate Tests for Type B Packages .......................................................... 42 8.3 Evaluation Findings ............................................................................................. 42 9.0 QUALITY ASSURANCE................................................................................................. 42 CONDITIONS ............................................................................................................................. 44 CONCLUSIONS ......................................................................................................................... 45 List of Tables Page Table 1.3.1-1 LTSS and Disposal Canisters Source Nuclides. ..................................................... 5 Table 1.3.1-2 Maximum Weight of Disposal Canisters Payload ................................................... 6 Table 1.3.2-1 Maximum Activity and Weight of IBL 437 Shielded Devices.................................. 6 Table 1.3.3-1 Hopewell Designs, Inc. Shielded Devices and Transport Shield ........................... 7 Table 3.1-1 Maximum Decay Heat of The Packages Payloads ................................................ 13 Table 3.4.2-1 Synopsis of painted and non-painted peak-bulk average gas temperature for the LTSS Payload model. ..................................................................................................... 17 Table 5.1.2-1 Estimated maximum dose rates for heavy-, medium-, and light-shielded canisters in respect to the package ................................................................................................ 23 iii

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SAFETY EVALUATION REPORT Docket No. 71-9355 Model No. 435-B Certificate of Compliance No. 9355 Revision 3

SUMMARY

By application dated November 19, 2018 (NNSA, 2018b), as supplemented on September 19, 2019 (NNSA, 2019b), the National Nuclear Security Administration (thereafter, NNSA or the applicant), requested that the U.S. Nuclear Regulatory Commission (NRC) approve changes to the Model No. 435-B package as a Type B(U)-96 package. The applicant requested adding of authorized contents as well as operational changes related to the package as discussed in the application and this safety evaluation report (SER). By letter dated April 11, 2019 (NNSA, 2019a), the applicant also requested the renewal of the certificate of compliance (CoC) for the Model No. 435-B. The applicant did not request any changes to the package design or its authorized contents as part of its renewal application. Therefore, the certificate has been renewed for a five-year term.

The package is designed with a leaktight containment that can be transported singly by air, ground, or water in non-exclusive use for most of its authorized contents. For the content in disposal canisters, the package is transported in a closed conveyance as exclusive use.

NRC staff reviewed the application, including its supplement, using the guidance in NUREG-1609, Standard Review Plan for Transportation Packages for Radioactive Material (NUREG-1609). Based on the statements and representations in the application, as supplemented, and the conditions listed below, the staff concludes that the package meets the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71, Packaging and Transportation of Radioactive Material.

1.0 GENERAL INFORMATION The applicant requested the following changes as part of this revision to the design of the Model No. 435-B package:

1) adding contents authorized contents such as Radium-bearing sources, disposal canisters and shielded devices;

2) changing the version of the American National Standards Institute (ANSI) N14.5, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, (ANSI, 2014) from the 1997 to 2014 version, and additional changes to Chapters 7.0, 8.0, and 9.0 of the application;

3) allow fabrication of rain shield bolts with security heads; and

4) change the orientation of the shipment of the GC-40 shielded device.

Enclosure 2

1.1 Packaging Section 1.2.1 of the safety analysis report (SAR) (also referred as the application in this document) provides a detailed description of the packaging. When loaded and prepared for transport, the 435-B package is 83 inches (in.) tall, 70 in. in diameter (over the lower impact limiter), and weighs a maximum of 10,100 pounds (lb.). The empty packaging weight is 4,940 lb. The maximum weight of the payload is 5,160 lb.

The packaging consists of a base, a bell cover (bolted to a base), an inner container (IC),

internal lodgments, and two internal impact limiters. Besides the lodgment to support the Long Term Storage Shield (LTSS), the applicant requested to adding lodgments for the Disposal Canisters and IBL 437 shielded devices. Unless noted, all elements of the packaging are made of Type 304 austenitic stainless steel in conformance with the American Society for Testing Materials (ASTM) A240.

For the LTSS and shielded devices, the package is designed to be transported singly, with its longitudinal axis vertical, by ground, air, or by water in non-exclusive use. For the Disposal Canisters, the package can transport one or two canisters in a closed conveyance as exclusive use.

1.1 Package The 435-B package provides leaktight containment of the radioactive contents under all normal conditions of transport (NCT) and hypothetical accident conditions (HAC). Also, the packaging does not provide shielding protection to its contents. Instead, the possible contents of the Model No. 435-B (i.e., LTSS, Disposal canisters, and shielded devices) contain lead, which provides shielding to its radioactive materials.

1.1.1 Lodgments for Disposal Canisters and IBL 437 Shielded Devices The applicant is requesting to add the following lodgments to the design of the Model No. 435-B:

(1) Disposal Canister lodgment, and (2) IBL 437 lodgment.

The purpose of the lodgments is to maintain the payload within the package payload cavity during NCT and HAC. Sections 1.2.1.5.2 and 1.2.1.5.3 of the application and licensing drawings Nos. 1916-01-04-SAR, 435-B Disposal Canister Lodgment SAR Drawing, and 1916-01-05-SAR, 435-B IBL 437 Lodgment SAR Drawing, include a discussion of the materials and weld specifications for the Disposal Canister and IBL 437 lodgments, respectively.

Further, the lodgments are components important to safety. The structural components of the packaging are primarily made from ASTM B209 or B221, 6061-T6 aluminum alloy. The center of the longitudinal ribs is a "hub" made from MIL-DTL-P25995, 6061-T6, schedule 40 aluminum pipe. The longitudinal ribs are spaced and stiffened by angles made from ASTM B308, 6061-T6 aluminum. In addition, two segments of the Disposal Canister and the three segments of the IBL 437 are connected to the lodgments using ASTM A193, Grade B8 stainless steel bolts.

Figure 2.7-15, Detail of Disposal Canister Lodgment Length Adjustment Feature, of the application includes details of the configuration features for the various Disposal Canisters and 2

Figure 2.7-16, IBL 437 Lodgment Central Component, of the application includes details of the IBL 437 Type 1 and 2 payload components.

1.1.1.1 Disposal Canister Lodgment The Disposal canisters are transported in a lodgment. The lodgment maintains Disposal Canisters in position during NCT and HAC. The Disposal Canisters lodgment is a weldment made from ASTM B209 or B221, 6061-T6 aluminum alloy and consists of an upper half and a lower half. The main structural components of the lodgment are 8 equally spaced longitudinal ribs and two circumferential ribs going around the canisters.

Section 1.2.1.5.2 of the application includes a more detailed description of the Disposal Canister lodgment.

1.1.1.2 IBL 437 Lodgment The IBL 437 lodgment is a weldment made from ASTM B209 or B221, 6061-T6 aluminum alloy. The assembled lodgment is 42.75 in. in diameter and 59.5 in. tall. Eight equally spaced ribs running longitudinally and two circumferential ribs going around the body of the IBL 437 constitute the main structural components of this lodgment. The lodgment is constructed with upper, middle, and lower segments. The applicant notes that the IBL 437 has a relatively large and flat base, which serves to stabilize the device.

Section 1.2.1.5.3 of the application includes a more detailed description of the IBL 437 lodgment.

1.2 Drawings The packaging is constructed in accordance with the following AREVA Federal Services LLC drawings:

a. 1916-01-01-SAR, 435-B Package Assembly SAR Drawing, sheets 1-7, Revision 7;

b. 1916-01-02-SAR, 435-B LTSS Lodgment SAR Drawing, sheets 1-2, Revision 4;

c. 1916-01-03-SAR, 435-B Inner Container SAR Drawing, sheets 1-2, Revision 4;

d. 1916-01-04-SAR, 435-B Disposal Canister Lodgment SAR Drawing, sheets 1-2, Revision 0; and

e. 1916-01-05-SAR, 435-B IBL 437 Lodgment SAR Drawing, sheets 1-2, Revision 1.

The staff reviewed the licensing assembly drawings (lodgments) and figures (payload containers) and finds the drawings (with exception of IBL 437, Type 1 and 2) contain a bill of materials, including appropriate consensus code information, that is, American Welding Society (AWS), American Society of Mechanical Engineers (ASME), and ASTM specification number(s) for the material(s) used in fabrication. Weld requirements were well-characterized in the licensing drawings, and standard welding symbols and notations in accordance with AWS Standard A2.4, Standard Symbols for Welding, Brazing, and Nondestructive Examination.

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The staff notes that the IBL 437 large shielded devices are not part of the containment boundary of the package, but shielded devices provide shielding. As a result, the applicant describes the materials of construction for the IBL 437 devices as a carbon steel shell filled with cast lead.

The staff finds that the applicant adequately described the shielding properties of the materials and demonstrated the structural integrity of the devices.

The staff reviewed the drawings and found them to be an adequate representation of the package. The drawings included dimensions, package markings, materials of construction, and the codes and standards used to design the package. Therefore, the staff finds the description of materials and fabrication in the drawings to be acceptable.

1.3 Contents The applicant requested to add radium-226 (226Ra) sources as authorized contents. Therefore, the Model No. 435-B package will be used to transport radioactive sources such as special form sources of cobalt-60 (60Co), cesium-137 (137Cs), strontium-90 (90Sr), 226Ra, americium-241 (241Am), iridium-192 (192Ir), selenium-75 (75Se), plutonium-238 (238Pu), and/or 239Pu (i.e., gamma, beta, and small neutron sources). These sealed sources can be transported using the lodgments or an inner container, which serves to keep shielded devices in position during NCT and HAC. The applicant requested to add the following as authorized contents of the Model No.

435-B:

a. Radium-bearing sources as part of the LTSS,

b. Disposal canister payloads,

c. IBL 437 shielded devices, and

d. Hopewell shielded devices.

The LTSS and the shielded devices provide shielding of the radioactive materials. The maximum amount of fissile materials (i.e., 239Pu) allowed in the package is 15 grams, which is exempted per 10 CFR 71.15(b). The following sections include descriptions of the proposed contents and the staffs evaluation.

1.3.1 Disposal Canisters (Heavy, Medium, and Light)

The Disposal Canisters are grouped into heavy, medium, and light versions, which represents the level of shielding present in the design. The canisters design includes welded steel shells, lead shielding, a drain port, a lid port, an elastomeric dust seal, and a bolted lid. The lids bolts are made of ASTM F3125, Grade A325, Type 1, galvanized alloy steel.

The purpose of the Disposal Canisters is for transport, and final disposal of sealed sources taken from shielded devices. Sections 1.2.1.6.2, Disposal Canisters, and 3.1.1.3, Design Features of the Disposal Canister Payload, and Figures 1.2-19 through 1.2-21 of the application include discussions about the materials and weld specifications for the heavy, medium, and light Disposal Canisters. The staff notes that each type of Disposal Canister is designed and fabricated with similar materials. The codes and standards related to the structural components of the Disposal Canister, including inner/outer shells and lid, are the following:

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(1) carbon steel, ASTM A516, Grade 60, 65, or 70; (2) ASTM A513 Tubing; (3) ASTM A333 Grade 6 pipe; and (4) ASTM A1008 Type B sheet.

Shielding is provided by thick cast lead built per ASTM B29 or Federal Specification QQ-L-171E, Grade A or C. Finally, the lid acts as a shield plug by covering the payload cavity, and the lid includes the following:

(1) carbon steel-encased compressed lead sheets, (2) a carbon steel structural plate retained to the shell using ASTM F3125, Grade A325, Type 1, galvanized alloy steel bolts, (3) a lid port, (4) an elastomeric ethylene propylene diene monomer (EPDM) dust seal, and (5) exterior epoxy paint.

Section 1.2.2.2, Disposal Canister Contents, of the application includes a description of the Disposal Canister contents. Table 1.2-1 of the application (Table 1.3.1-1 of this SER) includes the nuclides and corresponding maximum activities related to this revision request of the CoC for the Model No. 435-B. Table 1.3.1-2 of this SER includes the maximum weight of the Disposal Canister payloads and lodgment.

Table 1.3.1-1 LTSS and Disposal Canisters Source Nuclides.1, 2, 3, 4 LTSS 3, 4 Disposal Canisters 3 Nuclide Maximum Activity, Maximum Mass Maximum Activity, Ci Curies (Ci) 60Co 12,970 --- 12,970 137Cs 14,000 --- 27,000 90Sr 1,000 --- 1,000 226Ra (no Be) 5 20 --- 20 226Ra Be 5 1.2 --- 4.88 241Am (no Be) 6 1,000 ---

241Am Be 6 No Americium Allowed 6.6 ---

192Ir 200 --- 200 75Se 80 --- 80 238Pu (no Be) 7 --- 75 g Pu 239Pu (no Be) 7 --- 15 g Pu No Plutonium Allowed 239Pu Be 7 --- 15 g Pu Notes:

1. Physical form of all nuclides is solid material in a sealed capsule.

2. The maximum decay heat limit for the 435-B package is 200 watts (W).

3. The values in the table represent the absolute maximum activities allowed in the 5

435-B. Individual payload activity limits depend on the configuration of the LTSS or on the specific Disposal Canister used. Payload activity limits are specified in Section 7.1.4, Loading and Preparing the LTSS for Transport, and Section 7.1.5, Loading and Preparing the Disposal Canisters for Transport of the application

4. The total activity for the LTSS payload in this table is 86,732 A2, which bounds the value for the Disposal Canister. This value exceeds the maximum number of A2 that could be transported.

5. Impurities may include oxygen, carbon, sulfur, bromine, and chlorine (hydrous and anhydrous).

6. Impurities may include oxygen and chlorine.

7. Impurities may include oxygen.

Table 1.3.1-2 Maximum Weight of Disposal Canisters Payload Maximum Weight of Payload Content Type lb.

Lodgment 500 Heavy Disposal Canister 4,610 Medium Disposal Canister 4,630 Light Disposal Canister 4,165 Figures 1.2-19 through 1.2-21 of the application includes a graphical representation of the design and specification for the heavy, medium, and light Disposal Canister Baskets, respectively. Payload sources may be small (short cylinders), pencil (long cylinders), Nuclear Liabilities Management (NLM) capsule, and the Disposal Canister baskets made of Type 304 stainless steel, vary depending on the payload source type.

1.3.2 IBL 437 Large Shielded Device: Type 1 and Type 2 Sections 1.2.2.3, Shielded Devices, and 3.1.1.2, Design Features of the IBL 437, and Figures 1.2-22 and 1.2-23 of the application include a description of the shielded devices. The staff notes that the IBL 437 payload is a large shielded device with a fixed source, contained in up to 3 pencil sources and a rotating shielded sample chamber. In addition, the fabrication and material specifications for the IBL 437 are similar to the Disposal Canisters utilizing a carbon steel shell weldment filled with cast lead shielding. Further, the IBL 437, Type 1 and 2 payload devices differ in the way the source pencils are accessed. The staff notes that the IBL 437 is an irradiator that consists of a carbon steel-lined, lead-shielded, body enclosing a carbon steel-lined, lead-shielded, rotating drum. The drum is supported by ball bearings at the lower and upper ends and rotated by a motor beneath the unit. The exterior of the carbon steel body is epoxy coated. The sources are loaded in a Type 304 stainless steel source basket. Table 1.3.2-1 includes the information added to Table 3 of the CoC.

Table 1.3.2-1 Maximum Activity and Weight of IBL 437 Shielded Devices.1 Model Name/Type Maximum Activity Weight Sealed Source Ci lb. Drawer No.3 Other Devices IBL 437 (aka IBL 437C) 5,160 4,550 MA-0219-D-813-S 1

Taken from Table 1.2-2 of the application, Group 1, Group 3 and Other Shielded Devices.

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1.3.3 Hopewell Shielded Devices (including Transport Shield)

Hopewell devices consist of five devices (i.e., G10 series irradiators) and a transport shield manufactured by Hopewell Designs, Inc. Section 1.2.2.3, Shielded Devices, and Figures 2.7-3 through 2.7-9 of the application include a description of the materials and weld design specifications for the fabrication requirements of the five Hopewell Shielded Devices and one Transport Shield. In addition, Section 3.1.1.4, Design Features of Shielded Device Payload, includes a discussion of the design features of the shielded device payloads. Table 1.2-3 of the application includes the activities and weights associated with the Hopewell shielded devices.

The maximum weight of Type 1 or Type 2 Honeywell devices is 3,500 lb. Table 1.3.3-1 of this SER corresponds to Table 1.2-3 of the application and Table 4 of the CoC.

Table 1.3.3-1 Hopewell Designs, Inc. Shielded Devices and Transport Shield Maximum Activity Weight, Model Name lb.

Option a Option b Option c 530 Ci 137Cs G10-1-360 --- --- 1,200 or 2 Ci 60Co Shielded Devices 530 Ci 137Cs Total 530 Ci 137Cs Total 2 Ci 60Co in G10-2-360 and 2 Ci 60Co in 2 sources 2 sources 1,900 2800 Ci 137Cs G10-1-2600 or 5 Ci 60Co

--- --- 1,600 2800 Ci 137Cs 137 Total 2800 Ci Cs 60 Total 5 Ci Co in G10-2-2600 and 5 Ci 60Co in 2 sources 2 sources 2,200 2800 Ci 137Cs Total 2800 Ci 137Cs Total 5 Ci 60Co in G10-2-2600-BX and 5 Ci 60Co in 2 sources 2 sources 2,200 5,952 Ci 137Cs Transport and 1,945 Ci 60Co and SC2323-GC60 --- --- 3,500 shall not Shield exceed 30W decay heat 1.3.3.1 G10 Series Shielded Devices The model names of the G10 series irradiators to be transported in the Model 435-B are the following:

(a) G10-1-360 (b) G10-2-360 (c) G10-1-2600 (d) G10-2-2600 (e) G10-2-2600-BX The staff notes that the G10 series Shielded Devices have an axially-moving source located in a Type 304 stainless steel central tube surrounded by cast-pure lead 7

(corroding grade) and contained within a carbon steel shell coated with epoxy paint.

Within the stainless steel tube, the source is located between ASTM B777, Class 1 [90%

tungsten (W), 6% nickel (Ni), 4% copper/iron (Cu/Fe)] tungsten shield plugs and moved by a tungsten source rod. The source is shielded on all sides by lead or tungsten in the stored position and exposed to a lateral opening (beam port) in the active position. The beam port opening is filled with a lead [90% lead (Pb), 10% antimony (Sb)] shield plug for shipping. The carbon steel shipping cap is retained by four ASTM F568M, M6 Class, 12.9 screws.

1.3.3.2 Transport Shield The SC2323-GC60 is a Transport Shield for radioactive sources. A central ASTM B209, B221, and B308 aluminum, grade 6061-T6 carousel, rotated by a tungsten shaft, containing up to four sources is surrounded by pure lead (corroding grade) and contained within a carbon steel shell coated with epoxy paint. The carbon steel shipping cap includes an access port, which is closed by an ASTM B777, Class 1 (90% W, 6% Ni, 4% Cu/Fe) tungsten shipping plug. The shipping cap, independent of the top shield plug, retains both the top shield plug and the tungsten shipping plug, is attached using eight ASTM F568M, M14 Class, 10.9 alloy steel screws.

The staff reviewed the description of the contents and concludes that the information provided by the applicant meets the requirements of 10 CFR Part 71.

1.4 Evaluation Findings

The staff reviewed documentation provided by the applicant including package and packaging descriptions as well as design drawings to verify that statements presented by the applicant are acceptable for the review and approval of the revision of the CoC for the Model No. 435-B, as required by 10 CFR 71.33. Based on the review of the statements and representations provided by the applicant, the staff concludes that the package, packaging, and contents have been adequately described to meet the requirements of 10 CFR Part 71.

2.0 STRUCTURAL EVALUATION The purpose of this evaluation is to verify that the proposed changes to the 435-B transport package provide adequate protection against loss or dispersal of radioactive contents and to verify that the package design meets the requirements of 10 CFR Part 71 under NCT and HAC.

2.1 Description of Structural Design This revision request consists of adding authorized payloads to the CoC and changing associated structural elements to support those payloads within the Model No. 435-B packaging. The following aspects related to the structural analysis remained unchanged from revision 2 of the CoC of the Model No. 435-B:

a. Description of the structural design

b. Design Criteria

c. Loading and Load Combinations 8

d. Acceptance Criteria

e. Codes and Standards

f. General Standards for All Packages

g. Lifting and Tie-down Standards for All Packages

h. NCT Therefore, the staff only reviewed and evaluated the changes pertaining to this application.

2.2 Structural Evaluation 2.2.1 Hopewell Shielded Devices 2.2.1.1 G10 class shielded devices The applicant identified the following scenarios in the structural evaluation of the G10 class Hopewell shielded devices in which an exposure pathway for the source is present:

(a) Loss of a lead plug for the beam port, or Loss of the lead plug is only possible if fasteners for a retaining plate are loaded such that they fail in pure tension under an impact load. Calculations provided by the applicant for a 300g impact load demonstrate a positive margin of safety of 1.9 against fastener failure.

(b) Ejection of the source rod from the source tube.

Ejection of the source rod is only possible if the fasteners of the shipping cap fail under pure tension under an impact load. Calculations provided by the applicant for a 300g impact load demonstrate a positive margin of safety of 0.47 against fastener failure.

2.2.1.2 SC2323 - GC60 shielded device The applicant identified, in the structural evaluation of the SC2323 - GC60 shielded device, the loss of a shipping cap due to tensile failure of retention fasteners as the scenario in which an exposure pathway for the source is present. Calculations provided by the applicant for a 300g impact load demonstrate a positive margin of safety of 1.9 against fastener failure.

2.1.1.2 IBL 437 Shielded Device (Type 1 and Type 2)

The applicant identified the following scenarios in the structural evaluation of the IBL 437 shielded device in which an exposure pathway for the source is present:

(a) Loss of Type 2 shield plug (Type 1 does not have this feature) 9

Loss of Type 2 shield plug is only possible if upper shield flange attachment bolts fail as well as a structural fillet weld. The applicant demonstrated that a weld length of 14.3 in. is required to resist the entire load due to a 300g impact, which essentially ignores any contribution of the bolts. As an added measure of defense-in-depth, the applicant will apply 16 in. of weld prior to shipment.

(b) Loss of source basket in Type 1 configuration Loss of Type 1 source basket is only possible if three retaining screws fail to secure the basket. Calculations provided by the applicant for a 300g impact load demonstrate a positive margin of safety of 0.56 against fastener failure, when only conservatively considering two of the three screws as effective in resisting impact loading.

(c) Loss of rotating drum from shield body Loss of rotating drum can only occur if significant failures of the support bearing or baseplate. The applicant demonstrated during certification testing that no such failures occurred.

2.2.2 Disposal Canister The applicant identified the failure of attachment bolts for bolted lids as the potential type of failure for the three sizes of Disposal Canisters. The failure of attachment bolts for bolted lids of any type can be either from shear, tension, or a combination of both. Since the design of the Disposal Canisters included a recessed flange, tensile failure of the attachment bolts was the only mechanism evaluated. The applicant demonstrated, by calculation, that all the attachment bolts maintained positive margins of safety against tensile failure for each of the three types of canisters.

2.2.3 Disposal Canister Lodgment and IBL 437 Lodgment The applicant used similarity arguments to demonstrate that the Disposal Canisters and the IBL 437 lodgment will behave in a fashion consistent with the LTSS which was shown to have robust structural integrity during the certification tests. The applicant compared the following parameters to demonstrate the similarities: weight, overall dimensions, side plate thickness, materials of construction, and maximum decay heat. Furthermore, the applicant also made similarity arguments to demonstrate that the lodgment features for the tested LTSS were essentially the same for the Disposal Canister and the IBL 437 lodgment. The areas of similarity considered were materials of construction, design configuration, outer dimensional envelope, payload support, radial and axial gaps, and payload stabilization methods.

The applicant concluded that due to the similarities in the payload and payload lodgments, the structural behavior of the Disposal Canisters and IBL 437 lodgment will be essentially the same as the tested LTSS.

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2.2.4 Fabrication and Materials Requirements The staff confirmed that the application did not introduce any new structural materials or service conditions important to safety that the staff did not previously evaluate for the Model No. 435-B transportation package. As a result, the staff focused its materials review on new materials and design features that the applicant added to incorporate the following components and contents:

(1) Disposal Canister/IBL 437 lodgments, (2) Disposal Canisters (heavy, medium, and light),

(3) IBL 437 large shielded device (Type 1, Type 2), and (4) the Hopewell shielded devices (including the Transport Shield).

The staff verified that the applicant adequately described the materials and fabrication requirements consistent with the design codes and that the added contents do not introduce adverse reactions or changes in the properties of the materials used on the Model No. 435-B package.

2.2.5 Applicable Codes and Standards Section 2.1.4 of the application includes the codes and standards used for the design of the Model No. 435-B package. Section 1.0 of this SER includes a brief description of the main components of the package including applicable industry standards used to manufacture those components.

The applicant notes that the Model No. 435-B package qualifies as a Category I container. Per NUREG/CR-3019, Recommended Welding Criteria for Use in the Fabrication of Shipping Containers for Radioactive Materials, (NUREG/CR-3019) and NUREG/CR-3854, Fabrication Criteria for Shipping Containers, (NUREG/CR-3854) the package components are classified as follows:

(1) containment components are classified as ASME Code,Section III, Subsection NB, (2) non-containment structures are classified as ASME Code,Section III, Subsection NF, and (3) lodgment and the inner container are designated as other safety and classified as ASME Code,Section III, Subsection ND.

The staff reviewed the materials codes and standards for the added package contents and lodgments and finds them to be acceptable because the use of ASTM standards is consistent with NRCs recommendations for non-containment structures, systems, and components, and the standards are considered to provide adequate control of material chemistry, fabrication, and mechanical properties of the structures, systems, and components important to safety.

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2.2.6 Weld Design and Specification In Section 2.3.1, Brittle Fracture, of the application, the applicant notes that all welding procedures and welding personnel are qualified to ASME Code,Section IX. All welds are subject to visual examination per AWS D1.6, Structural Welding Code - Stainless Steel (AWS D1.6). All other welds on the packaging except seal, tack, and intermittent welds are liquid penetrant inspected on the final pass in accordance with the ASME Code, Subsection NF, Article NF-5000, and Section V, Article 6. The applicant notes that welds on the lodgment are subject to visual examination per AWS D1.2, Structural Welding Code - Aluminum, (AWS D1.2) and welds on the inner container are subject to visual examination per ANSI/AWS D1.6 and, when specified, to liquid penetrant inspection in accordance with ASME Section III, Subsection NF, and Section V.

The staff verified that that the weld design and inspections are in accordance with the recommendations in both NUREG/CR-3019 and -3854, which includes the use of ASME Code Section III, Subsection NB, for containment boundary welds, and Subsections NF for other code welds, as appropriate. Non-code welds are examined in accordance with ASME Code Section V, with acceptance criteria per Subsection NF. The staff concludes that the welded joints of the Model No. 435-B meet the requirements of the ASME and AWS Codes, as applicable.

The staff reviewed the drawings and welding criteria in the application to verify that welding code information is appropriately identified and that the design, inspection, and testing of the welds conform to the ANSI/ASME Code criteria. The staff finds that the codes, standards, and proposed alternatives identified by the applicant are acceptable to ensure that the appropriate material is used in the packagings components. The staff verified and confirmed that that the weld design and inspections are in accordance with the ASME and AWS Codes, as applicable.

2.3 Evaluation Findings

The staff reviewed documentation provided by the applicant including design drawings, detailed calculation packages, and test results to verify that statements presented by the applicant are accurate and within acceptable engineering practices. Based on the review of the statements, representations, and supplemental calculations in the application, the staff concludes that the structural design has been adequately described and evaluated and that the package has adequate structural integrity to meet the requirements of 10 CFR Part 71.

3.0 THERMAL EVALUATION The applicant requested the approval of several payloads as authorized contents of the Model No. 435-B package. These payloads include a family of lead-shielded Disposal Canisters, a large IBL 437 shielded irradiation device (IBL 437 Payload), and a series of shielded devices manufactured by Hopewell Designs, Inc.

The purpose of this thermal evaluation is to verify that the applicants proposed changes to the Model No. 435-B package design continues to:

1) provide reasonable assurance of adequate protection against the thermal tests specified in 10 CFR Part 71 under NCT and HAC, and

2) meet the thermal performance requirements of 10 CFR Part 71.

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Regulations applicable to the thermal review include 10 CFR 71.31, 71.33, 71.35, 71.43, 71.71, and 71.73. The following sections summarize the staffs thermal evaluation.

3.1 Description of Thermal Design This SER briefly described the thermal effects of the following proposed payloads to the Model No. 435-B package:

a. IBL 437 shielded devices,

b. Disposal Canister, and

c. Hopewell devices.

Section 1.2.2, Contents, of the application includes a description of the proposed contents of the package. Table 3.1-1 includes a summary of the heat loads of the current and proposed authorized contents.

Table 3.1-1 Maximum Decay Heat of The Packages Payloads Decay Heat Content Type W

LTSS shielded devices 200 Inner container shielded devices 30 IBL 437 (aka IBL 437C) Shielded Device 15 Heavy disposal canister 200 Medium disposal canister 159 Light disposal canister 144 The applicant stated in Section 3.1.1.4, Design Features of Shielded Device Payload, of the application that the Hopewell devices consists of the G10 series of irradiators and the SC2323-GC60 transport shield. The G10 series of irradiators consists of G10-1-360, G10-2-360, G10-1-2600, G10-2-2600, and G10-2-2600-BX. As stated in Section 3.5, Appendices, of the application, while the SC2323-GC60 dissipates 30 W of decay heat, all the G10 series of Hopewell devices dissipate a decay heat loading significantly lower than the 30-W decay heat used for the generic device.

The staff reviewed Sections 1.2.2, 3.1.1.4, and 3.5 of the application and concludes that the payload decay heat values of the Disposal Canister, IBL 437, and Hopewell Devices are appropriate. The staff also confirmed that the decay heat of the Disposal Canister bounds the decay heat values of the IBL 437 and Hopewell devices. Therefore, the Model No. 435-B package, loaded with the Disposal Canister, represents the bounding case for the thermal evaluation in this CoC revision request.

3.2 General Considerations for Thermal Evaluations The applicant presented the thermal properties of Type 304 stainless steel, 6061 aluminum, QQ-L-171E Grade A or C lead, tungsten, and brass in Table 3.2-1, Thermal Properties of Metallic Materials, of the application. The applicant calculated the properties of metallic material for temperatures between the tabulated values by linear interpolation within the heat 13

transfer code (i.e., Thermal Desktop and SINDA/FLUINT computer program). The applicant used the thermal properties for Type 304 stainless steel and 6061 aluminum from the ASME material properties database and the density from an on-line materials database.

The staff reviewed Table 3.2-1 of the application and confirmed that the material properties of the packaging components used in this amendment application are identical to those approved by the staff in the previous applications for the Model No. 435-B. Therefore, the material properties shown in Table 3.2-1 of the application are acceptable.

The applicant noted in Section 3.2.1, Materials Properties, of the application that stainless steel properties are used to model all stainless steel and steel components encasing lead. The applicant also noted that this approximation has a negligible effect on the thermal performance of the package because the shell is thin in comparison to the adjacent lead body, and the steel components encasing lead have a comparable allowable temperature to stainless steel.

The staff reviewed the Model No. 435-B package configuration. The thin shell, encasing the lead, has a limited effect on thermal evaluations when adjacent to the thicker lead. Therefore, the staff finds the use of stainless steel properties for all stainless steel and steel components in the thermal evaluations to be acceptable.

3.3 Thermal Evaluation under Normal Conditions of Transport 3.3.1 Heat and Cold The applicant noted in Section 3.3.1, Heat and Cold, of the application that the thermal performance of the LTSS Payload bounds the thermal performance of the Disposal Canister payload and the thermal performance of the shielded device payload bounds the thermal performance of the IBL 437 Payload under NCT because the Disposal Canister Payload has a decay heat load bounded by that of the LTSS Payload (200 W) and the IBL 437 Payload has the decay heat load bounded by the decay heat load of the shielded device payloads (30 W).

The staff reviewed the LTSS Payload and shielded device payloads for the 435-B package and found those payloads to be acceptable by the staff. Therefore, all components of the 435-B package, loaded with the Disposal Canister Payload or the IBL 437 Payload, remain below their allowable NCT temperature limits under the same heat load and pressure limit (5 psig) as presented in Tables 3.3-1, 3.3-2, and 3.3-3 of the application, respectively. The staff found these to be acceptable by the staff in the previous application for the Model No. 435-B.

Based on the thermal performance described in Section 3.3.1 of the application, the staff finds acceptable the thermal performance because the maximum NCT temperatures and pressures of the 435-B package loaded with the Disposal Canister Payload or the IBL 437 Payload are bounded by the predicted temperatures of the LTSS Payload and the other shielded device payload presented in Section 3.3.1.1, Maximum Temperatures, of the application and are therefore below their design limits, in compliance with 10 CFR 71.71.

3.3.2 Differential Thermal Expansion The applicant noted in Section 2.6.1.2, Differential Thermal Expansion, of the application that the calculations of differential thermal expansion under NCT demonstrated a positive clearance between the payload cavity and the lodgment of the 435-B package (including the LTSS, the 14

Disposal Canister and the IBL 437 lodgments), neglecting the expansion of the payload cavity itself.

The staff reviewed Section 2.6.1.2 of the application and finds it to be acceptable to apply the calculations of differential thermal expansion for the LTSS Payload (in the previous application) to the Disposal Canister Payload and the IBL 437 Payload. This is because the NCT packaging component temperatures for the Disposal Canister Payload and the IBL 437 Payload are bounded by those for the LTSS Payload. Therefore, the staff confirmed that like the LTSS Payload, there is a positive clearance between the 435-B package and the lodgment for the Disposal Canister Payload and the IBL 437 Payload under NCT.

3.4 Thermal Evaluation under Hypothetical Accident Conditions 3.4.1 Initial Conditions and Fire Test Conditions The staff reviewed Section 3.4.1, Initial Conditions, of the application and finds that there are no changes in the fire initial conditions and fire test conditions, as reviewed and accepted by the NRC in the previous application.

3.4.2 Maximum Temperatures The applicant noted in Section 3.4.3, Maximum Temperatures and Pressure, of the application that it did not model the HAC scenario for the head-down drop of the package because the side drop damage scenario results in the highest payload component temperatures, as reviewed and accepted by the NRC in the previous application (NNSA, 2018a).

The applicant noted in Appendix 3.5.5, Thermal Comparison of Lodgment Supported Payloads, of the application that the peak temperatures predicted for the Disposal Canister Payload bounds the peak temperatures for the Model No. 435-B package with the IBL 437 Payload under the HAC. Since the emissivity at the surface of the LTSS Payload and the Disposal Canister Payload are essentially and thermally identical, the applicant modeled the effect of the paint on the outer surface of the Disposal Canister Payload by performing a thermal analysis using the side drop damage case of the LTSS Payload thermal model described in Appendix 3.5.3, Analytical Thermal Model. The only difference was that the surface emissivity of the LTSS Payload was set to 0.8 (for a painted surface) instead of 0.46 for the LTSS stainless steel surface.

The staff reviewed the design features of the painted LTSS Payload and Disposal Canister Payload described in Section 3.1.1, Design Features, and descriptions of the modeling of the LTSS Payload provided in Appendix 3.5.3, Analytical Thermal Model, of the application and confirmed the following:

(1) the painted LTSS Payload model can represent the Disposal Canister Payload because both payloads are essentially thermally identical, and (2) an emissivity 0.8 used for the outer painted surface of the Disposal Canister Payload is appropriate.

The applicant noted the following in Section 3.4.3, Maximum Temperatures and Pressure:

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(1) the predicted peak temperatures of the lead and exterior surface for the increased surface emissivity case (Disposal Canister Payload) are 285 °F and 278 °F, respectively, which represents an increase of less than 10 °F above the peak lead and peak surface temperatures for the base LTSS Payload case reported in Table 3.4-1 (NNSA, 2018a), HAC Temperatures for Side Drop Damage with LTSS; (2) the maximum temperature on the upper torispherical head for the increased surface emissivity case also increased to 1,102 °F (i.e., a difference of +1 °F)

(the rest of the 435-B packaging components remain bounded by the temperatures in Table 3.4-1); and (3) the peak lead temperature of 285°F for the increased surface emissivity case is well below the limiting temperature for lead of 620°F and all component temperatures remain within the allowable temperature limits established in Section 3.2.2, Component Specification, of the application.

The staff reviewed the HAC initial conditions and the thermal evaluation for the side drop damage case (scenario) of the Model No. 435-B package for a painted LTSS Payload surface and finds the thermal evaluation acceptable because the following reasons:

(1) all peak packaging component temperatures are below their allowable HAC limits, (2) the temperature of the paint is bounded by the temperature of the Disposal Canister Payload shell (i.e., 278°F), and (3) the paint has no significant degradation below 392°F per thermogravimetric analysis as described in Section 3.4.3.7, Behavior of Non-metallic Contents Materials Under HAC, of the application (Williamson and Iams, 2004).

3.4.2.1 Maximum HAC Pressures The applicant noted in Section 3.4.3.6, Maximum HAC Pressures, of the application that the temperature achieved in the LTSS Payload pressure case bounds the HAC peak-bulk average temperature of gas achieved with the painted LTSS Payload model.

The painted LTSS Payload model represents the maximum decay heat of the Disposal Canister Payload (i.e., 362°F). Therefore, the pressurization of the package cavity will arise solely from ideal gas expansion.

The staff reviewed Section 3.4.3.6 of the application and compared the HAC peak-bulk average gas temperature with the painted LTSS Payload model to the peak-bulk average gas temperature with the non-painted LTSS Payload model, which the staff considered acceptable (NNSA, 2018a).

Table 3.4.2-1 of this SER includes a summary of the peak-bulk average gas temperature associated with different LTSS Payload models. The staff finds that the peak-bulk average gas temperature achieved with the painted LTSS Payload model, is lower than the side drop damage with the non-painted LTSS Payload model and the head down drop damage with the non-painted LTSS Payload model.

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Table 3.4.2-1 Synopsis of painted and non-painted peak-bulk average gas temperature for the LTSS Payload model.

Peak-Bulk Average Gas Model Temperature (°F)

Painted LTSS Payload model 362 Side drop damage with the non-painted LTSS 364 Payload model Head down drop damage with the non-painted 366 LTSS Payload model Based on the comparison of the peak-bulk average gas temperatures, the staff confirmed that the Model No. 435-B package loaded with the Disposal Canister Payload (200W), or IBL 437 Payload, has a HAC maximum pressure below the HAC design limit and meets the thermal requirements of 10 CFR 71.73.

3.4.3 Differential Thermal Expansion The applicant noted in Section 2.7.4.2, Differential Thermal Expansion, of the application that the calculations of differential thermal expansion under HAC, presented in Section 2.7.4.2 of the application, demonstrated a positive clearance under HAC between the Model No. 435-B payloads cavity and the LTSS payload, consisting of the lodgment or inner container.

The staff reviewed Section 2.7.4.2 of the application and the applicant justifies that the calculations of differential thermal expansion for the LTSS Payload in the previous application (NNSA, 2018a), approved by the NRC, are applicable to the Disposal Canister Payload and the IBL 437 Payload. The applicant explained that the HAC packaging component temperatures for the Disposal Canister Payload and the IBL 437 Payload are bounded by those for the LTSS Payload. Therefore, the staff confirmed that, like the LTSS Payload, there is a positive clearance between the 435-B package and the lodgment for the Disposal Canister Payload and the IBL 437 Payload under HAC.

3.5 Hopewell Device Payload The staff also confirmed that the maximum NCT and HAC temperatures, pressures, and thermal expansions of the Model No. 435-B package loaded with the Hopewell Device Payload are also below the corresponding NCT and HAC design limits, in compliance with 10 CFR 71.71 and 10 CFR 71.73, respectively. This is because the decay heat of the LTSS Payload (200W) bounds the maximum decay heat of the Hopewell Device Payload (30W) and the IBL 437 Payload (30W).

3.6 Evaluation Findings

Based on a review of the statements and representations in the application, the staff finds that the applicant adequately described and evaluated the proposed Disposal Canister Payload, IBL 437 Payload, and Hopewell Device Payload to the Model No. 435-B package, and that the package meets the thermal requirements of 10 CFR Part 71.

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4.0 CONTAINMENT EVALUATION The purpose of this evaluation is to verify that the proposed changes to the Model No. 435-B transport package provide adequate protection against radiation and to verify that the package design meets the requirements of 10 CFR Part 71 under NCT and HAC.

4.1 Description of the Containment System The applicant did not propose changes to the description of the containment system for the Model No. 435-B in Section 4.1, Description of the Containment System, of the application.

The Model No. 435-B continues to provide a single level of leaktight containment, defined as a leakage rate of less than 1 x 10-7 reference cubic centimeters per second (ref-cm3/s), air, per the American National Standards Institute (ANSI) N14.5, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, (ANSI, 2014). Figure 4.1-1, 435-B Package Containment Boundary, of the application continues to show the Model No.

435-B containment boundary.

4.2 Containment under Normal Conditions of Transport (Type B Packages)

Section 2.6, Normal Conditions of Transport, and Section 3.3, Thermal Evaluation for Normal Conditions of Transport, of the application include the results considering NCT, as required in 10 CFR 71.71, Normal conditions of transport. These results demonstrate that there is no release of radioactive materials, based on the leaktight definition in ANSI N14.5, under any of the NCT tests. The staff finds that this analysis meets the containment requirements in 10 CFR 71.51(a)(1).

4.3 Containment under Hypothetical Accident Conditions (Type B Packages)

Section 2.7, Hypothetical Accident Conditions, and Section 3.4, Thermal Evaluation for Hypothetical Accident Conditions, of the application include the results considering HAC, as required in 10 CFR 71.73, Hypothetical accident conditions. Through these results, the applicant demonstrated that there is no release of radioactive materials, based on the leaktight definition in ANSI N14.5, under any of the HAC tests. The staff finds that the analysis provided by the applicant meets the containment requirements in 10 CFR 71.51(a)(2).

4.4 Leakage Rate Tests for Type B Packages The applicant requested to use the latest revision of ANSI N14.5, changing from the 1997 to 2014 revision. The staff confirmed that the references to ANSI N14.5 in Chapters 1.0, General Information, 4.0, Containment, 7.0, Package Operations, and 8.0, Acceptance Tests and Maintenance Program, of the application were changed to reflect the 2014 revision. The staff confirmed that Chapters 7.0 and 8.0 of the application are referenced in the CoC. Sections 7.3 and 8.2 of this SER include additional discussions of the staffs evaluation of the leakage rate tests for the Model No. 435-B.

The staff confirmed that for the new contents, the lead-shielded Disposal Canisters, the Hopewell devices, and the IBL 437 devices (in addition to the LTSS) that the torque value for the vent port plug and its associated vent port sealing washer is applied in the operating procedures prior to the performance of the pre-shipment or periodic leakage rate tests. Based on the staffs review of the response to RAI-Co-2 (NNSA, 2019), this was accomplished by specifying that the vent port sealing washer and vent port plug are installed and tightened to 48 18

- 60 in. pound force (in-lbf) in Section 7.4.2, Determining the Test Volume and Test Time, of the application for the pre-shipment leakage rate test and Section 8.2.2.2, Maintenance/Periodic Leakage Rate Test, of the application for the maintenance and periodic leakage rate tests; therefore, the staff finds this to be acceptable. Providing the vent port sealing washer and vent port plug torque value in the operating procedures prior to performing either the pre-shipment or maintenance/periodic leakage rate tests provides reasonable assurance that the containment system is properly assembled in accordance with the pre-shipment leakage rate test, or has not deteriorated, during a period of use in accordance with the periodic leakage rate test. Both the pre-shipment and periodic leakage rate tests are described in ANSI N14.5-2014. The staff finds the leakage rate test descriptions for the Model No. 435-B meet ANSI N14.5-2014.

4.5 Evaluation Findings

Based on review of the statements and representations in the application, the staff finds that the applicant adequately described and evaluated the containment design and that the package design meets the containment requirements of 10 CFR Part 71.

5.0 SHIELDING EVALUATION The purpose of this evaluation is to verify that the proposed changes to the Model No. 435-B transport package shielding provide adequate protection against direct radiation from its contents and to verify that the package design meets the external radiation requirements of 10 CFR Part 71 under NCT and HAC.

Among the proposed changes to the certificate of compliance for the Model No. 435-B are adding several new contents, including a family of lead-shielded Disposal Canisters, a series of shielded irradiator bodies, and a large shielded device, the IBL 437. The following sections of the application include a summary of the shielding analysis for the new proposed contents:

1) Appendix 5.5.3, GC-3000 and GC-40 Shielded Device Evaluation, of the application includes the shielding analysis for the GC-3000 and GC-40 devices.

2) Appendix 5.5.4, IBL 437 and Hopewell Inc. Shielded Device Evaluation, of the application includes the shielding analysis for the IBL 437 devices (Types 1 and

2) and the Hopewell devices (i.e., G10-1-360, G10-2-360, G10-1-2600, G10 2600, G10-2-2600-BX, and SC2323-GC60).

3) Appendix 5.5.5, Disposal Canister Evaluation, of the application includes the shielding analysis for the light, medium, and heavy Disposal Canisters.

The following sections include the staffs evaluation of the proposed changes to the shielding evaluation as the result of the proposed changes to the CoC.

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5.1 Description of Shielding Design 5.1.1 Design Features The addition of the proposed contents, described in Chapter 1 of the application and Section 1.0 of this SER, resulted in the addition of shielding design features including the following components:

(1) GC-3000 and GC-40 Shielded Device, (2) IBL 437 shielded devices, and (3) the Hopewell devices (G10-1-360, G10-2-360, G10-1-2600, G10-2-2600, and G10-2-2600-BX) and transport shield SC2323-GC60.

The 435-B packaging is designed to transport the LTSS. The 435-B packaging itself offers little shielding. The outer shell of the 435-B is made of 0.5 in. thick steel.

5.1.1.1 GC-3000 and GC-40 Shielded Devices The GC-3000 and GC-40 shielded devices are addressed in this application. Both devices are heavily shielded with lead. This section of the SER provides additional descriptions of the shielding characteristics of the GC-3000 and GC-40 shielded devices:

(a) GC-3000 The GC-3000 is heavily shielded with lead. The lead thickness through the top lead plug is approximately 3 in. A source holder with approximately 2.35 in. [6 centimeters (cm)] of steel provides additional shielding at the top of the package. The minimum side lead thickness is approximately 4.5 in.

(b) GC-40 The GC-40 is also heavily shielded with lead. The GC-40 drawer provides an approximately 5.75-in. axial lead shield and a 1.3-in. axial steel shield. The GC-40 is highly asymmetrical in shape and it provides several inches of lead shielding.

5.1.1.2 IBL 437 and Hopewell Inc. Shielded Devices Section 1.3.2 of this SER includes a brief description of the IBL 437 Shielded Devices.

As previously mentioned in this SER, there are two types of IBL 437 devices. The IBL 437 Types 1 and 2 are similar in design. Both designs feature a source basket that is heavily shielded with lead. The IBL 437 design is complex in shape. The minimum lead thickness is approximately 5.3 in. for the Type 1 and 6.2 in. for the Type 2. The minimum lead thickness occurs on the side of the devices.

5.1.1.3 Hopewell Inc. Shielded Devices 20

Section 1.3 of this SER includes a brief description of the Hopewell Inc. shielded devices. This section of the SER provides additional descriptions of the shielding characteristics of the transport device and the Hopewell Inc. shielded devices:

(a) The SC2323-GC60 is heavily shielded with lead and tungsten. The lead thickness is 8.5 in. on the side and 8.875 in. at the bottom of the SC2323-GC60. The top plug features 9.0 in. of lead with a tungsten drive shaft and a tungsten shipping plug.

(b) The G10 Series devices are cylindrical in shape. Sources are placed in a vertical source tube and are axially shielded with tungsten plugs.

Lead in the body of the device provides bulk shielding. Each G10 Series device features a range of allowable lead thicknesses and tungsten plug lengths.

(c) The G10-1-360 features lead shielding on the sides and top and is designed to carry a single source. The minimum lead thickness is 1.25 in. on the top and 4.1 in. on the side. Sources are inserted into a vertical source tube, which is shielded by tungsten plugs on the top and bottom.

(d) The G10-2-360 features lead shielding on the sides and top and is designed to carry two sources. The minimum lead thickness is 1.25 in. on the top and 4.1 in. on the side. Sources are inserted into a vertical source tube, which is shielded by tungsten plugs on the top, bottom, and between the sources. The upper tungsten plug has a minimum length of 1.75 in., the lower tungsten plug has a minimum length of 3.0 in., and the middle tungsten plug has a minimum length of 2.0 in.

(e) The G10-1-2600 features lead shielding on the sides and top and is designed to carry a single source. The minimum lead thickness is 1.75 in. on the top and 4.584 in. on the side. The sources are inserted into a vertical source tube shielded by tungsten plugs at the top and bottom. The upper tungsten plug has a minimum length of 2.0 in., and the lower tungsten plug has a minimum length of 3.5 in.

(f) The G10-2-2600 features lead shielding on the sides and top and is designed to carry two sources. The minimum lead thickness is 1.75 in. on the top and 4.584 in. on the side. Sources are inserted into a vertical source tube, which is shielded by tungsten plugs on the top and bottom. The upper tungsten plug has a minimum length of 2.0 in., the lower tungsten plug has a minimum length of 3.5 in., and the middle tungsten plug has a minimum length of 3.0 in.

(g) The G10-2-2600-BX features lead shielding on the sides and top and is designed to carry two sources. The shipping orientation of this device is opposite to the other G10 Series devices because the base plate is in the positive z-direction. The minimum lead thickness is 4.0 in. at the end and 4.584 in. on the side. Sources are inserted into a vertical source tube, which is shielded by tungsten plugs above and 21

below the upper source (the lower source is not shielded by tungsten at the bottom direction). Both tungsten plugs have a minimum length of 3.5 in.

5.1.1.4 Disposal Canister The disposal canisters are the primary source of shielding and it varies widely between canisters. The disposal canister designs are addressed in this analysis, referred to as the heavy-, medium-, and light-shielded canisters.

(a) The heavy-shielded canister offers the lowest payload volume of the three types of disposal canisters, but it can shield high-activity 60Co sources. The heavy canister is also used to shield 226Ra sources.

The lead thickness through the top lead plug is approximately 8.41 in.

Structural elements add approximately 2.37 in. of steel shielding, which increases attenuation at the top of the canister. The side and bottom lead thicknesses are approximately 8.65 in. and 8.52 in.,

respectively, including lead shrinkage.

(b) The medium-shielded canister offers an intermediate level of shielding and payload volume. The purpose of the medium-shielded canister is to transport and dispose of 60Co sources. The lead thickness through the top lead plug is approximately 6.33 in. Structural elements add approximately 3.37 in. of steel shielding, which increases attenuation at the top of the canister. The thicknesses of lead for the side and bottom of the medium-shielded canister are approximately 7.54 in.

and 7.32 in., respectively, including lead shrinkage.

(c) The light-shielded canister offers the least shielding and largest payload volume of the three designs of disposal canister. The purpose of the light-shielded canister is to transport and dispose of 137Cs, 90Sr, 192Ir, and 75Se (60Co and 226Ra are excluded). The lead thickness through the top lead plug is approximately 4.30 in.

Structural elements add approximately 2.13 in. of steel shielding, which increases attenuation at the top. The side and bottom lead thicknesses are approximately 4.56 in. and 4.46 in., respectively, including lead shrinkage.

5.1.2 Summary Table of Maximum Radiation Levels Dose rates are computed for a Group 1 device (GC-3000) with a 3,840 Ci 137Cs pencil source and for the Group 3 device (GC-40) with a 2,250 Ci 137Cs point source. Table 5.5.3-1 and Table 5.5.3-2 of the application include the NCT dose rates for the GC-3000 and GC-40, respectively.

Because the GC-3000 maximum dose rate bounds the GC-40 maximum dose rate under NCT (based on the results in Tables 5.5.3-1 and 5.5.3-2 of the application), HAC dose rates are calculated only for the GC-3000. The HAC dose rates are computed at 1 m from the surface of the package and are provided in Table 5.5.3-3 of the application. The HAC dose rates are negligible compared to the limit of 1,000 millirem per hour (mrem/hr).

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Tables 5.5.4-1 through 5.5.4-3 of the application includes the maximum dose rates under NCT for IBL 437 Type 1 and Type 2, SC2323-GC60, and G10 Series devices. The dose rates provided for the IBL 437 are the maximum dose rates of both. The maximum dose rates provided for the G10 Series devices are the maximum values of the five G10 Series devices.

For heavy-shielded canisters, medium-shielded canisters, and light-shielded canisters, the isotope activities are limited to ensure that the dose rates do not exceed 95% of the dose rate limit. The limiting dose rate occurs on either the surface of the vehicle or 2 m from the outer lateral surfaces of the vehicle. The dose rates on the package surface and the normally occupied space are not limiting. Table 5.1.2-1 of this SER includes a summary of the maximum computed dose rates.

Table 5.1.2-1 Estimated maximum dose rates for heavy-, medium-, and light-shielded canisters in respect to the package.

Location Maximum Calculated Dose Rates surface a package 381 mrem/hr outer surface of the vehicle 190 mrem/hr 2 m from the outer lateral surfaces of the 9.5 mrem/hr vehicle any normally occupied space 1.8 mrem/hr The dose rates calculated using this method are below the NCT limits for non-exclusive use transportation. Therefore, the transport index (TI) will not exceed 9.5. Under HAC, dose rates are the same as the NCT dose rates at 1 m, or 9.5 mrem/hr. The staff verified that this was consistent with the condition of the package.

Under HAC, there is no damage to the canister, canister lodgment, or the 435-B package that affects dose rates in any significant manner. Because there is negligible change to the shielding under HAC, the HAC dose rates are bounded by the NCT dose rates on the surface of the vehicle.

The staff verified that radiation levels are within the regulatory limits. The staff also examined the variation of dose rates at different package locations and found them acceptable because most of the locations are near positions that can be streaming paths for gammas and neutrons.

5.2 Radiation Source Specifications 5.2.1 GC-3000 and GC40, Gamma Source The gamma sources contain only 137Cs. The decay of 137Cs is sufficiently simple to be treated explicitly. The decay of 137Cs/137mBa emits a 0.662 mega electron volt (MeV) gamma with an 85% probability.

Table 1.2-2 of the application includes the activities for the GC-3000 and GC-40. The GC-3000 contains up to 3,048 Ci 137Cs, although the maximum activity for Group 1 is 3,840 Ci 137Cs for both the Gammator M38 and GC-1000. Because the intent is to bound all Group 1 devices with the GC-3000 analysis, the larger 3,840 Ci activity is modeled in the GC-3000. Therefore, the as-modeled gamma source for the GC-3000 is the following:

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gamma source for the GC-3000 = [(Activity of GC-3000) x (0.85) x {3.7 x 1010[(decays/s)/Ci]}]

photon/s gamma source for the GC-3000 = [3,840 Ci x (0.85) x {3.7 x 1010[(decays/s)/Ci]}] photon/s gamma source for the GC-3000 = 1.208 x 1014 photon/s The GC-40 contains up to 4,200 Ci 137Cs. However, only the upper or lower module of a GC-40 will be transported within the Model No. 435-B package, and the maximum activity within a module is 2,250 Ci. Therefore, the as-modeled gamma source for the GC-40 is the following:

gamma source for the GC-3000 = [(Activity of GC-40) x (0.85) x {3.7 x 1010[(decays/s)/Ci]}]

photon/s gamma source for the GC-40 = [2,250 Ci x (0.85) x {3.7 x 1010[(decays/s)/Ci]}] photon/s gamma source for the GC-40 = 7.076 x 1013 photon/s 5.2.2 IBL 437, Gamma Source Sources are either 137Cs, 60Co, or combinations of these isotopes. The decay energies of each isotope are sufficiently simple to be treated explicitly. The decay of 137Cs/137mBa emits a 0.662 MeV gamma with an 85% probability. The intensity for a 137Cs source may be calculated using the following equation:

137Cs Intensity (photon/s) = 0.85 x A x {3.7 x 1010 [(decays/s)/Ci]}

where A is the 137Cs activity in Ci.

The decay of 60Co emits two gammas with energies of 1.332 MeV and 1.173 MeV. Because each decay emits two gammas, the intensity of a 60Co source may be calculated using the following equation:

60Co Intensity (photon/s) = 2 x B x {3.7 x 1010 [(decays/s)/Ci]}

B is the activity of 60Co in Ci. The various devices and the maximum activity of each device are summarized in Table 5.5.4-5 of the application.

For the SC2323-GC60, the source may be a mixture of 137Cs and 60Co sources limited to a total decay heat of 30 W. From Table 5.4-3 of the application, 60Co has a heat load of 1.542 x 10-2 W/Ci and 137Cs has a heat load of 5.040 x 10-3 W/Ci. Therefore, the limiting equation for the SC2323-GC60 (based on a maximum heat load of 30W) is the following:

A x (5.040 x 10-3 W/Ci) + B x (1.542 x 10-2 W/Ci) 30 W A is the total activity of 137Cs in Ci and B is the total activity of 60Co in Ci.

Therefore, the maximum activity of 137Cs allowed in the package is calculated assuming that A = maximum activity of 137Cs, B = 0, and the packages heat load = 30 W:

A x (5.040 x 10-3 W/Ci) = 30 W 24

A = maximum activity of 137Cs = [30W/(5.040 x 10-3 W/Ci)] = 5,952 Ci The maximum activity of 60Co allowed is calculated assuming that B = maximum activity of 60Co, A = 0, and the packages heat load = 30 W:

B x (1.542 x 10-2 W/Ci) = 30 W B = maximum activity of 160Co = [30W/(1.542x10-2 W/Ci)] = 1,945 Ci The applicant did not take credit for self-shielding of the source or by source encapsulation materials. The applicant modeled the source with the dimensions of the following devices:

(1) For the IBL 437, the source material is approximately 1.5 cm in diameter and 29 cm long. In the MCNP models, the source is conservatively modeled with a radius of 0.1 cm and a height of 20 cm.

(2) For the SC2323-GC60, the 60Co source is modeled with a diameter of 2.5 cm and a height of 3.5 cm. The 137Cs source is modeled with a diameter of 3.25 cm and a height of 6.0 cm.

(3) For the G10 Series devices, which feature many different source sizes, the radius of the source is modeled with a 0.2-in. clearance to the inside of the source tube. For the G10-1-360 and G10-2-360, the source is modeled with the maximum length of 2.1 cm. For the G10-1-2600, G10-2-2600, and G10-2-2600-BX devices, the source is modeled with the maximum length of 6.0 cm.

5.2.3 Disposal Canister, Gamma Source 5.2.3.1 Gamma Source Terms The applicant developed the gamma source terms for 60Co, 137Cs, 90Sr, 192Ir, 75Se, and 226Ra on a per Ci basis in Section 5.2.1 of the application. The following paragraphs summarize the results.

(a) 60Co. The decay of 60Co is sufficiently simple to be treated explicitly.

Each decay of 60Co results in two gammas, with energies of 1.173 and 1.332 MeV. Table 5.5.5-2 of the application includes the gamma source for 1 Ci of 60Co.

(b) 137Cs. The decay of 137Cs is sufficiently simple to be treated explicitly.

The decay of 137Cs emits a 0.662 MeV gamma with an 85%

probability. Table 5.5.5-3 of the application includes the gamma source for 1 Ci of 137Cs.

(c) 90Sr. Table 5.5.5-4 of the application includes the gamma source for 1 Ci of 90Sr calculated by the applicant using ORIGEN-S.

(d) 192Ir. Table 5.5.5-5 of the application includes the gamma source for 1 Ci of 192Ir calculated by the applicant using ORIGEN-S.

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(e) 75Se. Table 5.5.5-6 of the application includes the gamma source for 1 Ci of 75Se calculated by the applicant using ORIGEN-S.

(f) 226Ra.. 226Ra is from the used radium-based neutron sources. 226Ra is an alpha and gamma emitter. Table 5.5.5-7 of the application includes the gamma source for 1 Ci at the peak decay time of 0.3 years calculated by the applicant using ORIGEN-S.

The alpha radiation will lead to a neutron source when mixed with an (, n) target nucleus, such as beryllium.

5.2.3.2 Sum of Fractions The applicant performed all the MCNP calculations for a single source and isotope type.

However, in practice, two different 226Ra source types (i.e., with and without beryllium) may be present in the heavy canister. Once the single source activity limits are known, isotopes may be combined using the sum of fractions rule. This relationship may be expressed as follows:

1 where:

Si is the activity of each source in Ci, and Ai is the activity that results in a dose rate at 95% of the regulatory limits.

Values of Ai are developed for 226Ra with and without beryllium in the heavy canister.

These values are used in sum of fractions calculations when mixing 226Ra sources.

Sum of fractions are not needed for other canisters and source configurations.

5.2.4 Neutron Source In Section 5.2.2 of the application, the applicant shows the neutron source terms for 226Ra on a per Ci basis. A neutron source is generated by 226Ra due to (, n) [(alpha, neutron)] reactions.

Target nuclides that result in an (, n) source include oxygen, beryllium, and chlorine. The ORIGEN-S module of the SCALE6 code package is used to calculate the neutron sources.

The 226Ra sources exist either as a radium/beryllium (Ra/Be) mixture, or as radium with trace amounts of oxygen, carbon, sulfur, bromine, or chlorine (hydrous or anhydrous). Radium chloride (RaCl2) is used to bound all radium mixtures that do not include beryllium, and RaBe is treated separately. Table 5.5.5-9 of the application includes the maximum results for RaCl2 and RaBe. The neutron sources include neutrons from alpha-emitting daughter products of 226Ra.

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5.3 Shielding Model 5.3.1 Configuration of Shielding and Source 5.3.1.1 GC-3000 and GC-40 The applicant analyzed all allowable shielding and source configurations of the GC-3000 and GC-40 shielded sources. The GC-3000 transports pencil sources. The applicant assumed in its calculation that the GC-3000 source has a radius of 0.5 cm and length of approximately 24 cm, which is the length of the source capsule cavity. The GC-40 transports non-pencil sources, which are modeled as a point sources (radius of 0.1 cm to aid visualization). The applicant did not take credit for self-shielding within the source for either the GC-3000 or GC-40.

The applicant used dimensions in the MCNP models consistent with the dimensions of the engineering drawings in the application, except for the support tube under the GC-40 drawer. In the MCNP models, this tube has a thickness of 0.25 in., although the tube has been removed from the design. The drawer is now supported by a solid and square actuator rod with a side dimension of 1 in. The applicant conservatively neglected the actuator rod in the MCNP models. Replacing the support tube with the actuator rod in the MCNP models would decrease the dose rates because the actuator rod provides more shielding than the support tube. Therefore, the applicant kept the support tube in the models. For convenience, the applicant modeled all steel as stainless steel, although some items, such as the GC-3000 and GC-40 shells, are carbon steel. Each device is transported inside the inner container. The applicant did not model the IC explicitly in MCNP because it offers little axial or radial shielding, although the applicant took credit for radial placement of the device within the package.

The applicant developed models only for the GC-3000 under HAC as shown in Figure 5.5.3-10, HAC 435-B with GC-3000 MCNP Models, of the application. Under HAC, testing showed negligible deformation of the 435-B package. Therefore, the dimensions of the 435-B in the HAC model are the same as the NCT model. The foam is conservatively modeled as void in the HAC models. The applicant assumed damage to the blocking, resulting from the HAC impact, allowing the device to relocate within the IC.

Although the IC itself is not significantly damaged, any radial spacing provided by the IC is not credited, and the device is placed against the inner wall of the Model No. 435-B package.

5.3.1.2 IBL 437 Shielded Devices The applicant developed the models of the 435-B package as well as each device using the MCNP5 computer program. A lodgment specific to this device secures the IBL 437 within the 435-B cavity. The IBL 437 lodgment maintains a 10-in. clearance from the bottom of the IBL 437 to the bottom of the 435-B packages cavity. The applicant did not model the IBL 437 lodgment explicitly in MCNP because it offers little axial or radial shielding, although the applicant takes credit of the IBL 437 lodgment for placement of the device within the package.

The applicant determined the geometry of the IBL 437 Type 1 by physically cutting a device in half and measuring its dimensions. Figures 5.5.4-1 and 5.5.4-2 of the application include these measurements. The MCNP model of the IBL 437 Type 1 27

corresponds to these figures. The IBL 437 may transport up to 3 pencils, but the source is modeled as a single pencil, which increases the dose rates by concentrating the source. The source material is approximately 1.5 cm in diameter and 29 cm long. The applicant assumed in its MCNP models, a source with a radius of 0.1 cm and a height of 20 cm. The applicant did not take credit for self-shielding credit from the source material or source capsule.

The applicant noted that, in the MCNP models, the source is located within the boundary of the source holder. The applicant considered the following source orientations:

(a) Minimize Lead Shielding, Maximize Streaming Effects In the first orientation, the source is located as close as possible to the corner of the source cavity where the lead shielding is minimized. The source is shifted vertically to the top, and the locations of the screws are selected to perfectly align with the source to maximize any streaming effects (see Figures 5.5.4-13 and 5.5.4-14 of the application).

(b) Maximize Potential Streaming In the second orientation, the source is located at the bottom of the source cavity, centered in the y-direction, and as far to the left (negative x-direction) as possible within the source holder. This orientation is selected to maximize potential streaming through the steel at the drum interface in the bottom direction.

The geometry of the IBL 437 Type 2 has been determined by physically cutting a device in half and measuring its dimensions. Figures 5.5.4-3 through 5.5.4-5 of the application include these measurements. The applicant considered the following source orientations:

(a) Minimize Lead Shielding, Maximize Streaming Effects In the first orientation, the source is located near the corner of the source cavity where the lead shielding is minimized. The source is shifted vertically to the top, and the (x,y) coordinates of the source are selected to position the source directly under the steel interface between the plug and the body to maximize any potential streaming effects.

(b) Maximize Potential Streaming In the second orientation, the source is located at the bottom of the source cavity, centered in the y-direction, and as far to the left (negative x-direction) as possible within the source holder. This orientation is selected to maximize potential streaming through the steel at the drum interface in the bottom direction.

The source material is approximately 1.5 cm in diameter and 29 cm long. In the MCNP models, the source is modeled with a radius of 0.1 cm and a height of 20 cm. This is a conservative assumption because this model assumed a more concentrated source than the actual source distribution.

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For the HAC dose assessment, the Hopewell, Inc. devices are relocated to the inner surface of the Model No. 435-B containment shell. Likewise, the IBL 437 lodgment is used to secure the IBL 437 within the 435-B cavity. Because the IBL 437 lodgment is similar in design to the LTSS lodgment, it will not be significantly damaged under HAC, as demonstrated by the structural analyses.

5.3.1.3 Hopewell Inc. Shielded Device Figure 5.5.4-6 of the application depicts the design of the SC2323-GC60 transport shield. Table 5.5.4-6 of the application includes key dimensions used to develop the MCNP model. Sources are loaded into source rabbits. The 1-in. rabbits are for smaller sources and 2-in rabbits are for larger sources. The rabbits are modeled as void in the MCNP models. Up to four rabbits are loaded into a carousel insert. Sources of 60Co and 137Cs may be present in different recesses within the same carousel. MCNP models are developed with the device both at the bottom of the IC and shifted up so that the axial center of the device coincides with the axial center of the IC cavity. The applicant modeled the 60Co source with a diameter of 2.5 cm and a height of 3.5 cm, and the 137Cs-source with a diameter of 3.25 cm and a height of 6.0 cm.

Figures 5.5.4-7 through 5.5.4-12 of the application include descriptions of the five G10 Series devices, which are similar in design. Each G10 design has a range of possible lead thicknesses on both the top and side, as well as a range of source tube diameters.

In general, as the source strength increases, particularly for the 137Cs sources, the source becomes geometrically larger requiring a larger source tube diameter. For instance, while a 530 Ci 137Cs source may be transported in the G10-1-360 and the source tube diameter ranges from 1.25 to 1.75 in., a source of this magnitude would not physically fit within the 1.25-in. source tube.

The applicant states that no credit is taken for self-shielding by the source. The applicant also modeled the following:

(a) The steel source holders, which provide some axial shielding, as void.

(b) The radius of the source with a 0.2-in. clearance to the inside of the source tube.

(c) G10-1-360 and G10-2-360, the source with a maximum length of 2.1 cm.

(d) G10-1-2600, G10-2-2600, and G10-2-2600-BX devices, the source with a maximum length of 6.0 cm.

(e) Sources at the maximum length to maximize the distance between the tungsten plugs, which increases streaming from the lead plug gap on the side of the device.

Each MCNP model features either one 137Cs or 60Co source. For the two-source devices, 137Cs and 60Co are modeled in separate MCNP input files to improve model convergence.

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5.3.1.4 Disposal Canister The MCNP5 models are developed for all limiting configurations of source and shielding.

Six distinct source isotopes are considered. These source isotopes are 60Co, 137Cs, 90Sr, 192Ir, 75Se, and 226Ra. Because 226Ra may generate neutrons due to (, n) reactions, RaCl2 is used to bound 226Ra compounds not containing beryllium, while RaBe is treated separately.

The applicant states that sources may be divided into two general categories, small and pencil. Small (point) sources fit into a 1.25-in. diameter and 2.0-in. long cylindrical envelope. In the MCNP models, small sources are modeled as a cylinder with the height equal to the diameter. Pencil (line) sources fit into a 1.0-in. diameter and 18.5-in. long cylindrical envelope. Pencil sources are limited to 60Co and 137Cs. In the MCNP models, pencil sources are modeled as a cylinder with a minimum height equal to 6 in. Pencils with a shorter length are treated as small sources. For the same source isotope, small sources bound pencil sources because the source is concentrated into a more compact volume. Limited credit is taken for self-shielding provided by the source material. All sources are sealed and encapsulated.

The heavy canister is designed to transport large activity 60Co sources. 226Ra is also allowed in the heavy canister, but 137Cs, 90Sr, 192Ir, and 75Se are not allowed. There are two baskets that may be used in the heavy canister:

(a) small source basket and (b) pencil source basket.

The NLM capsules use a side spacer. The 60Co limit in the Model No. 435-B is 12,970 Ci.

The medium canister is also designed to transport 60Co sources. No other source types are allowed in the medium canister. There are three baskets that may be used in the medium canister:

(a) small source basket, (b) pencil source basket, and (c) NLM basket.

The applicant addressed each basket separately. The 60Co in the medium canister is dose-rate limited. The source is modeled as a single source, and self-shielding credit is taken for 3,500 Ci 60Co. The source is modeled at different locations within the cavity, including the top near the lid and the bottom near the drain tubing.

In the light canister models, 137Cs, 90Sr, 192Ir, and 75Se are modeled as small sources at the limiting location. Sources containing 60Co and 226Ra are not allowed. Only one basket is available for the light canister, and this basket accepts small sources, pencil sources, and NLM capsules.

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The staff verified the dimensions of the sources and the packaging used in the shielding model. The applicant showed and the staff confirmed that the source was positioned at various locations in the cavity of the package. The applicant calculated the dose rates at the distance of the package surface, at one meter from the package and 2 m from the package. Location and physical properties on the contents, used in the shielding evaluation, resulted on the maximum external radiation levels.

5.3.2 Material Properties The applicant used the following materials in the MCNP models for the Model No. 435-B:

(1) stainless steel, Table 5.5.5-20 of the application includes the composition and density of stainless steel. This stainless steel specification is a generic specification for use in shielding applications. Stainless steel is used in the Model No. 435-B shell, canister baskets, and canister drain tube/blocks.

(2) carbon steel, Table 5.5.5-21 of the application includes the composition and density of carbon steel. Carbon steel is used as the primary structural material of the disposal canisters.

(3) lead, Lead is modeled as pure with a density of 11.35 g/cm3. Lead is the primary shielding material in the disposal canisters.

(4) Aluminum, and Aluminum with a density of 2.7 g/cm3 is used in the 435-B internal impact limiter plates that form the top and bottom of the 435-B cavity. It is modeled as pure.

(5) polyurethane foam.

Polyurethane foam with a density of 14.0 pounds per cubic foot (lb/ft3)

(0.224 g/cm3) is used in the lower assembly of the 435-B. This bounds the actual density of 15 lb/ft3. The foam composition used in the models is provided in Table 5.5.5-22 of the application.

The applicant modeled the source isotopes with the materials and densities indicated in Table 5.5.5-10 of the application. Each source isotope is modeled as a pure concentration of the primary source isotope, neglecting daughter products and inert materials in the source matrix.

The staff reviewed the information provided by the applicant about the material properties used in the shielding models of the packaging and found them appropriated, since these are materials commonly used by the industry for developing this type of package.

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5.4 Shielding Evaluation 5.4.1 Methods The staff reviewed and evaluated the method used by the applicant to perform the shielding analysis. The applicant utilizes several conservative assumptions, throughout the shielding calculations, to provide assurance that the actual dose rates will be below the regulatory limits.

The applicant used MCNP5, version 1.51, for the shielding analysis.

The lodgment holding the LTSS suffered negligible damage from any of the free drops and the position of the LTSS inside the package was essentially unchanged. The LTSS did not experience any lead slump or damage to the closure doors, which ensures that the radioactive sources will stay in position relative to the lead shielding. Since there were no loadings or evidence of damage to the LTSS end door closures, the radioactive sources within the LTSS could not change their position relative to the lead shielding. Due to the similarities between the LTSS, the Disposal Canisters, and the IBL 437 payloads and their respective lodgments, these conclusions also apply to the Disposal Canisters and the IBL 437.

5.4.1.1 GC-3000, GC-40 Shielded Devices To analyze the dose rates, the applicant calculated the dose rates by tallying the gamma fluxes over surfaces (or volumes) of interest and converting these fluxes to dose rates.

The applicant placed the mesh tallies at the top, bottom, and side surfaces of the 435-B, as well as at 1 m from these surfaces. Because the top surface of the 435-B is curved, and the mesh tally is flat, the top surface mesh tally is placed as close as possible to the top of the package. The 1 m top tally is located at 1 m from the axial center of the 9-in.

head (i.e., the axial center of the head is approximately 4.5-in. below the top of the head) to bring this dose location closer to the package surface. The bottom mesh tally is at the bottom surface of the impact limiter, and the 1 m bottom tally is located at 1 m from this surface. The top and bottom mesh tallies are rectangular 32 x 32 grids, with mesh dimensions of 10 cm x 10 cm. Therefore, the top and bottom mesh tallies extend approximately 1 m from the side surface of the side thermal shield. This is a conservative assumption because the calculated dose rate is inside the surface of the top end of the package surface and yields a larger than the actual dose rate due to reduced shielding thickness and distance from the detector to the source.

5.4.1.2 IBL 437 and Hopewell Inc. Shielded Device The applicant noted that only one source was present per MCNP model. For devices that accept two sources such as G10-2-360, G10-2-2600, and G10-2-2600-BX, the applicant developed separate MCNP models for the source in either the upper or lower source cavity and added the results of the two models. To minimize the number of MCNP input files required for the G10 Series analysis, the applicant performed an initial run for each the models with the device shifted upward to the maximum allowed extent.

Based on these results, the applicant only ran the case with the maximum bottom dose rates with the device shifted to the bottom of the IC. Because the 435-B experiences little damage under HAC, the 1 m dose rate is about the same as the dose rate at 1 m from the surface of the package under NCT. The applicant used the NCT models that resulted in the maximum dose rates at 1 m from the top, side, and bottom of the 435-B package as the basis of the HAC analysis. In the HAC analysis, the device is allowed to shift axially and radially within the 435-B package cavity. The applicant assumed that 32

the foam at the bottom of the impact limiter burns in the fire and leaves a void, although the impact limiter foam offers little shielding against gamma radiation.

5.4.1.3 Disposal Canisters The applicant performed the analysis for six source isotopes and three canister designs.

Tables 5.5.5-17 through 5.5.5-19 of the application includes a summary of the general configurations considered in the shielding analysis of the disposal canisters.

According to the applicant, each configuration of source and shielding may or may not be limited by dose rate. If the source/shielding configuration is not limited by dose rate, the applicant modeled the source at the maximum activity, depicted in Table 5.5.5-1 of the application, in a double package configuration. The goal of this configuration is to demonstrate that the package will not exceed the dose rate regulatory limits of 10 CFR 71 for exclusive transport.

The applicant notes that for some configurations, both single and double package, the maximum activities included in Table 5.5.5-1 of the application could result in dose rates that exceed the regulatory dose rate limits for exclusive use transportation. Therefore, the maximum activities for these configurations are going to be limited by dose rate.

These limits are conservatively reduced to 95% of the regulatory values. The activities that result in dose rates 95% of the regulatory limits are then:

Equation 1 Apackage = Mx950/Dmax_package_surface Equation 2 Avehicle = Mx190/Dmax_vehicle_surface Equation 3 A2m = Mx9.5/Dmax_2m Equation 4 Aoccupied = Mx1.9/Dmax_occupied

where, Dmax is the maximum dose rate at the specified location (computed by MCNP) and M is the activity of the source in the MCNP input file. For the dose rate limited isotopes, M = 1 Ci. The activity limit Ai = MIN (Apackage, Avehicle, A2m, Aoccupied). In all cases, either Avehicle or A2m is limiting.

The applicant calculated the Ai values for dose rate limited and non-dose rate limited configurations and used these to develop a sum of fractions rule when mixing 226Ra with and without beryllium in the heavy canister. For dose rate limited isotopes, Ai is the true maximum allowed activity in the canister. For non-dose rate limited isotopes, Ai represents the theoretical activity at which the dose rate would be at 95% of the regulatory limits without regard to heat load limits or administrative limits. The Ai value could be large for low energy gamma sources that are heavily shielded (e.g., 192Ir).

33

To capture the maximum dose rate, the applicant modeled each configuration of the source in several different locations within the canisters cavity. For each of the three canister designs, the region of weakest shielding is at the top (near the lid). The top region features steel structural components that penetrate the lead shielding, axial and radial gaps due to lead shrinkage, and a radial lid clearance gap between the lid and the canister body.

The heavy disposal canister is used only to transport 60Co and 226Ra. Devices or sources containing 60Co and 226Ra are not allowed to be present in the same canister.

The small source is modeled with an activity of 12,970 Ci. The applicant modeled 226Ra in two forms, RaCl2 and RaBe. The RaBe source is modeled as a single small source, in the most limiting configuration, and dose rate limited. Due to the strong neutron source, the maximum RaBe activity is determined so that 95% of the dose rate limit is achieved.

The applicant calculated and added the primary gamma and neutron dose rates and only considered a single small source, as pencil sources are not available for this isotope. Because RaBe is dose rate limited in the heavy canister, the RaBe sources are modeled with a source strength of 1 Ci per Table 5.5.5-10 of the application.

The medium disposal canister is used only to ship 60Co sealed sources. The applicant modeled all configurations for the medium disposal canister as dose rate limited.

The light disposal canister is used only to ship 137Cs, 90Sr, 192Ir, and 75Se. None of these isotopes are dose rate limited. In an actual shipping configuration, a Disposal Canister will likely contain many source isotopes distributed throughout the volume of the basket.

However, the applicant modeled the source as a single small (point) source as the bounding source in the limiting location. The limiting location is generally near the interface of the lid with the side of the canister cavity due to streaming through the steel structural members. This configuration bounds the more typical distributed source. For 137Cs, the small source configuration also bounds pencil (line) sources (pencil sources are not available for 90Sr, 192Ir, and 75Se).

In MCNP, the applicant used mesh tallies to calculate the dose rates. Mesh tallies are placed at the top, bottom, and side surfaces of the 435-B. The top and bottom 435-B package surface tallies are used for the top and bottom of the vehicle. Separate tallies are used for the vehicle side, 2 m from the vehicle side, and the occupied location.

When developing the mesh tallies, the applicant also assumed the following:

(a) the width of the vehicle is along the x-axis, (b) the length of the vehicle is along the y-axis, and (c) the height of the package/vehicle is along the z-axis.

The same mesh tallies are used in the single and double package models.

The staff verified the dimensions of the sources and packaging used in the shielding model and found them acceptable because the applicant positioned the contents at various locations. The staff also found that the locations and physical properties of the contents used in the evaluation are those resulting in the maximum external radiation levels. The staff examined all changes in configurations under NCT and HAC and found them appropriate. The staff verified that the dose points shown in Sections 5.5.3, 5.5.4, 34

and 5.5.5 of the application were chosen to identify the location of the maximum radiation levels. The staff finds the Disposal Canisters appropriate to be transportable in the 435-B package.

The applicant provided sample ORIGEN-S and MCNP input files in Appendix 5.5.2, Sample Input Files for LTSS Evaluation, of the application. For Shielded Devices, Appendix 5.5.3.5.2 of the application includes a sample input file.

The applicant used accelerate model convergence (the importance of the cells is increased radially and axially through the shield). This produces an accurate and converged flux distribution in all regions of the model. Problem convergence is accelerated by dividing the device into layers and splitting the particles as particles traverse outwardly through these layers.

The Monte Carlo uncertainty associated with the limiting dose rate is less than 5% as shown in the output file provided by in the application.

For IBL 437 and Hopewell devices, Appendix 5.5.4.5.2 of the application includes a sample input file. The Monte Carlo uncertainties associated with the limiting dose rate are less than 5%

for most cases, with a few in the range of 5-8%. This is acceptable because according to NUREG-CR-6802, Recommendations for Shielding Evaluations for Transport and Storage Packages, (NUREG-CR-6802) recommended detector statistics are less than 5-10% standard deviation for a point, surface, or volume detector. (For Disposal Canisters, Appendix 5.5.5.5.2 of the application includes a sample output file.)

The staff reran the input file using MCNP6 to verify that the results from the Monte Carlo uncertainty associated with the limiting dose rate location resulted to be less than 2%. The results from the reran showed that the uncertainty associated with the limiting dose rate location The staff verified that Sections 5.5.3, 5.5.4, and 5.5.5 of the application includes key input data needed for the shielding calculations. The application includes a representative input file used in the analyses. The staff examined the input file to verify that voids, streaming paths, and irregular geometries are included in the model.

5.4.2 Flux-to-Dose Rate Conversion Table 5.4-1 of the application includes ANSI/ANS-6.1.1-1977 flux-to-dose rate conversion factors used in this analysis.

The staff verified that the accuracy of the flux-to-dose rate conversion factors, which were tabulated as a function of energy group structure and finds it to be acceptable based on the guidance provided in NUREG-1609, Standard Review Plan for Transportation Packages for Radioactive Material.

5.4.3 External Radiation Levels As described in Section 5.5.3.4.1 of the application, the applicant calculated the dose rates at the surface and at 1 m from the surface of the 435-B package using mesh tallies. For non-exclusive use transportation, the dose rate is limited to 200 mrem/hr on the surface of the package and 10 mrem/hr at a distance of 1 m from the surface of the package. For Type B package, the dose rate limit for package under HAC is 1,000 mrem/hr at 1 m from the surface.

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Tables 5.5.3-4 and 5.5.3-5 of the application includes a summary of the limiting dose rate results for the GC-3000 and GC-40, respectively. Table 5.5.3-5 includes the results for the GC-40 device in a vertical and horizontal orientation. Table 5.5.3-5 of the application indicates that the side dose rates for a vertically-oriented device are applied to the top and bottom of the package, when the device is transported horizontally. The staff found this acceptable because the dose rate limit for non-exclusive use package is at any point of the external surface of the package per 10 CFR 71.47(a) regardless of the position of the package. For both the GC-3000 and GC-40, the dose rates are far below the limits of 200 mrem/hr on the surface and 10 mrem/hr at a distance of 1 m from the surface.

Table 5.5.4-18 of the application includes a summary of the results for the IBL 437 Types 1 and

2. The magnitude of the dose rate changes with source location, although in all cases the dose rates remain well below the limits of 200 mrem/hr on the surface and 10 mrem/hr at a distance of 1 m from the surface. The maximum IBL 437 package surface and 1 m dose rates occur for the Type 2 when the source is in the upper-right location, directly under the top plug streaming path, with dose rates of 1.65 mrem/hr and 0.56 mrem/hr, respectively.

Table 5.5.5-17 of the application includes a summary of the heavy disposal canister configurations. Calculations are performed for the small source basket and NLM capsule. The small source basket bounds the pencil source basket. The Ai results for the small source basket with a 60Co sources are summarized in Table 5.5.5-23 of the application. The applicant performed calculations for 226Ra sources. 226Ra with beryllium (RaBe) and 226Ra without beryllium (chlorine target, or RaCl2) are addressed separately. The dose rate from RaBe and RaCl2 is due to neutrons and gamma. The 226Ra activity limit per package is 20 Ci. The Ai values provided in Table 5.5.5-25 of the application indicate that RaBe is dose rate limited, while RaCl2 is not dose rate limited.

The staff confirmed that the external radiation levels under NCT and HAC agree with the summary tables as discussed in Sections 5.5.3.4.4, 5.5.4.4.4, and 5.5.5.4.4 of the application.

The staff verified that the analysis shows the locations selected are those of maximum dose rates. Also, the staff verified that the external radiation levels meet the regulatory requirements and their variation with location are consistent with the geometry and shielding characteristics of the package.

5.5 Evaluation Findings

The staff reviewed the documentation provided by the applicant, including the maximum dose rates for NCT and HAC, to verify that the statements presented by the applicant were accurate, within acceptable engineering practices, and in compliance with shielding requirements in 10 CFR Part 71. As part of its review, the staff also performed confirmatory analyses on source terms using the ORIGEN-S in SCALE 6.1 depletion code. The staff evaluated the adequacy of the description, methods, and analyses of the package design bases related to the shielding evaluation of the 435-B package and found them acceptable. The staff also determined that the reported dose values were below the regulatory limits in 10 CFR 71.47 and 71.51.

Based on the review of the statements, representations, and supplemental information to the application, the staff concludes that the shielding design for the proposed GC-3000, GC-40, IBL 437 and Hopewell shielded devices, and Disposal canisters have been adequately described and evaluated. Therefore, the package has adequate structural integrity to meet the external radiation requirements of 10 CFR Part 71.

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6.0 CRITICALITY The changes requested by the applicant did not impact the previous criticality review findings.

Therefore, the staff did not perform a criticality review.

7.0 PACKAGE OPERATIONS The purpose of this evaluation is to verify that the operating controls and procedures of the Model No. 435-B transport package, loaded with the Disposal Canister Payload, the IBL 437 Payload, or the Hopewell Device Payload, meet the requirements of 10 CFR Part 71.

Besides proposing the addition of payloads to the CoC for the Model No. 435-B and changing the version of ANSI N 14.5 from 1997 to 2014, the applicant proposed the following changes relate to the GC-40 shielded device:

1) Ship the device horizontally, which makes it easier to handle and to emplace into the IC.

2) Do not require to cut off the lower base weldment, which reduces the workers exposure.

3) Due to security upgrades that have been made to GC-40 units in the field, it is no longer possible to place the shipping fixture into the device.

The shipping fixture is not necessary to ensure that the source remains shielded during shipping. Section 5.0 of this SER includes the shielding analysis for the package.

7.1 Package Loading Chapter 7.0 of the application includes loading and unloading instructions depending on the type of payload. There are three new types of payloads in the 435-B:

a. Disposal Canisters,

b. IBL 437 shielded devices, and

c. Hopewell shielded devices.

The applicant revised Chapter 7.0 of the application to include procedures related to the new payloads. Disposal canisters and IBL 437 shielded device are transported with a lodgment as specified in drawing Nos. 1916-01-04-SAR and 1916-01-05-SAR, respectively. Hopewell devices are transported in the IC.

Regardless of the payload, the loading instructions include the maximum weight of source devices, heat load per isotope, and source capacity that can be loaded in the Model No. 435-B package. Additionally, the instructions require visual inspections for identifying and addressing damage that could impair the integrity of the containment or foreign material in the cavity of the packaging. Foreign material must be removed prior to shipment and damage repaired. The 37

package has a tamper-indicating lockwire on two adjacent rain shield bolts located on the same shield half.

The applicant proposed to make the rain shield bolts with security hears in order to enhance the security of the package. All other characteristics and operational requirements for the rain shield bolts remained unchanged. The applicant removed the designation as socket head cap screws of the security bolts and clarified at least two of them are cross-drilled for the tamper-indicating lockwire.

The loading instructions for the Model No. 435-B must be provided by the applicant. Also, the package loading instructions includes torque values for components of the package.

7.2 Loading of Contents 7.2.1 Disposal Canisters The three types of Disposal Canisters and the IBL 437 shielded device(s) are transported using lodgments. The Disposal Canister lodgment must be properly adjusted for the canister type to be transported. Section 7.1.5, Loading and Preparing the Disposal Canisters for Transport, of the application includes a description about loading the Disposal Canisters. Different types of basket are used to accommodate small sources, pencil sources, and NLM capsules. Sources shall be loaded with the appropriate basket/spacer/divider corresponding to the source type.

Tables 7.1-6 to 7.1-8 include the activity limits for isotopes transported in the Disposal Canisters. Table 7.1-9 of the application includes the different basket configurations for the Disposal Canisters. Disposal canisters are transported as exclusive use shipments.

7.2.2 IBL 437 Shielded Device The IBL 437 Shielded Device is loaded with a lodgment. Section 7.1.2.4.2 includes a description of the process to load the IBL 437 into the 435-B packaging. Visual inspections are performed to ensure that the integrity of the package is maintained prior to shipment.

7.2.3 Hopewell Shielded Devices Hopewell devices are transported using the IC. During loading, the shipper shall ensure that the cavity of the inner container is clean, dry, free of foreign material, and protect it from entry of precipitation.

7.3 Preparation for Transport As part of the preparation for transport, the package is checked to ensure that all conditions of the CoC are met and that the tamper-indicating lockwire is in place. Section 7.1.4 of the application includes the process for preparing the 435-B package for transport. The procedures also require compliance with 49 CFR 172, Hazardous Materials Table, Special Provisions, Hazardous Materials Communications, Emergency Response Information, Training Requirements, and Security Plans, and 49 CFR 173, ShippersGeneral Requirements for Shipments and Packagings.

Also prior to transport, the shipper must perform a pre-shipment leakage rate test to confirm proper assembly of the package and demonstrate containment integrity. The acceptance criterion for the pre-shipment leakage rate test is no detected leakage when tested to a 38

sensitivity of less than or equal to 1 x 10-3 ref-cm3/s air, per ANSI N14.5, Section 7.6, Pre-shipment leakage rate test, 2014 (ANSI, 2014).

The applicant proposed to use the more recent 2014 version of ANSI N14.5 instead on the 1997 version. Section 4.4 and 8.2 of this SER includes additional information about the staffs evaluation of this change.

7.3.1 Preparation for Transport - Inner Container The applicant revised procedures related to the IC, to include aspects related to the Hopewell shielded devices. The radiation levels at the surface of the entire device cannot exceed 200 mrem/hr at the surface and 1,000 mrem/hr at 1 meter from the surface. Specifically, the procedures highlight that failure to meet these requirements disqualifies the device for transport.

Presence of defects that can significantly decrease the structural or shielding integrity of the device would also disqualify the device for transport. Sections 7.1.2.2.1 and 7.1.2.2.2 of the application includes the main steps for preparing the inner container devices for transport.

7.4 Package Unloading Section 7.2.1.1 of the application includes the unloading procedures for the Disposal Canisters and IBL 437 shielded device payloads. Section 7.2.1.2 includes a description of the process to unload the inner container, which would contain the Hopewell Devices. The consignee must also follow the procedure delineated in Section 7.1.1, General Lifting and Handling, when unloading the 435-B package. When received, the condition of the tamper-indicating lockwire is checked and recorded. After tamper-indicating devices are removed, the cavity of the package must be sampled to determine if contamination is present, and the cavity equalized to atmospheric pressure before continuing unloading the contents.

7.5 Preparation of Empty Package for Transport The applicant specified in Section 7.3 of the application that it would follow the requirements of 49 CFR 173.428, Empty Class 7 (Radioactive) Materials Packaging, when preparing and transporting the 435-B package.

7.6 Evaluation Findings

Based on review of the statements and representations in the application, the staff concludes that the operating procedures meet the requirements of 10 CFR Part 71 and that these procedures are adequate to assure the package will be operated in a manner consistent with its evaluation for approval.

8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM REVIEW 8.1 Material Properties 8.1.1 Mechanical/Thermal Properties The staff notes that Revision 5 to the safety analysis report (i.e., the application) does not introduce any new structural materials important to safety or service conditions that were not previously evaluated by the staff for the Model 435-B transportation package. The new 39

material(s) (e.g., carbon steel) were introduced as part of the Disposal Canisters (heavy-,

medium-, and light-), IBL 437 large shielded device (Type 1 and Type 2), and the Hopewell Shielded Devices (including the Transport Shield) non-containment boundary components.

The staff previously evaluated the mechanical and thermal properties of these materials as part of the review of other transportation packages.

The staff notes that Sections 3.2.2, 3.3.1.2, and Table 3.2-1 of the application include a discussion of component specifications, minimum temperatures, and thermal properties, respectively. In addition, Section 3.1.3 of the application discusses the summary of temperatures. The applicant notes that heat loads of the previously-approved LTSS and the shielded device payloads bound the IBL 437 and Disposal Canister payloads.

8.1.2 Fracture Resistance Section 2.1.2.3 of the application includes a discussion of structural failure modes. The applicant points out that, with exception of the closure bolts, all structural components of the Model 435-B package are fabricated of austenitic stainless steel or aluminum. These materials do not undergo a ductile-to-brittle transition in the temperature range of interest (i.e., down to -40 ºF), and, therefore, these do not need to be evaluated for brittle fracture.

The staff notes that the structural components added to the Model No. 435-B package (i.e., the Disposal Canister and IBL 437 lodgments) are fabricated from aluminum and do not undergo a ductile-to-brittle transition in the temperature range of interest (i.e., down to -40 ºF). In addition, the staff finds that brittle fracture for the added payload containers (i.e., Disposal Canisters, IBL 437 and shielded devices) is not a failure mode of concern because they are non-containment boundary systems, structures, and components, and are fabricated from structural carbon steel intended for improved notch toughness in low temperature service. The staff concludes that the added susceptibility to brittle fracture of the Model No. 435-B packages structures, systems, and components is acceptable based on the discussion above.

8.1.3 Corrosion and Chemical Reactions Section 2.2.2 of the application discusses chemical, galvanic, and other reactions. The applicant states that the materials of construction of the 435-B package will not have significant chemical, galvanic, or other reactions in air or water environments. In addition, these materials have been previously used, without incident, in radioactive material packages for transporting similar payload materials such as the RH-TRU 72-B (Docket No. 71-9212) and the BEA Research Reactor (BRR) package (Docket No. 71-9341). The applicant noted that a successful radioactive material packaging history combined with successful use of these fabrication materials, in similar industrial environments, ensures that chemical, galvanic, or other type of reactions do not compromise the integrity of the 435-B package.

The applicant also noted that the lead gamma shielding in the Disposal Canisters and shielded devices is fully encased in a carbon steel or stainless steel weldment and cannot be affected by water or atmospheric moisture. In addition, the Disposal Canisters and IBL 437 shielded devices rest on an aluminum plate covered with a layer of neoprene rubber attached to the plate using multiple screws. Per the applicant, there is nominally no contact between lodgment ribs and the canisters.

The staff reviewed the design drawings and applicable sections of the application to evaluate the effects, if any, of intimate contact between various materials of construction related to the 40

Model No. 435-B package during all phases of operation. The staff evaluated whether these contacts could initiate a chemical or galvanic reaction that could result in corrosion or combustible gas generation that could adversely affect safety. A review of the Model No. 435-B package, its contents, and its operating environments has been performed to confirm that no operation will produce adverse chemical or galvanic reactions. The staff notes that the 435-B internals will not be subject to moisture during normal operation. Further, visual inspections are to be performed of the payload cavity prior to loading and following off-loading and provide reasonable assurance against any considerable corrosion occurring unnoticed. The Model No. 435-B package and packages constructed of similar materials (e.g., aluminum, stainless steel, and coated carbon steel) have been previously approved and successfully transported.

Based on the above discussion, the staff does not expect significant material interactions or galvanic reactions in the 435-B package.

8.1.4 Protective Coatings Section 3.4.3.7 and Table 3.4-4 of the application include a discussion about the behavior of non-metallic contents materials temperature limits and sources. The applicant notes that paint is used on the outer surfaces of most shielded devices and on the Disposal Canisters for corrosion protection. In addition, the Disposal Canisters shell temperature of 278°F bounds the temperature of the paint. The applicant also noted that thermogravimetric analysis (TGA) for various paint types show that significant degradation, measured as weight loss, does not occur below 200°C to 300°C (392°F to 572°F, respectively). The minimum margin of safety is 114°F.

The staff reviewed the applicants use of protective coatings and finds them to be acceptable based on their ability to prevent oxidation and withstand the maximum normal service temperatures.

8.1.5 Radiation Effects Section 2.2.3 of the application includes a discussion about the effects of radiation on materials.

The applicant states that the radiation associated with the source payload will have no effect on the containment or other safety components comprising the Model No. 435-B package. In addition, since the payload of the Model No. 435-B package is heavily shielded, the radiation exposure of the package materials (including the butyl rubber containment seal) is negligible.

The applicant states for the above reasons, there will be no deleterious radiation effects on the packaging.

The staff notes that the lead and carbon steel shells of the Model No. 435-B payload canisters provide shielding between the Model No. 435-B shielded sources and the exterior surface for the attenuation of gamma radiation. Further, the radiation associated with the decay of shielded sources will have no effect on the stainless steel and aluminum comprising the primary structural components of the 435-B package during transportation. In addition, since the payload of the 435-B package is heavily shielded, the radiation exposure of the materials is negligible and the containment seal, which is also located outside of the gamma shielding, likewise receives a negligible exposure. The staff finds, based on the above reasons, there will be no deleterious radiation effects on the packaging, and the requirements of 10 CFR 71.43(d) are met.

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8.2 Leakage Rate Tests for Type B Packages The staff verified that the applicant would require personnel with an American Society for Nondestructive Testing (ASNT), nondestructive testing (NDT), Level III certification to develop and approve leakage rate testing procedures related to the Model No. 435-B, as described in Sections 8.5, Leakage rate testing, and 8.8, Quality assurance, of ANSI N14.5-2014. Based on the staffs review of those Sections of ANSI N14.5-2014 and the response to RAI-Co-1 (NNSA, 2019) that described that the requirement in the application to follow ANSI N14.5-2014 includes the requirement that leakage rate testing procedures must be approved by personnel certified as an ASNT NDT Level III, the staff finds the applicants response to be acceptable.

The staff also verified the leakage rate testing would be performed by personnel who are qualified and certified in accordance with ASNT Recommended Practice No. SNT-TC-1A, Personnel Qualification and Certification in Nondestructive Testing, (ASNT, 2006) as described in Sections 8.5, Leakage rate testing, and 8.8, Quality assurance, of ANSI N14.5-2014. Based on the staffs review of those Sections of ANSI N14.5-2014 and the response to RAI-Co-1 (NNSA, 2019) (that described that the requirement in the application to follow ANSI N14.5-2014 includes the requirement that leakage rate testing must be performed by personal qualified and certified in accordance with ASNT Recommended Practice No. SNT-TC 1A), the staff finds the applicants response to be acceptable. Sections 4.4 and 7.3 of this SER include additional discussion on the leakage rate testing for the Model No. 435-B.

8.3 Evaluation Findings

Based on review of the statements and representations in the application, the staff concludes that the acceptance tests for the packaging meet the requirements of 10 CFR Part 71, and that the maintenance program is adequate to assure packaging performance during its service life.

9.0 QUALITY ASSURANCE The applicant is an organization under the U.S. Department of Energy (DOE). The Los Alamos National Laboratory (LANL) Operations Support-Packaging and Transportation (OS-PT) organization provides quality assurance (QA) oversight and AREVA provides technical support for the 435-B package. The DOE Order 460.1C, Packaging and Transportation Safety, contains the QA requirements for using NRC-certified packagings. The 435-B will be under the LANL SD330, Los Alamos National Laboratory Quality Assurance Program. Regarding the 435-B package technical services (i.e., licensing documentation, design, and certification expertise), the staff notes a company name change from AREVA Federal Services to ORANO Federal Services, with continued QA oversight by the LANL Operations Support-Packaging and Transportation (OS-PT) organization. The staff notes that ORANO has an established QA program compliant to 10 CFR 71, Subpart H, and DOE Order 414.ID and is documented on the LANL Institutional Evaluated Suppliers List (IESL).

Table 9.2-1 of the application includes a cross-map of the QA Program requirements of 10 CFR Part 71, Subpart H; LANL QA Program; and ORANO Federal Services QA Program. The packages users would adhere to their QA programs. As this package is originally certified in the United States, it meets the NRC QA requirements. The staff notes a component category change from LTSS Lodgment to lodgment in order to encompass two new lodgments. The staff notes that the change did not impact the quality categories of the Subcomponent as categorized in Table 9.2-2, QA Categories for Design and Procurement of 435-B Subcomponents, of the application.

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REFERENCES (ANSI, 2014) American National Standards Institute, ANSI N14.5-2014, American National Standard for Radioactive Materials -

Leakage Tests on Packages for Shipment, ANSI, New York, NY.

(ASNT, 2006) American Society for Nondestructive Testing, Recommended Practice No. SNT-TC-1A, Personnel Qualification and Certification in Nondestructive Testing, ASNT, Columbus, OH.

(AWS D1.6) American Welding Society, AWS D1.6, Structural Welding Code - Stainless Steel, AWS\ANSI, Danvers, MA.

(AWS D1.2) American Welding Society, AWS D1.2, Structural Welding Code - Aluminum, AWS\ANSI, Danvers, MA.

(NNSA, 2018a) Rahimi, Meraj, U.S. Nuclear Regulatory Commission (NRC) letter to Al-Daouk, Ahmad M., National Nuclear Security Administration (NNSA), January 26, 2018, ADAMS Package Accession No. ML18026A874.

(NNSA, 2018b) Al-Daouk, Ahmad M., National Nuclear Security Administration (NNSA), letter to Michael Layton, U.S. Nuclear Regulatory Commission (NRC) (Attn: Document Control Desk), November 19, 2018, ADAMS Accession No. ML19038A112.

(NNSA, 2019a) Al-Daouk, Ahmad M., National Nuclear Security Administration (NNSA), letter to U.S. Nuclear Regulatory Commission (NRC)

(Attn: Document Control Desk), April 11, 2019, ADAMS Accession No. ML19227A299.

(NNSA, 2019b) Al-Daouk, Ahmad M., National Nuclear Security Administration (NNSA), letter to U.S. Nuclear Regulatory Commission (NRC)

(Attn: Document Control Desk), November 1, 2019, ADAMS Accession No. ML19305A276.

(NUREG-1609) U.S. Nuclear Regulatory Commission, Standard Review Plan for Transportation Packages for Radioactive Material, NUREG-1609, March 1999.

(NUREG/CR-3019) U.S. Nuclear Regulatory Commission, Recommended Welding Criteria for Use in the Fabrication of Shipping Containers for Radioactive Materials, NUREG/CR-3019, UCRL-53044, March 1984.

(NUREG/CR-3854) U.S. Nuclear Regulatory Commission, Fabrication Criteria for Shipping Containers, NUREG/CR-3854, UCRL-53544, March 1985.

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(NUREG/CR-6802) U.S. Nuclear Regulatory Commission, Recommendations for Shielding Evaluations for Transport and Storage Packages, NUREG/CR-6802, May 2003, ADAMS Accession No. ML031330514.

(Williamson and Iams, 2004) Williamson, C., and Iams, Z., Thermal Assault and Polyurethane Foam - Evaluating Protective Mechanisms, (General Plastics Manufacturing Company, Tacoma, WA),

PATRAM International Symposium, Berlin, Germany, 2004.

CONDITIONS The certificate of compliance includes the following condition(s) of approval:

1) Condition No. 3.b., Title and Identification of Report or Application, includes the date of the consolidated application.

2) Condition No. 5.(a)(2)(iii) is revised as follows to include the new authorized contents:

Internal lodgments, made of aluminum, which support the Long Term Storage Shield (LTSS), Disposal Canisters, or the IBL 437 (aka IBL 437C) Shielded Device (Type 1 or Type 2)

3) Condition No. 5.(a)(2)(v) is revised as follows to include the new authorized contents:

An inner container, which supports shielded devices (including Hopewell Designs, Inc. Shielded Devices and Transport Shield)

4) Add Disposal Canisters in the last paragraph of Condition No. 5.(a)(2).

5) Condition No. 5.(a)(3), Drawings, contains the latest revision of the licensing drawings that the package must be fabricated to as well as the new drawings related to the proposed contents.

6) Condition No. 5.(b)(1), Type and Form of Material, contains the approved sources that can be shipped in the 435-B package. This condition was revised to include all tables including the description and activities of the sources.

7) Condition No. 5.(b)(2)(i), LTSS, was revised to add 226Ra sources to Table 1 as authorized contents of the LTSS and revised the notes to the table to be in alignment with the request in the application.

8) Condition No. 5.(b)(2)(ii), Inner Container-Shielded Devices, was revised to add the following information:

a. the IBL 437 (aka IBL 437C) device to Table 3 as authorized contents to be transported with the inner container, and

b. Table 4, Hopewell Designs, Inc. Shielded Devices and Transport Shield 44

9) Condition No. 5.(b)(2)(iii), Disposal canisters, was added to include Table 5, Disposal Canisters Source Nuclides.

10) Condition No. 5.(b)(3), Maximum Weight of Contents, was revised to reflect the current the weight of the package, the LTSS, IBL 437 (aka IBL 437C), Disposal Canisters, and Honeywell Inc. Shielded Devices (Type 1 and Type 2) that are allowed to be shipped in the Model No. 435-B.

11) Condition No. 5(b)(4), Maximum decay heat, consolidated the maximum decay heat for the packages authorized payloads into Table 8.

12) Condition 10 was removed, since the certificate was renewed. This caused the previous Condition No. 11 to be renumbered as Condition No. 10.

13) The certificate was renewed for 5 years, therefore, the expiration date of the certificate was changed to February 28, 2025.

The staff also made some editorial changes to the certificate. The References section includes the consolidated application provided as part of the review process.

CONCLUSIONS Based on the statements and representations contained in the application, as supplemented, and the conditions listed above, the staff concludes that the design has been adequately described and evaluated, and the Model No. 435-B package meets the requirements of 10 CFR Part 71.

Issued with Certificate of Compliance No. 435-B, Revision 3 on 03/12/20.

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