ML18208A260

Letter to R. Boyle Certificate of Approval No. CDN/2094/B(U)-96, Revision 1, for the Model No. F-522 Package Dkt. No. 71-3091 (W/Enclosure 1)

ML18208A260
Person / Time
Site:	07103091
Issue date:	07/26/2018
From:	John Mckirgan Spent Fuel Licensing Branch
To:	Boyle R US Dept of Transportation (DOT)
Garcia-Santos N
Shared Package
ML18208A259	List: ML18208A260
References
CAC L25212, EPID: L-2017-LLA-0022
Download: ML18208A260 (30)

Text

Official Use Only-Security-Related Information UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 26, 2018 Mr. Richard W. Boyle Radioactive Materials Branch U.S. Department of Transportation 400 Seventh Street, S.W.

Washington, D.C. 20590

SUBJECT:

CERTIFICATE OF APPROVAL NO. CDN/2094/B(U)-96, REVISION 1, FOR THE MODEL NO. F-522 PACKAGE (DOCKET NO. 71-3091)

Dear Mr. Boyle:

This is in response to your letter dated March 15, 2017 [Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML18137A496], and as supplemented on September 20, 2017 (ADAMS Accession No. ML18151A700), November 16, 2017 (ADAMS Package Accession No. ML18143B680), January 16, 2018 (ADAMS Package Accession No. ML18151A695), April 4, 2018 (ADAMS Accession Nos. ML18143B343 and ML18151A701), and May 30, 2018 (ADAMS Package Accession No. ML18151A695), requesting our assistance in evaluating the Model No. F-522 transport package, authorized by Canadian Certificate of Approval No. CDN/2094/B(U)-96, Revision 1.

Based upon our review, the statements and representations contained in the application and its supplements, and for the reasons stated in the enclosed Safety Evaluation Report, we recommend revalidation of the Canadian Certificate of Approval No. CDN/2094/B(U)-96, Revision 1, for the Model No. F-522 package, with the following additional conditions:

(A) Shielding The impurities within the 99Mo fission product content must meet one of the three impurity profiles in document No. R119.017.SUR[1], Shielding Analysis Report for F522 Mo-99 Impurities, Table 6. (Charbonneau et al., 2018, and DOT, 2018a)

Upon removal of Enclosure 3, this document is uncontrolled Official Use Only-Security-Related Information

Official Use Only-Security-Related Information R. Boyle If you have any questions regarding this matter, please contact me or Norma García Santos of my staff at (301) 415-6999.

Sincerely,

/RA Bernard White Acting for/

John McKirgan, Branch Chief Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-3091 CAC No. L25212 EPID L-2017-LLA-0022

Enclosures:

1. Safety Evaluation Report

2. Certificate from Competent Authority (Redacted)

3. Certificate from Competent Authority (Official Use Only-Security-Related Information)

Official Use Only-Security-Related Information

Official Use Only-Security-Related Information R. Boyle

SUBJECT:

CERTIFICATE OF APPROVAL NO. CDN/2094/B(U)-96, REVISION 1, FOR THE MODEL NO. F-522 PACKAGE (DOCKET NO. 71-3091), DOCUMENT DATE:

July 26, 2018 Distribution: SFM r/f, DMarcano http://fusion.nrc.gov/nmss/team/sfst/sfst-licensing/10_cfr_71/f-522_reval/Shared Documents/F-522 SER_7-2018_2_SF_NGS_CLEAN.docx This closes CAC No. L25212 and EPID L-2017-LLA-0022.

ADAMS Package Accession No.: ML18208A259 LTR: ML18208A260 OFFICE: DSFM DSFM DSFM DSFM DSFM NAME: NGarcía Santos SFigueroa JChang YKim VWilson by email by email by email by email DATE: 7/6/18 7/9/18 7/12/18 7/12/18 7/16/18 OFFICE: DSFM DSFM DSFM DSFM DSFM NAME: TAhn by email YDíaz Sanabria TTate MRahimi BWhite for by email by email JMcKirgan DATE: 7/10/18 7/23/18 7/18/18 7/26/18 7/26/18 OFFICIAL RECORD COPY Official Use Only-Security-Related Information

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3091 Model No. F-522 Package Certificate of Approval No. CDN/2094/B(U)

Revision 1 Enclosure 1

Table of Contents Page

SUMMARY

................................................................................................................................... 1 1.0 GENERAL INFORMATION ............................................................................................ 1 1.1 Package Description....................................................................................................... 1 1.1.1 Packaging ....................................................................................................................... 1 1.1.2 Contents ......................................................................................................................... 2 2.0 STRUCTURAL EVALUATION ....................................................................................... 3 2.1 Description of Structural Design ..................................................................................... 3 2.2 Mechanical Analysis ....................................................................................................... 4 2.2.1 Stacking Test .................................................................................................................. 4 2.2.2 Lifting Test ...................................................................................................................... 4 2.2.3 Penetration Test ............................................................................................................. 4 2.2.4 Tie-Down Test ................................................................................................................ 5 2.3 Structural Evaluation under Normal Conditions of Transport and Hypothetical Accident Conditions ..................................................................................................................................... 5 2.3.1 Drop Tests ...................................................................................................................... 5 2.3.2 Water Immersion Test .................................................................................................... 6 2.4 Evaluation Findings ........................................................................................................ 6 3.0 MATERIALS EVALUATION .......................................................................................... 6 3.1 Packaging Materials ....................................................................................................... 6 3.2 Fabrication/Welding ........................................................................................................ 6 3.3 Chemical, Galvanic, or Other Reactions ........................................................................ 7 3.3.1 Radiolysis ....................................................................................................................... 7 3.3.2 Pyrophoricity ................................................................................................................... 8 3.3.3 Stress Corrosion Cracking ............................................................................................ 10 3.4 Thermal Properties and Effects on Materials ............................................................... 10 3.4.1 O-Rings ........................................................................................................................ 10 3.4.2 Foam ............................................................................................................................ 10 3.5 Evaluation Findings ...................................................................................................... 10 4.0 THERMAL EVALUATION ............................................................................................ 11 4.1 Description of the Thermal Design ............................................................................... 11 4.2 Thermal Evaluation under Normal Conditions of Transport and Hypothetical Accident Conditions ................................................................................................................................... 12 4.2.1 Thermal Evaluation under Normal Conditions of Transport ......................................... 12 4.2.2 Thermal Evaluation under Hypothetical Accident Conditions ....................................... 14 4.2.3 Maximum Normal Operating Pressure ......................................................................... 15 4.3 Evaluation Findings ...................................................................................................... 16 5.0 CONTAINMENT EVALUATION ................................................................................... 16 5.1 Description of the Containment Boundary .................................................................... 16 5.2 Helium Leakage Rate Testing ...................................................................................... 16 5.2.1 Special Form Sealed Source ........................................................................................ 16 5.2.2 F-248 Leak Proof Insert ................................................................................................ 17 5.3 Evaluation Findings ...................................................................................................... 17 6.0 SHIELDING .................................................................................................................. 18 6.1 Description of the Shielding Design .............................................................................. 18 6.2 Evaluation Method ........................................................................................................ 18 6.2.1 Package Modeling ........................................................................................................ 19

ii 6.2.2 Source Modeling ........................................................................................................... 20 6.2.3 Flux-to-Dose Rate Conversion Factors ........................................................................ 21 6.3 Hypothetical Accident Conditions ................................................................................. 22 6.4 Confirmatory Evaluations ............................................................................................. 22 6.5 Evaluation Findings ...................................................................................................... 22 7.0 CONDITIONS ............................................................................................................... 23

8.0 REFERENCES

............................................................................................................. 23 CONCLUSION ............................................................................................................................ 24 List of Tables Table 1.1. Radioactive contents of the Model No. F-522 transportation package. ...................... 2 Table 3.1. Temperature limits of seals used with the Model No. F-522. .................................... 10 Table 4.1. Heat load limit per Model No. F-522 allowable content. ........................................... 11 Table 4.2. Load cases analyzed under NCT using 99Mo as content. ......................................... 13 Table 4.3. Maximum temperatures of the package, containment O-rings, and outer surface of the package under NCT for the cases described in Table 4.2 of this SER. ................................ 14 Table 4.4. Load cases under HAC: maximum temperatures of containment O-rings and depleted uranium. ....................................................................................................................... 14 Table 5.1. Leak testing requirements for the main components of the Model No. F-522. ......... 17 Table 6.1. Applicable IAEA, SSR-6, 2012 Edition, shielding requirements to evaluate the adequacy of a package design [Type B(U) Package]. ................................................................ 18

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Docket No. 71-3091 Model No. F-522 Package Certificate of Approval No. CDN/2094/B(U)-96 Revision 1

SUMMARY

By letter dated March 15, 2017 (DOT, 2017), and as supplemented on September 20, 2017 (Nordion, 2017a), November 16, 2017 (Nordion, 2017b), January 16, 2018 (Nordion, 2018),

April 4, 2018 (DOT, 2018a) and May 30, 2018 (DOT, 2018b); the U.S. Department of Transportation (DOT) requested the review and recommendation from the U.S. Nuclear Regulatory Commission (NRC or staff) regarding the revalidation of the Canadian Certificate of Approval No. CDN/2094/B(U)-96, Revision 1, for the Model No. F-522 package (referred as Model No. F-522 or F-522). The Model No. F-522 is a Type B(U) package.

The NRC reviewed the information provided to the DOT by Nordion (Canada) Inc. (the applicant) in its application for the Model No. F-522 package and its supplements against the regulatory requirements of International Atomic Energy Agency (IAEA) SSR-6, Regulations for the Safe Transport of Radioactive Material, 2012 Edition (IAEA, 2012a). Based on the statements and representations in the information provided by DOT and the applicant, the staff recommends the revalidation of Certificate of Approval No. CDN/2094/B(U)F, Revision 1, Model No. F-522 package, for the contents included in Section 1.1.2, Contents, of this safety evaluation report (SER).

1.0 GENERAL INFORMATION The F-522 package consists of an F-522 overpack, a UK-201 shielding vessel, and a containment system. The F-522 overpack consists of a double walled stainless-steel cylindrical vessel. Section 6.1 of this SER includes a description of the UK-201 shielding vessel.

1.1 Package Description 1.1.1 Packaging The packaging consists of an inner container (flask) and an impact limiting fire shield overpack.

The inner container is a stainless steel encased lead cylinder, with a removable top plug attached by eight 5/8-inch (in.) diameter bolts. The container is sealed by a silicone O-ring.

The overpack consists of a capped, double carbon-steel wall cylinder mounted on a disk base.

External fins are welded to the outer skin to provide heat transfer and impact energy absorption.

2 Lifting lugs are integral with four of the heat transfer fins. The cylinder is attached to the base disk by eight 1-in. diameter bolts. The inner container is mounted on the disk of the overpack by four steel brackets and eight 3/4-in. diameter bolts. The overall dimensions of the overpack are approximately 60-in. high by 49.5-in. diameter. Section 2 and 6 of this SER includes additional information about the description of the package.

1.1.2 Contents The radioactive contents of the package are in solid, liquid, or special form. The table below summarizes the radioactive contents of the F-522/UK-201 package. [Appendix 1 of the Certificate of Approval include a description of the authorized contents for the Model No. F-522.

(DOT, 2017)]

Table 1.1. Radioactive contents of the Model No. F-522 transportation package.1 Package Configuration (Maximum Activity) Physical Isotope Characteristics F-522/UK-201 Form F-522/UK-201 with F-248 Irradiated cobalt metal and Solid, Special Cobalt-60 (60Co) 1,000 GBq2 ---

associated Form products Aqueous NaOH solution or aqueous Molybdenum-99/ 37 TBq3 NaOH with up to 1 Liquid Technetium-99m --- M NH4OH or up to (99Mo/99mTc) 0.4% NaOCl Aqueous NH4OH 8.4 TBq Liquid solution 366 TBq 2 of 99Mo with a maximum 99 99m a impurity level Mo/ Tc --- Fission product Solid equivalent to 1,850 GBq1 of 132I.

6 TBq 2 (total) of 82Sr, 83Sr, 85Sr, Proton irradiated 82Rb, 83Rb, 84Rb, rubidium-based Strontium-82

--- target material and Solid (82Sr b) 55Co, 48V, 52Mn, associated target and other shells radionuclides.

a. Including impurities

b. Targets 1

Iodine-132 (132I), Rubidium-82 (82Rb), Rubidium-83 (83Rb), Rubidium-84 (84Rb), Vanadium-48 (48V),

Manganese-52 (52Mn), Sodium hydroxide (NaOH), Ammonium hydroxide (NH4OH), and Sodium hypochlorite (NaOCl).

2 Gigabequerels 3

Terabequerel

3 For 99Mo, the maximum ratio of shipped activity to A (A1 or A2) limit is 610. For all other shipments the ratios are less than 100. These ratios are all less than the limit of 3,000 prescribed in IAEA No. SSR-6, 2012 Edition (IAEA, 2012a). The staff concludes that the maximum ratio of shipped activity is within the maximum allowed limit based on the discussion above.

2.0 STRUCTURAL EVALUATION The purpose of the structural evaluation is to verify that the structural performance of the package meets the regulatory requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a). A summary of the staffs structural evaluation is provided below.

2.1 Description of Structural Design The proposed F-522/UK-201 transportation package is a containment vessel to transport Type B quantities of radioactive materials in solid or liquid form. It is designed to protect the radioactive materials from normal conditions of transport (NCT) and hypothetical accident conditions (HAC) as required by IAEA SSR-6, 2012 Edition (IAEA, 2012a).

The F-522/UK-201 transportation package is made of three principal structural components:

a. containment Drawings F552201-001 and F652201-001 in Appendix 1 of the application (DOT, 2017) includes the details of the structural design for the containment body. The containment body is a cylindrical body made of stainless steel (SS) Type 304. The minimum wall thickness of the containment is 0.25 centimeters (cm) and a minimum internal volume is 194 milliliters (ml). It is built to withstand a pressure of 4.9 megapascals (MPa) and a temperature of 232 degrees Celsius (°C) without leakage.

b. shielding vessel Drawings F5224801-001 and F652201-001 in Appendix 1 of the application (DOT, 2017) provide details of the structural design for the shielding vessel (i.e., UK-201). The UK-201 shielding vessel is a depleted uranium-filled cylindrical vessel, and is made of SS Type 321 encased depleted uranium. A minimum thickness of the vessel is 11 millimeters (mm) and has a nominal cavity size of 63 mm in diameter by 118 mm in height. The shielding vessel plug retained with 8 screws M10 x 20 SAE J1199 Class 8.8 and plug is sealed with silicone O-ring. A nominal weight of the UK-201 shielding vessel is 206 kilograms (kg).

c. insulated overpack Drawings F5224801-001 and F652201-001 in Appendix 1 of the application (DOT, 2017) provide structural design details of the F-522 overpack. The F-522 overpack provides impact and fire protection for the UK-201 shielding vessel and its contents during accident conditions of transport. The dimension of the over pack is 45 cm in diameter and 57 cm in height. The nominal weight of the empty F-522 overpack is 52 kg. It is made of SS Type 304L and is equipped with four lifting apertures. A closure lid is bolted

4 to the container body and is fitted with a gasket. The space between the double skins of the container is filled with cell foam.

The applicant provided the general assembly drawings of the F-522/UK-201 transportation package in Appendix 1 of the application (DOT, 2017), where it identified the major components of the package.

The NRC staff reviewed the drawings for completeness and accuracy, and finds that the applicant adequately incorporated information related to the geometry, dimensions, material, components, notes, and relevant details of the major components of the Model No. F-522.

2.2 Mechanical Analysis The following sections include a discussion of the information provided by the applicant related to the mechanical analysis of the Model No. F-522.

2.2.1 Stacking Test The IAEA SSR-6, 2012 Edition, (IAEA, 2012a) requires to subject the F-522/UK-201 the one of the following approaches to calculate the greater compression stress of the package:

(1) 5 times the mass of the package or (2) 13 kilopascals (kPa) time the vertical projected area of the package.

Using the approach that results in the greatest compressions stress of these two, the applicant calculated a compressive stress of 531 pounds per square inch (psi), which is negligible in comparison with the compressive strength of 35,000 psi for SS Type 304L. The applicant concluded that the F-522/UK-201 meets the regulatory requirements of the stacking test.

2.2.2 Lifting Test The applicant performed a lifting test to meet the requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a). From the test, the applicant observed that the opening of the F-522 overpack and moving the UK-201 shielding vessel out of the F-522 overpack were acceptable, and welding integrity was maintained. The applicant concluded that all test results were acceptable because there was no deformation or any other damages on the package.

2.2.3 Penetration Test The applicant used the results of the penetration test on the F-458 overpack, which were previously reviewed and accepted by the NRC staff. The applicant stated that the designs of the F-522 and F-458 overpacks were similar to each other and both overpacks include a foam backing for the shell. The major design difference between these two overpacks is the thickness of the shell (see Table 2.1 below).

Table 2.1. Comparison of the overpack shell thickness for the Model Nos. F-522 and F-458.

Model No. Overpack Shell Thickness (mm)

F-522 2.5 F-458 1.5

5 The thickness of the F-522 overpack is larger (2.5 mm) than the F-458 overpack (i.e., 1.5 mm).

The applicant noted that because the F-522 has a thicker shell, it offers greater resistance to the penetration test and the same or better performance is expected from the F-522 overpack.

Based on the results and the thickness of the F-458 overpack, the applicant concluded that the penetration test will cause no significant damage to the F-522 overpack.

2.2.4 Tie-Down Test The applicant assessed the handle openings of the F-522 overpack with respect to the tie-down loads specified in Table IV.2 of the IAEA Specific Safety Guide No. 26 (SSG-26), Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, 2012 Edition (IAEA, 2012b). The applicant performed a container tie-down test with a hanging mass of 2,217 kg attached to two of the tie-downs, each at a 45-degree (45°) angle. The force applied to each of the tie-downs was 1,108.5 kg, which is greater than the required 1,038 kg from the calculated resultant load. The applicant observed that there was no sign of deformation or yield following the test.

The NRC staff reviewed the analysis and test results submitted by the applicant and finds that these meet the regulatory requirement of IAEA SSR-6, 2012 Edition.

2.3 Structural Evaluation under Normal Conditions of Transport and Hypothetical Accident Conditions The following sections include a discussion of the information provided by the applicant related to the structural analysis under normal conditions of transport for the Model No. F-522.

2.3.1 Drop Tests The applicant performed a series of full scale model tests to demonstrate compliance with the free drop test requirements in the IAEA SSR-6, 2012 Edition (IAEA, 2012a). A single test specimen described in Appendix 5 of the application (DOT, 2017) was subjected to the cumulative effects of two 9 meters (m) drops, three 1 m pin drops and a free drop from 1.2 m in height. The specimen consisted of an F-522 overpack, an UK-201 shielding vessel, and an F-248 leak proof insert (LPI). All of the drops were performed on the same F-522/UK-201/F-248 specimen. The drop test orientations were chosen to cause maximum damage to the top of the package and to provide maximum inertial loads to the fasteners.

Based on the results of the model tests, the applicant concluded the following:

(1) The radiation dose rates from the F-522 were less than 10 millisieverts per hour (mSv/h) at 1 m from the surface of the package; (2) The post-test radiation survey results were unchanged and showed a maximum radiation level less than 0.6 mSv/h on contact with the F-522 and a transport index (TI) less than 3. Both of these requirements satisfy the requirements for normal conditions of transport; (3) There were no fractures of the UK-201 shielding vessel; and (4) The F-248 LPI was leaktight.

6 2.3.2 Water Immersion Test The applicant analyzed the performance of the Model No. F-522/F-UK-201 package by calculation and reasoned argument instead of performing the water immersion test. The applicant noted that the package is not used to transport fissile material, and, therefore, criticality is not a concern. If submerged under a head of 15 m of water, the F-522 overpack may slowly leak water. Over a period of eight hours, the F-522 cavity may fill with water. The UK-201 shielding vessel is watertight, to ensure water will not leak into the shielding vessel.

Considering the effect of a 15 m head of water of the pressure on the UK-201 shielding vessel, the applicant calculated a tensile stress of 15.47 MPa. Since the ultimate tensile strength of SS Type 321 is 515 MPa, a large margin of safety exists against failure. Based on its calculation, the applicant concluded that the shielding or containment system of the Model No.

F-522/UK-201 package would not suffer damage if it was subjected to the water immersion test.

2.4 Evaluation Findings Based on the review of the statements and representations contained in the application, the NRC staff agrees that the F-522/UK-201 transportation package meets the requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a).

3.0 MATERIALS EVALUATION The staff reviewed the adequacy of the package materials of the Model No. F-522. The shipping container is welded construction made of stainless steel. A summary of the applicable guidance IAEA SSR-6, 2012 Edition, (IAEA, 2012a) requirements related to the materials performance of the package, and the ability of the package design to meet the applicable requirements are provided below.

3.1 Packaging Materials The packaging materials comply with consensus standards (e.g., ASTM Standards). The packaging materials are stainless steel Types 304, 316, 321, and 416 including capsules that have a special form radioactive material certificate. The package also includes depleted uranium for shielding purposes and a closed-cell polyurethane foam for impact and thermal protection. The staff determined that the drawing details along with information related to the package [IAEA SSR-6, paragraph 306, (IAEA, 2012a)] are appropriate.

3.2 Fabrication/Welding The shielding vessel and F-522 container are welded. Welding will be in accordance with ASME (American Society of Mechanical Engineers) Boiler and Pressure Vessel (B&PV) Code,Section IX, or (equivalent) CSA (Canadian Standards Association) Standard W59, or other approved equivalent standards. The staff determined that the use of the ASME consensus welding methods is appropriate.

7 3.3 Chemical, Galvanic, or Other Reactions 3.3.1 Radiolysis The applicant assessed the radiolysis of aqueous forms of the content at STP [standard temperature and pressure [ i.e., 0 ºC and 1 atmosphere (101.3 kPa, approximately),

respectively]. The applicant determined a gas generation factor, GX (ml/TBq, ml/Ci), in each case by measuring the volume of gas generated by the water radiolysis of 99Mo over time. In each case of two solutions, the applicant scaled the generation factor by using the maximum isotope activities. GX is an equivalent factor to the G value (particles produced per 100 eV irradiation in radiolysis). GX was converged to be a steady-state (constant) value with time. For example, 99Mo gas generation became constant at near 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />. The applicant assessed the gas pressure in STP with space volume and temperature. The pressures measured in each case were within the safe range of leak proof pressure limit (at least 4.9 MPa, 715 psig).

3.3.1.1 Hydrogen Generation The applicant described in Appendix 10.2, Assessment of Containment for Stoichiometric Gas Mixture Containing Hydrogen, of document No. IS/TR-2650-F522 (DOT, 2017) its assumptions to verify the impact of radiolytic decomposition to demonstrate compliance with paragraph 644 of the IAEA SSR-6, 2012 Edition. Even though hydrogen gas generated during radiolysis has the potential for resulting in an explosion that may damage the containment boundary, the conditions for an explosion to occur depend on other factors such as the concentration of hydrogen in the containment boundary, the presence of an oxidizing source (e.g., oxygen), and an ignition source. For the purposes of a flammable or explosive gas mixture, the applicant assumed the following:

(a) a stoichiometric ratio 2:1 of hydrogen and oxygen (i.e., water molecular formula) could exist in the headspace of the LPI.

(b) The applicant stated in Appendix 10.2 (DOT, 2017) that the gas mixture in the LPI can be safe under the following conditions:

(i) There are no potential sources of ignitions, (ii) Sparks are not allowed to enter or occur, and (iii) The temperature is maintained below the auto-ignition temperature; or (iv) The containment system is explosion-proof.

The applicant noted in Appendix 10.2 (DOT, 2017) that the F-522 package meets all of the four criteria mentioned above, and no auto-ignition was observed with the vessel temperature up to 346oC.

The applicant did not find published and reliable data that provided the gas generation rate (particularly the hydrogen generation rate) for medical isotopes within a containment vessel and receptacles as used in Model No. F-522. For this reason, the applicant conservatively assumed a worst-case hydrogen generation rate of 90 volumetric percent [%(v/v)] of the evolving gas

8 hydrogen. Under these conditions, the LPI exceeds 5%(v/v) hydrogen in a matter of hours and days.

3.3.1.2 Combustion Testing The applicant conducted combustion testing at Fike Corporation (Missouri) to assess the effects of hydrogen combustion. Experiments were performed to establish the auto-ignition temperature and explosion pressure of stoichiometric mixture of hydrogen and oxygen, in the presence of inert nitrogen. Water was present to mimic the liquid isotope inside the vessel. The vessel temperature and pressure exceeded the limits of the test apparatus, 346ºC and 426 psi, respectively. The applicant concluded that a stoichiometric mixture of hydrogen and oxygen will not auto-ignite under normal and accident conditions of transport. The temperature and pressure assessed even under conservative assumptions of explosion are within safety limits.

At spark ignition tests, there was short-duration (~ 0.2 seconds) pressure rise of about 600 psia.

A safety margin greater than approximately 115 psi exists under these conditions because the F-248 LPI can withstand 715 psig static pressure.

The staff reviewed Appendix 6 of document No. IS/TR-2650-F522 (DOT, 2017) and finds that the calculated maximum temperatures of the LPI for liquid contents under NCT and HAC are below 130oC, which is below the auto-ignition temperature limit of 346oC (based on combustion test performed at Fike Corporation, Missouri) with a safety margin of greater than 200oC. Based on the information provided by the applicant about the temperatures of the LPI for liquid contents, the staff concludes that the gas mixture in the F-522 package will not auto-ignite and the containment of the LPI and O-ring seal is maintained under NCT and HAC, which is in compliance with IAEA SSR-6, 2012 Edition (IAEA, 2012a).

3.3.1.3 Conclusion In Section 3.3.1 of this SER, the staff evaluated the rationale used by the applicant to determine the gas generation factor (GX) and extrapolation to standard temperature and pressure (STP).

The applicant assessed the following conditions:

(a) testing of pressure increase by radiolysis, and (b) testing of temperature and pressure increases by ignition temperatures.

The applicant used acceptable chemical reaction principles available in the literature and considered conservative assumptions to evaluate the worst-case for a hydrogen generation rate. The data obtained are reasonable. The key parameters involved in the radiolysis and the ignition tests were well encompassed. The staff determined that the applicants assessment considering gas generation in the containment boundary due to radiolysis and the effects of hydrogen combustion are acceptable.

3.3.2 Pyrophoricity The proposed radioactive contents include pyrophoric target materials of primarily solid Rubidium (Rb) (i.e., 82Rb, 83Rb, 84Rb) and 82Sr. The applicant assessed the amount of hydrogen generation assuming the following:

9 (1) three potential chemical reactions with water and oxygen coming from air intrusion with moisture (the hazard analysis in the application, Appendix 13, Section 3) (DOT, 2017),

(2) the amounts of water and oxygen were conservatively assumed 20% oxygen in air moisture of 100% relative humidity, and (3) the hydrogen generated by the pyrophoric reaction may result in explosion and containment damage.

The staff evaluated the conservative assumption of air intrusion, and determined that the applicants assessment of pyrophoric reactions was appropriately based upon comparable standard ambient air chemistry and chemical reactions. The staff estimated by calculation that the hydrogen generated (using this assumption) by the pyrophoric reactions was less than 5%(v/v). The staff determined that the applicants assessment did not require experimental support because this analytical assessment was based on well-established chemical principles.

The applicant estimated the amount of energy released from chemical reactions for Rb to be 3.14 kilo Joules (kJ). The applicant noted the following:

(a) the temperature would rise by 2ºC for the LPI made of 1 kg of stainless steel.

(b) the energy generated could result in a temperature rise of the remaining nitrogen (after oxygen consumption) to 3,083°C (per the applicant, this is a conservative assumption).

The applicant does not intend using an additional risk label on hydrogen ignition because the calculated 2ºC rise would not affect the emergency actions related to this package (see application, Section 2.4).

The high temperature rise to 3,083ºC of the nitrogen is conservative for the purposes of the pressure calculation. This temperature is not realistic and is instead chosen to provide an extreme safety margin for the pressure calculation. This calculation does not take into account the following factors:

(a) the mass of the target shell, (b) the target holder, or (c) the LPI.

When just the mass of the LPI is taken into account, the temperature would increase by 2ºC as discussed below.

The applicant also assessed the pyrophoricty of 82Sr targets (see SAR, Section 2.5). The applicant stated that the chemical reaction would result in a temperature increase of 2ºC to the LPI, which would not cause damage to the package insert. This is based on the assessment made for Rb, which would have greater susceptibility to pyrophoricity. The staff concludes that no damage will occur based on the assessment made for Rb which is conservative and bounding for 82Sr. Rubidium is pyrophoric and reacts exothermically with water.

10 3.3.3 Stress Corrosion Cracking The aqueous forms of content product are 99Mo/99mTc in aqueous sodium hydroxide (NaOH) and 99Mo/99mTc in aqueous ammonium hydroxide (NH4OH). In steady-state conditions, the maximum temperature of the NaOH solution is 93C within the LPI. This temperature is below the susceptibility temperature of stress corrosion cracking (SCC) of the LPI made of 300 series stainless steel. The applicant (Nordion, package designer, Nordion, Canada, Inc.) has a history of shipping NaOH solutions within the LPIs for over 30 years without incident or sign of SCC.

3.4 Thermal Properties and Effects on Materials 3.4.1 O-Rings The applicant assumed a maximum O-ring temperature of 190°C [with a 40 watt (W) heat load]

as the bounding case for solid radioactive material. Depending on the content, transported, the applicant uses two different seal materials as depicted in Table 3.1 below.

Table 3.1. Temperature limits of seals used with the Model No. F-522.

Content Physical State Seal Type Temperature Limit Solid Silicone O-ring 232°C Liquid Neoprene O-ring 149°C The staff determined that the use of these two types of O-rings is safe for the contents based on package temperature limits.

3.4.2 Foam Regarding the thermal stability of the foam, the applicant assessed the effect on the density of the foam and the depth of char during regulatory fire conditions based on tests performed by General Plastics (Appendix 14) (DOT, 2017). The applicant concluded that the density change is minimal and would not affect drop tests.

Foam (i.e., polymer) is thermally stable in terms of strength (such as elastic modulus) at less than approximately 120ºC (Sanchez et al., 2000). The staff determined that the applicant provided appropriate data on mechanical properties within 121ºC (General Plastics) based on ASME B&PV Code, Section IID. The applicant provided the thermal conductivity and heat capacity for various materials in the SAR, Section 2.2.2. The staff determined that the values used are appropriate based on ASME B&PV Code, Section IID, and public literature (e.g.,

https://www.engineeringtoolbox.com/thermal-conductivity-d_429.html).

3.5 Evaluation Findings The staff finds that the F-522 transportation package is composed of materials appropriate for the intended application. Regarding radiolysis and pyrophoricity, the safety assessment complies with the IAEA regulation associated with IAEA SSR-6, 2012 Edition, (IAEA, 2012a).

Based on review of the statements and representations in the application, the staff concludes that the applicant adequately described the materials aspects of the design and the package meets the IAEA requirements of SSR-6, 2012 Edition (IAEA, 2012a).

11 4.0 THERMAL EVALUATION The purpose of the thermal review is to verify that the package design satisfies the requirements for the thermal evaluation under the IAEA SSR-6, 2012 Edition. The staff reviewed the thermal properties of the materials used for the Model No. F-522 package and the description of the thermal analysis against the standards in the IAEA SSR-6, 2012 Edition (IAEA, 2012a). A summary of the staffs review is provided below.

4.1 Description of the Thermal Design The Model No. F-522 is a Type B(U) package designed to transport a variety of isotopes in solid or liquid form. The certificate for the F-522 does not allow to transport fissile material. The staff reviewed the following information related to the thermal evaluation for the F-522:

a. thermal material properties,

b. the descriptions of the thermal modeling,

c. the assumptions used in the thermal analyses, and

d. the calculations related to the thermal models for normal transport and accidental transport.

The applicant stated in Appendix 3, Heat Generation, and Appendix 6, Thermal Analysis of the F-522/UK-201, of document No. IS/TR-2650-F522 that for shipments of 60Co in special form, the heat load limit is 26.2 W in order to maintain the temperature at the surface of the package below 50°C for air transport (DOT, 2017). Up to 120 TBq of 60Co could be shipped without exceeding the heat load limit of 26.2 W. For material other than special form, the heat load limits are 40 W for 99Mo (solid) and 4.0 W for 99Mo (liquid), respectively, contained in the LPI. The maximum heat load for a single target is about 0.7 W, as listed in Table 5 of Appendix 3 of document No. IS/TR-2650-F-522 (DOT, 2017). For shipments of 99Mo (solid) with allowable heat load of 26.2 W, silicone O-rings are used. For shipments of 99Mo (liquid) with heat load generated less than 4.0 W, neoprene O-rings are used. Table 4.1 includes the heat load limits for the allowable contents of the Model No. F-522.

Table 4.1. Heat load limit per Model No. F-522 allowable content.

Content Heat Load Limits, W 60 Co (special form, 120 TBq) 26.2 99 Mo (solid) (not special form) 40 99 Mo (liquid) (not special form, in a leak proof insert) 4.0 single target 0.7 The staff reviewed Appendix 3 of document No. IS/TR-2650-F522 (DOT, 2017) and determined that the applicant adequately described the thermal features of the F-522 package and its heat generation for the allowed contents and that this information is in alignment with the packages thermal evaluation.

12 4.2 Thermal Evaluation under Normal Conditions of Transport and Hypothetical Accident Conditions The applicant performed tridimensional (3-D) analyses of the F-522 package using the thermal analysis code Flow Simulation to verify if the thermal design of the F-522 package (under NCT and HAC) was in compliance with the regulatory safety requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a). The applicant provided the thermal properties of the materials in Table 1 of Appendix 6 of document No. IS/TR-2650-F522, including stainless steel, depleted uranium, neoprene, and FR-3708/3712 Last-a-Foam (DOT, 2017) (see Section 3.4 of this SER).

The staff reviewed the following information pertaining to the F-522:

a. description of the package,

b. input parameters,

c. boundary conditions of the thermal analysis,

d. thermal properties of the packaging material,

e. RAI response, and

f. temperature distributions under NCT and HAC.

The staff confirmed that an F-522 package loaded with 60Co in special form, solid 99Mo in the LPI, or liquid 99Mo in the LPI meets the requirements for thermal performance in IAEA SSR-6, 2012 Edition, for the air transportation of the radioactive isotopes based on the following information:

a. the package is subject to a hot ambient temperature of 38°C for NCT;

b. the initial temperature of the package is set as 73oC for the HAC fire test;

c. the package is exposed to an 800°C fire for a period of 30 minutes with fire directly applied to package outer surface for the HAC fire analysis; and

d. the package is exposed to a hot ambient temperature of 38°C (natural convection and solar insolation) for post-fire cool down.

4.2.1 Thermal Evaluation under Normal Conditions of Transport The applicant assessed the load cases for NCT shown in Table 4.2 and provided the predicted NCT maximum temperatures of the entire package, O-ring seal, and package outer surface for each load case (see Table 4.3 of this SER).

The staff reviewed the descriptions and the modeling calculations of the F-522, verified the thermal properties used in the analyses, and evaluated the thermal performance under NCT.

The staff confirmed the following:

13 (1) The peak temperatures of silicone O-rings used in load cases Nos. 1 to 3 and 6 to 8, are within the service range of -40 to 232°C (Hannifin, 2007, and DOT, 2017).

(2) The peak temperatures of neoprene O-rings, used in load cases Nos. 4 and 5, are below the allowable limit of 150°C for neoprene [i.e., a neoprene O-ring can tolerate 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of exposure at 150°C (Hannifin, 2007, and DOT, 2017).

(3) The containment will be maintained for all load cases under NCT based on the applicants calculated temperatures of the containment O-rings for all load cases (Hannifin, 2007, and DOT, 2017).

(4) For F-522 package under case No. 2 (i.e., heat load of 26.2 W, no solar insolation, 38°C ambient, and transported by air) the maximum outside temperature of the package is 50°C, which is the maximum allowable temperature of 50°C required by the regulations for air transport (paragraph 619 in IAEA SSR-6, 2012 Edition) and a heat shield or screen may be used to decrease the outer surface temperatures.

(5) For F-522 package under case No. 7 (i.e., materials in LPI and heat load greater than 26.2 W, no solar insolation, 38°C ambient, and transported by air) the maximum outside temperature of the package is 55°C, which is above the maximum temperature of 50°C required by the Regulations for air transport (paragraph 619 in IAEA SSR-6, 2012 Edition) and a heat shield or screen shall be used to make the hot surface inaccessible.

Table 4.2. Load cases analyzed under NCT using 99Mo as content.

Ambient Load Heat Load Physical Form of Containment Solar 99 Temperature Cases (W) Mo O-ring Types Insolation

(°C) 1 26.2 Solid in special form Silicone4 Yes 38 2 26.2 Solid in special form Silicone No 38 3 26.2 Solid in special form Silicone No 55 4 4 Liquid in LPI Neoprene1 Yes 38 5 4 Liquid in LPI Neoprene No 55 6 40 Solid in LPI Silicone Yes 38 7 40 Solid in LPI Silicone No 38 8 40 Solid in LPI Silicone No 55 4 The silicones maximum service temperature is 232°C and the neoprenes maximum service temperature is 149°C.

14 Table 4.3. Maximum temperatures of the package, containment O-rings, and outer surface of the package under NCT for the cases described in Table 4.2 of this SER.

Load Maximum Temperature Peak O-ring Maximum Outer Surface Cases of Entire Package (°C) Temperature (°C) Temperature (°C) 1 156 134 108 2 115 92 50 3 < 142 108 > 50 4 < 108 94 NA 5 < 69 64 NA 6 191 190 111 7 155 118 55 8 169 169 > 55 4.2.2 Thermal Evaluation under Hypothetical Accident Conditions The applicant performed a thermal analysis for the package for the following conditions:

(1) A 30-minute fire test at 800°C. The applicant performed a thermal analysis for the package under an 800°C fire for a period of 30 minutes as described in Appendix 6 of document No. IS/TR-2650-F522 (DOT, 2017) to demonstrate the performance of the package under HAC. The test included applying a flame directly to the outer walls of the package until reaching a temperature of 800°C and, then, leaving the package engulfed in the fire for 30 minutes. The applicant simplified the model to an internal flow analysis.

(2) Post-fire cooldown. For post-fire cooldown, the applicant performed a transient analysis using temperature data from the end of 30-minute fire, a 38°C ambient temperature, solar insolation, and natural convection between the package surface and its surroundings. The applicant imported the initial temperatures of HAC load cases Nos. 1 to 3 from the NCT load cases Nos. 1, 4, and 6 with heat loads of 26.2 W, 4.0 W, and 40.0 W, respectively. The applicant assumed an emissivity for all package outer surfaces of 0.8. The applicant predicted the peak O-ring and depleted uranium temperatures for heat load cases Nos. 9, 10, and 11 under HAC as listed in Table 4.4 below.

Table 4.4. Load cases under HAC: maximum temperatures of containment O-rings and depleted uranium.

Max. Depleted Heat Peak O-ring Load Containment Uranium Load Contents Temperature Cases O-ring Types Temperature (W) (°C)

(°C) 1 26.2 60Co in special form Silicone 168 172 2 4 Liquid 99Mo in LPI Neoprene 128 <172 3 40 Solid 99Mo in LPI Silicone 220 196

15 The staff reviewed the following information in Appendix 6 of document No. IS/TR-2650-F522 (DOT, 2017) related to the temperature distributions:

(1) Figures 26 and 27 for HAC load case No. 1, (2) Figures 28 and 29 for HAC load case No. 2, and (3) Figures 30 and 31 HAC for load case No. 3.

Table 4.4 of the SER includes a summary of maximum temperatures of containment O-rings and depleted uranium. The staff verified the following:

(1) the predicted peak temperatures of silicone O-rings used in HAC load cases No.

1 and 3 are within the service range of -40 to 232°C (DOT, 2017),

(2) the peak temperature of neoprene O-rings used in HAC load case No. 2 is within the service range up to 149°C for 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (DOT, 2017), and (3) the predicted maximum depleted uranium temperatures in all three cases have no significant impact to the shielding.

The staff concluded that the containment will be maintained under all three HAC load cases.

4.2.3 Maximum Normal Operating Pressure In Appendix 10.2 of document No. IS/TR-2650-F522 (DOT, 2017), the applicant noted that the pressure buildup (over time) in the LPI was primarily due to generation of hydrogen and oxygen gas, from radiolysis. The pressure rise in the LPI due to radiolysis is a function of many parameters such as (this list does not include all the parameters):

(1) isotope, (2) diluent solution, (3) packaging materials and geometry, and (4) temperatures.

As described in Appendix 10.1 of document No. IS/TR-2650-F522 (DOT, 2017), the applicant calculated maximum normal operating pressures (MNOPs) in the LPI of 642 kPa in aqueous solutions for liquid 99Mo in the LPI and 157 kPa for solid 99Mo.

The staff reviewed Appendix 10.1 for the determination of MNOP and Appendix 10.2 for the assessment of hydrogen generation (DOT, 2017). The staff finds that the pressure of 642 kPa from the loading of the liquid 99Mo contained in LPI with a heat load of 4 W is greater than the pressure of 157 kPa from the loading of the solid 99Mo contained in LPI with a heat load of 40 W. Therefore, the pressure of 642 kPa for radionuclide in aqueous solutions is the bounding MNOP, and this is mainly because of the huge pressure rise due to radiolytic reaction of liquid 99Mo.

16 4.3 Evaluation Findings Based on the review of the statements and representations in the application for the F-522 package, the staff concludes that the applicant adequately described and evaluated the thermal design of the F-522 package. Therefore, the package meets the thermal requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a).

5.0 CONTAINMENT EVALUATION The purpose of the containment review is to verify that the package design satisfies the requirements for the evaluation of the containment boundary as required in the IAEA SSR-6, 2012 Edition. The staff reviewed Section 1.1, Package Description, and Figures 1 and 2 in Section 1.0 of document No. IS/DS-2650-F522 and Section 2.0, Package Description, of document No.IS/DS-2651-F522 and confirmed that the containment system of the F-522 package is well described for revalidation. A summary of the staffs review is provided below.

5.1 Description of the Containment Boundary The F-522 is designed to transport Type B quantities of radioisotopes in solid or liquid form (see Section 1.1.2 of this SER). The F-522/UK-201 is not approved for transporting fissile material.

The package consists of a primary containment (surrounded by a shielding vessel) and a shielding vessel (surrounded by an insulated overpack).

For special form of radioactive materials or a Model No. F-248 LPI, welded stainless steel sealed sources provide containment for the special form of radioactive materials and the LPI requires a helium leakage rate test. In general, the LPI consists of the following features:

a. a cylindrical body made of stainless steel Type 304,

b. a cap made of stainless steel Type 416, and

c. a neoprene or silicon O-ring or equivalent for containment.

The LPI demonstrated to withstand a pressure of 4.9 MPa and a temperature of 232°C without leakage as described in Section 2.0 of document No. IS/DS-2651-F522.

5.2 Helium Leakage Rate Testing 5.2.1 Special Form Sealed Source The applicant noted in Section 3.2, Manufacturing Requirements, of document No. IS/DS-2651-F522 that sealed sources shall be leakage tested according to the International Organization for Standardization (ISO) 9978, Radiation protection -- Sealed radioactive sources

-- Leakage test methods. Leakage testing by hot-liquid bubble test per ISO 9978, paragraph 6.2.2, or a test of equivalent or better sensitivity. Table 5.1 includes a summary of leak testing parameters for the Model No. F-522.

17 Table 5.1. Leak testing requirements for the main components of the Model No. F-522.

Leak test requirement Process Package Component ref-cm3/s LPI (dry) < 1x10-7 Fabrication LPI stainless steel shell < 10-7 UK-201 cavities 1x10-3 Pre-shipment LPI (after final lid closure at site) 10-3 5.2.2 F-248 Leak Proof Insert The staff reviewed Section 3.2 for special form sealed source and Section 4.4 for the LPI and accepts the helium leakage rate testing for F-522 package.

Section 4.4.1, Pressure Test, of document No. IS/DS-2651-F522 includes a description of the hydrostatic test for the LPI. The applicant noted in Section 4.4.1 that the LPI shall be subjected to a hydrostatic pressure test at a minimum gauge pressure of 1,070 kPa (155 psi) for at least 5 minutes at room temperature of 20°C.

The staff reviewed Appendix 8.3, Hydrostatic Leak Test on F-320 Leakproof Insert, of document No. IS/TR 2650 F522 and confirmed that the pressure test conditions approved for F-320 LPI are also applicable to F-248 LPI. This is because F-248 LPI has a smaller outside diameter, the same base thickness, but lesser wall thickness than the F-320. Therefore F-248 LPI is capable of withstanding the same test pressure as the F-320, as the stresses are less than in the F-320.

The applicant stated in document No. IS/DS-2651-F522 that the leak test of each LPI is performed after the pressure test and after the LPI has thoroughly dried. According to ISO 12807, Safe transport of radioactive materials -- Leakage testing on packages, the fabrication leak rate for the LPI shall be less than 1x10-7 reference cubic centimeters per second (ref-cm3/s). The stainless steel shell shall be leak tested using a test sensitive to 10-8 ref-cm3/s and shall have leakage rate less than 10-7 ref-cm3/s. In addition, the UK-201 cavities shall be leak tested using a test sensitive to 10-4 ref-cm3/s as described in Section 4.4.2, Leak Testing, of document No. IS/DS-2651-F522 (Nordion, 2017a). Per the applicants SAR, the leakage rate from assembled units of UK-201 cavity shall be less than 1x10-3 ref-cm3/s during manufacturing at shop as described in Section 4.4, Manufacturing Testing, of document No. IS/DS-2651-F522. The LPI shall undergo additional leak testing after final lid closure at site to ensure a leakage rate less than 10-3 ref-cm3/s as described in Section 6.2 of document No. IS/DS-2651-F522 (Nordion, 2017a).

The staff reviewed Appendix 8.1, Results of Leak Tests on F-320 LPI of document No. IS/TR 2650 F522 and confirmed that the leak test conditions approved for F-320 LPI are also applicable to F-248 LPI. This is because the O-ring compression on the F-248 LPI is greater than F-320 LPI, resulting in better sealing.

5.3 Evaluation Findings Based on review of the statements and representations in the F-522 package application, the staff concludes that the applicant adequately described and evaluated the containment system for the F-522 package and that the package meets the containment requirements of the

18 IAEA SSR-6, 2012 Edition (IAEA, 2012a). The staff recommends revalidation of the Canadian Certificate of Approval No. CDN/2094/B(U)-96, Revision 1.

6.0 SHIELDING The staff reviewed the application to ensure that the shielding is adequate to meet the radiation level requirements within the IAEA SSR-6, 2012 Edition (IAEA, 2012a) for protecting people and the environment, for this type of package. Specifically, the paragraphs depicted in Table 6.1 below.

Table 6.1. Applicable IAEA, SSR-6, 2012 Edition, shielding requirements to evaluate the adequacy of a package design [Type B(U) Package].

Paragraph No. of the Brief Description of the Requirement SSR-6, 2012 Edition The radiation level cannot exceed 0.1 mSv/hr (10 mrem/hr) at 523 1 meter from the package.

526 The transport index (TI) shall not exceed 10.

527 (for non-exclusive The maximum radiation level at the surface of the package does use packages) not exceed 2 mSv/hr (200 mrem/hr).

652 (for Type B(U) The package shall meet the requirement in paragraph 648 of packages) IAEA SSR-6.

The package must not experience more than a 20% increase in 648(b) (under NCT) the maximum radiation level.

The package does not exceed 10 mSv/hr (1,000 mrem/hr) at 659(b)(1) 1 meter under accident conditions.

6.1 Description of the Shielding Design The UK-201 shielding vessel consists of a body and a lid made of 75 mm thick depleted uranium encased in stainless steel. Per the Canadian certificate, all sources, with the exception of the special form 60Co source because it is special form, must be shipped within the F-248 LPI.

6.2 Evaluation Method The applicant evaluated the shielding capability for the F-522 package by calculating the external radiation level for the following three unique sources using a point source geometry approximation:

a. 60Co, 1TBq - The applicant also measured the external radiation level from 60Co (normalized to 500 GBq) using a prototype.

b. 99Mo, 370 TBq with 1,850 GBq 132I - This amount of 99Mo bounds the amount of the other two liquid contents, and since the applicant used a point source approximation, this is conservative with respect to the form of the source material. With respect to the impurity content for the non-aqueous fission product 99Mo, the applicant submitted three impurity profiles. This was submitted by an email to DOT dated April 4, 2018 (Charbonneau et al., 2018, and DOT, 2018a).

19

c. Activated Rb - Activated rubidium represented by various isotopes totaling 6 TBq.

The applicant calculated the external radiation level of the package using the MicroShield code, version 10.0. The MicroShield code uses a point kernel method, which is a deterministic method for evaluating the attenuation of gamma rays. The calculations using this method are quick to execute in comparison to a Monte Carlo method. However, there are approximations to geometry that need to be handled in a conservative way because complex geometries cannot be handled explicitly. The staff reviewed the applicants MicroShield evaluation as discussed in Appendix 4 of the Engineering Assessment of the Ability of the F-522/UK-201 Type B Package to Meet IAEA SSR-6, 2012 Edition (DOT, 2017). The staff also reviewed the MicroShield output files submitted on September 20, 2017 (Nordion, 2017a).

Based on the geometry of the F-522 overpack and UK-201 shielding vessel, the staff finds that there are no streaming paths and using the MicroShield code is acceptable for approximating the geometry of this package.

6.2.1 Package Modeling The applicant evaluated the external radiation level through the radial side of the package.

Based on the calculations in Appendix 4 (DOT, 2017) of the Engineering Assessment report, the staff found that the external radiation level are roughly consistent with the thickness of components depicted in following drawings:

a. Drawing No. F524801-001, F-248X Leakproof Insert,

b. Drawing No. IS/SS 1786 F248, F-248 Leakproof Insert, and

c. Drawing No. F552201-001, F-522 Transport Package.

These drawings do not specify the minimum thicknesses of these components. The applicant submitted them in a letter to DOT on November 16, 2017 (Nordion, 2017b). The staff found that the dimensions chosen for the MicroShield model (as compared to the minimum dimensions) are within 1 mm and are, therefore, acceptable.

The applicant credits the presence of the LPI (about 4 mm stainless steel) when it may not be present for the special form sources. This is a non-conservative assumption. The staff used MicroShield, version 11.20, and compared external dose rates with and without the LPI and found that the external dose increases by about 17% for the 60Co source, which is allowable in a special form radioactive material. The applicants evaluation shows a 35% margin to the regulatory limit. Therefore, the staff found that this modeling assumption is acceptable for this application.

For the 99Mo fission product and impurities, the applicant modeled minimum dimensions consistent with those described in the November 16, 2017, letter to DOT (Nordion, 2017b). The staff found this acceptable.

Another approximation within the MicroShield code comes in the form of the buildup factor, which is used to account for scattering and generation of secondary gammas within the shield medium. In the MicroShield code, this buildup factor is only applied to one material within the code and its application can change the results of the radiation level significantly. As the

20 buildup factor represents additional scattering into the detector, the MicroShield developers recommend to use the outermost layer with significant shielding. The applicant applied the buildup factor to the depleted uranium shield. The staff found this approach appropriate and acceptable.

6.2.2 Source Modeling The applicant assumed a point source geometry. This is a very conservative estimate because it does not account for any self-shielding from the source and, therefore, acceptable to the staff.

For the 60Co source, the MicroShield output files show that the applicant used actual photon energies, versus energy groups, as the source. Staff found this appropriate considering that 60Co is dominated by two gamma energies and grouping these energies often ends up misrepresenting the source.

For the other contents the nuclide energy appears to be grouped by the Standard Indices option in MicroShield, which is a group structure adopted from ANSI/ANS 6.4.3, Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials Standard. Grouping photons may not necessarily give a conservative estimate of external dose rate as photons of a higher energy may be represented by a lower average group energy depending on the energy of the photons and their location with respect to the group boundaries. For the sources which the applicant has represented as grouped sources within MicroShield (i.e., 99Mo with impurities and activated Rb), the staff performed a sensitivity study where it grouped the source within MicroShield using both the standard indices and the linear options. The staff found that the standard indices option gave the more conservative calculation of dose rate (i.e., a higher dose rate). Although this may still be non-conservative as compared to using actual photon energies, the staff has reasonable assurance that it is representative of the activated Rubidium gamma spectrum and that any additional uncertainty attributed to the group structure is likely bounded by other conservative assumptions such as the point source approximation.

The fission product 99Mo content within the CDN/2094/B(U)-96, Revision 1, certificate does not specify the impurities of the 99Mo content except to say that it is equivalent to 1,850 GBq of 132I. The staff found this language to be ambiguous because it does not provide further specification of the impurities and how it is to be determined equivalent to 132I. The impurities are also the dominant contributor to external dose for this content. In addition, the impurities for the 99Mo content (when generated as a fission product) are varied and depend highly on the purification process, which the applicant noted was still under development. By letter to DOT dated March 29, 2018 (Charbonneau et al., 2018 and DOT, 2018a), the applicant submitted a table of three impurity profiles including the evaluations it performed to demonstrate that these impurity profiles meet regulatory radiation levels. The staff used MicroShield, version 11.20, and confirmed that the three impurity profiles meet regulatory radiation levels in paragraphs 526 and 527 of SSR-6 and recommended revalidation of the Model No. F-522 with the following condition:

The impurities within the 99Mo fission product content must meet any one of the three impurity profiles in document No. R119.017.SUR[1], Shielding Analysis Report for F522 Mo-99 Impurities, Table 6. (Charbonneau et al. 2018, and DOT, 2018a)

The applicant modeled the activated Rb using Table 3 of Appendix 4 of the application (DOT, 2017). The applicant noted that this nuclide profile is for a typical irradiation history. The staff found the specification of the activated Rb to be vague. However, the staff concluded that proton activated Rb would produce a source with similar activation characteristics. In a letter to

21 DOT dated November 16, 2017 (Nordion, 2017b), the applicant submitted data from 11 separate targets shipped between November 2016 and April 2017, which is reasonably representative. In addition, the staff performed an evaluation of the activated Rb source using the following:

(1) the MicroShield code, version 11.20, (2) the source information from both Table 3 of Appendix 4 of the application (DOT, 2017), and (3) the maximum activity from the data representing the 11 separate targets in the November 16 letter (Nordion, 2017b).

The staff scaled the activity up to 6 TBq, as is allowed in the Canadian certificate, and found that the source meets all regulatory radiation levels as specified in paragraphs 526 and 527 in the IAEA SSR-6, 2012 Edition (IAEA, 2012a).

6.2.3 Flux-to-Dose Rate Conversion Factors In the letter to DOT dated November 16, 2017 (Nordion, 2017b), the applicant stated it used the default flux-to-dose rate conversion factors within MicroShield for the contents. These conversion factors are based on ICRP-51 and are non-conservative as compared to the ANSI/ANS 6.1.1-1977 (ANSI/ANS, 1977) flux-to-dose rate conversion factors. The staff used the MicroShield output files provided by the applicant and converted the fluence rate in mega electron volts per square centimeter per second (MeV/cm2/s) to a dose rate by applying the ANSI/ANS 6.1.1-1977 (ANSI/ANS, 1977) flux-to-dose rate conversion factors to these values.

The staff calculated about a 20% higher dose rates than the applicant for 60Co. The staff further evaluated the two contents: 60Co and activated Rb targets [as represented by Table 3 of Appendix 4 of the application, (DOT, 2017)] using MicroShield, version 11.20, using the ANSI/ANS 6.1.1-1977 (ANSI/ANS, 1977) flux-to-dose rate conversion factors within the code and these calculations further show that the external dose rate as calculated using the default (ICRP-51) flux-to-dose rate conversion factors is less than the dose rate calculated using the ANSI/ANS 6.1.1-1977 (ANSI/ANS, 1977) flux-to-dose rate conversion factors by about 20% for these 2 contents.

The staff found the use of the ICRP-51 flux-to-dose rate conversion factors acceptable given the following:

(1) The margin to the regulatory radiation limits calculated by the applicant. As calculated by the applicant, the dose rate closest to the regulatory limit is the surface dose rate [see paragraph 527 of the IAEA SSR-6, 2012 Edition (IAEA, 2012a)] for 60Co with about 35% margin to the limit. Some of this margin is needed to off-set the non-conservative assumption of including the LPI material for special form 60Co.

(2) Conservative modeling assumptions. Since the applicant used other conservative modeling assumptions such as using the point source model (which concentrates the source and neglects self-shielding), the staff found the applicants use of the ICRP-51 flux-to-dose rate conversion factors acceptable.

The staff has additional assurance that the package will meet regulatory dose rate limits in paragraph 527 of the SSR-6 as its confirmatory evaluation using

22 MicroShield, version 11.20, of 60Co source, without the LPI and the ANSI/ANS 6.1.1-1977 (ANSI/ANS, 1977) flux-to-dose rate conversion factors still show that the package is within regulatory radiation level limits.

For the 99Mo fission product content with impurities, the applicant noted that it did use the ANSI/ANS 6.1.1-1977 (ANSI/ANS, 1977) flux-to-dose rate conversion factors, and the staff found this acceptable.

6.3 Hypothetical Accident Conditions For the 60Co source, the applicant also performed radiation level measurements on a prototype before and after the drop tests to demonstrate its compliance with external radiation levels set forth within the IAEA SSR-6, 2012 Edition (IAEA, 2012a). The drop tests were described in Appendix 5 of the application (DOT, 2017). The applicant shows the measured radiation levels before and after the drop tests in Figures 8 and 7 of Appendix 4 of the application (DOT, 2017),

respectively, and these are essentially the same. The staff found that this provides reasonable assurance that the package meets the radiation level requirements in paragraphs 648(b) and 659(b)(1) in IAEA SSR-6, 2012 Edition (IAEA, 2012a).

6.4 Confirmatory Evaluations As discussed in the previous paragraphs, the staff performed independent evaluations of the F-522 package. The staff used MicroShield, version 11.20, and represented the package using the minimum dimensions from the letter dated November 16, 2017 (Nordion, 2017b). The staff used the ANSI/ANS 6.1.1-1977 flux-to-dose rate conversion factors in all of its calculations (see Section 2.2.3).

For the 60Co content, the staff modeled the package without the LPI. For the 99Mo fission product content, the staff used the 99Mo source in addition to the three impurity profiles provided by the applicant in the document dated March 29, 2018 (Charbonneau et al., 2018 and DOT, 2018a). For the activated rubidium content the staff scaled the nuclide profile up to 6 TBq, as allowed within the certificate, from Table 3 of Appendix 4 of the application (DOT, 2017). In addition the staff modeled the maximum nuclides from the letter dated November 16, 2017 (Nordion, 2017b), and scaled this up to 6 TBq. All of the staffs calculations were below the radiation level limits in SSR-6 paragraphs 526 and 527 (IAEA, 2012a). The staff found that this provided additional assurance that this package would meet the regulatory radiation level limits.

6.5 Evaluation Findings Based on review of the statements and representations in the F-522 package application and as discussed in the paragraphs above, the staff has reasonable assurance that the F-522 package meets the requirements in paragraphs 526, 527, 648(b), 659(b)(1) in IAEA SSR-6, 2012 Edition.

The staff recommends revalidation of Canadian Certificate of Approval No. CDN/2094/B(U)-96, Revision 1, for the F-522 package with the following condition:

The impurities within the 99Mo fission product content must meet one of the three impurity profiles in document No. R119.017.SUR[1], Shielding Analysis Report for F522 Mo-99 Impurities, Table 6. (Charbonneau et al., 2018 and DOT, 2018a).

23 7.0 CONDITIONS The staff recommends the revalidation of Canadian Certificate of Approval No. CDN/2094/B(U)-

96, Revision 1, for the Model No. F-522 package, with the following additional condition:

The impurities within the 99Mo fission product content must meet one of the three impurity profiles in document No. R119.017.SUR[1], Shielding Analysis Report for F522 Mo-99 Impurities, Table 6. (Charbonneau et al., 2018 and DOT, 2018a)

8.0 REFERENCES

(ANSI/ANS, 1977) American National Standards Institute/American Nuclear Society, ANSI/ANS 6.1.1-1977, Neutron and Gamma-Ray Fluence-To-Dose Factors, ANS, LaGrange Park, IL.

(IAEA, 2012a) International Atomic Energy Agency, IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material, 2012 Edition, https://www-pub.iaea.org/MTCD/Publications/PDF/Pub1570_web.pdf.

(IAEA, 2012b) International Atomic Energy Agency, IAEA SSR-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, 2012 Edition, https://www-pub.iaea.org/MTCD/publications/PDF/Pub1586web-99435183.pdf.

(Charbonneau et al., 2018) Charbonneau, S. et al. R119.017.SUR[1], Shielding Analysis Report for F522 Mo-99 Impurities, Nordion, Redacted, March 29, 2018, ADAMS Accession No. ML18151A701.

(DOT, 2017) Boyle, Richard W., U. S. Department of Transportation (DOT),

letter to U. S. Nuclear Regulatory Commission (NRC) (Attn: Mr.

Michael Layton), March 15, 2017, ADAMS Package Accession No. ML18137A496.

(DOT, 2018a) Conroy, Michael, U. S. Department of Transportation (DOT), email to John Vera, U. S. Nuclear Regulatory Commission (NRC),

F-522 Endorsement-Response to NRC question 3, April 4, 2018, ADAMS Accession No. ML18143B343 and ML18151A701.

(DOT, 2018b) Conroy, Michael, U. S. Department of Transportation (DOT), email to John Vera, U. S. Nuclear Regulatory Commission (NRC),

Request for Redacted Documents, May 30, 2018, ADAMS Accession No. ML18151A695.

(Hannifin, 2007) Parker O-ring Handbook Parker Hannifin Corporation, Lexington, KY, 2007.

(Nordion, 2017a) Fulford, Greg, Nordion (Canada) Inc. (Nordion), letter to U. S.

Department of Transportation (DOT) (Attn: Mr. Michael Conroy),

September 20, 2017, ADAMS Accession No. ML18151A700.

24 (Nordion, 2017b) Fulford, Greg, Nordion (Canada) Inc. (Nordion), letter to U. S.

Department of Transportation (DOT) (Attn: Mr. Richard Boyle),

November 16, 2017, ADAMS Package Accession No. ML18143B680.

(Nordion, 2018) Fulford, Greg, Nordion (Canada) Inc. (Nordion), letter to U. S.

Department of Transportation (DOT) (Attn: Mr. Michael Conroy),

January, 16, 2018, ADAMS Package Accession No. ML18151A695.

(Sanchez et al., 2000) Sanchez, E.M.S., C.A.C. Zavaglia, and M.I. Felisberti.

Unsaturated Polyester Resins: Influence of the Styrene Concentration on the Miscibility and Mechanical Properties, Polymer 41, pages 765-769, 2000.

CONCLUSION Based on the statements and representations contained in the documents referenced above, and the conditions listed above, the staff concludes that the Model No. F-522 package meets the requirements of IAEA SSR-6, 2012 Edition (IAEA, 2012a).

Issued with letter to R. Boyle, U. S. Department of Transportation, on 7/26/18.