Archived Web Page: Draft Regulatory Document RD-364Joint Canada - United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages

4.0 CONTAINMENT

This section of the application should identify the package containment system and describe how the package complies with the containment requirements of 10 CFR 71.43(f) and 10 CFR 71.51 or Paragraphs 646, 656, 658, and 659 of TS-R-1 which are incorporated in Subsection 1(1) of the PTNS Regulations by reference to Paragraph 650 of TS-R-1.

The section should address the structural and thermal effects on the packaging and its content under normal and hypothetical accident conditions and their effects on the containment system of the package. Any operational, fabrication, and maintenance requirements with respect to containment for the package should be included in the application in Section 7, Package Operations, and Section 8, Acceptance Tests and Maintenance Program.

4.1 Description of the Containment System

This section should define and describe the containment system. The containment boundary of the package should be explicitly identified, including the containment vessel, welds, drain or fill ports, valves, seals, test ports, pressure relief devices, lids, cover plates, and other closure devices. If multiple seals are used for a single closure, this section should identify the seal defined as the containment system seal. Detailed drawings of the containment system should be included.

Packaging design features important for containment include the following:

1. Materials of construction of the containment system;
2. Welds;
3. Applicable codes and standards (e.g., ASME code specifications for the vessel);
4. Bolt torque required to maintain positive closure;
5. Maximum and minimum allowable temperatures of components, including seals; and
6. Maximum and minimum temperatures of components under the tests for normal conditions and hypothetical accident conditions of transport.

All containment boundary penetrations and their method of closure should be adequately described. Performance specifications for components such as valves, O-rings, and pressure relief devices should be identified and documented; no device may allow continuous venting.

The containment evaluation should show that compliance with the containment requirements does not rely on any filter or mechanical cooling system, as specified in 10 CFR 71.51(c) or Paragraph 659 of TS-R-1 which is incorporated in Subsection 1(1) of the PTNS Regulations by reference to Paragraph 650 of TS-R-1.

If the design includes valves or similar devices, the application should demonstrate that these are protected against unauthorized operation and, except for a pressure relief valve, have an enclosure to retain any leakage.

Demonstration that no galvanic, chemical, or other reactions will occur between the seal and the packaging or its contents, and that the seal will not degrade due to irradiation, should be addressed. If penetrations are closed with two seals, specification of which of the two seals is defined as the containment boundary should be addressed.

Specifications of the seal grooves and the type and size of seals should be provided. The temperature of containment boundary seals should be shown to remain within the specified allowable limits under both normal conditions and hypothetical accident conditions of transport.

Demonstration of how the containment system is securely closed with a positive fastening device that cannot be opened unintentionally or by pressure that may arise within the package and a description of the features that ensure that continuous venting is precluded should be included.

Scale model testing is not a reliable or acceptable method for qualifying the leakage rate of a full-scale package. If compliance is demonstrated by analysis, the structural evaluation should show that the containment boundary, seal region, and closure bolts do not undergo any inelastic deformation and that the materials of the containment system (e.g., seals) do not exceed their maximum allowable temperature limits.

For SNF packages, the material used for the containment system and the design, fabrication, examination, testing, inspection, and certification should be in accordance with Section III, Division 3, of the ASME Boiler and Pressure Vessel Code [6]. This includes an agreement with an authorized inspection agency to provide inspection and audit services for the design owners, packaging owners, and Class W certificate holders. Justification for the use of other codes should be provided in the application. Also, the codes, standards, and criteria for the inner containment system should generally be the same as those of the outer containment system. Justification for differences should be presented in the application.

4.1.1 Special Requirements for Damaged Spent Nuclear Fuel

The determination of the fuel condition should be based, as a minimum, on review of fuel records. Damaged fuel may consist of fuel assemblies with either cladding or structural defects. Fuel with damaged cladding should be contained to facilitate handling and to confine gross fuel particles to a known subcritical configuration under normal and hypothetical accident conditions. Use of a canister for the fuel may be an option for consideration. The application should include justification for the material specifications and the design/fabrication criteria for the can. These specifications and criteria should be the same as those for containment or criticality support structures.

4.2 Containment under Normal Conditions of Transport

This section should include the evaluation of the containment system under normal conditions of transport, using the methods in American National Standards Institute (ANSI) N14.5-1997, American National Standard for Radioactive Materials—Leakage Tests on Packages for Shipment [1], or International Standard Organization (ISO) 12807, Safe Transport of Radioactive Materials—Leakage Testing on Packages [11]. This section should demonstrate that the package meets the containment requirements of 10 CFR 71.51(a)(1) or Paragraph 656(a) of TS-R-1 which is incorporated in Subsection 1(1) of the PTNS Regulations by reference to Paragraph 650 of TS-R-1 under normal conditions of transport. The evaluation should be performed for the most limiting chemical and physical forms of the contents. Significant daughter products should be included. The constituents of the releasable source term, including radioactive gases, liquids, and powder aerosols, should be identified. If less than 100% of the contents are considered releasable, the justification for the lower fraction should be included. The containment evaluation should not rely on the blockage of a leakage path by particulate contents to meet the containment criteria in the regulations. Any seal demonstrating a leakage rate of 1x10^-7 reference-cm³/s, as defined in ANSI N14.5, may be considered to be leak-tight.

The evaluation under normal condition of transport should include the following:

1. The maximum internal pressures, including any gases generated in the package during a period of one year;
2. The structural performance of the containment system, including seals, closure bolts, and penetrations; and
3. The leak testing of the containment system.

Combustible gases should not exceed 5% (by volume) of the free gas volume in any confined region of the package. No credit should be taken for getters, catalysts, or other recombination devices.

For Type A fissile material packages, the evaluation should show that there is no loss or dispersal of radioactive material under normal conditions of transport, as specified in 10 CFR 71.43(f) or Paragraph 646 of TS-R-1 which is incorporated in Subsection 1(1) of the PTNS Regulations by reference to Paragraph 650 of TS-R-1. For Type B packages, the evaluation should show that there is no release under normal conditions of transport to the required sensitivity. In both cases, there should be no significant increase in external radiation levels.

For SNF packages, the releasable source term is composed of crud (surface contamination) on the outside of the fuel rod cladding that can become aerosolized and the fuel fines, volatiles, and gases that are released from a fuel rod in the event of a cladding breach. Bounding values for the effective surface activity density in Bq/cm² (Ci/cm²) of the crud on fuel rod cladding are based on experimental determinations. A computer code, such as ORIGEN-S, can be used to identify the radionuclides present for a given percent fuel enrichment, burnup, and cooldown time. Using the individual A₂ values for the crud, fines, gases, and volatiles individually, the effective A₂ of the releasable source term mixture can be determined by using the relative release fraction for each contributor and the methods from ANSI N14.5 [1] or ISO 12807 [11]. The release fractions and effective specific activities for the various releasable source term contributors for SNF with an initial enrichment of 3.2%, a burnup of 33 gigawatt-days per metric tonne of initial heavy metal (GWD/MTIHM), and a cool-down time of 5 years are given in Table 4-1. The release fractions presented in Table 4-1 have been developed from reasoned argument and experimental data (NUREG/CR-6487, Containment Analysis for Type B Packages Used to Transport Various Contents [17]). These values may be considered as default values for light water reactor SNF with a burnup below 45 GWD/MTIHM. The release fractions and specific activities should be justified in the application as appropriate.

Table 4-1: Release Fractions and Specific Activities for the Contributors to the Releasable Source Term for Packages Designed to Transport Irradiated Fuel from Commercial Pressurized and Boiling-Water (PWR and BWR) Reactors for Burnups below 45 GWD/MTIHM

Variable	PWR		BWR
	Normal Conditions of Transport	Hypothetical Accident Conditions	Normal Conditions of Transport	Hypothetical Accident Conditions
Fraction of crud that spalls off rods, fC	0.15	1.0	0.15	1.0
Crud surface activity, SC (Ci/cm2)*	140 x 10-6	140 x 10-6	1254 x 10-6	1254 x 10-6
Mass fraction of fuel that is released as fines due to a cladding breach, fF	3x10-5	3x10-5	3x10-5	3x10-5
Specific activity of fuel rods, AR (Ci/g)	0.60	0.60	0.51	0.51
Fraction of rods that develop cladding breaches, fB	0.03	1.0	0.03	1.0
Fraction of gases that are released due to a cladding breach, fG	0.3	0.3	0.3	0.3
Specific activity of gas in fuel rod, AG (Ci/g)	7.32 x 10-3	7.32 x 10-3	6.28 x 10-3	6.28 x 10-3
Specific activity of volatiles in a fuel rod, AV (Ci/g)	0.1375	0.1375	0.1794	0.1794
Fraction of volatiles that are released due to a cladding breach, fV	2 x 10-4	2 x 10-4	2 x 10-4	2 x 10-4

*Values for the crud activity are for the time of reactor discharge and should be corrected for radioactive decay.

The maximum permissible release rate and the maximum permissible leakage rate should be based on the mass density, effective specific activity, and effective A₂ of the releasable source term and should be calculated in accordance with the methods specified in ANSI N14.5 [1] or ISO 12807 [11].

4.3 Containment under Hypothetical Accident Conditions

This section should include the evaluation of the containment system under hypothetical accident conditions, considering factors given in Section 4.2 above. This section should demonstrate that the package meets the containment requirements of 10 CFR 71.51(a)(2) or Paragraph 656(b) of TS-R-1 which is incorporated in Subsection 1(1) of the PTNS Regulations by reference to Paragraph 650 of TS-R-1 under hypothetical accident conditions. In particular, the structural performance of the containment system should be addressed, including seals, closure bolts, and penetrations, as well as leakage testing of the containment system. The evaluation should consider differences relevant to the accident conditions (e.g., pressurization of the containment system under fire test conditions, a possible increase in the releasable source term, and possible changes in containment system performance due to package damage).

4.4 Leakage Rate Tests for Type B Packages

This section should describe leakage tests that are used to show that the package meets the containment requirements of 10 CFR 71.51 or Paragraph 656 of TS-R-1 which is incorporated in Subsection 1(1) of the PTNS Regulations by reference to Paragraph 650 of TS-R-1. These may include the following:

1. Fabrication leakage rate test;
2. Maintenance leakage rate test;
3. Periodic leakage rate test; and
4. Pre-shipment leakage rate test.

Fabrication, maintenance, and periodic leakage rate tests should be included in Section 8, Acceptance Tests and Maintenance Program, of the application. The pre-shipment leakage rate test for assembly verification should be included in Section 7, Operating Procedures, of the application..

Sample analyses for determining containment criteria for Type B packages are provided in NUREG/CR-6487. If these analyses are used, demonstration that the assumptions used in NUREG/CR-6487 are applicable to the package under consideration is required.

Methods for leak testing of all containment seals and penetrations, including drain and vent ports, should be described. If fill, drain, or test ports use quick-disconnect valves, demonstration that these do not preclude leak testing of the containment seals should be provided. The maximum allowable leakage rate and the minimum test sensitivity should be specified for each type of test (e.g., fabrication, maintenance, periodic, and pre-shipment tests).

A method to determine the maximum permissible volumetric leakage rates based on the allowed regulatory release rates under both normal conditions of transport and hypothetical accident conditions can be found in ANSI N14.5. The smaller of these air leakage rates is defined as the reference air leakage rate. ISO 12807 also presents a way of calculating the release rate.

4.5 Appendix

The appendix should include a list of references, applicable pages from referenced documents, supporting information and analysis, test results, and other appropriate supplemental information.