Certification of Packages for Transportation of Unirradiated Material
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Interagency Transportation Coordination
The NRC and the U.S. Department of Transportation (DOT) co-regulate the transportation of radioactive material. The two agencies delineated their respective responsibilities in a memorandum of understanding signed July 2, 1979. The DOT regulates packages for small quantities of radioactive material (Type A quantities and less) whereas the NRC regulates package approval for Type B and fissile material packages. The threshold for Type A quantities of radioactive material are defined for numerous radioisotopes in Appendix A to 10 CFR Part 71, Table A-1. DOT also regulates carriers and international transport, as the U.S. Competent Authority.
Transportation Package Requirements
A transportation package consists of the protective overpack, the container holding the fissile material or radioactive material, and the radioactive material or fissile material to be transported. The NRC reviews applications for package designs. If the package design meets NRC requirements, the NRC typically issues a Radioactive Material Package Certificate of Compliance (CoC) (see Nuclear Materials Transportation Package Certification for more information) to the organization requesting approval of a package, with the organization becoming a CoC holder. NRC licensees are authorized to ship radioactive material in a package approved for use under the general license provisions of 10 CFR 71.17 (see Shipping Requirements for more information).
While the accident tolerant fuel (ATF) assemblies may not look any different than conventional fuel containing uranium dioxide (UO2) pellets surrounded by zirconium (Zr)-alloy cladding, there are several differences that are being considered for ATF that will make transportation certification different. These differences are increased enrichment above 5 weight percent uranium-235 (U-235), different pellet material other than UO2, and different cladding material other than Zr-alloy.
Typically, when fuel vendors (who are typically the CoC holders) are developing new fuel assembly designs, they request a limited approval to transport lead test assemblies (LTAs) in transportation packages from the fabrication facilities to reactors. After sufficient post-irradiation tests of the LTAs are completed and the fuel data is analyzed, the fuel assembly design is finalized by the vendor. Subsequently, the industry prepares for the batch loading of ATF into the reactor, prompting fuel assembly vendors to request approval of transportation packages that allow large-scale (i.e., batch) shipment of unirradiated ATF assemblies.
Typically, the approval for the transport of LTAs is through a letter authorization, rather than revising the package's CoC. Letter authorizations are documents that authorize a licensee to transport a limited quantity of material (i.e., a limited number of fuel assemblies) for a short duration. The NRC reviews letter authorizations against the same regulations and gives the application the same level of scrutiny as a CoC. Once a fuel design has been finalized for batch loading, the CoC holder will request revision of the CoC for the new fuel assembly design. The CoC typically does not limit the number of shipments a licensee can make.
Packages for unirradiated fuel assemblies typically rely on the fuel assembly structural components (cladding, top and bottom nozzles, and grid straps) to maintain the fuel assembly structure during the evaluation of the tests for normal conditions of transport in 10 CFR 71.71 and hypothetical accident conditions in 10 CFR 71.73. Maintaining fuel assembly geometry under all conditions is important for criticality safety during transportation.
Different pellet material
In evaluating different pellet material other than conventional UO2 pellets, one of the key evaluation items is material interactions between the pellet and cladding. The NRC reviews for unfavorable material interactions that could degrade the performance of the cladding. If the NRC finds that the degradation of the fuel cladding during normal conditions of transport or hypothetical accident conditions may cause failure of the package, then additional evaluations, design changes, or restrictions on the package use may be necessary.
Different cladding material
With different cladding materials, knowing the mechanical properties for the new cladding becomes vitally important in assessing the structural performance of new fuels during hypothetical transportation accident conditions. If the new cladding material properties are not provided, an applicant may not be able to show that the fuel assembly structure will remain intact after hypothetical accident conditions tests and, therefore, the applicant may need to consider potential reconfiguration of the fuel assembly.
When considering the safe transportation of material for unirradiated material, an important technical issue associated with increased enrichment is maintenance of nuclear criticality safety for UF6 feed material and fresh fuel assemblies.
Currently for fuel enriched to 5 weight percent or less U-235, the feed material is transported from the enrichment facility as UF6 in 30-inch-diameter cylinders and deconverted from UF6 into UO2 at the fuel fabrication facility. These 30-inch cylinders hold just over 5,000 pounds of UF6 and are placed in a protective overpack. The package is tested and analyzed for the hypothetical accident condition tests described in 10 CFR 71.73. The CoC holders have shown that after the sequence of tests (30-foot drop, 40 inch puncture, and 30-minute fire test) the cylinders remain leak tight (essentially no radioactive material escapes and no water can enter the cylinder). Since the cylinders are shown to be leaktight, they do not have to be evaluated with water per 10 CFR 71.55(g), so long as UF6 enriched to a maximum of 5 weight percent U-235 and they meet the other requirements as well.
Packages for feed material enriched above 5 weight percent U-235 that will be transported as UF6 will have to either be evaluated with water inside the cylinder or, applicants and licensees will need to request an exemption.
In addition to challenges for approval of transport of UF6 at greater than 5 weight percent U-235, non-NRC regulatory requirements present challenges to the vendors. One non-NRC applicable regulation is the DOT regulation in Title 49 of the Code of Federal Regulations (49 CFR) 173.420, "Uranium hexafluoride (fissile, fissile excepted and non-fissile),"that states UF6 packaging (whether fissile, fissile excepted, or non-fissile) must be designed, fabricated, inspected, tested and marked in accordance with the American National Standard N14.1 that was in effect at the time the packaging was manufactured. American National Standards Institute (ANSI) N14.1, "Nuclear Materials – Uranium Hexafluoride Packagings Transport," only applies to enrichments up to 5 weight percent U-235 for the 30-inch-diameter cylinders. In addition to NRC regulations, transportation packages must also meet all other applicable federal regulations. Further, DOT regulation 49 CFR 173.417, "Authorized fissile materials packages," provides requirements for shipment of UF6 heels without a protective overpack and also limits the enrichment of 30-inch-diameter cylinders to 5 weight percent U-235. In addition to an NRC approval for shipment in a package using a 30-inch cylinder, CoC holders will require a special permit from DOT to transport increased enrichment UF6.
Fresh Fuel Assemblies
Criticality analyses for fresh fuel assemblies with enrichments greater than 5 weight percent U-235 that rely on intact fuel assembly structure under accident conditions may not change, as discussed above. However, evaluating the accuracy of the codes used to determine whether the package/shipment will maintain criticality safety (called benchmarking analyses) for fissile material enriched to greater than 5 weight percent U-235 presents a challenge due to the limited number of critical experiments in that range. Applicants for package approval could potentially address this challenge by:
- performing new critical experiments to validate criticality calculations for 5 to 8 weight percent U-235,
- relying on sensitivity and/or uncertainty analysis methods to determine whether existing critical experiments are applicable to enrichments above 5 weight percent U-235 (i.e., extrapolation beyond the enrichment range of applicability),
- adding an appropriate margin to account for lack of sufficient validation on new materials and higher enrichments, or
- using some combination of the above options.