Active Standard ASTM C1431 | Developed by Subcommittee: C26.13
Book of Standards Volume: 12.01
Historical (view previous versions of standard)
Significance and Use
Disposition of aluminum-based spent nuclear fuel will involve:
Removal from the existing storage or transfer facility,
Characterization or treatment, or both, of the fuel or the resulting waste form, or both,
Placement of the waste form in a canister,
Placement of the canister in a safe and environmentally sound interim storage facility,
Removal from the interim storage facility and transport to the repository,
placement in a waste container,
Emplacement in the repository, and
Repository closure and geologic disposal. Actions in each of these steps may significantly impact the success of any subsequent step.
Aluminum-based spent nuclear fuel and the aluminum-based waste forms display physical and chemical characteristics that differ significantly from the characteristics of commercial nuclear fuels and from high level radioactive waste glasses. For example, some are highly enriched and most have heterogeneous microstructures that include very small, uranium-rich particles. The impact of this difference on repository performance must be evaluated and understood.
The U.S. Nuclear Regulatory Commission has licensing authority over public domain transportation and repository disposal (and most of the interim dry storage) of spent nuclear fuels and high-level radioactive waste under the requirements established by 10 CFR 60, 10 CFR 71, and 10 CFR 72. These requirements outline specific information needs that should be met through test protocols, for example, those mentioned in this guide. The information developed from the tests described in this guide is not meant to be comprehensive. However, the tests discussed here will provide corrosion property data to support the following information needs.
A knowledge of the solubility, leaching, oxidation/reduction reactions, and corrosion of the waste form constituents in/by the repository environment (dry air, moist air, and repository relevant water) (see 10 CFR 60.112 and 135).
A knowledge of the effects of radiolysis and temperature on the oxidation, corrosion, and leaching behavior (see 10 CFR 60.135).
A knowledge of the temperature dependence of the solubility of waste form constituents plus oxidation and corrosion products (see 10 CFR 60.135).
Information from laboratory experiments or technical analyses, or both, about time dependence of the internal condition of the waste package (see 10 CFR 60.143 and 10 CFR 72.76).
Laboratory demonstrations of the effects of the electrochemical differences between the aluminum-based waste form and the candidate packaging materials on galvanic corrosion (see 10 CFR 71.43) or the significance of electrical contact between the waste form and the packaging materials on items outlined in 4.3.1-4.3.4 (see 10 CFR 60.135), or both.
Information on the risk involved in the receipt, handling, packaging, storage, and retrieval of the waste forms (see 10 CFR 72.3).
Information on the physical and chemical condition of the waste form upon repository placement so that items outlined in 4.3.1-4.3.4 can be evaluated (see 10 CFR 60.135).
Knowledge of the degradation of the waste form during interim storage so that operational safety problems with respect to its removal from storage can be assessed, if such removal is necessary (see 10 CFR 72.123).
Knowledge of the condition of the waste form prior to repository placement so that items outlined in 4.3.1-4.3.4 are properly addressed (see 10 CFR 60.135).
Conditions expected during each stage of the disposition process must be addressed. Exposure conditions anticipated over the interim storage through geologic disposition periods include dry and moist air, and aqueous environments. The air environments are associated with interim storage and the early stages of repository storage while the aqueous environments arise after water intrusion into the repository has caused corrosion-induced failure of the waste package.
1.1 This guide covers corrosion testing of aluminum-based spent nuclear fuel in support of geologic repository disposal (per the requirements in 10 CFR 60 and 40CFR191). The testing described in this document is designed to provide data for analysis of the chemical stability and radionuclide release behavior of aluminum-based waste forms produced from aluminum-based spent nuclear fuels. The data and analyses from the corrosion testing will support the technical basis for inclusion of aluminum-based spent nuclear fuels in the repository source term. Interim storage and transportation of the spent fuel will precede geologic disposal; therefore, reference is also made to the requirements for interim storage (per 10 CFR 72) and transportation (per 10 CFR 71). The analyses that will be based on the data developed are also necessary to support the safety analyses reports (SARs) and performance assessments (PAs) for disposal systems.
1.2 Spent nuclear fuel that is not reprocessed must be safely managed prior to transportation to, and disposal in, a geologic repository. Placement in an interim storage facility may include direct placement of the irradiated fuel or treatment of the fuel prior to placement, or both. The aluminum-based waste forms may be required to be ready for geologic disposal, or road ready, prior to placement in extended interim storage. Interim storage facilities, in the United States, handle fuel from civilian commercial power reactors, defense nuclear materials production reactors, and research reactors. The research reactors include both foreign and domestic reactors. The aluminum-based fuels in the spent fuel inventory in the United States are primarily from defense reactors and from foreign and domestic research reactors. The aluminum-based spent fuel inventory includes several different fuel forms and levels of 235U enrichment. Highly enriched fuels (235U enrichment levels >20 %) are part of this inventory.
1.3 Knowledge of the corrosion behavior of aluminum-based spent nuclear fuels is required to ensure safety and to support licensing or other approval activities, or both, necessary for disposal in a geologic repository. The response of the aluminum-based spent nuclear fuel waste form(s) to disposal environments must be established for configuration-safety analyses, criticality analyses, PAs, and other analyses required to assess storage, treatment, transportation, and disposal of spent nuclear fuels. This is particularly important for the highly enriched, aluminum-based spent nuclear fuels. The test protocols described in this guide are designed to establish material response under the repository-relevant conditions.
1.4 The majority of the aluminum-based spent nuclear fuels are aluminum clad, aluminum-uranium alloys. The aluminum-uranium alloy typically consists of uranium aluminide particles dispersed in an aluminum matrix. Other aluminum-based fuels include dispersions of uranium oxide, uranium silicide, or uranium carbide particles in an aluminum matrix. These particles, including the aluminides, are generally cathodic to the aluminum matrix. Selective leaching of the aluminum in the exposure environment may provide a mechanism for redistribution and relocation of the uranium-rich particles. Particle redistribution tendencies will depend on the nature of the aluminum corrosion processes and the size, shape, distribution and relative reactivity of the uranium-rich particles. Interpretation of test data will require an understanding of the material behavior. This understanding will enable evaluation of the design and configuration of the waste package to ensure that unfilled regions in the waste package do not provide sites for the relocation of the uranium-rich particles into nuclear critical configurations. Test samples must be evaluated, prior to testing, to ensure that the size and shape of the uranium-rich particles in the test samples are representative of the particles in the waste form being evaluated.
1.5 The use of the data obtained by the testing described in this guide will be optimized to the extent the samples mimic the condition of the waste form during actual repository exposure. The use of Practice C1174 is recommended for guidance. The selection of test samples, which may be unaged or artificially aged, should ensure that the test samples and conditions bound the waste form/repository conditions. The test procedures should carefully describe any artificial aging treatment used in the test program and explain why that treatment was selected.
2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.
C1174 Practice for Prediction of the Long-Term Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geological Disposal of High-Level Radioactive Waste
Government Documents10CFR72 US Code of Federal Regulations Title 10, Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear and High-Level Radioactive Waste
ICS Number Code 27.120.30 (Fissile materials and nuclear fuel technology)
UNSPSC Code 15131500(Nuclear fuel)