12 research outputs found
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Potential mechanisms for corrosion and stress corrosion cracking failure of 3013 storage containers composed of 316 stainless steel
The degradation of 316 stainless steel (SS) storage container materials is a potential problem for radioactive waste disposition. Container materials will be exposed to significant ionizing radiation, elevated temperatures, embrittling and/or alloying agents (e.g., gallium), chloride-containing compounds (as much as 20 wt% Cl or Cl{sup {minus}}), oxidizing compounds, and a limited quantity of moisture. Additionally, containers will contain welds that have heterogeneous composition due to solute segregation and that may retain significant residual stress. All of the above-listed environmental and material conditions have been shown to be deleterious to material integrity under certain conditions. Unfortunately, the precise conditions within each container and environment is unknown and may vary widely from container to container. Thus, no single test or set of tests will be able mimic the broad range of storage container conditions. Additionally, material behavior cannot be predicted because the synergistic effects of temperature, time, chloride, moisture, sensitization, weldments, salt formation, etc., have not been fully studied. The complexity and uncertainty of storage conditions precludes any detailed recommendations. This document attempts to detail selected previous studies and to suggest some general guidelines for storage of radioactive waste. Because of the voluminous research in this area, this review cannot be considered to be comprehensive. Readers are directed to references that contain detailed reviews of particular processes for more information. Note that the effect of gallium on the degradation of SS storage containers has been discussed elsewhere and will not be discussed here
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ENVIRONMENTALLY-ASSISTED CRACKING OF 316 L STAINLESS STEEL EXPOSED TO GALLIUM
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The United States pit disassembly and conversion project -- Meeting the MOX fuel specification
The US is actively involved in demonstrating the disassembly of nuclear weapons pits to an unclassified form readied for disposition. The MOX option is the most likely path forward for plutonium that originated from nuclear weapon pits. The US demonstration line for pit disassembly and conversion is known as ARIES, the advanced recovery and integrated extraction system. The ARIES demonstration line is being used to gather data in an integrated fashion of the technologies needed for pit disassembly and conversion. These activities include the following modules: pit bisection, hydride-dehydride, oxide conversion, canning, electrolytic decontamination, and nondestructive assay (NDA). Pit bisection swages in a pit in half. Hydride-dehydride converts the pit plutonium metal to an unclassified metal button. To convert the plutonium metal to an oxide the US is investigating a number of options. The primary oxide conversion approach involves variations of combining plutonium hydriding and subsequent oxidation. Another approach is to simply oxidize the metal under controlled conditions-direct metal oxidation (DMO). To remove the gallium from the plutonium oxide, a thermal distillation approach is being used. These pyrochemical approaches will substantially reduce the wastes produced for oxide conversion of weapon plutonium, compared to traditional aqueous processing. The packaging of either the plutonium metal or oxide to long term storage criteria involves the canning and electrolytic decontamination modules. The NDA suite of instruments is then used to assay the material in the containers, which enables international verification without the need to open the containers and repackage them. All of these processes are described
The microstructural, mechanical, and fracture properties of austenitic stainless steel alloyed with gallium
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An assessment of the validity of cerium oxide as a surrogate for plutonium oxide gallium removal studies
Methods for purifying plutonium metal have long been established. These methods use acid solutions to dissolve and concentrate the metal. However, these methods can produce significant mixed waste, that is, waste containing both radioactive and chemical hazards. The volume of waste produced from the aqueous purification of thousands of weapons would be expensive to treat and dispose. Therefore, a dry method of purification is highly desirable. Recently, a dry gallium removal research program commenced. Based on initial calculations, it appeared that a particular form of gallium (gallium suboxide, Ga{sub 2}O) could be evaporated from plutonium oxide in the presence of a reducing agent, such as small amounts of hydrogen dry gas within an inert environment. Initial tests using ceria-based material (as a surrogate for PuO{sub 2}) showed that thermally-induced gallium removal (TIGR) from small samples (on the order of one gram) was indeed viable. Because of the expense and difficulty of optimizing TIGR from plutonium dioxide, TIGR optimization tests using ceria have continued. This document details the relationship between the ceria surrogate tests and those conducted using plutonia