[en] Spent fuel is being considered as a waste form for disposal in a repository located in salt. To adequately model spent fuel performance as a waste form that may be contacted by brine in a repository, it is necessary to describe the leach (dissolution) behavior of spent fuel and the chemical interactions of the released radionuclides with their environment. To this end, leach tests were conducted on: UO₂ in Permian Basin salt brine or deionized water at test temperatures of 25, 75, and 150⁰C. Some tests were done in the presence of ductile cast iron, which is a representative overpack material, and/or oxidized Zircaloy, which is the usual fuel cladding material. Spent fuel (H.B. Robinson) in Permian Basin salt brine at 25 and 75⁰C. Some of the tests were conducted in the presence of ductile cast iron. The release values for leach periods up to 60 days were determined for systems utilizing both UO₂ and spent fuel. This report is based upon data obtained during 1982 and 1983. The larger temperature dependence of the leach behavior that was observed for deionized water than was observed for brine is difficult to interpret on the basis of our knowledge. Differences in UO₂ alteration products may be involved. For example, the existence of sodium uranates is likely in the brines but not possible with deionized water. Differences in ionic strength may also play a role. Observations derived from the leach tests performed in brine include: (1) the presence of iron coupons had no effect on total release of uranium from either spent fuel or UO₂ but did reduce solution concentrations, and (2) 100 to 200 times more uranium was released from spent fuel than from UO₂ per unit of geometric surface area