The dissolution behavior and fission-product release from irradiated thoria-urania fuel was studied by immersing fuel samples in J-13 well water at 90°C. The samples are from the Shippingport Light Water Breeder Reactor and consist of binary solid solutions of (U,Th)O2 with UO2 contents varying from 2.0 to 5.2 Wt.%. The post-irradiation U isotopic composition of the samples used in our experiments is: 87.3% 233U, 10.4% 234U, 1.8% 235U, and <0.5% 238U, 236U, 232U. Burn up values for the samples range from 22.3 to 40.9 megawatt-days per kg-metal. Our tests were performed on polished disks and on crushed and sieved samples in stainless-steel reaction vessels with air-filled head-space. After 196 days of reaction, samples showed no evidence for corrosion at the micrometer scale. Concentration ranges (μgL-1) of key radionuclides in filtered (∼5 nm pore size) leachates were: 0.1 – 15 90Sr, 0.9 – 7.0 99Tc, 0.1 – 35.2 137Cs,<0.2 – 0.8 233U, <0.1 – 0.7 232Th. Concentrations of 237Np, 239Pu, 240Pu and 241Am were all <0.2 μgL-1. The relatively high concentrations of the fission products 90Sr and 137Cs occur early during leaching and decrease for later samplings. Matrix dissolution rates for the irradiated thoria-urania samples range from ∼3x10-3 to <3×10-5 mg m-2day-1 and are at least two orders of magnitude lower than those measured for UO2 spent fuels under similar experimental conditions.