Three head-end processes that culminate in decontamination and recovery of uranium and plutonium by solvent extraction were evaluated on a laboratory scale, with unirradiated UC and with UC and UC-PuC fuel specimens irradiated to burnups of up to 20 000 MWd/t. The most promising process was reaction with air-free steam (pyrohydrolysis) at 750°C followed by dissolution of the resulting oxide (UO2 or UO2-PuO2) in nitric acid. Cesium was the principal fission product volatilized, but the amount was very low (about 0.5%). The oxide and fission products were dissolved in 6.5 M HNO3, yielding solutions suitable as feeds for Purex solvent extraction. Uranium and plutonium recoveries were greater than 99.9% in batch extraction tests, being separated from fission products by a factor of at least 104. An alternative but less desirable process was direct dissolution in 13 M HNO3 followed by partial oxidation with acid permanganate of the soluble organic species formed. Plutonium losses of up to 0.4% occurred when the uranium and plutonium were stripped with dilute nitric acid after solvent extraction. Reaction of the carbides with water followed by dissolution of the oxides in nitric acid was an attractive process when tested with unirradiated materials, but this scheme is not feasible for irradiated carbides since they are relatively inert to boiling water.