Titanosilicates, layered manganese oxides and an antimonysilicate were tested for their ability to take up the reprocessing spent fuel transuranium isotopes americium-241 and plutonium-236. This was performed in different concentrations of acid, and sodium and calcium nitrate solutions to represent the conditions of nuclear waste treatment. The purpose was to assess the waste types (acidic or low-medium-high-salt) most suited to treatment by these materials. The experiments were carried out by the batch method and the results obtained expressed as distribution coefficients. Variation in the magnitude and mechanism of actinide sorption by the materials was attributed to the differences in the frameworks, physical properties (porosity, crystal size and chemical composition) and surface charges of the ion exchangers. In nitric acid solutions, the H-antimonysilicate proved to be a better sorbent than H-birnessite for both actinide elements. The void space of AM-4 layered titanosilicate material made it appropriate for actinide removal in sodium salt solutions. The differences in selectivity of Ca-antimonysilicate and Ca-birnessite for americium-241 and plutonium-236 in 0.01 M calcium nitrate solution reflected the preference of each material for actinides over calcium ions.